IR 05000334/1985010
| ML20209E917 | |
| Person / Time | |
|---|---|
| Site: | Beaver Valley |
| Issue date: | 06/21/1985 |
| From: | Dante Johnson, Keller R, Kister H NRC OFFICE OF INSPECTION & ENFORCEMENT (IE REGION I) |
| To: | |
| Shared Package | |
| ML20209E911 | List: |
| References | |
| 50-334-85-10, NUDOCS 8507120349 | |
| Download: ML20209E917 (49) | |
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U. S. NUCLEAR REGULATORY COMMISSION REGION I OPERATOR LICENSING EXAMINATION REPORT EXAMINATION REPORT N /85-10 (0L)
FACILITY DOCKET N FACILITY LICENSE N DPR-66 LICENSEE: Duquesne Light Company Post Office Box 4 Shippingport, Pennsylvania 15077 FACILITY: Beaver Valley 1 EXAMINATION DATES: April 30 - May 3,1985 CHIEF EXAMINER:
actor Engineer (Examiner)
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' Dafe D. JohnsonVLea~
REVIEWED BY:
R. 'M. Keller, %i6f, Project Section 1C
[ fI Date APPROVED BY: I H. B. Kiste U hief, Project Branch No. 1 ( D' ate SUMMARY: Oral, written and simulator exams were administered to ten Reactor Operator (RO) candidates. In addition, the oral portion only was given to one RO retake candidate. All candidates passed all portions of their exams and will be issued license ~
8507120349 850627 PDR ADOCK 05000334 G PDR
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4 REPORT DETAILS TYPE OF EXAMS: Replacement EXAM RESULTS:
l R0 l l Pass / Fail l l l l l l IWritten Exam l 10/0 l l 1 I I I I l Oral Exam l 11/0 l l l l I I I l Simulator Exami 10/0 l l 1 I I I I-l0verall l 11/0 l l l l Chief Examiner at Site: D. Johnson, NRC Other Examiners: R. Keller, NRC R. Schreiber, PNL
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R. Gruel, PNL R. Cochran, NRC Consultant Generic Deficiencies From Oral and Simulator Exam: Malfunctions provided examiners are not specific and in some cases in error. This vagueness in expected responses to particular incidents and malfunctions resulted ' in difficulties preparing adequate scenario The radiation monitoring system does not function properly, i.e.,
during a RCS cold leg break no associated containment radiation alarms or indications are receive .During operation of the CVCS when performing a dilution proper indication and subsequent actions do not always occu During a blackout scenario unrelated and unexplained responses occur such as steam demand signals, and Tavg - Tref oscillation There are no instructions in the control room for computing a QPT . . . - -
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3 The index for annunciator alarms has several errors, i.e. references the wrong pages, difficult to locate proper procedure, The remote shutdown panel indication for pressurizer pressure has only a narrow range indicato . Improvements in Training Program as a Result of Prior Operator Licensing Examinations:
The examiners noted that there was an _ increased knowledge level overall
.regarding control board familiarity, manipulation of controls and ability to adequately utilize reference material such as P&ID's, Technical Spec-ifications and facility procedures, as compared to the group performance on the exams administered in February 198 This apparent increase in performance for this group of candidates was probably attributed to the use of a site specific simulator as part of the training for this grou . Personnel Present at Exit Interview:
NRC Personnel
D. Johnson, Lead Reactor Engin,eer (Examiner)
D. Coe, Reactor Engineer (Examiner)
R. Cochran, Consultant Facility Personnel J. Seiber, Senior Manager T. Jones, General Manager, Operations T. Burns, Training Director L. Schad, Simulator Coordinator T. Kuhar, Operations Training Coordinator S. Rovin, Simulator Operator 6. Summary of NRC Comments Made At Exit Interview:
Generic deficiencies during oral / simulator exams were discussed. Training improvements and interface with plant staff were summarized a's wel Preliminary results of all candidates clearly passing the oral / simulator portions of the exam were presente . Examination Review:
At the conclusion of the written examinations, the examiners met with the following licensee personnel to review the exam and answer keys to iden-tify any inappropriate questions relative to plant specific design and to ensure that the questions will elicit the answers in the key and that the reflect the most current plant condition .
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T. Kuhar, Training Coordinator
.T.-Burns, Director, Nuclear Training P. Russe.ll, Operations Instructor R._ Ferrie, Training Instructor R. Brooks, Operations & Maintenance Instructor 8. ~R0 Exam Comments and Resolution:
question 2.01, Part b. was delete The question _was not clear, and as a result of possible interpretation could be answered either true or fals .03, Part Answer was modified to accept other plausible responses as indicated by review of reference operating procedure .05, Part Answer was modified to accept partial credit for boron
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mixing and higher flux per reference from Procedure BVPS 0.M. 1. .0 Answer was modified to accept additional possible responses of 1ack of preheating, seal flow problems and pressurizer level maintenance per BVPS 0.M. 1. .08 Part b. (iii). Answer was changed to " primary drain tank" due to an error in facility supplied reference materia .08 Part c.(ii). Answer was modified to accept additional response of bearing temperature indications per reference supplied by license .1 Due to plant modification, the answer was modified to accept " con-densate storage tank" as a correct response in addition to " auxiliary river water and water treatment system".
3.03, Part a.(1). Clarification was added to the answer key by inserting
" loop Tavg" rather than merely "Tavg".
