Text

Exam Rept 50-271/88-01OL on 880119-20.Exam Results:All Reactor Operator Candidates Passed Exams
	ML20196H656
Person / Time
Site:	Vermont Yankee File:NorthStar Vermont Yankee icon.png
Issue date:	02/29/1988
From:	Lange D, Lumb T; NRC OFFICE OF INSPECTION & ENFORCEMENT (IE REGION I)
To:	;
Shared Package
ML20196H652	List: IR 05000271/1988001; ML20196H650;
References
	50-271-88-01OL, 50-271-88-1OL, NUDOCS 8803140009
	Download: ML20196H656 (71)
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U.S. NUCLEAR REGULATORY CORMISSION REGION I OPERATOR LICENSING EXAMINATION REPORT EXAMINATION REPORT NO. 83-01 (OL) FACILITY DOCKET NO. 50-271 FACILITY LICENSE NO. ORP-28 LICENSEE: Vermont Yankee Nuclear Power Corporation RD 5, Rox 169 Ferry Road Brattleboro, Vermont 05301 FACILITY: Vermont Yankee EXAMINATION DATES: January 19-20, 1988

CHIEF EXAMINER: - d A. Lumb, Senior Operations Engineer _ A 29 -W

    ~Date APPROVED BY:  %__  .2 .21 - P P David J. Lancje, Chief ? BWR Section, Date Operations Branch, Division of Reactor Safety SUMMARY: Written examinations were administered to two (2) reactor operator (RO) candidates and operating tests were administered to two (2)

reactor operator (RO) candidates. All candidates passed the examinations, f,[[I$0bOf: h k 71 y DCD

. . DETAILS TYPE OF EXAMINATIONS: Replacement EXAMINATION RESULTS: l R0 l l Pass / Fail l l l 1 I I l Written 1 2/0 l l l l l l l l Operating i 2/0 l l l l l l 1 l0verall l 4/0 l l I I 1. CHIEF EXAMINER AT SITE: T. Lumb, Senior Operations Engineer 2. OTHER EXAMINERS: D. Lange, Chief, BWR Section K. Brockman, Chief, Operator Licensing Section 2, Region II 3. Due to the number of candidates there were no generic strengths or deficiencies noted on the operating tests administered January 20, 1988.

4. The following is a summary of generic strengths and deficiencies noted from the grading of the written examinations. This information is be:ng provided to aid the licensee in upgrading license and requalification training programs. No licensee response is required.

STRENGTHS a. Understanding of plant condition effects on Critical Power - Question 1.04 b. Understanding of the effects of condenser vacuum ircrease on efficiency - Question 1.07 c. Knowledge of thermal limits - Question 1.08 d. Understanding of the effects of Control Rods on reactor power - Question 1.09 e. Knowledge of Backup Scram Valve operation - Question 2.01

-      i f. Diagnosis of Recire Pump seal failures - Question 2.04 g. Knowledge of Main Generator instrumentation and controls -

Question 2.05 h. Knowledge of Feedwater pump start interlocks - Question 2.09 1. Knowledge of the Tip Shear Valve System - Question 3.02 j. Understanding of the operational implications of a Group I isolation

- Question 3.05
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k. Knowledge of the immediate actions for a Shutdown Using Alternate Shutdown Methods - Question 4.01 1. Understanding of the bases for steps in the Recirculation Pump Trip l Procedure - Question 4.02 DEFICIENCIES a. Understanding of Control Rod Worth changes during a startup - Question 1.02 b. Understanding of the effects of vessel parameters on Recirc pump NPSH - Question 1.05 c. Understanding of the Heat Balance equation - Question 1.06 d. Knowledge of APRM functions and alarms - Question 3.10 5. Personnel Present at Exit Interview, January 20, 1988: NRC Personnel D. Lange, Chief BWR Section, DRS K. Brockman, Chief, Operator Licensing Section 2, Region II G. Grant, Senior Resident Inspector

Facility Persoi.nel J. Pelletier, Plant Manager , R. Spinney, Training Manager E. Lindamood, Operations Training Supervisor R. Devercelly, Operations Instructor L. Cantrell, Operations Instruc;or G. Johnson, Operations Supervisor G. LeClair, Assistant Operations Supervisor ' 6. Summary of NRC Comments Made at Exit Interview: There were no delays in access to the plant. ,

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i Simulator performance was good. There was a problem with B EDG tripping when a Turbine trip caused an undervoltage spike on the grid. Vermont Yankee acknowledged the problem and intend to repair it.

' There were no generic strengths or weaknesses noted on the operating tests.

The results of the examinations would not be discussed at the exit meeting but would be contained in the Examination Report. Every effort would be made to send the candidates' results in approximately two weeks.

7. Summary of Facility Comments Made at Exit Intarview: A quality, straight forward written examination was administered.

K. Brockman was asked about the material provided for examination preparation. Mr. Brockman said that more instructor guides and notes are needed to relate the learning objectives to test items.

Attachments: 1. Written Examination and Answer Key (RO) 2. Facility Comaents on Written Examinations after Facility Review 3. NRC Response to Facility Comments 4. Simulation Facility Fidelity Report

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U. S. NUCLEAR REGULATORY COMMISSION

-  REACTOR OPERATOR LICENSE EXAMINATION       ,

FACILITY: _ygBMQNI_ YON [gE__________ ; REACTOR TYPE: _DWB-@g!_________________ ! DATE ADMINISTERED _p0f91/12________________ ' EXAMINER: _2GOCOC.^fj _Ez_3c _ majE___ CANDIDATE: __[ h M ________________ INSI69QIIgNS_IQ g80QJp$Ig1 Use separate paper for the answers. Write answers on one side only. 1 Stcple question sheet on top of the answer sheets. Points for each i quastion are indicated in parwntheses after the question. The passing I grede requires at least 70% in each category and a final grade of at least 90%. Examination papers will be picked up six (6) hours after tho examination starts.

% OF CATEGORY % OF CANDIDATE'S CATEGORY ,__YeLUE_ _IDIeb ___EG9BE___ _YeLUE__ ______________G91E90BY 2Ez99_- _2Et99 ____..______ ________ 1. PRINCIPLES OF NUCLEAR POWER PLANT OPERATION, THERMODYNAMICS, i HEAT TRANSFER AND FLUID FLOW

t t _2E199__ 29199 ___________ ________ 2. PLANT DESIGN INCLUDING SAFETY ! AND EMERGENCY SYSTEMS _2Et99__ _2Et99 ___________ ________ 3. INSTRUMENTS AND CONTROLS 1 2 I_3Et99 _ 29199 ___________ ________ 4. PROCEDURES - NORMAL, ABNORMAL, l EMERGENCY AND RADIOLOGICAL l CONTRDL i 199z99__ ___________ ________% Totals Final Grade

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All work done on this examination is my own. I have neither given 7 nor received aid.

___________________________________ Candidate's Signature

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NitC flut.ES AND 3UIDELINES FDit LICIINSE EXAMINATIONS Dur t-no the administratton of this examination the following rules applyi 1. Cheating on the examination means an automatic denial of your application cnd could result in more severe penalttes.

2. R:stroom trips are to be limited and only one candidate at a time may leave. You must avoid all contacts with anyone outside the examination j room to avoid evan the appearance or possibility of cheating. '

3. Une black ink or dark pencil gely 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 for answers.

7. Print vaur name in the upper right-hand corner of the first page of eaqb section of the answer sheet.

D. Consecutively number each answer sheet, write "End of Category __" as appropriate. start each category on a gew page. write 90 1y 90 goe 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.

I 10. Skip at least tbter lines between each answer.

l 11. S;parate answer sheets from pad and place finished answer sheets face t down on your desk or table.

l i 12. Uce abbreviations only if they are commonly used in facility litecatute.

" I 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. l 14. Show all calculations, methods, or assumptions used to obtain an answer l to mathematical problems whether indicated i s, the question or not. l

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 enamicet 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 af ter the examination has , been completed.

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' 1 11. When you completa your examtnatton, you shall:
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c. Assemble your exami nation as follows:

 (1) Exam questions on top.

(2) Exam aids - figures, tablen. 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 ar.swer l

3 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, vou are found in this area while the examination is still in progress, your license may be denied or revoked.

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1....ERINCIPLCD_QE_NUg(EA8_ POWER _P(ANI_QPgRAllON, PAGE 2 !

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i r i QUESTION 1.01 (1.00)

Which one of the f ollowing radioactive isotopes f ound in the rcactor coolant WOULD NOT indicate a leak through the fuel l cicdding? (1.0) ! a. Co - 60 i b. Xe - 133 ' c. I - 131 d. Kr - 87 I i I QUESTION 1.02 (2.50) I

A Reactor startuo i s i n progress. In WHICH ONE of the f ollowing i conditions would Control Rod Worth be the greatest? JUSTIFY , your choice. Consider the effects of moderator temperature. I local neutron flux and core average flux in your answer. (2.5) ; a. Cold Shutdown

; b. Heatup in Progress (* 1% Reactor Thermal Power)

I i l c. Heatup Complete (* 1% Reactor Thermal Power) I i

d. 50% Reactor Thermal Power i

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100% Reactor Thermal Power

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QUESTION 1.03 (2.00) Tho reactor was operated at 60% power f or one (1) week. The CRO has juct increased power f rom 60% to 90X by increasing recirc pump speed to 85% of rated speed. During the next four (4) to six (6) hours, thJ operator will decrease recirculation pump speed to mainatin power ct 90%. After approximately 40 hours, the recirculation pump speed will be greater than 85X if the reactor power is mainatined at 90%. a. WHY does the operator initially have to decrease the recirc ! pump speed to maintain power at 90%"' (1.0) b. WHY is the recirculation pump upeed approximately 49 hours r after the power was raised greater than the pump speed just l after the power was raised? (1.0) l

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QUESTION 1.04 (2.00) I o. DEFINE "Critical Power". (1.0) j b. Which one of the following conditions would INCREASE the Critical j Power level assuming all other variables remain unchanged? (1.0) )' NOTE: ASSUME NORMAL FULL-POWER OPERATING CONDITIONS

1. Inlet subcooling is DECREASED j 2. Reactor pressure is DECREASED 3. The axial power peak is RAISED l 4. Coolant flow rate is DECREASED ,

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QUESTION 1.05 (3.00) For each of the following paramater changes, indicate whether the cvcilable NPSH at the suction of the recirculation pumps would INCREASE / DECREASE / REMAIN THE SAME and JUSTIFY your choice.

