IR 05000470/1985001
| ML20135A108 | |
| Person / Time | |
|---|---|
| Site: | 05000470 |
| Issue date: | 08/22/1985 |
| From: | Dudley N, Dante Johnson, Kister H NRC OFFICE OF INSPECTION & ENFORCEMENT (IE REGION I) |
| To: | |
| Shared Package | |
| ML20135A107 | List: |
| References | |
| 50-470-85-01, 50-470-85-1, NUDOCS 8509100004 | |
| Download: ML20135A108 (73) | |
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O. S. NUCLEAR REGULATORY COMMISSION REGION I OPERATOR LICENSING EXAMINATION REPORT EXAMINATION REPORT N /85-01 s
' .- , FACILITY 00CKET NO. 50-470 LICENSEE: Combustion Engineering 1000 Prospect Hill Road Windsor, Connecticut 06095 FACILITY: Combustion Engineering Training Center EXAMINATION DATES: June 18-20, 1985 CHIEF EXAMINER: //
N. Dudley, Red: tor Engineer Examiner 8-/f-8 8 Date
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REVIEWED BY _ lef,' P4 ject Section 1C - 7[,/fr Date APPROVED BY: s CWief, Podfe n 3 ranch No. 1 L [[
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SUMMARY: Examinations were administered to twelve Instructor Certification candidates, and nine Instructor Certificates were issued. Continuing defic-tencies in training material were identifie / t
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REPORT DETAILS TYPE OF EXMS: Initial' Replacement X Requalification EXM RESULTS: l Inst. Cert l l Pass / Fail l y I l l 1 l l Written Exam i 10/2 l ' r I I I I I i 10ral Exam i 10/0 l l l l l 1 I l Simulator Examl 7/3 l ; I I I I I I l0verall l 9/3 l l l l
I I i CHIEF EXMINER AT SITE: J. W. Upton, Jr. PNL OTHER EXMINERS: J. D. Smith, PNL ' R. G. Clark, PNL i . O
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3 Summary of Training Deficiencies Observed in Simulator or Written Examinations: There is an inconsistency between normal plant parameters as presented in Emergency Operating Procedures, the lesson plan on Reactor Coolant System, and the lesson plan on Reactor Theory. None of the parameters correspond to the values for the core model used for the simulator as presented in CEN-128, vol 1. For example, the value for delta T across the core varies by 5% between all references. This results in many different " correct" values for exact parameters which would be unacceptable at an operating facili t The simulator malfunction sheet provided for preparation of the simulator i examinations was inadequat The single line for a malfunction does not provide sufficient detail to an examiner to determine what type of failure is simulated, the expected plant response, nor the expected operator respons Some candidates did not understand the effect that the Flow Dependent Setpoint Selector Switch has on the Reactor Protection System setpoint Some candidates were unfamiliar with the technical specifications relating to the spent fuel building ventilation syste . Personnel Present at Exit Interview: NRC Contractor Personnel J. W. Upton, Jr. , PNL J. D. Smith, PNL R. G. Clark, PNL , Facility personnel W. E. Burchill R. E. Price P. J. De11 arco
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4 Summary of NRC Comments made at exit interview: The individuals who had clearly passed the oral and simulator examinations were identified. The examiners made the following observations: Knowledge and understanding by the candidates in the areas of
, radiation limits, protection and instrumentation was weak, Verbal communication between team members during the simulator examinations was noticeably inadequat Knowledge and understanding by the candidates in the areas of neutron detectors and the excore/incore safety and control systems was good, The Center staff assigned to run the simulator examinations were competent and helpful.