3.06, Part a.(2). Answer was changed from " Nuclear Power (N44)" to
" turbine first stage pressure (446 or 447)" typographic error correctio .03, Part Answer was modified to accept "12 rem" if the NRC limit is considere .0 Opposite responses in answer would be accepted if candidate assumes a control channel failure which was not clear in wording of questio .-- . . -- . - . _ _ . . _ _ . -. . .
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Attachments: Written Examination and Answer Key Facility Comments on Written Examinations and Resolutions
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U.S. NUCLEAR REGULATORY COMMISSION
REACTOR OPERATOR LICENSE EXAMINATION
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Facility: BEAVER VALLEY 1 Reactor Type: Westinghouse PWR
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Date Administered: April 30,1985
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Examiner: Schreiber/ Gruel Candidate: tt A$7E f(. A F V Iiv 'CoPP INSTRUCTIONS TO CANDIDATE:
Use separate paper for the answers. Write answers on one side onl Staple question sheet on top of the answer sheet. Points for each question are indicated in parenthesis after the questio The passing grade requires at
. .least 70% in each category and a final grade of at least 80%. Examination papers will be picked up six (6) hours after the examination start ; Category % of Candidate's % of
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Value Total Score Cat. Value Category 2.6. o 2 .- 1. Principles of Nuclear Power Plant Operation, Thermo-dynamics, Heat Transfer and Fluid Flow
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2. .+.o 99- M 2. Plant Design Including Safety and Emergency Systems 2.+. 5 25 B&r4 3. Instruments and Controls 2. f.s i
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26 E4 4. Procedures: Normal, Abnormal, Emergency, and Radiological Control
, 1 a MP:9' TOTALS Final Grade .%
All work done on this examination is my own; I have neither given nor received
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1.0 PRINCIPLES OF NUCLEAR POWER PLANT OPERATION, THERMODYNAMICS, (26.5)
HEAT TRANSFER AND FLUID FLOW
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Points Available OUESTION 1.01
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The effective value of delayed neutron fraction changes over core lif Explain how it changes (increase, decrease) and give a brief reason wh (1.5) What is the effect of the change on reactor operation? (1.0)
ANSWER ~1.01 Beta bar effective decreases over core life because of the decrease in U-235 (and also U-238) which has a relatively large delayed neutron fraction and the buildup of fissionable Pu isotopes (239 and 241) with their lower delayed neutron fractio The reactor period decreases for a given reactivity addition, meaning a faster response (SUR).
~ Reference (s)
BVPS Reactor Theory Manual, chapter 5, Reactor Kinetics, pp.15-16 for part a. and p. 26, equation 5-6 for part .
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Points Available 00ESTION 1.02 ,
If the initi'al neutron count rate is 117 cps and withdrawal of Bank D a few steps gives a steady state neutron count rate of 420 cps, calculate the final value of Keff. Assume the initial value of Keff is (2.0)
ANSWER 1.02 The basic equation relating count rate to multiplication constant is given by CR1 (1 - K1 ) = CR2 (I - K2 ) (1.0)
This may be solved by setting Cry = 117 and Cr2 = 42 K,ff = 0.9721 (accept 0.972). (1.0)
Reference (s)
BVPS Reactor Theory Manual, chapter 5, Reactor Kinetics, p. 48.
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. Points Available 00ESTION 1.03
Briefly explain the difference between MODERATOR TEMPERATURE COEFFICIENT and MODERATOR REACTIVITY DEFECT. Consider the case where the ppm boron concentration is constant and only the aver-age moderator temperature is changing. Make sketches if that will hel (2.0)
l ANSWER 1.03 MTC is the rate of change of reactivity with changes in moderator temperature, pcm/deg. The moderator defect is the accumulated negative reactivity, pcm or % delta rho, that is built in as the average moderator temperature is increased during power ascensio .
Reference (s)
BVPS Reactor Theory Manual, chapter 6, Inherent Reactivity, pp. 26 and 3 QUESTION 1.04 Why is some boron in the core in the form of burnable poison rods and the remainder of the baron is dissolved in the coolant? (1.0)
ANSWER 1.04 The total amount of boron in the core is needed to hold down the excess reactivity of the fue If it was all dissolved in the coolant, the MTC would be positiv Reference (s)
BVPS Reactor Theory Manual, chapter 8, Power Control, pp. 40, 4 Section 1 continued on next page -
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QUESTION 1.05 ,
a. What two isotopes are mainly responsible for the DOPPLER COEFFICIENT of reactivity? (0.5) Briefly explain how the DOPPLER COEFFICIENT affects reactor power operatio (1.5)
ANSWER 1.05 U-238 and Pu-240 (the Doppler effect is a broadening and flat-tening of the absorption cross section resonance peaks as fuel temperature increases) . The fact that makes the Doppler coefficient particularly impor-tant is that fuel temperature immediately increases following an increase in reactor power, and fuel temperature changes always precede changes in either moderator temperature or void conten In the event of a large reactivity addition to the reactor, the moderator temperature coefficient cannot come into play for several seconds and would have little effect on the termination of the power excursion. The fuel temperature coefficient, on the other hand, acts instantly and represents the primary negative coefficient for a fast power rise transien Reference (s)
a. BVPS Reactor Theory Manual, chapter 6, Inherent Reactivity, p. 4 b. _ BVPS Reactor Theory Manual, chapter 6, Inherent Reactivity, p. 3 .