NOTE: CONSIDER ONLY THE VARIED PARAMETER - DO NOT ADDRESS ANY POTENTIAL EFFECTS OF INTEGRATED PLANT RESPONSES a) The Feedwater Flow Rate is INCREASED (1.0) b) The Recirculation Flow Hate is INCREASED (1.0) c) The Vessel Pressure is INCREASED from 200 psig to 800 psig (1.0) QUESilON 1.06 (3.00) A reactor heat bal ance was perf ormed (by hand) during the 00-08 shift due to the Process Gnmputer being OOC. The GAF's were computed, but the APRM GAIN ADJUSTMENTS HAVE NOT F4EEN MADE.

STATE whether the Actual Power will be HIGHER THAN, LOWER THAN, or THE SAME AS the Calculated Power for each of the followings a. The steam flow used in the heat balance calculation was HIGHER than the actual steam flow. (0.75) b. The reactor recirculation pump heat input used in the heat balance calculation was OMITTED. (0.75) c. Tim RWCU return temperature used in the heat balance calcu-lation was LOWER than the actual RWCU return temperature. (0.75) d. One string of low pressure feedwater heaters were not in service during the conduct of the heat balance. (0.75) r (***** CATEGORY 01 CONTINUFD ON NEXT PAGE *****)

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l OUESTION 1.07 (1.50) ; Given that condenser vacuum is INCREASED from 26" Hg to 28" Hg. STATE whether the following efficienties wi11 INCREASE /DECREASC/ REMAIN THE SAME.

c. Cycle Efficiency (0.75) b. Turbine Efficiency (0.75) DUESTION 1.08 (3.00) MATCH the appropriate Thermal Limit (a-c).

a. Linear Heat Generation Rate (LHGR) (1.0) 6. Averamp Planar Linear Heat Generation Rate (APLHGR) (1.0) c. Mini Critical Power Ratio (MCPR) (1.0) to each FAIcuRE MECHANISM AND to each LIMITING CONDITION given below: FAILURE MECHANISM LIMITING CONDITION F1. Clad melting caused by L1. Coolant transition decay heat & stored heat boiling following a LOCA F2. Clad cracking from the surface L2. Clad plastic strain becoming vapor "blanketed" < 1% F3. Clad cracking caused by L3. Maximum clad temp-high stress from pellet erature of 2200 deg F expansion

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QUESTION 1.09 (2.00) Concerning the effects of Control Rods on reactor powers c. EXPLAIN how a control rod withdrawal at certain power levels could recult in a reactor power decrease (reverse power effect). (1.0) b. Which one F the following would be the rod movement sequence most likely to cause the reverse oower effect? (1.0)

(1) Deep Rod - 10 notch movement (2) Deep Red - 1 or 2 notch movement (3) Shallow Rod - 10 notch movement (4) Shallow Rod - 1 or 2 notch movement OUESTION 1.10 (2.00)

For each of the f oll owi ng transients, STATE which of the reactivity confficients (Moderator, Void, or Doppler) would respond first.

JUSTIFY your choice.

c. Rod Drop Accident from 10% thermal power. (1.0) b. M3?V Closure from 90% thermal power. (1.0) QUESTION 1.11 (2.00) Racctor power drops to approximately 7.5% instantly after a reactor ceram. Given that the delayed neutron fraction is less than 1%,- EXPLAIN why reactor power doesn't instantly drop to less than 1%. F ADDRESS THE IMPORTANCE OF DELAYED NEUTRONS AND SUBCRITICAL , MULTIPLICATION IN YOUR RESPONSE. (2.0) I l

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IUE 600 Q YNO d LGS t _UE 61_188N S EE 8_GN Q _E( U 1D _E(QW

QUESTION 1.12 (1.00) i Attached Figure #404 illustrates the "Combined Head / Pressure Curves for Two Pumps." Select from the figure the appropriate system operating point (numbered 1 through 6) for each of the f oll owing condi ti on< . c. Pumps A and B running in SERIES with the pump discharge valve (0,5) throttled shut from the initial condition.

b. Pumps A ar.d B running in PARALLEL with the pump discharge valve (0.5) fully open.

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QUESTION 2.01 (1.00) B ckup Scram valves provide a redundant means of venting air from tho scram pilot valves and scram discharge valves. These backup volves are ...(CHOOSE ONE) (1.0) a. ...normally energized and will de-energize upon a RPS Scram signal.

b. ... aligned such that two valves in series, one from each RPS trip channel, must actuate to vent the s c r a.n 6tr header.

c. ... designed such that both RPS channels must trip in order for any one of the valves to actuate.

d. ... powered from the RPS Buses A and B.

QUECTION 2.02 (3.00) The RCIC System automatically initiated and is injecting into the vessel. For each of the situations listed below. STATE whether RCIC i nj ec ti on into the reactor would CONTINUE or STOP. If injection is STOPPED, DESCRIBE any AUTOMATIC or OPERATOR ACTION that would be required to reinitiate injection.

a) The RCIC Test Bypass Valve to the CST FAILS OPEN. (0.75) b) A 125% Overspeed Trip is received due to low control oil pressure. Control oil pressure is then returned to normal. (0.75) c) After decreasing to 50 psig, RCIC Steam Line (i nl et ) pressure increases to 150 psig. (0.75) i d) After increasing to +180 inchea, Reactor Vessel Water Level decreases to +80 inches. (0.75) i r l i l (***** CATEGORY O2 CONTINUED ON NEXT PAGE *****) ,. _

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QUESTION 2.03 (2.00) Fill in the blankn to describe the main turbine's response to a SLOW OVERSPEED condition. Choose from the component actions listed below.

SPEED RESPONSE 102% a) ________ b) __________

     (1.0)

110% c) ________

     (0.5)

111.5% d) ________

     (0.5)

COMPONENT ACTIONS 1. Emergency 2. Backup Governor Overspeed Trip Trip 3. Intermediate 4. Intermediate Stop Valves Stop Valves Full Open Full Closed S. All Intercept 6. All Intercept Valves Full Valves Full Open Closed b3o k' OA% o^C c.ogomV eLdion is gdreA Or ecck, ce spons c QUESTION 2.04 (1.00) Th2 plant is operating norina11y at power when Recirc Pump A Controlled Lockage (FS "A") alarms LO (0.1 gpm). No other alarms are present.

You note an INCREASE in No.2 Recirc Pump seal pressure with NO CHANGE in No. 1 Recire Pump seal pressure. Which ONE of the following failures would cause these i ndi cati ons? NOTE: NO OTHER ALARMG ARE PRESENT. Att ached is Recirculation Pump Soc 1 Assembly Sketch - Transparency 4 - LOT 03-007 f e your>'ref erence.

a. Failure of No. 1 seal - b. Failure of No. 2 seal c. Plugging of the No. 1 internal restricting / breakdown orifice d. Plugging of the No. 2 internal restricting / breakdown orifice (***** CATEGORY O2 CONTINUED ON NEXT PAGE *****)

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QUESTION 2.05 (1.00) Whnn synchronizing the Main Generator to the grida c. TRUE or FALSE? The synchroscope circuitry provides an INTERLOCK to preclude the Output Breaker from closing, unless the frequency is cpproximately in phase, as shnwn by the synchroscope being between 11 and 1 O' Clock. (0.5) b. TRUE or FALSE 7 After the Output Breaker is closed, load is picked up on the generator by going to RAISE on the Speed Load Changer. (0.5)

QUESTION 2.06 (2.00) A LOCA and a Loss of Offsite Power (LOSP) have occurred simultaneously.

The Diesel Generators have started and AUTOMATICALLY energized their respective emergency busses. The sequential loading relays will delay the automatic ntarting of emergency loads for O, 5, or 10 seconds. LIST the Essential Bus 4 1 cads which wi11 be automatical1y started at EACH of these time intervals. (2.0) L Sb.k c 4.15WuM) r (***** CATEGORY O2 CONTINUED ON NEXT PAGE *****)

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QUESTION 2.07 (2.50) Concerning the Reactor Water Cleanup (RWCU) Systems c) Each of the events listed below would cause the Reactor Water Cleanup System to ISOLATE. For each event, STATE whether the Inboard Isolation Valve (CU-015) AND/OR the Outboard Isolation Valve (CU-Ol8) would automatically CLOSE.

1. A loss of cooling water to the NRHX, resulting in an NRHX RWCU outlet temperature exceedino 140 deg F. (0.5) 2. A MANUAL initiation of the Standby Liquid Control System (SLC). (0.5) b) LIST, the interlocks, if any, which exist between each of the below litt Mi valves and the RWCU pumps. (i . e. , what valve position, if applicable, would result in an RWCU pump trip) If none, state NONE.