, CHANGES MADE TO WRITTEN EXAM DURING EXAMINATION REVIEW: SEE ATTACHED SHEETS (6) -
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' Written Examination and Answer Key (SRO) l Facility Comments on Written Examinations made after Exam Review :
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EASikl11 RK3tlEM DE 185_1312158 22hM18h2198 C-E TRAINING CENTER WINDSOR, C JUNE 18, 1985 Attendees Upton, J Chief Examiner, PNL R. G. Clark PNL J. D. Smith PNL 1. E. Price C-E Section 8 E. S. Ressort C-E Section 7 J. A. Magennis C-E Section 4 P. J. De11aroo C-E Section 3 raaiiiiv c namani la-QMERI1QN L.Ai The roaster operator would never oatoulate KVA. One sound expect him to determine MW and MVARS from the Plant Physics Book, but not EV Rameluting Any training with respect to power and reactive power would teach the relation between watta, vars and voltamps. No change to the answer ke raattilw Cammant te_OME111Qg S.07h We haveNPS " actual" removed in our traininS all reference to " required" vs The candidates may just comment that the surve in the suppliers manual would show an increase in NPS ISagluting Any phraseology that says that the NPSH would increase with flowrate will be acceptable LAgi1ttv C2masM.,,,1g_Q1lERZlDU2_2.91_,53d L 01 Some spread in the numertaal answer should be accepted because the sandidates may use the simulator values for T_C and T ave at 100% power which are lower than the published value Mll1R ,
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A spread in the numertosi values wn!! be accepted ,
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i Facility Co==ent to QUESTION 6.01 } QUESTION 6.01 was modified prior to the taking of the examination by the candidates such that the second sentence read, " Figure
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6.01 shows the * FLOW DEMAND' signal being routed to the main feedwater valve controller." The facility reviewers provided a
I copy of the flow diagram during the examination revie i
! Ramalution The answer will be based on the flow diagram provided by the reviewer In a list format, the answer is now: . level . ! . feedwater flowrate ; . steam flowrate
! . level setpoint .
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the level signal is passed through a lead / lag circuit and then compared to the level setpoint i
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the feedwater flowrate is passed through a lead / lag circuit and then compared to the steam flowrate '
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these two signals are added and used to control the main feedwater control valve i (+0.5 for each bullet, +3.0 max) i
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Faeility Co- ent to OUESTION 6.02 The answer to part "a." is incomplete and should include as
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"use the other excore channels". The answer to "b." should be corrected to read, "Psr pressure and T.H". Part "c." is also incomplete and should be modified to include a "4." which would cover the fact that there are 2 channels of the RRS and the operator could choose the other channel
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Resolution The answer key will include the additions to parts "a." and
"o." and athe change to part "b.". '
following manner: The answer will read in the 4 . _T power i feedwater flowrate, feedwater temperature and S/O pressure (secondary ottorimetric) 3, incore neutron detectors T H. T_C and primary-coolant flowrate (primary calorime-
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! other excore neutron detectors
- (+0.5 each, +1.0 max) Pressuriser pressure and T_H (+0.5)
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l l . T_H and T_C S/G pressure incore T/Cs other channel of the RRS (+0.5 each, +1.0 max) Emellity Co==ent to OUESTION 6.08 The answer to part "b." should include, " Select the other RRS channel after taking local control of the Pressuriser-level set point."
. Resolution The statement above waill be accepted as an alternate answe Facility Co== ant to OUES?f0N 7.01 The answer to part "d " should accept "(8.)"
as an alternate
answer because during the blowdown of S/G A the RTDs would behave irraticall Resolution Either "(6.)" or "(8.)"
will be accepted as an answer to "d.". Facility , Co== ant to OUESTION 7.02 l T
'the he reviewers showed documentation that supports " increases" as answer to parts "a." and to "b.".
Resolution The answer to parts "a." and to "b."
will be " increased".
Facility Co==ent to OUE9?fCN 7.04 The sen reference is correct, but it is not the reference that was It is an old EOP and hence the answer is wrong. There are 4 success CEA not inserted.paths and the statements should refer to greater than 1 This question is not a good question as it i requires more memorisation than should be required,
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i Resolution The question will be retained but modified to be consistent with the latest version of the EO The corrected answer is the following: CEA insertion (+0.7) used when there is greater than 1 rod ' bottom light not lit (+0.2) or the equivalent on the metrascope (+0.1) borate via the CVCS (+0.7) used when there is greater than
- 1 CEA not inserted and the reactor power is >1% or increas-ing (+0.1) and the RCS pressure is >1250 psia (+0.2) borate via the HPSI (+0.7) used as in "2." above (+0.1)
but with the RCS pressure (1250 psia (+0.2) , CEA drive-down if path 3 fails (+1.0)
(+3.0 max)
Facility Co==ent to OUESTION 7.05
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This is not an appropiate question as it is beyond what an operator should know by memor An operator would review the procedur Resolution The question will be retained; but in the grading, a general description of the situation will be accepted for (+1.4).