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Points Avail able QUESTION 1.06
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a. Which dropped rod at BOL is likely to cause the greatest radial flux tilt? Refer to the attached Control Rod Location figur (1.0)
(a) The Control Bank A rod at position F (b) The Shutdown Bank SA rod at rod at position G (c) The Control Bank C rod at position H b. Estimate the relative rod worth of two rodded positions in the core which contain assemblies of the same enrichment and burnup, but whose relative assembly power is 1.2 (1.0)
ANSWER 1.06 a. [This is because it is in a position of higher _ power (flux) than F2 and thus will influence a larger area of the core, but is farther out on the radius than H6, and thus has more flux " leverage" to cause radial tilt.] Relative rod worth is 1.44. (The same enrichment and burnup (same
" region" of fuel), but different power means that the condition exists at BOL on fresh or similar power history fuel. The ratio of power under these conditions is the same as the relative flu The relative rod worth is the square of the relative flux.)
Reference ( s)
BVPS Cycle Core 5 Core Physic's, Figure 3.3A, pp. 3-14, Assywise l Power at BOL, HFP, AR0, No Xe. Also, BVPS Reactor Theory, chapter 8, p.1 '
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CONTROL ROD LOCATIONS R P N M L 'K J H GFE DC 8A wo-
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.: sa* * 'b O C C 0 SA SP SS 55 SA 9 A E O C 0 9 A 10 55 SP 55 SP !!
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Points Available QUESTION 1.07 ,
In the sketch ~telow, the reactor power is shown varying with tim Draw the corresponding Xenon reactivity transient curve, showing how it varies'with time in response to the power level change Label the points on the Xenon curve to correspond to the points on the power profile. Assume that reactor power was at 0% before step change at (1) to 50%. ,
(3.0)
(1)
100 .
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%P 50 (6) (8)
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(2) (4) Time ANSWER 1.07 7) g,g Xe (4) (6) 8)
(2 () (9)
(5)
(1)
Time
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BVPS Reactor Theory Manual, chapter 7, Poisons, pp.13-1 Section 1 continued on next page -
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Avail able QUESTION 1.08 ,
With the reactor at power and control rods in manual, a makeup flow problem causes a 25 ppm boration of the reactor coolant system. The only action taken by the operator is to stop the boratio Choose the correct explanation of what has happened to the SHUTDOWN MARGI (1.0) The Shutdown Margin is increased because more boron is present and T-avg has decreased, causing less temperature defect on tri The SDM is decreased because rods have automatically stepped out to compensate for the boron increas The SDM is increased because the effect of boron addition is to reduce power without affecting T-av . The SDM is decreased because power remains the same after the boron addition and T-avg has increased to offset the effect of boro ANSWER 1.08 Reference (s)
BVPS Reactor Theory Manual, Chapter 9, p. 2 .
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Points Avail able QUESTION 1.09 ,
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Match the tenn in column A with the single expression in column B that best relates to i (2.5) I NPSH Pump motor overload Runout Very steep characteristic curve
! Shutoff head P-P sat Positive displacement Zero flow Flow meter Enthalpy Elbow tap Entropy
ANSWER 1.09 ,
a.3, b.1, c.4, d.2, Reference ( s)
BVPS Thermo Manual, chapter 4, Systems Pumps and Valves, pp. 9b-17 '
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Points Available QUESTION 1.10 ,
Mark the following statements about centrifugal pumps in closed "
systems TRUE or FALS Reducing the pump speed by 40% reduces the flow by the same amoun (0.5) A small increase in the inlet pressure produces a large increase in the outlet pressur (0.5) Increasing the pump speed by 30% increases the power require-ments by about 120%. (0.5)- Decreasing the power supplied to a pump by 10% causes a decrease in the differential pressure developed by the pump of less than 5%. (0.5)
ANSWER 1.10 True Fal se True Fal se Reference (s)
BVPS Thermo Manual, chapter 4, Systems Pumps and Valves, p. 33.
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Points Avail able QUESTION 1.11 ,
Consider a large gas storage tank at room temperature and 200 psi You are told it contains exactly one pound iaass of gas. Sometime later you notice that the tank has sprung a leak and is now down to only 20 psia. Use the gas equation, PV = MRT, to find the final amount of gas, M, remaining in the tan (1.5)
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ANSWER 1.11 V, T, and R are constant, so M changes directly a3 P changes (1.0).
The numerical value is exactly 0.1 (0.5).
Reference ( s)
BVPS Thermo Manual, chapter 5, Behavior of Steam & Gases, p. 2 QUESTION 1.12 The main purpose of the Moisture Separator Reheater is: (1.0) To significantly improve plant thermal efficienc To redirect steam from the impulse (HP) turbine to the reaction (LP) turbin To provide a high temperature source for the reheater To protect the LP turbines fran blade errosion by moistur .
ANSWER 1.12 Reference ( s)
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Points Avail able-QUESTION 1.13 J
Tube fouling af the main condenser causes T-avg in the primary to increase, which in turn causes steam flow rate in the secondary to increase. TRUE or FALSE (0.5)
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ANSWER 1.13 False (T-avg does increase, but steam flow decreases).