1. CU-015 - Inboard isol ati on Val ve (0.5) 2. CU-018 - Outboard Isolation Valve (0.5) 3. CU-068 - Outlet Valve (0.5) N0h ' l. d c oh>e poIM 05 Ul W , E kW cp A , U C.W d oc OvoM OUESTION 2.08 (3.00) Concerning the CRD System: a. LIST the two (2) causes of a CRD Accumulator Trouble alarm (Sstpoints NOT required) and DESCRIBE the action which must be teken to determine the cause. (1.0) b. Explain HOW you can determine the position of the inlet and outlet scram valves using indications in the Control Room. (1.0) c. Shortly after resetting a reactor SCRAM, it is reporte&' that Cooling Water flow is LOW; however, the CRD System flow indicator is reading FULL SCALE. EXPLAIN ti.s apparent diccrepancy (in terms of the CRD System flow). (1.0)

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QUESTION 2.09 (2.00) When starting a Feedwater pump, four (4) interlock conditions must be m;t, when the control switch is taken to START, for the pump to begin running. LIST the four interlock conditions which must be met. Include catpoints if applicable. (2.0) DUESTION 2.10 (3.00) Reactor level is at 75" and reactor pressure i s at 300 psi g. The RHR system has automatically aligned to the LPCI Mode.

a. When RHR automatically aligns to the LPCI Mode, the open signal to RHR-27A, "RHR Isolation Valve to Recirc Loop A", is delayed for two (2) seconds. WHAT is the reason for this time delay? (0.5) b. LIST the conditions and/or actions required to throttle RHR-27A

"RHR Isol ati on Val ve to Recirc Loop A" after it opens. (1.0)

c. LIST the conditions and/or actions required to OPEN the Recirc Pump Discharge and Discharge Bypass valves after they shut due to LPCI initiation. (1.0) d. The RHR Service Water pumps trip when LPCI i ni ti at es. LIST the conditions or actions required to restart the pumps. (0,5) DUESTION 2.11 (2.50) Concerning the RPT/ARI Trips o. LIST the two (2) si gnal s, excluding MANUAL PB, which will initiate the RPT/ARI trip . (1.0) NOTE: INCLUDE SETPOINTS, TIME DELAYS, ETC. p b. TRUE or FALSE 7 (0.5) Lons of 125 vdc ECCS power will cause the RPT/ARI solenoid valves to open, thereby "dumping" the scram air header.

c. LIST two (2) places where RPT/ARI interacts (interf aces) with CRD Hydraulics. (1.0)

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QUESTION 2.12 (2.00) An automatic HPCI initiation has occurred. Subsequently, HPCI injection was automatically terminated due to high reactor water lovel.

c. LIST the valve (s) which will receive an AUTO CLOSE signal on this high water level. (1.0) b. HOW will HPCI respond to a subsequent decrease in water level to below the initiation setpoint? (0.5) NOTE: ASSUME NO OPERATOR ACTIONS OCCUR c. Assume that 6.he HPCI system had switched suction sources from the CST to the Suppression Pool due to a Low CST level. The CST level is later restored to normal.

WILL the HPCJ system AUTOMATICALLY switch back to CST suction? (0.5)

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QUESTION 3.01 (2.00) Whct are two reasons for resetting the Reactor Protection System cfter a scram? (2.0) OUESTION 3.02 (1.50) c) WHAT i s the purpose of the 11P Shear Valve System? (1.0) b) Briefly DESCRIDE how a shear valve is actuated. (0.5) OUESTION 3.03 (1.50) Assume that a low water l evel trip has actuated the ECCS systems: a) Where is the water level with respect to the top of the active fuel at the time of actuation? (0.5) b) Why was that l evel chosen? (two reasons recuired) (1.0) DUESTION 3.04 (3.50) a) Under what plant conditions is the ROD BLOCK MONITOR (RBM) required to be in service? (0.5) b) What adverse condition is the RBM designed to prevent? (1.0) c) What are the four types of input signals to the RBM system? (2.0)

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OUESTION 3.05 (2.50) A Group I isolation has occurred: c) What valves should have closed as a direct result of the isolation signal? (valve No.'s not required) (1.5) b) Why is the reset system designed so that some of the valve control switches have to be manually operated to complete a Group 1 isolation reset action? (1.0)

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QUESTION 3.06 (3.00) If fuel rods were broken during handling in the fuel pool causing high iodine release and a subsequent Group III isolation: a) Specifically WHAT three automatic actions should occur? (1.5) b) WHAT instrument signals cause a Group III isolation? (1.5)

(include setpoints)

QUESTION 3.07 (3.00) Assuming that actual vessel water level and the indicated water level is initially the same for the narrow range GEMAC water level instruments at full power. HOW and WHY would the indicated level change with respect to the actual level.if actual level remained constant and (Consider each change separately.)

a) the drywell temperature increased? (1.5) b) the reactor pressure decreased? (1.5) QUESTION 3.08 (2.00) For the following situations, state whether the Automatic Depres-curization System (ADS) relief valves will OPEN. CLOSE or REMAIN AS IS.

Consider each set of conditions separately.

o) Drywell pressure greater than 2.6 psig and reactor water level less than 75 inches for 2.5 minutes, ADS valves closed . . . . then CS pump A discharge pressure increases to 195 psig. (0.5) b) ADS initiating signal sealed in, ADS valves open . . . . reactor water level then rises to 1SO inches. (0.5) y c) ADS initiating signal sealed in, ADS valves open . . . . then a DC power failure occurs that affects all busses supplying ADS valves. (0.5) d) ADS initiating signals sealed in, ADS valves open. . . . , drywell pressure decreases to 2.1 psig. (0,5) l

     )
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4t__1NB18QdgNJS_6NQ_CQN18QLS PAGE 16

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QUESTION 3.09 (3.00) Tha reactor is operating at 30% power. Recirculation Pump A is running with flow control in local manual. Recirculation pump B io secured.

Explain HOW and WHY Recirculation Pump A responds to the f ollowino conditions. Where applicabl e, provide specific values or limits.

Consider each condition separatel y.

Note: Transparency 16 LOT-03-OO7, "Speed Control System" is attached for your reference.

a) Master Controller output f ails low (1.0)

(signal from M/A transfer station)   I b) Recirc MG set A tachometer output f ails to zero  (1.0)

c) Startup signal generator fails (1.0) i QUESTION 3.10 (3.00) Consider the following information

- the Reactor is at 100% power
- APRM CHANNEL D is reading 102%
- FLOW UNIT A is reading 90%
- FLOW UNIT B is reading 98%
- 8 LPRM inputs to APRM CHANNEL D are bypassed STATE whether each of the following will occur. (YES or NO) occur.

c) RPS DIV I tripped (1/2 scram) (0.5) b) Cor. trol Rod Withdrawal Block (0.5) c) APRM D Inop j- (0.5) d) APRM D upscale high (0.5) o) APRM D upscale Hi Hi (0.5) f) Flow Refe ence Off Normal (0.5) i (***** END OF CATEGORY 03 *****) J l

l 4t__EB0gGDUBES_:_NQBdGLi_OpNpBdGL3_EdEBGENgy_8ND PAGE 17 I BOD 1960GigeL_ggN1BOL l

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I OUESTION 4.01 (2.00) A fire in the cable vault has rendered some safety equipment inoperable from the control room. STATE four of the immediate actions the control room crew should take per OP 3126, Shutdown using Alternate Shutdown M3thods, before they leave the control room? OUESTION 4.02 (2.00) In accordance with OT 3118 Reci rcul ati on Pump Trip - Procedure, if one recirculation pump trips: a) Procedure directs the operator to close the pump's discharge valve. WHY? (1.0) b) After several ini nut es . the procedure directs the operator to reopen the pump's discharge valve. WHY? (1.0) OUESTION 4.03 (3.00) For each of the following situations STATE which, if any, Emsrgency Operating Procedure (s) should be entered. If none, state NONE. Consider each set of conditions separately. Assume all automatic systems function as designed and no operator cctions have been taken. State any additional assumptions.

a) The reactor was operating 9 100% power when reactor pressure increased to 1O75 psig. One SRV opened and stuck open. Torus water temperature is 90^F. Torus volume is 69,500 cubic feet. (1.0) b) A severe fir e in the switchyard has caused extensive damage to the AC Electrical Distribution System.

Multiple instrument f ailures due to power f ailure and bus grounds have made it impossible to determine reactor power or reactor vessel water level. (1.0) c) Drywell pressure is 2.0 psig. Drywell Average Return Air Temperature is 169^F. Unidentified leakage has increased to the extent that continuous Floor Drain Sump Operation ; i s required. (1.0) l (***** CATEGORY 04 CONTINUED ON NEXT PAGE *****) l

4t__PBOCEllyOES_:_NORMOL t_QQNOBdGL,_FUEBQENCY_AND PAGE 18

. R AD I Ot.OG I CAL,_CONIBOL.

. QUESTION 4.04 (3.00) You are performing a lineup of the Service Air (SA) system per procedure AP 155,"Current System Valve and Breaker Lineup and Identification". You find thats a) a valve is required to be open per the lineup. You find the valve closed with a white tag attached. WHAT INFORMATION should you place on the lineup sheet? b) a valve is required to be open per the lineup. You find the valve closed with a caution tag attached. Should you open the valve? (answer YES or NO) c) there is a valve with a White and Caution tao attached. This is NOT in accordance with the tagging procedure (Answer TRUE or FALSE) d) a valve required to be posi tioned has an illegible i denti f i cati on tag.