Facility Co==ent to OUESTION 7.06 i The RCP-seal high-temperature alarm setpoint is 200 Resolu+1on The alarm setpoint is not required for full credit.
i i Facility Co==ent to OUEETTON 7 08 The answer is essentially correct as far as AOP-3 is concerne , However, an operator would also consider the possibility that the float valve on the head tank was stuck open.
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j Resolution ( The answer key will be modified to accept as an answer operator actions to verify the status of the valves and to restore CC Reference to the concern about a RCS leak will be required for full credi . The answer key is the following: Locate and isolate the leak from the RC ' Verify the status of the valves to the head tank and restore the CC Check for indications of RCS leakage into the CC ' C+1.5 for either answer)
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Facility Co==ent to OUESTION 7.09 QUESTION 7.09 is not a good question as it requires an unnecess-ary amout of memorizatio , Lesolution There is-no change to QUESTION 7.0 Facility Comment to OUESTION 8.01 An answer to part "b." that says, "2 independent HPSI pumps are required to be operable" is indeed the correct answer, but it is often assumed that "2 HPS! pumps" means that they are being powered off of seperate l'use should be differen Part "c."
Hence the distributuin of credit
is not c omp l e te as Tech-Specs also specifies that the plant should be in Cold Shutdown within the next 24 hour Resolution The point credit for part "b." will be (+1.0) sentence and (+0.5) for the second sentenc for the first The sentence, "Re in Cold, Shutdown wilthin the next 24 hours. (+0.25)" will be i added to tan answer to part "c.", . The maximum credit for part
. "c." will remain as (+1.5 max).
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Faci *ity Comment to OUESTION 8.02
'l QUESTION 8.02 asks only for "the maximum time" and hence an answer of 2 hours should be sufficien Resolution ,
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l Facility Co==ent to OUESTION 8 05
"We take exception to QUESTION 8.05 hour action statements should not need to be committed to memor All that an operator needs to know is that there is an action statement that pertains to this situation."
Rasolution No change to QUESTION 8.0 Facility Co==ent to OUESTION 8 06 , The answer to part "c." is correct according to Tech-Specs, but in the Administrative Procedures it specifies that the' Shift Supervisor should approve of the chang Resolution No change to QUESTION 8.0 Facility Co==ent to OUESTION 8.07
"We fail to see the intent of this question and feel that it is not appropiat The Shift Supervisor's approval is not required in any of these situations. Acccrding to AOP-8 the HP Supervisor and HP Department Head are the;same perso Resolution HP Supervisor and HP Department Head will be accepted interchang-embly in the answe Facility Co==ent to OUES?f0N 8 09 Same as 8.07 in that there is no need for the Shift Supervisor to know thi Resolution No change to OUESTION 8.0 '
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U.S. NUCLEAR REGULATORY COPNISSION SENIOR REACTOR OPERATOR LICENSE EXAMINATION Facility: C-E TRAINING CENTER (470) Reactr Tyse: PWR-CE Data Amisistered.- June 18, 1985 Exminer: Joe Upton et al . , PNL Candidate: M ASTC INSTRUCTIONS TO APPLICANT:
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Print your name on the line above marked " Candidate." The grade points available for each question are indicated within parentheses after each ques-ti on. The passing grade is at least 70% in each of the four (4) categories and is at least 80% for the total grad Use separate paper for your answers and write on only one (1) side of the paper, unless a specific question instructs you otherwis Staple this question package to your answer sheets. The exami-nation questions and answers will be picked up six (6) hours after the examina-tion was started. Read the statement at the bottom of this page. When you have finished this examination, affim the statement by signing your nam Category % of Applicant's % of Value Total Score Cat. Value Category
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25 75 5. Theory of Nuclear Power Plant Operation, Fluids and Thermo-dynamies 25 25 _ 6. Plant System Design, Control and Instrumentation 25 25 7. Drocedures - Normal, Abnomal,
- Emergency and Radiological Contro:
25 25 6. Acministrative Procedures, Conditions and Limitations 100 100 TOTALS Total Grade % All work done on this examination is my own; I have neither given nor received * ai . Candidate's Signature - - -- - ._ -
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' ' . 1 C-E TRAINING CENTER , June 18,1985