Reference ( s)
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BVPS Thermo Manual, chapter 6, The Steam Water Cycle, p. 2 QUESTION 1.14 Departure from Nucleate Boiling can result in the destruction of fuel, but the same phenomenon cannot destroy tubes in the Steam Generator. TRUE or FALSE (0.5)
ANSWER 1.14 True Reference ( s)
BVPS Thermo Manual, chapter 7, Nuclear Power Plant Character-istics, p.1 .
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Points Avail able QUESTION 1.15 ,
Briefly explain how a tube rupture would interfere with long-term cooling by natural circulation in the affected loo (1.0)
ANSWER 1.15 Initially the tube rupture would not greatly affect natural circu-lation, but as the primary system pressure is reduced to minimize flow into the secondary, the feed flow and steam flow from the affected 'S/G are isolated. With the S/G no longer serving as a heat sink (unless the secondary safeties lift or atmospheric dumps are operated), natural circulation in that loop of the primary cease Reference ( s)
BVPS Thermo Manual, chapter 7, Nuclear Power Plant Characeris-tics, p. 2 .
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QUESTION 1.16 /
Explain why it-is necessary to use two stages of Steam det Air
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Ejector to maintain vacuum in the Main Condense (2.0)
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ANSWER 1.16 Condenser back pressure must be maintained at about 1 psia (2 i Hg abs) . A SJAE in each stage uses two convergent-divergent (c/d)
nozzles to achieve close to the maximum practical pressure dif-ferential, AP=9 psid, at which the Critical Pressure Ratio is approached. Because it is not possible to increase the aP across a single stage SJAE to 13.7 psid (14.7-1), it is necessary to use two stage Reference ( s)
BVPS Thermo Manual, chapter 6, Steam Water Cycle, p. .
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2.0 PLANT DESIGN INCLUDING SAFETY AND EMERGENCY SYSTEMS (25.0)
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QUESTION 2.01.,
Answer the following TRUE or FALS The Train A protection system de-energizes the Train B bypass breaker and the Train A reactor trip breaker undervoltage coil (0.5) Assuming that the Train A reactor trip breaker and bypass a,,, M.7 ofeJote breaker are both closed, opening of either the Train B A u-e - co- og,j,j reactor trip breaker oja_ the bypass breaker will cause a oc * i+ke'-
reactor tri (0.5) The main control room is isolated from the outside atmosphere on a Containment Isolation Phase B (CIB) signa (0.5) Feedwater lines are isolated by a safety injection signal (SIS). This is done to prevent excessive cooldown of the RC (ac q $25 u ro./ 3 peea A I y,) (0.5)
ANSWER 2.01 True Fal se True True Reference (s)
BVPS 0.H 1.1.1, pp. 3 and BVPS 0.M, Figure' 1- .
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Points Avail able QUESTION 2.02 ,
Answer the following regarding Residual Heat Removal System (RHS)
operatio Reactor coolant pressure and temperature should be less than psig and 'F before the RHS is placed in operatio (1.0) How is decay heat removed from the RHS? ,,(1.0) How is the cooldown rate of the RCS controlled using the RHS? (1.0) What is the purpose of the RH pump miniflow recirculation line? (0.75)
ANSWER 2.02 psig, 350* Through the RH heat exchanger to the reactor plant component cooling wate , By manual control of the flow control valve (MOV-1RH-758) down-stream of the RH heat exchanger To prevent overheating of the RH pumps at shutoff head (low l
flow) conditions.
l Reference ( s)
l BVPS 0.M. 1.10,.pp. 2-5.
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Points Avail able QUESTION 2.03 , What indications alert the operator to flow through a pressurizer safety valve? (1.0) Why are the pressurizer safety valves' loop seals insulated to maintain the seal water at an elevated temperature? (1.0)
ANSWER 2.03 vaw6 T=, o, + erature, s':chama tunifald prectuC l o ^ I+4 ch* I A ..* A . o,3 Valve Wr#$lp temp?J Preu & Iswa
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BVPS 0.M. 1.6.1, p. 3 .
QUESTION 2.04 If both low head SI pumps become unavailable during a LOCA, how can water be supplied to the high-head SI pumps for recirculation? (1.5)
ANSWER 2.04 The suction of the high-head SI pumps can De cross-connected to the discharge of the Outside Recirculating Spray Pumpsiby manual opening (of RS-157 and RS-159)
Reference (s)
BVPS 0.M. 1.11.4, p. 7 BVPS 0.M., Figures 11-1 and 13- .
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Points Available QUESTION 2.05 ,
Some 7% of re~ actor core flow, deemed bypass flow, is essentially ~
unavailable to remove heat generated by the core. While most bypass flow is undesirable, briefly explain why each of the following flowpaths of bypass flow are intentionally allowe Flow through rod-guide-thimble (1.0) Flow through holes drilled in the upper core barrel lip and top support plat (1.0)
ANSWER 2.05 To allow a means for water to leave the rod-guide thimbles to decelerate the rod after it has been droppe To allow cold leg water to enter the head area to cool the upper head regio Reference ( s)
BVPS 0.M. 1.6.1, p. 1 & Pe <+:n I e ved o'+ apas~ So# be '* ~ " ' " ; "7 *
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Points Available QUESTION 2.06 ,
State two reasons for heating the charging flow with the regenerative heat exchange (1.5)
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ANSWER 2.06 (1) To eliminate reactivity effects due to insertion of cold wate (2) To reduce thermal shock on the charging penetrations to the RC NOTE: Will accept "to improve overall plant efficiency" for 0.5 point Reference (s)
BVPS 0.M. 1.7.1, p. 1 .