1. You shoul d posi ti on/ veri f y the positi on of the valve.

(answer YES or NO) 2. What information should you put on the tag? QUESTION 4.05 (3.00) a. Under what two (2) plant conditions is Primary Containment intergrity required to be maintained. (1.5) b. When primary containment i s required, per the Technical Specifications: 1. what in the maximum torus water temperature permited? (0.5) 2. what is the minimum and maximum water level limits f or the torus ? (1.0)

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4.__P80CEDU6ES_ s _NQ6001_,_O!WQ800L t_EdIJ8GENCY_OND PAGE 19

. 60DIOLOGIG01,_GLWIBOL l
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l l l QUESTION 4.06 (3.00) i Bsred on the VY administrative control limits a) What are the whole body exposure control limits per week and per quarter without a completed NRC Form 4 for VY pccsonnel? (1.0) b) What is the whole body exposure limit per week after a NRC Form 4 is completed and proper documentation is in place (0.5) c) Who can authorize exceedino administration control limits on a temporary basis? (0.5) d) What are the suggested emergency exposure limits to:

  (whole body)

1. save life? (0.5) 2. protect equipment or save propertv? (0,5) QUESTION 4.07 (3.00) STATE the entry conditions for a) OE 3101 - Reactivi ty Control Procedure (two reouired) (1.0) b) OE 3102 - RPV Level Control (two required) (1.0) c) OE 3103 - Drywell Pressure and Temperature Control (1.0)

 (two required)

QUESTION 4.08 (1.50)

)*

Dafine each of the terms listed belows a) Locked Door High Radiation Area (0.5) b) Radiation Area (0.5) c) Contaminated Area (0.5)

 (***** CATEGORY 04 CONTINUED ON NEXT PAGE *****)

42-_P6pCEDUBES_ _NORdOL,_OljNOOMOL3_EMESGENgY_GND PAGE 20

BODIPLDGIGOL_CONIBOL

OUESTION e4 ., 09 (2.00) ThD Reactor Scram Emergency Procedure, DE3100, directs the operators a) . . . to verify that reactor power is less than two percent. For WHAT two reasons is this value selected? b) . . . to establish and maintain reactor water level between 127 and 177 inches. WHY are these two values selected? OUESTION 4.10 (2.50) Reactor Startup to Criticality, OPO100, presents a number of steps or precautions designed to protect plant equipment or assure safe operation . a) IDENTIFY the equipment protected AND the adverse affects avoided by EACH of the following procedure requirements: 1) Neither core flow nor reactor power should be increased if the difference betweer the bottom vessel drain and the steam dome saturation temperature is greater than 145^F (1.00) 2) Automatic level control should be started early in the cycle. (1.00) b) Per procedure OPO100, define "slightly supercritical".

J'

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l t __ EG INC IEL EM _QE _UUGI,E G B _ E OWE 8_ EL GUI_ QEE B Ol igN 3 PAGE 21

IUE;6M00yNOMICS t _UEGI_lB8NSEEB_ONQ_Elyly_E(QW ANSWERS -- VERMONT YANKEE -88/01/19-BROCKMAN. K.

. ANSWER 1.01 (1.00) a REFERENCE BFNP BFN Mitigating Rx Core Damage, pp 17 - 18: RO 85/02/01 GGNS: MCD. SECTION 4.4 EIH L-RQ-540 (MCD) VTY : LOT-01-111, CRO Obj's 4, 5 &6 BWR K/A: 2.7/2.8 292OO6K101 ...(KA'S) ANSWER 1.02 ( 2. ".0) c (0. 5) Control rod worth is proporti onal to thermal neutron diffusion length (moderator temperature) (0.5) and to the local-to-average flux r atio (0,5). During startup and heatup, local Flux and moderator temperature are increasing at a faster rate than core average flux (0.5).

After heatup is complete, the increases in local neutron flux and moderator temperature are at a slower rate than the increase in core cverage flux (0.5).

REFERENCE General El ectric Reactor Theory. Chapter 5 BFNP Rx Theory, Chapter 5 L.O. 2.4 VTY : LOT-02-016, CRO Obj's 5 & fLt CNR K/A: 2.5/2.6 292OO5K109 ...(KA*S) J' l

 . _ .

__ __ - __ _ ____-____ 1. __C8.I NC I El,ES_OE _ NUCLE 06_EOWER_El.ON1_OPEROT I QN, PAGE 27

. ]HE8MODYNAM1CS, ., HEAT _T RANSFER ()N!}_Fidjl D_ELOW ANSWERS -- VERMONT YANKEE  -88/01/19-BROCKMAN. K.

. l ANSWER 1.03 (2.00) l c. After a power increase, xenon burnup is greater than xenon production (0.5), therefore power increases due to less negative reactivity in the core (0.5) (and recirc pump speed must be decreased to maintain the same power level.)

Alternate answer - Xe buildup is initially less than before thus recirculation flow must be decreased to balance the reactivity change.

b. After the power change, xenon is being produced at the equilibrium l evel for 60% power (0.25). After the xenon transient, xenon is being produced at the equilibrium l evel for 90% power (0.25). The 90% equilibrium xenon concentration adds more negative reactivity than the 60% equili bri um concentrati on (0.25). More posi ti ve reactivi ty must be added by increasing recirc pump speed (0.25).

REFERENCE VTY : LOT-02-018. CRO Obj's 1. 2 L3 BWR K/A: 2.9/3.2 292OO6K105 ...(KA'S) ANSWER 1.04 (2.00) a. The assembly power which would cause the onset of transition boiling at some point in the assembly. (1,0) b. 2 (1.0) REFERENCE BFNP: TRANSITION BOILING & ATLAS TESTING LP,P.5-6 GEXL CORRELATION & CRITICAL POWER LP,P.3 GGNS: MCD, THERMAL LIMITS, P.26,32-33 r EIH : L-RO-672 VTY : LOT-02-312, CRO Obj's i.e & 3 BWR K/A 3.3/3.73 2.9/3.3; 2.9/3.23 2.7/3.2; 2.7/3.2 293OO9K117 293OO9K122 293OO9K123 293OO9K124 293OO9K125

...(KA*S)
        ,

l

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i l 1.__EBINCIELES_QE_NGCLEOB_EQW68_E(ON1_QEEBOllON, PAGE 23

IUE600DYUGUICSt_UEGI_160NSEE8_OND_ELy1D_ELQW 1 i ANSWERS -- VERMONT YANKEE -88/01/19-BROCKMAN, K. ) l

ANSWER 1.05 (3.00) l c. INCREASE (0.5) More Subcooling at the pump suction (0.5) (1.0) l b. REMAINS THE SAME (0.5) Available suction head is independent of pump speed (0.5) (1.0) c. INCREASE (0.5) Further to saturation temperature and increased (1.0) density causing less static head (0.5) REFERENCE General Electric Heat Transfer and Fluid Flow. Chapter 6; BFNP: HTFF, Chapter 6, L.O.'s 7.7 8.1, & 8.2 VTY : LOT-02-203. CRO Obj 6 BWR K/A: 2.7/2.8 293OO6K110 ...(KA'S) ANSWER 1.04 (3.00) c. Lower than (0.75) b. Lower than (0.75) c. Lower than (0.75) d. the same as (0.75) REFERENCE General Electric Heat Transf er and Fluid Flow, Chapter 7 EIH: L-RQ-667, p 10 BFNP: Rx Heat Balance LP RQ 85/03/05 ' HTFF, Chapter 7, L.O.'s 5 & 6.43 HTFF, Chapter 8, L.Qc 5.1 VTY : LOT-02-115, CRO Obj 1 BWR K/A 2.6/3.1 293OO7K111 ...(KA'S) l

l L t _ _EB I NCIE(E S _QE _NUCL,E GB _ CQWE B _ EL ON1_QEE BOIlQN i PAGE 24 )

. ItiE8dODyN0dICS . UEGl,180NSERB_8NQ_E(ylO_E(QW ANSWERS -- VERMONT YANKEE  -88/01/19-BROCKMAN. K. l
.

l l ANSWER 1.07 (1.50) To c. D# crease (0.75 each) b e.

b. Jdcrease REFERENCE General Electric HTFF. Chapter 5 BFNP: HTFF - Chapter 5, t.O. 5 VTY : LOT-02-114 3 CRO F.tufent Objective 3 & 4 BWR K/A: 2.8/3.1 293OO3K123 ...(KA'S) ANSWER 1.08 (3.00) F1. b (0.5) F2. c (0.5) F3. a (0.5) L1. c (0.5) L2. a (0.5) L3. b (0.5) REFERENCE Gsneral Electric NEDO 24810-B (September 1983) EIH: GPNT, Vol VII, Chapter 10.2-23 VTY : LOT-02-312, CRO Obj's 9, 10, & 11 BWR K/A: 2.8/3.63 3.0/3.43 2.8/3.61 2.9/3.53 2.8/3.6 293OO9K107 293OO9K108 293OO9K111 293OO9K112 293OO9K119

...(KA'S)
     )*

i l __

L t _ _C8 ] NC IELC S ,.QE _NUCL E80_EOWE8_ELONI _OEE8811gN . PAGE 25

. I H E GMO DX NOM I C S t _UEQ 1_ISON SEfjB _8N D _ELVID _ EL OW ANSWERS -- VERMONT YANKEE   -88/01/19-BROCKMAN. K.

. ANSWER 1.09 (2.00) c. The negative reactivity added by the increased voids generated by tho rod withdrawal is greater than the positive reactivity added by tha reduced rod absorbti on. (1.0) b. (4) (1.0) REFERENCE General Electric Reactor lheorv. Chapter 5 BFNP: Rx Theory, Chapter 5, L.O. 3 VTY : LOT-02-016, CRO Obj 5 BWR K/A: 3.5/3.5 292OO5K104 ...(KA'S) ANSWER 1.10 (2.00) c. Doppler Coefficient (0.5) The rod insertion will be first "seen" by the adjacent fuel rods. (The immediate flux changes will cause fual temperature to change before the effects are seen in moderator tcmperature or void fraction.) (0.5) b. Void Coefficient (0.5) The MSIV Closure will immediately cause e pressure increase which will result in a change in the void fraction.

(This change will be "seen" before any changes in moderator temperature or fuel temperature.) (0.5) REFERENCE General Electric Reactor Theory, Chapter 4 BFNP: Rx Theory, Chapter 4, L.O.'s 4.3 & 6.3 /' VTY : LOT-02-207, CRO Obj 4 BWR K/A 3.3/3.31 2.5/2.6 292OO4K111 292OO4K114 ...(KA*S) ,

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it_ EBINCIPLES_DE_NUCLEGB_EOWE8_PLON1_OEEB9110N 1 PAGE 26

. IHEBMQQyNOM,ICS 3 _Ug8I_IBONSEgB_6ND_ELUIQ_ELQW ANSWERS -- VERMONT YANKEE  -88/01/19-BROCKMAN, K.