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5.0 THEORY OF NUCLEAR POWER PLANT OPERATION, FLUIDS AND (25) l THERMODYNAMICS i Points Available l The following statenests apply ta A 5.3:,, 5.02, 5.03, 5.04 and 5.0 The C-E Training Facility " power plant" has operating con-tinuously at a steady 100% of full power for 10 days. All control rods (CEAs) are fully withdrawn from the nuclear reactor core (ARO). All controlled parameters are equal to their respective programmed values. The fuel-burnup status is that the core has reached 12K MWD /T in cycle 6. The present boron concentration is 200 ppm. The generator is operating with a pf of 0.95 laggin Use any of the provided figures and tables. Show your work and your procedures for arriving at your answer ESTION 5.01 What is the expected value for the AT across the core? (0.5) What is the expected axial neutron-flux shape; i.e., sketch (qualitatively) the themal neutron flux as a function of axial distance. Explain your sketch; i.e., provide the rationale for the shape of your sketc (2.5) If tne flowrate of the primary coolant through the core was recuced, explain the effect this would have on the AT across the cor (1.0) If the flowrate of tne primary cociant tnrough the core was reduced, explain the effect tnis would nave or the axial neutron-fiux snac (1,0) l .
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Points Available QUESTION 5.02 a. ilhat is the temperature in me Prmwizer? (0.5) b. iturt is De maryte af suecas'W te v for the coolant in the hot leg and for the casiast te me cold leg? (1.0) l c. Plot the point on the figure, h Coolant System Pres-sure Temperature t. imitations," on page 2 of the Technical Data Book that corresponds to the present operating condi-tions, and detensine the margin in psi to the 50*F Subcooled margin limi (1.0) If the power level was reduced from 100". to 0*., does the mass of the primary coolant increase or decrease? Explain your answe (1.5) OUESTION 5.03 If the control rods (CEAs) were inserted while in the manual sequential (MS) mode until Bank.6 is at 54 inches L.L3, w eTH ### and if the baron concentration was adjusted to maintain the
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present power level of 100%, what would be the new concen-tration of boron required for steady operation? Neglect any effect fras changes in the zenon or samaritmi concentration (1.5) How many gallons of Doric acic o' makeup water would have to be added? (1.0) According tc tne CEA Insertion Limits criteria, is this operation allowaole? If not, wnat insertion would be allowaole? (1.0)
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; June 18,1985 d Points Available QUESTION 5.04 The " dispatcher * calls and informs you that your power plant is
, going to haue to cavy a lagim5 pf of 0.85 due to changes in the - other paarsta=s fumetag tan lay (lefinite) gri (The initial -- plant conditions are states prior to QUESTION 5.01.)
' List the adjustments that must be made in the control room for the power plant to meet this new requiremen (1.5) After the adjustments have been made, what will be the raal or true output power, the reactive output power and the apparent output power? (1.5) OUESTION 5.05 The power plant is to be taken from 100% to 80% of full powe Asstne Tay,= Tref at both 100% and at 80%. What would be the magnitude of the change in reactivity (in
% ao) due to the change in the temperature in the power plant? Specify whether this change would add or take away reactivity from the cor (1.0) What is the change in reactivity (in % ao) due to the change in the moderator temperature? Specify magnitude and directio (1.5) If the core were a fresh core; i.e., zero MWDR on cycle 6, l how would your answe- to "b." aDove change? Expl ai (1.5) !
l Asstzing tnat this maneuve" takes 20 minutes, make a sketch i on the prov1oed graon (Figure 5.05) of tne xenon wortn as a l function of time. Show time from tne point in time of decreasing powe* and for the next 50 hour (1.5)
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Points Available QUESTION 5.06 For SESTIon 5.06, refer to Figure 5.06 which shows a condenser - hetnell and cameensate pum Recall that 1 ata = 29.92 i og . 31.4 P. p -. Select the letter designation below that most accurately gives the absolute pressure in psia in the condense (a.) 1.g3 psia (b.) 2.31 psia (c.) 12.77 psia (d.) 16.91 psia (1.0) OUESTION 5.07 Why is *NPSH" important; i.e., what concern does it address? (1.0) Why is the following statement true: "If the flowrate through a centrifugal pump was increased, then the ' required NPSH' would be increased."? (1.0) OUESTION 5.08 While cooling the power plant on natural circulation, how can the cooldown rate De controlled? How can the cocidown rate be increased? (1.0) i
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6 C-E TRAININE CDITER
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. l 6.0 PLANT SYSTEM DESIGN, CONTROL AND INSTRUMENTATION (25)
Points l Available QUESTION 6.01 .