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Points Available QUESTION 2.07 State three probleas that result fecm remaining on excess let-down instead of using the normal letdown lin (2.25)
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ANSWER 2.07 (1) Bypass' of the demineralizers and filters (the activity of the coolant and impurities will increase much faster using the excess letdown line only).
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(2) Lownr flow (borations and dilutions will take much longer, limiting the rate of change of power at certain times).
(3) No hydrogen addition to the coolant (due to bypassing of the volume control tank during normal system lineup).
Reference (s)
BVPS 0.M. 1.7.1, p. 1- Also addmss/ : o5 p s Hcd*y
$y p ,_ pwn %A 0. '1 S r~ok-P>< a4 n'~+~ ~
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Points Avail able QUESTION 2.08 ,
Answer the following regarding the Reactor Coolant Pumps: Under normal conditions, how much total injection water flows directly into the RCS through the RCP Thermal Barrier /
Heat Exchanger? (0.5) What supplies the backpressure for each of the following:
(i) Number 1 seal (0.5)
(ii) Number 2 seal (0.5)
(iii) Number 3 seal (0.5) During a saf ety injection system (SIS) actuation, the #1 seal leakoff valve (MOV-CH-378) close (1) Under these conditions, how does the Injection and Seal i.edoff System react to maintain sufficient cooling flow through the lower radial bearing? (1.0)
(ii) How can the operator tell if insufficient cooling flow is provided to the lower radial bearing? (1.0)
ANSWER 2.08 a. 15 gpm (i) VCT (and line resistance)
(ii) Standpipe (iii) None (atmosphere),an PGT (06f A/c l') e r: q D u ts EA (1) The #2 seal will open slightly allowing sufficient Cos ~w)
cooling flow (1 gpm minimum)
(ii) The #1 seal leakoff header relief valve (RV-CH-382A)
opens (indicated by increasing PRT pressure and erratic leakoff flow indication). ( W ;< p ,6.4s y e e, ps e e,1 )
Ala. w eeff w.7 N Reference ( s)
BVPS Multigating Core Damage, LOCA, p.12 Section 2 continued an next page - -
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QUESTION 2.09
. Explain p, the availability of steam generator feedwater and
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>of a large break LOCA, but may have a significant effect upon
-'a small break LOC (2.0)
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ANSWER 2.09 On a large break the major cooling mechanism is the injection or recirculation of coolant through the core and energy removed out the break. Howevec, on a small break, the break size is not adequate to remove the decay heat. To insure an adequate heat sink, one must maintain the feedwater inventory in the steam generator and insure primary flow thFough the U-tube l[
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BVPS Mitigating Core Damage, LOCA, p.122- . r
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Points Avail able QUESTION 2.10 ,
The primary and backup water sources for the Auxiliary Feed Pumps (FW-P-2, 3A, 38) are the Primary Plant Demineralized Storage Tank (WT-TK-10) and the Reactor Plant River Water system. What sources of water can be used in the unlikely event that both primary and back-up water sources are unavailable? (1.0)
ANSWER 2.10 (1) Service water system of Unit (1) Engine driven fire pump (1-FP-P-2). s at
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(s) % we to w -v.rc- BVPS 0.M. 1.24.1, pp. 5, 1 '
BVPS 0.M. 1.30.1, p. QUESTION 2.11 Under what conditions is the residual heat release valve
! (HCV-1MS-104; located in a common steam header supplied from ( upstream of the main steam trip valves) used? (1.5)
l l ANSWER 2.11 1 It is used for core decay heat removal one-half hour after a reactor trip when the condenser is unavailabl NOTE: Will accept '.'use during physics testing" or "use during operator training" for half credi Reference ( s)
BVPS 0.M. 1.21.1, pp. 2, .
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Points Avail able QUESTION 2.12 ,
Why does each -main steam trip valve (TV-1MS-101A, B, C) have three (3) solenoid actuation valves? ( 1.5 )
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ANSWER 2.12 Train A and B redundant trip capability (1.0 pts) and for partial stroking of the valve during operation (0.5 pts).
Reference (s)
BVPS 0.M 1.21.1, p. 1 End of Section 2 -
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3.0 INSTRUMENTS AND CONTROLS (25.0)
, Points Available QUESTION 3.01 -
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Answer the following TRUE or FALS There are no protection functions related to the startup-rate indicatio (0.5) Both Intermediate Range Channels are required to be above the permissible (P6) level before the Source Range Level trips are blocke (0.5) Intermediate Range Level trips are automatically reactivated when 2 out of 4 Power Range Channels drop below the permissive (P10) level . (0.5) A " Power Range High Setpoint Flux Deviation or Auto Defeat" alarm will occur if an upper ion chamber fails low during operation at 100% powe (0.5)
ANSWER 3.01 True False (either 3R channel) False (3 out of 4 PR channels) True (3 channels will be higher than the average)
Reference (s)
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List the reactor trip signals that are automatically made operable as the reactor increases to 10% powe (2.0)
ANSWER 3.02 (1) Pressurizer Low Pressure (2) Pressurizer High ' Water Level (3) Low Reactor Coolant Flow (Breaker open, UV or UF)
(4) ' Turbine Trip Reference ( s)
BVPS 0.M. 1.1.1, pp. 5-1 BVPS 0.M. 1.2.4, p. .