. ANSWER 1.11 (2.00) Th3 delayed neutrons behave as a source (0.5). K-eff after a scram will still.be well above 0.93 therefore, subtritical multiplication of the delayed neutrons will take place (1.0). Also, the delayed neutrons cro generated as a function of higher power conditions (and the longer half-life precursors will arti ficially keep power high). (0.5) REFERENCE CR3 : ROT-1-33 VTY LOT-02-OO7 CRO Student Objective 3 LOT-02-010 CRO Student Objective 2 BWR K/A: 2.9/3.04 3.3/3.3 292OO3K101 292OO3K107 ...(KA'S) ANSWER 1.12 (1.00) C. 1 b. 4 (0.5 each) REFERENCE BFNP PUMP CHARACTERISTICS. PUMP HEAD. PUMP LAWG LP.P.4 GGNS: OP-NP-504; OP-NP-514 l EIH : L~RO-655 ' VTY : LOT- s 2 205, CRO Obj's 2, 3, &4 ! l BWR K/A: 2.6/2.7 293OO6K113 ...(KA'S) r l

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7t__ELONI_QESIGN_lNCLUQ1NQ_SGEEly,,0NQ_EdEEGENCY_Sy@ldMS PAGE 27 A'NSWERS -- VERMONT YANKEE -88/01/19-BROCKMAN. K.

. ANSWER 2.01 (1.00) c REFERENCE EIH GPNT, Vol V, Chapter 2.5-7 & Chapter 4.2-11,14 VTY : LOT-03-108, CRO Obj's 1 &6 BWR K/A: 3.4/3.43 3.5/3.6; 3.5/3.7: 3.6/3.6 201001K107 201001K203 201001K402 201001K404 ...(KA'S) ANSWER 2.02 (3.00) c) Continues (0.75) b) Stopped (0.25) (LOCAQ Operator Action required to reset the overspeed trip (0.5) c) Stopped (0.25) CONTROL ROOM Operator Action to reset the isolation (0.5) d) Stopped (0.25) Reinitiates AUTOMATICALLY (0.5) REFERENCE EIH: GPNT, Vol VI, Chapter 8.1; L-RQ-737 VTY : LOT-05-OO2, CRO Obj's 2 & 3 *L LaT - oS - 3 01 * " 5 , 4 *1 BWR K/A: 3.8/3.7: 3.8/3.7; 3.4/3.3; 3.6/3.5; 3.3/3.3 217000A201 217000A202 217000A203 217000A301 217000K402

...(KA'S)
    /*

2t__CL,0N1_ DESIGN _ LNG (yQ1NQ_g8Egly_@NQ_gdgRQENGy_SYSlgdS PAGE 28

ANSWERS -- VERMONT YANKEE -88/01/19-BROCKMAN, K.

. ANSWER 2.03 (2.00) SPEED RESPONSE c) 102% ___3._et S (0.5) b) 102% __ -e _ }_ o r 6 (0,5) c) 110% ee L> (0.5) __ _1._,_tf d) 111.5% __ _-2 _l,4 e r b (O.S) REFERENCE EIH: GPNT, Vol. VI, Chapter 5.5-2; Vol. VII. Chapter 9.4-19, 20; L-RQ-705 VTY : LOT-OO7, CRO Obj's 2 & 3 BWR K/A: 3.3/3.3 3.0/3.1 241000A208 241000K403 ...(KA'S) ANSWER 2.04 (1.00) d REFERENCE BFNP: LPH7.P. 28 EIH L-RO-714 Figure 714-6; L-RO-714 GGNS: SD B33-1, pp 5. 6; OP-B33-1-501, p 5: ARI B33-FAL-L603A l VTY : LOT-05 101, CRC Ob; 2; LOT-03-OO7, CRO OBj's 3, 5, &6 ; i BWR K/A: 3.5/3.93 3.0/3.1 202OO1A210 202OO1K404 ...(KA'S) 1 I l ANSWER 2.05 (1.00) /* c. FALSE (0.5) b. TRUE (0.5) l REFERENCE l

EIH EIH Simulator; VTY : LOT-05-202, CRO Obj 3 VTY : OP-2160, pg 13

BWR K/A 3.2/3.23 3.1/2.9 241000A415 245000A402 ...(KA*S)

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2t__ELONI_QES10N_lNGLUDING_EGEEIY_QNQ_EdEBGENCY_@YSlEdS PAGE 29 A'NSWERS -- VERMONT YANKEE -88/01/19-BROCKMAN. K.

. ANSWER 2.06 (2.00) O seconds - RHR Pump A Service Water Pump A or C 5 seconds. - RHR Pump B 10 seconds - CS Pump A (4 & O.5 each) REFERENCE BSEP: SD 50.1, p 3: SD-17, pp 9, 33 HD 21-2/3-D. Secti on 3. 3. 6, pp 50 - 52 VTY : LOT-05-208: LOT-05-209, CRO Obj 4 BWR K/A: 3.2/3.5: 3.5/3.6 264000K405 295003K303 ...(KA'S) ANSWER 2.07 (2.50) a. 1. INBD - Closes OTBD - Closes 2. INBD - Closes OTBD - Cl oses (0.25 each) b. 1. CU-015 Not Full Open 2. CU-018 Not Full Open 3. CU-068 Full Closed (0.5 each) REFERENCE BSEP: 11-06-A, Section 3, pp 15, 18 VTY : LOT-03-OOO, CRO Obj 3 BWR K/A 3.2/3.43 3.6/3.6; 3.1/3.33 3.3/3.5 20COOOA201 204000A303 204000K601 204000K607 ...(KA*S) J'

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2t__PLON1_ DESIGN _ INCLUDING _ SAFETY _OND_ EMERGENCY, SYSTEMS PAGE 30 dNSWERS -- VERMONT YANKEE -88/01/19-BROCKMAN. K.

. i l ANSWER 2.08 (3.00) c. Low Nitrogen Pressure (of 965 psig) (0.25) l High Water Level in the instrument block (of 60 ml) (0.25) At the local control panel, the back lit button muat be depressed.

(If the light goes out, the cause is waters if the light stays lit, thm cause i s gas pressure. ) (0.5) L W,k * b. The Full-Core Display on the center panel has a DEUE scram light for each control rod (.5) which when illuminated indicates that both the inlet and outlet scram valves for that rod are open (.5). (1.0) c. The CRD FCV is downstream of the flow element. (0.25) All of the indicated flow is going through the Charoing line to recharge the accumulators. (0.5) The sensed high flow is sendi eg a signal to close the FCV, (and thus Cooling Water flow is low). (0.25) REFERENCE EIH: L-RQ-718 i BFNP: LP#29.P.6 EIH: GPNT, Vol VII, Chapter 9.2.1, Chapter 9.3 i VTY : LOT-03-108, CRO Obj 2 VTY : LOT-03-OO5, CRO Obj's 3 & 5 BWR K/A: 3.4/3.4; 3.5/3.6; 3.0/3.0: 3.4/3.4 3.9/3.83 3.5/3.6 201001A106 201001A210 201001A301 212OOOA108 212OOOA304 212OOOK106 ...(KA'S) ANSWER 2.09 (2.00) i LO Pressure adequate (0.25) > 10 # (0.25) , Condensate avail abl e (0.25) >= 1 CS pump running (0.25) J' Suction Valve Open (0.5) Suction Pressure adequate (0.25) >= 200 # (0.25) REFERENCE VTY : LOT-05-109: CRO Obj. 3 VTY : OP-2172 BWR K/A: 3.6/3.6 3.9/3.7 259001A203 259001A402 ...(KA'S)

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2t__PL@N1_phS!GN_lNQLUDINO_SOEEly_8ND_EMER@ENQY_SYSIEMS PAGE 31 ANSWERS -- VERMONT YANKEE -88/01/19-BROCKMAN, K.

. t ANSWER 2.10 (3.00) c. To prevent overloading of UPS (0.5) b. 5 minute TD L o r- Se gd CM d * UPS Feeder Trip Initiation Block Sw in the BLOCK position (0.5 each) c. Initiation signal cleared (0.3) and reset (0.3) Take the CS to OPEN (0.4) (1.0) d. Pump switch (S-19) to MANUAL OVERRIDE (0.5) REFERENCE VTY : LOT-03-306, CRO Obj's 3, 4 &6 BWR K/A: 3.6/3.6 3.9/3.9; 3.6/4.1 203000A404 203OOOA406 203OOOK410 ...(KA'S) ANSWER 2.11 (2.50) c. Reactor High Pressure (0.25) - 1150 psig (0.25) Lo Lo Level (0.25) - 82.5" w/ 10 sec TD (0.25) (0.5 each) b. FALSE (0.5) c. Individual HCU Scram Air Header Scram Discharge Instrument Volume (2 O O.5 each) REFERENCE VTY : LOT-03-108, CRO Obj's 3 & 5 BWR K/A: 3.4/3.41 4.1/4.23 4.1/4.3 201001K107 212OOOA207 212OOOA208 ...(KA'S)

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7t__L'LONI_PESJON_]NGLUDINQ_g8 Eely _8ND_EMEBQgNgy_gygIgd5 PAGE 32 dNSWERS--VERMONTYANKEE -88/01/19-BROCKMAN, K.

. ANSWER 2.12 (2.00) c. HPCI Turbine Stop Valve (0.5) and the Minimum Flow Bypass Valva (to the Suppression Pool) (MOV-25) (0.5) (1.0) b. Automatic restart will occur on the Low-Low Level (0.5) c. No (0.5) REFERENCE BFNP LP-042. pp 12 & 13 VTY : LOT-03-303, CRO Obj's 3 & 5

OWR K/A: 4.3/4.4 !,5/3.6 3.7/3.8; 3.8/3.9; 3.9/4.0 l r

( E

* i

6

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Ut__INSIBUMGNIS_AND_CONIBOLS PAGE 33 MNSWERS -- VERMONT YANAEE -88/01/19-DROCKMAN, K.