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Figure 6.01 shmes part of the Fee 6sater Control System (FWCS).
Figure 6.01 shows the * FLOW DEMAm" s.ignal being routed to F*d #### # . t' :; i;; *;- --
:- ' W ^ *-M 0,. .; '2 } "" 't. ." Describe how this FLOW DEMAND signal is generate (Asstme that the caster controller is in AUTO.) In particular, list the sensor signals that are used as inputs, the inputs that are manually set, and specify how these inputs are processed and combine Do not describe test or calibration functions and operation (3.0)
QUESTION 6.02 What alternate sets of instrumentation can be used to verify each of the following indications? nuclear excore instrtmentation (TWO sets required) (1.0) subcooled-margin meter (ONE set required) (0.5) Tayg (TWO sets required) (1.0)
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8 C-E TRAINING CENTER
.- June 18,1985
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.5 FLOW DEMAND TO FAVCS 2 (1)
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\ 7 sw,!E t- [-;M/A4FEEDWATER To syrAss SIGNAL y8* g r .- ,
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9 C-E TRAINING CENTER , June 18, 1985 g Points Available QUESTION 6.03 Sketch the flow paths associated with Safety Injection Tank (SIT) - 11A as directed below. Indicate valve types but not valve nueer ersipuuttano -- The sketch should show the SIT and the f1_ow path (s) to the Loop 11A cold leg, including all valve (1.0) Include on the sketch the flow path (s) that connect the AUX HPSI, MAIN HPSI and LPSI to the Loop 11A cold leg. Show all valves and flow sensors associated with this portio (1.0) Include on the sketch the flow path (s) that allow for a measurement of leakage through the loop-entry check valve Show all valves and flow sensor (1.0) If the level indicator for the SIT initiates a LO alana, specify two (2) additional Control Room indications that you would check as verification that the SIT level is indeed lo (1.0) l
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June 18, 1985 ,j Points l
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Avail able I QUESTION 6.04
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Asse 22 or FALSE to each of the following statements concern-tag tse Cawtainment Iodine Removal Syste .. Each filter mit consists of a moisture seperator, a high-efficiency particulate air filter and a fa (0.5) The filter-unit fan starts on either a CIS or a SIA (0.5) c. niith two (2) filter units operating and with the other postulated, conservative asseptions associated with a LOCA, the two (2) hour dose at the site boundary and the total dose at the outer perimeter of the low-population zone are within the limits of 25 rm to the whole body and 300 rm to the thyroi (0.5) Number 12 and 13 filter units can be lined up to receive electric power from either vital bus through the use of remote-operated disconnect (0.5) l l l
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!a Points Avail able QUESTION 6.05 Specify the letter designation of the most correct statement '
from those given belo (1.0)
( A. ) The purpose of the Auxiliary Feedwater System is to pro-vide feedwater for the removal of heat fra the primary coolant after a Steam-Generator tube-rupture event has been diagnose (B.) The auxiliary feedwater pep has a capacity of 5% of rated feedwater flow at rated hea (C.) At a level of -40 inches or less in a Steam Generator, the staan inlet valve automatically opens providing stem to the Auxiliary Feedwater Pep Turbine and automatically opens the auxiliary feedwater inlet valv (D.) During plant cooldown, the preset (for the Auxiliary Feed-water Ptarp) turtine speed is greater than that required to supply feedwater for decay-heat removal and cooldown and the speed must be lowered by means of the local manual speed contro .