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Points Avail able QUESTION 3.03 , What three signals are inputs to the Overtemperature AT set-point, and how does an increase in each input effect the set-
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point (raise, lower, or remain the same)? (2.4) . By what percent is the Overtemperature aT turbine runback setpoint lower than the trip setpoint? (0.5) How far will the turbine run back if the turbine runback setpoint is exceeded? (0.5)
ANSWER 3.03 unf (1) oT-avg--lower (2) Pressurizer Pressure--raise (3) Del ta-fl ux--lower ( era, % p_s a f)
Point Value: Input (0.5), Effect (0.3) . 9%
' Until the runback setpoint is cleared (on 1.5 seconds, off 30 seconds, at 200% per minute) .
Reference (s)
BVPS Technical Specifications, Table 2.2-1, Note '
BVPS 0.M. 1.1.1, p. 18 BVPS 0.M. 1.1.2, pp. 9-10
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Points Available
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QUESTION 3.04
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Assume that Pr.essurizer Pressure Channel 455 (PT-RC-455) has failed high during at-power operation. For each of the follow-ing,-failure of which other pressurizer pressure channel (s)
will produce the noted result? Each answer may require multiple channel s, Actuation of pressurizer relief valve (PCV-1RC-455C). (0.5) Actuation of ' pressurizer relief valve (PCV-1RC-456). (0.5)
c .. Reactor Trip (0.5) Safety Injection (0.5)
ANSWER 3.04 Channel 444 (failure of Channel 455 immaterial) Channel 445 (failure of Channel 455 immaterial) Channels 456 or 457 (0.25 each) None; there is no SI on high RCS pressure ; 14 a. 2 -.. de /s win. m.i t: ele. +.:4 u , m ,p .og.,
Reference ( s)
BVPS FSAR Figure 7.2-1, sheets 6, 1 BVPS 0.M. 1.6.1, p. 5 .
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Points Available QUESTION 3.05 ,
The reactor has been operating at 65% power with all control systems in automatic. For each of the following, give (1) the direction of initial rod motion (if any), (2) a brief explanation of why the rods moved, and (3) the resultant steady-state reactor powe A Steam Generator PORV fails ope (2.0) A lower detector of a power range channel (N44) fails hig ~
(2.0)
c. "B" RCP trips of (2.0)
ANSWER 3.05
. (1) Rods move ou (2) SG PORY opening increases secondary steam flow 'and hence heat removal. T-avg decrease (3) 68% (based on 10% increase in steam flow in 1/3 SG). (1) Rods move i (2) NI failure causes increase in nuclear power signal to rate comparator indicating increasing reactor power relative to turbine loa (3) 65% (Rate mi smatch signal) . (1) Rods move i (2) Low flow on single loop causes reactor trip (3) 0%.
Point value: (1) 0.5 pts (2) 1.0 pts (3) 0.5 pts
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BVPS 0.M.1.1.1, p.12,. Figure 1-1 BVPS FSAR Subsection 7.7.1.1, pp. 7.7-3, Section 3 continued on next page -
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Points Avail able QUESTION 3.06 , What is used to determine the programmed level setpoint for the control systems for each of the following valves?
' Steam Generator Water Level Control (FCV-1FW-478,488,498) (0.5) Low Power Steam Generator Control (FCV-1FW-479,489,499) . (0.5) Which one of the following depicts the response of actual steam generator water level to a low failure of feed flow Channel 47 Assume that the level control system remains in automatic, and Channel 476 remains the controlling channel (it is not selected out). (1.0) Increases and remains ther . Increases, then returns to its original valu . Decreases and remains ther . Decreases, then returns to its original valu ANSWER 3.06 . Turbine first stage pressure (446 or 447). vec, -
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hNticlea b 8* No-er ("44) P = cm m 94)
t s*rre . (It goes up due to controller response to steam flow greater than feed flow, then returns to its original level due to the integral level error signal.)
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l BVPS FSAR Subsection 7.7.1 BVPS FSAR Figure 7.7- BVPS 0.M.1.24.1, pp.17-20, Figure 24-1 .
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Points Avail able QUESTION 3.07 ,
What five (5) -signals will block automatic starts of the motor driven auxiliary feed pump (FW-P-3A,38)? (2.5)
ANSWER 3.07
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(1) Placing the control switch in PULL-TO-LOCK or STO (2) IAE Bus undervoltage (loss of voltage) APk *
(3) Diesel loading sequencer (not timed to start yet). fe u ,- r (4) Motor electrical protectio (5) SG main feed pumps (FW-P-1A,1B) control switches in the } mdle AFTER STOP positio C og br, % mn +r4 Reference ( s)
BVPS 0.M. 1.24.1, pp. 5,1 QUESTION 3.08 Identify the normal, alternate, and backup power supplies (specific bus or MCC) for 120V AC Vital Bus #1? ;(2.1)
ANSWER 3.08 Nonnal - 480V AC MCC1-E9 (through UPS 1)
Alternate - 125V DC Bus #1 (through UPS 1)
Backup - 480V AC MCC1-E13 (through Static Line Voltage Regulatory 1015)
Point Value: Order (0.4 each), Source (0.3 each).
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BVPS 0.M. 1.38.1, p. 3, Figure 38- _
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Points Available QUESTION 3.09 ,
Which one of the following radiation monitors does not have an associated automatic action? (1.0)
~ Component Cooling Water Monitor (RM-1CCR-100). Auxiliary Feedwater Area Drain Tank Monitor (RM-1DA-100). Gaseous Waste Particulate Monitor (RM-1GW-108A). Gaseous Waste Gas Monitor (RM-1GW-1088) .