. ANSWER 3.01 (2.00) a) Restore the normal crd valve lineup (i . e. shut the scram valves which restores the primary system boundary at the scram val ves) (1. 0) . b) Allow the Scram Dump Volume to drain (i . e. Open the SDV vent and drain val ves) (1. 0) l Or * W ~f k " *) LOT-03-108 Pg. 30 LOT-03-108H. Pg 1.

Student Objective 4 212OOOK412 ...(KA*S) ANSWER 3.02 (1.50) a) The shear valves provide a means to isolate the TIP guide tubes (.5) in the event a TIP cannot be withdrawn (outside of the ball valve which normally would provide isolation) when containment isolation is required.(.5) (1.0) b) A shear valve i s actuated by means of its keylock switch located on the control panel.(.5) (0.5) REFERENCE LOT-03-105, Pg 14 Student Objective 3 715001K4ul ...(KA'S) ANSWER 3.03 (1.50) c) 82.5" above the top of the active fuel (.5) (0.5) r h) Lcw enough tn prevent spurious actuation (.5) and high enough to r.* ovide assurance that the fuel temperature can be maintained below 2200 degrees F (.5). (1.0) 9EFERENCE LOT-03-OO2,Pg. 23 Student Objective 3 216000K103 ...(KA'S) _

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_ _. _ Et__lNB16VMENIS_OND GQNIROLS PAGE 34 dNSWERS -- VERMONT YANKEE -OG/01/10-BROCKMAN. K.

. ANSWER 3.04 (3.50) c) When the reactor is operating above 30 percent of rated power. (0.5) b) Limit fuel rod power in eases to within predetermined limits 9xpected not to damage the fuel ( s)( by i mposing rod bl ocks 6,44.

(exact words not rwquired) (1.0) c) 1. LPRM detector signal currents (0.5) 2. Rod Select Signal (0.5) 3. Reference APRM signal (0.5) P 4. Recirculation flow unit signals (0.5) REFERENCE VY LOT-03-107 Rod Block Munitor, pgs 5,6 & 7 CRO Student Objectives 1 &5 215002K101 215002K102 215002K403 ...(KA*S) ANSWER 3.05 (2.50) a) The (eight) Main Steam Isolation valves (.5), the (two) reci r cul ati on system sample valves (.5), and the (two) steam line drain valves (.5).

(1.5) b) The system i s designed to pravent inadvertent or autcmatic opening of the isolation valves (when resetting an inolation signal). (1.0) REFERENCE VY LOT-03-211, Primary Containment Isol ation System, pgs 12,14,15, 37 & 38 CRO Student Objective 4,10 223002K101 223OO2K406 ...(KA'S) r

 .
  -- .,,

dt _.INSIBUMENIS_6ND_ggNIBQLD PAGE 35

. .
        ,

ANSWERS -- VERMONT YANKEE -88/01/19-BRDChMAN, K.

. ANSWER 3.06 (3.00) o) 1. Standby Gas Treatment System starts (0.5 2. Reactor building ventilation system isolates (0.5) 3. Primary containment vent and purge valves isolate (0.5) b) 1. High radiation on refuel floor ( 0.25) Set point 100 mr/hr (. o . n25) 4f6 2. High radiation in the reactor building ventilation exhaust duct (O?5) Set point 14 mr/hr '^ 25)

3. b w vene.A W - 5C" (.c.3%-) g

% % %.   [ 3g REFERENCE 3 s \Me" ~ W sy f (O. M$~)

LOT-03-209 Pg 9 Student Objective 7 261000K401 ...(KA'S) ANSWER 3.07 (3.00) , c) Indicated will be higher than actual (.5). The temperature in the reference leg increases (.5)and the sensed differential pressure becomes less (.5). (1.5) b) Indicated will be higher than actual (.5). Due to calibration differences (.5) the sensed differential pressure becomes less (.5).

(1.5) REFERENCE VY LOT-03-OO2, Reactor Vessel Instrumentation, pgs 25-27 1' CRO Student Objective 7 216000K501 216000K507 216000K510 :. . . ( KA ' S )

p

 ,   , - - - , . - . - - , . . - . . - . , ,

Es__lNBIByDENJS_ANQ_GQNIBQLS PAGE 36

$NSWERS -- VERMONT YANKEE  -88/01/19-BROCKMAN, K.

. ANSWER 3.08 (2.00) o) ADS valves open (0.5) b) ADS valves remain as i s (0.5) c) ADS valves cl ose (0.5) d) ADS val ves r.A-ese- h.s a b (0.5) REFERENCE VY LOT-03-304 Automatic Depressurization System, pgs 12 - 14 CRO Student Objectives 2 & 4 218000A105 21BOOOA205 2190COK404 218000K501 210000K602

...(KA'S)

ANSWER 3.09 (3.00) a) No akk oo (67) J.

Recire Pump A e!r"r L. _ Am de"- e -J h tu- C e b. h o A*k e a Sh ld b

  ' 53 due to the duw, ccmco spurd

_ (1.0) _drmano i.2e uutti toe ivw l1cw rtep rn th. ocmep twLe 1. . i te

+h- dec.::;; ' ??.
 . '9;te si dec ees; as lim.i d by ihr Er e ur Sign =1 Li-! tin; Nminorx.,

b) Recire Pump A speeds up (.5) due to the mismatch between speed (1.0) demand and feedback signal (.2) until the high flow stop on the scoop tube limits the increase (.3) (Rate of ircrease is limited by the Error Signal Limiting Network.)

c) No effect on Recirc Pump A (.5) because the Startup Signal (1.0) Generator is only placed in the circuit when the MG Set Field Breaker is open (.5) REFERENCE VY LOT-03-OO7, Reactor Recirculation System, pgs 22 & 23 CRO Student Objective 8 /' 202OO2K305 ...(KA'S) ) , l

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33.__1NB16UMEN1E_@ND_GONI6DLO PAGE 37 dNSWERS -- VERMONT YANKEE -80/0,' /19-BROCkMAN. N. i

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      .

ANSWER 3.10 (3.003 a) NO (0.5) b) YES (0.5) c) NO ,e M ( 4 w,w op erah , a, d. . - rt g *,A Wg TsA sp eu) (0.5) l d) NO (0.5) e) NO (0.5) f) YES (0.5) REFERENCE VY LOT-03-106. Average Power Range Monitor (APRM), pgs 11,12 &13 Trcnsparencies 6 & 7 . CRO Student Objective 3 215005K101 ...(KA'S) l

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00l!!O100 J COLGON1 ROl, ANSWEl~~1000 DPM/100 sq cm Beta, gamma, or~~

~~>100 D?M, alpha, or any leak / spill which is potentially contaminated (0.5) l' (cxact wording not required) (1.5)~~
~~REFERENCE LOT-06-305, pg 13,14, CRO Student Objective 3 294001K103 ...(KA'S)~~
~~$1-_P80GEDUGES_ _NOBMOL a _OON06M6L s _gMCIf 0EjNQy ONI)~~
~~_ PAGE 42~~
~~, 80DIOLOW COL CON 160t_~~
ANSWERS -- VERMONT YANKEE -08/01/19-BROCKMAN. K.
. '
,
i ANSWER 4.09 (2.00) c) Two percent power is selected because: 1. It is easily measurable by verifying the APRM downstale lights.(.5) 2. A thermal power of less than 2% can be easily removed by normal plant systems.(.5) ,
;
b) The values are selected because: 1. Establishing l evel above 127 inches allows resetting the scram and restoring CRD cooling.(.5) 2. Establi shing l evel bel ow 177 inches assures a level less than the component trip points.t.5) (2.00) REFERENCE ! LOT-09-OO1, pg 27.28, CRO Student Objective 2 295031K206 295031K211 295037A106 ...(KA'S)
ANSWER 4.10 (2.50) c) 1) Vessel Bottom Head (.5) - thermal shock (.5)
~~2) Feedwater flow nozzles (.5) - thermal cycling (.5)~~
b) When neJtron flux rises with a const t (stable) period without any positive reactivity ad61..on,(such as additional control rod withdrawal). (.5) (2.5) REFERENCE LOT-07-OO1, pg 20, CRO Student Objective 3 OPO100, pgs 2 and 7 /' 202OO1K117 259001K101 ...(KA'S)

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TEST CROSS REFERENCE PAGE 1 'OUESTION VALUE REFERENCE ________ ______ __________
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01.01 1.00 TRLOOOO673 01.02 2.50 TRLOOOO663 L ) 01.03 2.00 TRLOOOO669 01.04 2.00 TRLOOOO672 ; 01.05 3.00 TRLOOOO662 i 01.06 3.00 TRLOOOO667 01.07 1.50 TRLOOOO693 , 01.08 3.00 TRLOOOO668 01.09 2.00 TRLOOOO6s4 01.10 2.00 TRLOOOO666 01.11 2.00 TRLOOOO692 > 01.12 1.00 TRLOOOO671 ______ 25.00 02.01 1.00 TRLOOOO650
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~~25.00 03.01 2.00 TRLOOOO641 03.02 2.00 TRLOOOO642 04.03 3.00 TRLOOOO643 03.04 3.00 TRLOOOO644 04.05 3.00 TRLOOOO645 04.06 3.00 TRLOOOO646 i 04.07 3.00 TRLOOOO688 i~~
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~~TERT CROSS REFERENCE PAGE 2 UUENTION VALUE REFERENCE __.,___ ___ ___._ __ _ 04.00 1.50 TRLOOOO609 00.09 2.00 1RLOOOO690 04.10 2.50 TRLOOOO691 _._ 25.00 __ ___ ____ . 100.00 DOCKET NO 271~~
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A licc h b 2 E.) . VERMONT YANKEE co bo- L>t b , NUCLEAR POWER CORPORATION ee 6 afM g d. k O' eu
~~, RD 5. Box 169. Ferry Road, Brattleboro, VT 05301 ,~~
V
.
~~January 26, 1988 TOL88.5 (602)257 5271 David Boiling Water Reactor Section Division of Reactor Safety Nuclear Regulatory Commission Region I 631 Park Avenue King of Prussia, PA 19406~~
~~Dear Mr. Lange:~~
In accordance with NUREG 1021, we are submitting a detailed review of the U.S.N.R.C. Reactor Operator License Exams which were conducted at Vermont Yankee on January 19, 1988. The details of this review are contained in the attach-ments and should clearly delineate our concerns with the examinations.
If you or your staff have any questions about this material, please do not hesi-tate to call.
Very truly yours, VERMONT YANKEE NUCLEAR POWER CORPORATICN f> M$ N Warren P. Murphy a Vice President and Manager of Operations
^*tachments .. )*
~~EVL880125.1.3 I m ? .": C ~ , i,~~
~~- - ' - , ..~~
~~. QUESTION 2.02~~