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Points ! Avail able QUESTION 6.06
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Describe the construction and operation of the letdown process radiation monitor. In particular: } How and from dere is the sample obtained that is to be measured for its radioactive level? -
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(1.0) What type of detectors are used and dat type of circuitry is used? What information or output is the result of the design of these detectors and associated circuitry? (1.0) Describe the impact that ABNORMAL flowrates of the Component Cooling Water would have on the operation of the process radiation monitor. Give reasons for any stated impact (1.0) How would you use the process radiation monitor infonsation available in the Control Roan to determine that you have had a sudden " crud-burst?" (1.0)
OUESTION 6.07 If, during reactor plant operations at 957, power, a feedline rupture were to occur inside the contairunent, what are ALL the Engineering Safety Features (ESFs) that could possibly be actuated and what signals will cause these actuations?
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Points Avail able ) QUESTION 6.08 t i
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The power plant is in equilibrium (steady) operation at 100% of ! full power. RRS Channel-Y is selected to calculate Tm from
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Loop-2. Ists Channel-X is selected to calculate Tm from Loop- l The Pressurizer level control is selected to Channel-X. Tem-perature element TE-111X (Loop-1 hot-leg temperature element) fails and reads 150*F lo ; Explain how and why the charging punps and letdown valves should respond to the failed TE-111 (2.0) , b. What actions should the operator take to compensate for the failed tanperature indication? (1.0) OUESTION 6.09 Will the plant trip as a result of the following situation Explain your answer Consider each situation separately, volt vital bus A is de-energized and channel B pressurizer pressure indication fails hig (0.8) The Flow Dependent Setpoint Selector Switch is placed in the 3 pumo position while at 70% power and Axial Shape Index is
- (0.8) SG 1 pressure enannel A fails to 400 psia and SG 2 oressure channel B vails to 450 osia dile at 60% powe (0.8) - End of Section 6 -
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14 C-E TRAINING CENTER
.- June 18,1985 7.0 PROCEDURES - NORMAL, ABNORMAL, EMERGENCY AND RADIOLOGICAL CONTROL (25)
Points
- Availabl e QUESTION 7.01 ..
Assume that the scenario descrfbed below has ace 9p aus cf the four (4) conditions (a, b, c, and d). - Designate the most s probable cause by choosing from the list of possible cause Consider each of the four conditions separatel Scenario: The power plant has been operating at 100% of full power for two (2) months. All four (4) RCPs trip, and after 15 minutes your operators infom you that natural circulation flow has NOT develope The wide-range levels for both Steam Generators are indicating less than 0% and TC is increasing above the Tsat for the Steam Generator (0.75) Both Tc and the pressures in the Steam Generators are increasin (0.75) The RCS loop AT is increasing, and the Pressurizer level is errati (0.75) The Pressurizer level is low, and there is a mismaten between tne loop RTDs and the core exit themocouple (0.75) Possible causes:
(1.) Condensible voids developing in RCS flow path.
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(2.) Failure of both Main Feedwater Ptanp l
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(3.) Tne Mxiliary Feeowater Pumos are operating a . maximur fl owra t (4.) Inadeouate secondary steam flowrat j
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/b i Points Avail able QUESTION 7.01 (contd)
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5.) pressarizar relief valves stuck close ti.) h E3 inventor (7.) tre than one CEA stuck in the out posi. tio (8.) Staas-line rupture on Steam Generator QUESTION 7.02 At the moment a bubble is fomed in the Pressurizer, how should each of the following parameters respond? the letdown flowrate (0.5) the Pressurizer pressure (0.5) the Pressurizer temperature (0.5) the Pressurizer level (0.5) OUESTION 7.03 Specify the letter designation of the most correct phrase with which to complete the following sentence concerning the Pres-surizer Pressure Control Syste To increase tne heat supplied by the proportional heaters with the manual / automatic station for the Pressurizer heaters and spray in the Manual Mode, you must . . .