ANSWER 3.09 Reference (s)
BVPS 0.M. 1.43.1, pp. 4, 9-12, 2 _
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QUESTION 3.10 ,
Answer the following TRUE or FALS '
a. Proper operation of the meteorological tower warning lights is automatically monitore (0.5)
b. The discharge piping of each Main Steam Safety Valve and Steam Generator Atmospheric Dump Valve is monitored for radiatio (0.5)
c. A flashing annunciator alarm will return to normal if the abnormal input signal disappears within 5 second (0.5)
d. Redundant electric traced (ET) heat circuits on safety-related piping are powered from separate emergency buse (0.5)
ANSWER 3.10 a. True (current sensed, alarmed in control room)
b. False (only lowest lift setpoint safety in each header)
c. False (only " Group 3")
d. True a
Reference (s)
BVPS 0.M. 1.43.1. p. 2 BVPS 0.M. 1.45.1, pp. 1, S-6, 1 BVPS 0.M.1.45.5, Figures 45-7,8, .
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4.0 PROCEDURES: NORMAL, ABNORMAL, EMERGENCY, AND RADIOLOGICAL (26.0)
CONTROL
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QUESTION 4.01 Consider the following regarding a chlorine leak at the Beaver Valley Power Statio What indication or warning do you have if you are suddenly ,,
immersed in a cloud of escaping chlorine gas? (0.5)
' What action do you personally take while leaving the area affected by the leaking gas? (0.5) What kind of firs't aid would you give to someone who is overcome by the chlorine fumes? (0.5)
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ANSWER 4.01 Warning is always given by the odor and color of the escaping gas. The eyes will also be irritate The gas is greenish-brown color and has a suffocating odor. It is nonflammable but supports combustion similar to ai Avoid panic and try hard not to cough. This will minimize irri-tation of the respiratory system. Keep mouth closed and avoid deep breathing. Keep head high as you leave the area; chlorine gas is heavy. When fresh air area is reached, exhale deeply, before inhaling to clear nose of chlorin Move the person to fresh air. Give anti-spasmodic first aid medication or other mild stimulants, such as black coffee or peppennint solution (one teaspoonful in a glass of water).
If person is unconscious, start artificial respiration at onc Obtain medical help as soon as possibl v.1c. : Firs t 44 8 t-a i 37 no+ y.# ; ;m Reference ( s)
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Points Avail able QUESTION 4.02 ,
If you are an on-duty Nuclear Control Operator (part of a minimum shift crew), what duties do you have in an EMERGENCY (abnormal "
si tuation)? (1.0)
ANSWER 4.02 You will deal with abnormal 0Ps, instrument alarm response pro-cedures, and ops per the BVPS Op Manual, including immediate actions and monitoring automatic response Reference ( s)
BVPS Emergency Preparedness Plan, Issue 7, Rev. O, paragraph 6.1.
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Points Avail able QUESTION 4.03
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a. As a permanent employee of Duquesne Light you are allowed the following maximum radiation exposures: (Complete the table.) (2.0)
Part of body Cum. Dose, rems Quarterly dose, rems Whole body Skin of whole body XXXXXXXXX Extremities XXXXXXXXX Your whole body exposure. limit in any calendar year is .
(0.5)
ANSWER 4.03 (N-18) 3 .75 rem - 1 r-e :+ f
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Reference ( s)
BVPS Radiation Control Manual, Chap 1, Issue 3, Rev. 8, page .
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Points Available QUESTION 4.04 ,
A " Radiation Area" is a controlled area in which personnel may be exposed to a maximum of mr/hr, or, mr in consecutive days. A "High Radiation Area" is a controlled area where personnel could receive an exposure of mr/hr, or mor (2.0)
ANSWER 4.04
Si ***" ****d 100 5 100 N -- A.A a u d k/:. .
ne A4 , 94 6. i oew w e i n.- r . /., ..
Reference (s) soo ~,A 4, ww #:.. . w,.uy .+
y a a.sw~ us 10 CFR 20.20 no~ sn J A , o a.as w -s a s, f,.- 4.,
s o f . , ,,,, ~ 9 :s fe.A. w) - e QUESTION 4.05 Answer the following TRUE or FALS According to Station Admin-istrative Procedures, the Nuclear Control Operator is allowed to do the following with no SR0 present: Manually trip the reactor if he verifies that an uhsafe condition exist (0.5) Reduce power by inserting rods or adding boro (0.5) Increase power by withdrawing rods or dilutin (0.5)
ANSWER 4.05 True True Fal se Reference (s)
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Points Avail able '
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QUESTION 4.06 ,
During refuel 4ng operations, the reactor operator on duty notes that the cavity refueling water has started to leak past the '
seal around the flange of the reactor. ,The sump is filling at a rate that projects to 2 hours2.314815e-5 days <br />5.555556e-4 hours <br />3.306878e-6 weeks <br />7.61e-7 months <br /> before the water level has dropped to flange level . What two actions are crucial to avoid excessive radiation exposure to personnel? (2.0)
ANSWER 4.06 All highly radioactive material, such as burned fuel and core components must be returned to the core or to the spent fuel poo As soon as possible, the gate valve on the spent fuel pool side of containment must be closed. (The water level in the spent fuel pool should be restored as soon as possible, and the refueling cavity should be drained to the Refueling Water Storage tank to conserve as much water as possible and reduce the amount to be recovered and processed by the waste collection system.)