The RCIC System automatically initiated and is injecting into the vessel. For each of the situations listed below, STATE whether RCIC injection into the reactor would CONTINUE or STOP. If injection is STOPPED, DESCRIBE any AUTOMATIC or OPERATOR ACTION that would be required to reinitiate injection, b) A 125% Overspeed Trip is received due to low control oil pressure.
Control oil pressure is then returned to normal.
ANSWER 2.02 b) Stopped. LOCAL Operator Action required to reset the overspeed trip REFERENCE EIH: GPNT, Vol VI, Chapter 8.1; L-RQ-737 VTY: LOT-05-002, CR0 Obj's 2 + 3 BWR K/A 3.8/3.7; 3.8/3.73 3.4/3.3 3.6/3.5 3.3/3.3 217000A201 217000A202 217000A203 217000A301 217000K402
~~...(KA'S)~~
FACILITY COMMENT: 2.02 b) Question asks for the required operator action, not location of this action. "Local" should not be required for full credit.
QlESTION 2.03 Fill in the blanks to describe the main turbine's response to a SLOW OVERSPEED condition. Choose from the component actions listed below.
SPEED RESPONSE 102% a) _ , _ _ _ , b) _____ 110% c) _____ 111.5% d) _____ COMPONENT ACTIONS j, 1. Emergency 2. Backup Governor Overspeed Trip Trip l 3. Intermediate 4. Intermediate Stop Valves Stop Valves Full Open Full Closed 5. All Intercept 6. All Intercept Valves Full Valves Full Open Closed-1-l
.
_.
.
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~~ANSWER 2.03~~
*
SPEED RESPONSE a) 102% 3.
b)1024 _ _6._ _ ' c) 1105 1.
d) 111.5% _ _2._ _ REFERENCE EIH GPNT, Vol. VI, Chapter 5.5-2 Vol. VII, Chapter 9.4-19, 20 L-RO-705 VTY: LOT-007, CR0 Obj's 2 + 3 BWR K/A 3.3/3.3: 3.0/3.1 - 241000A208 241000K403 ...(KA'S) FACILITY COMMENT: The following responses are corrects a. 3 or 5 b. 3 or 5 c. 1 or 4 or 6 d. 1 or 4 or 6 See attached drawings from LOT-05-004 - , QtESTION 2.08 Concerning the CRD Systems i b. Explain HOW you can determine the position of the inlet and outlet scrata , l valves using indications in the Control Room.
ANSWER 2.08 b. The Full-Core Display on the center panel has a B.UE scram light Nor each control rod which when illuminated indicates that both the inlet and outlet scram valves for that rod are open.
l l REFEREtCE
EIH L-RQ-718 BFPF: LP*29,P.6 EIH: (PNT, Vol. VII, Chapter 9.2-1, Chapter 9.3 l VTY: LOT-03-108, CR0 Obj,2 l VTY: LOT-03-005, CR0 Obj s 3 + 5 BWR K/A 3.4/3.As 3.5/3.6 3.0/3.0, 3.4A.4, 3,9/3.8 3.5/3.6 201001A106 201001A210 201001A301 212000A108 212000A304 212000K106 ...(KA's)
~~>- - - _ _ _ - _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _~~
~~. _ . - .~~
~~FACILITY COMENT:~~
.
The Vermont Yankee full core display has white scram lights, not blue.
LOT-03-108 is in error. See attached LOT-03-006 T/P3.
! QL.ESTION 2.10 , Reactor level is at 75" and reactor pressure is at 300 psig. The RHR system i has automatically aligned to the LFCI Mode.
b. LIST the conditions and/or actions required to throttle RHR-27A "RHR Isolation Valve to Recirc Loop A" after it opens.
ANSWER 2.10 b. 5 minute TD UPS Feeder Trip Initiation Block Sw in the BLOCK position FACILITY COMMENT: 2.10.b Should also accept "Initiation signal clears and operator resets." See LOT-03-306.
j Q!ESTION 3.08
For the following situations, state whether the Automatic Depressurization
System (ADS) relief valves will OPEN, CLOSE, or REMAIN AS IS. Consider each set of conditions separately. ' d. ADS initiating signals sealed in, ADS valves open. . . . drywell pressure decreases to 2.1 psig. , ' ANSWER 3.00 ADS Valves Close REFERENCE VY LOT-03-304, Automatic Depressurization System, pgs 12 - 14 CR0 Student Objectives 2 + 4 , 218000A105 218000A205 218000K404 218000K501 218000K602
~~...(KA'S) #~~

FACILITY COMENT: l Answer Key is incorrect. Valves will remain open. The high drywell pressere l signal is a "seal in" signal, once initiation has occurred, a drop in drywell } pressure will have no effect. See pages 14 and 15 of LOT-03-304 and the i attached drawing. '
l

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.
~~Qt.ESTION 3.09~~
~~. The reactor is operating at 304 power. Recirculation Pump A is running with flow control in local manual. Recirculation Pump B is secured.~~
Explain HOW and WHY Recirculation Pump A responds to the following conditions.
Where applicable, provide specific values or limits. Consider each condition separately.
Note: Transparency 16, LOT-03-007, "Speed Control System" is attached for your reference, a) Mastei Controller out put fails low (signal from M/A transfer station) ANSWER 3.09 a) Recire Pump A slows down due to the decreased speed demand until the low flow stop on the scoop tube limits the decrease. (Rate of decrease is limited by the Error Signal Limiting Network.)
! REFERENCE VY LOT-03-007, Reactor Recirculation System, pgs 22 + 23 CR0 Student Objective 8 i 't 202002K305 ...(KA'S) FACILITY COMMENT: Answer Key is incorrect. If the recire pump is in M/A Station Control, master controller failures will have no effect on speed. See attached drawing.
Qt.ESTION 3.10 Consider the following information
- the Reactor is at 100% power - APfN CHANNEL 0 is reading 1025 ' - FLOW UNIT A is reading 905 - FLOW UNIT B is reading 985 - 8 LPRM inputs to APRM CHANtEL D are bypassed STATE whether each of the following will occur. (YESorNO)occy.
~~c) APRM D Inop ANSWER c) NO REFERENCE VY LOT-03-106, Average Power Range Monitor (APRM), pgs.11, 12 + 13 l Transparancies 6 + 7, CR0 Student Objective 3 l 215005K101 ...(KA'S)~~
~~-4- i i --__ , - - - . __ , , _ . - , - - - - - , _ _, --~~
~~. FACILITY COMMENT:~~

Question does not make a distinction between automatic actions and operator initiated actions. Due to 8 LPINS being bypassed, APRM 0 is in fact inop I requiring an operator initiated declaration of an inoperable channel. See Technical Specification Section 3.1.
QlESTION 4.06 Based on the W administrative control limits: d) What are the suggested emergency exposure limits tos (whole body) 1. save life? 2. protect equipment or save property? , ANSWER 4.06 d) 1. 100 kem whole body (AP 0501) or 75 Rem whole body (AP 3507) 2. 25 Rem whole body REFERENCE LOT-06-305. pgs 9 + 10, CR0 Student Ubjective 1 294001K103 ...(KA'S) FACILITY COH4ENT:
.
AP 3507 provides an additional limit of 12.5 REM to protect equipment if prior planning can be performed. See attached Table from AP 3507. j QLESTION 4.07
STATE the entry conditions for , c) OE 3103 - Drywell Pressure and Temperature Control j (two required) ANSWER 4.07 # c ) 1. Drywell RRU average return air temperature is greater than 160'F or
2. Drywell pressure is greater that 2.5 psig l REFERENCE LOT-09-001, pg 13, CR0 Student Objective 3 ' OE 3101, 3102, 3103 295024K101 ...(KA'S)
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~~FACILITY COMMENT:~~
'
~~, A recent procedure change now has the entry condition reading "Drywell~~
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Atmospheric Temperature rather than "Drywell average return air temperature."
, QUESTION 4.10 Reactor Startup to Criticality, OP0100, presents a number of steps or pre-cautions designed to protect plant equipment or assure safe operation.
, ' a) IDENTIFY the equipment protected AND the adverse affects avoided by EACH of the following procedure requirements: 1) Neither core flow nor reactor power should be increased if the dif-ference between the bottom vessel drain and the steam dome saturation
,
temperature is greater than 145'F f ANSWER 4.10 a) 1) vessel Bottom Head - thermal shock REFERENCE LOT-07-001, pg 20, CR0 Student Objective 3 0F0100, pgs 2 and 7 " 20200lK117 25900lK101 ...(KA'S) FACILITY COMMENT:
;
~~OP 2110 contains a precaution stating the limit is based on damage to CR0 stub~~
~~tubes and In core housing welds. This answer should also be accepted. See~~
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_.. Emergenc) Goveenor; i Load Limit 4 110-111% ' Trip Piston (CRP 9-7)SW-18 Emergency Governor Lockout Valve Emer- ~ Acceleration gency Trip Handle (011 Trip) Relay MTS-3 , , Emergency PB-1 3 J ' Otmp _ OP 9-7 / Valve Master Su-17 Trip 0F 9-7 MTS-2 ,
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~~- _ _ _ _ .~~
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LOT-03-304 - Rev. 2, 6/67 Page 14 of 17 OUTLINE NOTES (1) At this point the 120 second timer is energized c) Low pressure ECCS pump running at 100 psig discharge pressure from any one RHR or CS pump, d) 120 second timer timed out e) Once energized the K6A relay seals in around the high drywell and low-low reactor water level contacts C. To reset ADS blowdown 1) Timer not timed out a) Any of the following will stop the timer:
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~~' (1) Low-low reactor water leve'i clears ,~~
~~I (2) High drywell pressure contact ' clears and is manually reset l~~
! (3) Depressing the timer reset pushbuttons will reset the timer to Zero, but it will restart when the buttons are released. ; (4) Deenergizing the 125 VOC power supply b) Stopping all low pressure ECCS pumps will not reset the timer but it will prevent AOS actuation # 'J u-2w) r~:-
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~~depressingthetimerresetbuttons) hilf~r~eset ~the ADS;18gTcT[cjoie[the/ ~ V e 1 Fei"and~~
- -p '" top - t h's ' t i mer.s ,;"v.] -.___ '. b )' If high drymll'presiFre~ahd liw%1o D ' r e akifoiCwa_ t e r ' l eve 1[a r e' no p .61iiaR depressing the timer. reset buttons T ivill rese't.the timers'and the' ADS' l -. _ ___ . _ _
~~, ' LOT-03-304 Rev. 2, 6/87 Page 15 of 17 OUTLINE NOTES~~
~~'yalyes._w t1]_c lose. Valves will s .~~
~~reopen when the timer times out. / c) Stopping all low pressure ECCS pumps will not affect the timer but it will close the ADS valves, d) Deenergize the 125 VOC power supp'y 3. ADS Power Monitor~~
.
~~a. Upon a loss of the normal power supply (DC-2C) CWO 750-755 to the "B" AOS Logic or any one of the four SRV's, an automatic swap to the alternate Annunciator: '~~
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power supply (OC-1C) will occur. BLOWOOWN CONTROL POWER FAILURE b. The "A" AOS Logic does not have an alternate (Common to all) power supply . NOTE: Only 1 ADS Logic needed for ADS Initiation.
c. Pushbutton and keylock switches 1) Provided for I&C Surveillance of AOS Power Monitor Relays 2) Two pushbuttons provide annunciator test 9-32 only for power loss to "A" or "B" Logic 9-33 3) Five keylock switches provide power swap 9-32 , testing and annunciator a) 1 "B" ADS Logic b) 4 - 1 for each SRV d. ADS Status lamps # 1) Indication light for high drywell pressure / low low level is provided en 9-32 and 9-33 2) Individual indication for ADS relay actuation is provided by individual lamps.
V. OPERATIONAL SUMMARY A. .When ADS is actuated, the flow of steam through the valves provides a maximum energy removal rate while minimizing the corresponding fluid mass loss from the reactor vessel. l
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. EMERGENCY 00SE LIMITS * . Oose limit Criteria (NCRP 39) 1. 5 Rem to the whole body, or its Oose limit applied to emergency equivalent to any part of the body, center personnel and center habitability.
2. 12.5 Rem to the whole body, or its_f Dose Ifsit applied to in-plant i equivalent to any part of the body. I activities required to. correct ' or prevent plant degradation.
3. 25 Rem to the whole body, or its Maximum allowable dose to an equivalent to any part of the body. emergency worker for the duration of the accident.
4 75 Rem to the whole body, or its Immediate evaluation and action equivalent to any part of the body. required for saving of life. When efforts are completed, revert to limits 1 through 3 above, as appropriate.
~~(Information Notice No. 84-40)~~
~~NOTE: If the limit specified in 4 is involved, the following considerations should be made:~~
~~....................~~
1. Female employees of child-bearing age should not be allowed to participate; 2. Volunteers above the age of 45 years should be given priority; 3. The individual (s) awareness of the biological consequences that such an exposure can haves 4: All practical protective measures to limit such an exposures 5. Concurrence of individual (s) involved (i.e., voluntary risk acceptance): 6. The probability of success should be balanced against the exposure limit; 7. The individual's familiarity with the task to be performfd and; 8. The speed with which the individual can perform the task.
~~NOTE:~~
All occupational doses including emergency doses are required to be included as part of a worker's exoosure history, and hence can affect the worker's allowable exposure during the current quarter and subsequent quarters.
....................
~~- . _ _ _ ,. _ _,__-~~
~~. _ _ -_ ________~~
0 0. allo Rev. 16 ' Page 5 of 15 the following wie can be used: The motor windings can be assumed to have returned to rated temperature after 45 min-utes de-energized or after 15 minutes running at rated speed.
3. During normal operation, both PC sets should be operated at speeds that are within 54 of each other and in no instance operated beyond the limits of Figure I.
4. Prior to and during startup of an idle recirculation loop at tem-peratures >200 F, log reactor coolant temperatures on VYOPF 2110.01.
5. Do not place the master flow controller on CRP 9-5 to AUTO.
6. For single loop operation, do not exceed pump notor current of 363 amps, drive motor current of 450 amps, or 704 speed.
C. Other , 1. %K-17436 2. EX-9609 3. OwG 191162, Misc. Systems - Recirc IC Lite Oil 4. OWG 191167, Nuclear Boiler 5. OwG 191159, Sh. 5, Recirc Pump 6. 0.P. 0102, Power Operations (Manuevering at Power) l 7. R.P. 2182, RBCCW 8. 0.P. 2111, Control Rod Drive System 9. SIL 406, Incore Instrumentation Protection 10. E Information Report Allowable Recirculation Flow Rates for In-Core Protection, Dated 9/85 11. 0.P. 2427, 6tnitoring Reactor Stability 12. 0.P. 2428, Single Loop Operation 13. SIL 368 14. SOER 84-7 15. G.E. Document 23A1487, Single Loop Operation Precautions: 11._ ,To avoid overstressing .ths;MstWtube and.in-core housing; welds nthe , reactor vessel botton drain toeparefurfddilt'he~irithiW145frr o rtfil r*7 ' seturationi reture before reg Treactor y g _ S g
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2. The pung in an idle recirculation loop shall not be started unless the temperatures of the coolant within the idle and operating loops are within 50*F of each other. If this limit is exceeded, a normal reactor cooldown shall be conducted or the idle loop warmed up by reverse flow until this condition is met.
CONTROL CCPY VAUD ONI.Y '# HEN STAMP iS RED __ - -. .
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ATTACiGENT 3
;
NRC RESPONSE TO FACILITY CCP.MENTS r The following represents the NRC resolution to the facility comments (listed ! in Attachment 2) made as a result of the current examination review policy . Only those comments resulting in significant changes to the master answer key, or those that were "not accepted" by the NRC, are listed and explained below. Comments made that were insignificant in nature and resolved to the satisfaction of both the examiner and the licensee during the post examination review are not listed (i.e.: typographical errors, relative acceptable terms, minor set point changes).
, Question 2.02: Comment acceptad.
j Question 2.03: Comment accepted.
Question 2.08: Comment accepted. The error in the training material should t i be corrected.
' Question 3.08: Comment accepted.
i ' Question 3.09: Comment accepted. The error in the answer key was due to a typographical error 'n the question, i Question 3.10c: Comment accepted. "Yes" is an acceptable answer if the candidate assumed that the required operator actions had been taken, j Question 4.06: Comment noted. Additional limit will not be considered incorrect, but the limit provided in the answer key is required for a complete answer. !
;
~~Question 4.07: Comment accepted. Recent procedure change should be~~
~~incorporated into training material.~~
~~' Question 4.10: Comment accepted. Alternate answer is acceptable. This answer was not reflected in the lesson plan used to develop~~
~~the question.~~
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b ATTACHMENT 4 SIMULATION FACILITY FIDELITY REPORT Facility Licensee: Vermont Yankee Nuclear Power Corporation Facility Licensee Docket No.: 50-271 Facility Licensee No.: DRP-28 Operating Tests administered at: Vermont Yankee Operating Tests Given On: January 20, 1988 During the conduct of the simulator portion of the operating tests administered January 20,19S8, the following apparent performance and/or human factors discrepancies were observed:
-
B Diesel Generator tripped when a Main Turbine trip caused an undervoltage spike on the grid.
- Local operations required for simulating preparations for testing can be performed for A Diesel Generator only, This detracted from the reality of the scenario.
The performance of the simulator on January 20, 1988 was greatly improved over the performance during the previous operating examinations, i l l
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v t e Inspection Report - Vermont Yankee - 1988001
80 \| 81 \| 82 \| 83 \| 84 \| 85 \| 86 \| 87 \| 88 \| 89 \| 90 \| 91 \| 92 \| 93 \| 94 \| 95 \| 96 \| 97 \| 98 \| 99 00 \| 01 \| 02 \| 03 \| 04 \| 05 \| 06 \| 07 \| 08 \| 09 \| 10 \| 11 \| 12 \| 13 \| 14 \| 15 \| 16 \| 17 \| 18 \| 19 \| 20 \| 21 \| 22 \| 23 \| 24 \|
1988 Quarterly Integrated 1Q (0) 2Q (0) 3Q (0) 4Q (0) Assessments Mid Cycle End of Cycle 7(0) 8(0) 9(0) 11(0) 12(0) 13(0) 14(0) 16(0) 17(0) 18(0) Security 200(0) Operator Exams Emergency Planning Other

IR 05000271/1988001

Text

Dear Mr. Lange:

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