Reference (s)
BVPS Refueling Procedure, Reactor Cavity Seal, p.11. Al so ,
Cavity Water Seal Failure incident at Haddam Neck (Connecticut Yankee) on August 21, 198 Cavity lost 200,000 gallons in 20 minutes, only 40,000 gallons were returned to the RWS No fuel was exposed. The valve to the Spent Fuel Pool was closed to prevent significant change in leve .
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Points Avail able QUESTION 4.07 ,
What is your first automatic indication (alarm) that the control rods are not moving in AUTO during a turbine load increase? (1.0)
ANSWER 4.07 T-avg - T-ref deviation alarm. (This will most likely happen before the pressurizer pressure or level alarms come in.)
Reference ( s)
BVPS Abnormal Operating Procedures, A0P-1, Issue 2, Rev. QUESTION 4.08 If the Pressurizer Pressure Control is failing high and you have verified the auto actions and indications on Benchboard Section B, what three Operator Actions should you take to avoid a reactor trip? (3.0)
ANSWER 4.08 (1) Secure any pressurizer heaters that may be energize "'" 7'# *// * '#'
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gradually modulate it open to reduce pressur (3) If pressure is increasing rapidly, open PORV, being sure that valve is unbic:ke ,c,,,g,J , P
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Reference (s)
BVPS Abnormal Ops Procedures, A0P-9, Issue 2, Rev. 3, p. 2 .
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Points Avail able QUESTION 4.09 ,
You are operating at power with no changes in progress. Suddenly a large number of process alarms on several annunciator-panels come in simultaneously. You notice there are no changes in the electrical load. Before you can take any action, there is a reactor trip fol-lowed by a turbine trip. What do you first suspect has happened? (2.0)
ANSWER 4.09 Most likely 120V AC vital bus one, three, or four has been los Reference ( s)
BVPS Abnormal Operating Procedures, AOP-17, -21, -23, Issue 2, Rev. 3, pp. 52, 63, 6 s
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Points Avail able QUESTION 4.10 ,
A general operating precaution is that while the reactor is criti-cal at minimum power, any plant change should be avoided that '
produces a sudden change of order 10*F in the ' reactor coolant temperature or -10 ppm dilution of coolant boron concentratio Give three reasons why these changes should be avoide (1.5)
ANSWER 4.10 (l> A sudden dilution or cooling could put the reactor on a fast ramp of increasing power when the rest of the plant is not prepared for it.(2)When the plant is at minimum power, safety system set-points are set low and a power increase may cause a reactor trip or challenge a safety system.Q)A sudden increase in temperature may reduce power and cause an upset in plant systemss(b Cooling or dilution may cause a violation of required Tech. Spec. Shutdown Margi Reference ( s)
BVPS General Operating Instructions, precaution 5, Issue 2, Rev. 1, p. 1.
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Points Available QUESTION 4.11 ,
If reactor poker is greater than 90%, with Rod Control in AUTO, Bank "D" must be withdrawn 215 steps or greater. Briefly explain wh (2.0)
ANSWER 4.11 If a control rod drops, the bank will step out in an effort to maintain power. This could lead to a flux tilt that violates Tech Spec Reference ( s)
The limitation statement appears in BVPS General Operating Instructions, Issue 2, Rev.1, p. QUESTION 4.12 At several points in the Load Following procedure, the operator must request Chemistry to make an Iodine Isotopic analysis on the coolan Briefly explain why that must be don (2.0)
ANSWER 4.12 If there is failed fuel, it should show up as as Iodine in the coolant within 2 to 6 hours6.944444e-5 days <br />0.00167 hours <br />9.920635e-6 weeks <br />2.283e-6 months <br /> after a fairly rapid power change (15%/hr) is complete. Tech Specs limit the amount of iodine allowed in. the coolant, depending on power level and time. The time limit is to accommodate iodine spiking phenomena and allow the cleanup system time to bring the iodine level in the coolant back to equilibriu Reference ( s)
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BVPS General Operating Instructions for Load Following, Issue 2, Rev.16, pp. 38, 39 (and other power change procedures in the .
same ref.). Tech Spec Table 4.4-12, 4b, sets the requiremen The Bases discuss iodine spikin Section 4 continued on next page -
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'UESTION 4.13 , After a Safety Injection has been initiated and Reactor Cool-ant System pressure has decreased due to a LOCA, it is neces-sary to shut off the Reactor Coolant Pumps. What are the two ,
conditions, either one of which is controlling, that the operator must look for to determine when the RCPs must be stopped? (2.0) Explain what it means to RESET S (2.0)
ANSWER 4.13 RCS pressure <1350 psi, CCW to RCPs isolate RESET s. hall mean the act of resetting the initiating signal, but does not mean that ECCS or Containment Isolation equipment should be restored to normal operating configuration. RESET will be perfonned to allow termination of operating equipment not required at this point in time (e.g., diesel generators, low head SI pumps et Reference (s) BVPS Emergency Operations, Diagnostics Figure E-1, Issue 2, Rev. 1, p. BVPS Emergency Operations, Summary Description, note, Issue 2, Rev. 1, p. 5.
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