(0.65) ( A. ! incmase tna controlle- outpu (B.' adjust tne cressure setooint higne '(C.) decrease the controller outpu (D.) adjust the pressure setpoint lowe . - Section 7 continued on next os ge - ,
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June 18,1985 / 7/ Points Avail able QUESTION 7.04
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la ac ardance vith E0P 7A, Reactivity Control, what are the TMEE (3) reactivity control success paths and inal would auca se mac
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(3.0)
_ OUESTit3s 7.05 In preparation for reactor plant heat up, RCP 11-A has been started. If both RCP 11-B and RCP 12-A had tripped on their initial start, what should be the pump starting sequence to establish three running pumps in the shortest time if all subsequent RCP starts are successful? Explain your reasonin (2.0) OUESTION 7.06 In each of the following situations, specify the conditions which would require that the power plant be triope decreasing condenser vacuum, while at 80% power (0.7) loss of an operating Component Cooling Water ptznp while at 50% power (1.1) a reactor coolant system leak wnich is slowly increasing wnile at 50% power (0.7) OUESTION 7.07
011owing a eacto* trip, wnat FOUR actions must be taker Desides verifying orocer automatic functions i' al', systems operate nonnally' 'Z.5) OUESTION 7.08 In general, dat actions need to be taken if Component Cooling
- Water is last because of a high-level in the head tank? (1.5)
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Points I Avail able ! QUESTION 7.09 l Explain dy each of the following events would or would not' rewire suspension of movement of irradiated fuel inside the
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causainmen j a. A maintenance person opens the inner door-of the air lock to exit the contairunent. The outer door of the air lock is shu (0.75) Securing of the operating shutdown cooling loop which leaves both loops secured but operabl (0.75) The damper of a spent fuel ventilation exhaust fan fails shut due to a los5 of instrissent air to the dampe (0.75) OUESTION 7.10 under dat conditions would a fonnal ALARA review be required? (1.0) OUESTION 7.11
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Indicate by title the person wio: will act as the Manager of Control Roan Operations during an emergency (0.6)
' is responsibile for the assessment, classification, and declaration of emergencies (0.6) Initially assumes the resconsibilities of tne Director o#
Station Emergency Doeration ,
(0.6)
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June 18,1985 O Points Avail able QUESTION 7.12 ; i Assume that a Steam Generator tube rupture has been verified ' at a leak rate of 2 gp Emergency Operating Procedure EDP-4 cautions the operator to reduce the ES Tg to less than 5254 before isolating the affected Steam Generator. What is the reason for this caution? - -
(1.0)
QUESTION 7.13 Following a Steam Generator tube rupture, the operator is instructed to control the RCS pressure, maintaining it below 1000 psia. What three (3) systems are available to the operator to effect this control of the pressure? (1.8)
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c2 8.0 ADMINISTRATIVE PROCEDURES, CONDITIONS AND LIMITATIONS (25) Points
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Available QUESTION 8.01 For each of the following situations indicate est QIT.1Derr, l if any, applies and what ACTION, if any, should be take Con-sider each situation separatel Diesel Generator A's operability load test, which is required every 31 days, is scheduled for today. The last three tests were ccmipleted 36, 68, and 102 days ago, respectivel The plant is at 100% powe (1.5) The plant is at 295'F and heating up at 1*F per minute, when an HPSI pump is found to be inoperabl (1.5) c. The plant is at 100% power den it is detemined that the heat tracing circuits for both boric acid storage tanks are inoperable and cannot be repaired for 4 day (1.5) i
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Points Avail able QUCSTION 8.02 a. Complete the following table to indicate the minimum shift crew composition in the applicable mode (2.0) MINIMim 5HIFT CREW COMPOSITIO4 #
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LICENSE CATEGORY APPLICABLE MODES 1, 2, 3, & 4 5&6 SOL OL Non-Licensed Shift Technical Advisor b. If the minimum shift crew composition of Question 8.02 above cannot be met, what is the maxima time allowable to restore the shift crew composition to within the minime requirements? (1.0)
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June 18,1985 s-x -- Points Available QUESTION 8.03 -
~ If a bypass device is to be used on a system and it is detemined that using the bypass device WILL cause adverse environmental impai:t, dut form of approval is morired to use the bypass device? (1.01 Can safety tags be lifted for any reason other than clearing the tags? EXPLAI (1.0)
00ESTION 8.04 Classify the following conditions according to the " Emergency Pl an" in E0P-9. Consider each part of the question, each event, as separate and unrelated to the other event Specify each as Unclassified, Unusual Event, Alert, Site Area Emergency or General Emergenc (3.0) Power: 100% All Ts: (Tc , Th , Tave, Tfeed, Tsteam, ...): nomal Pzr level: nomal Pzr pressure: nomal Letdown process monitor: alams Ig i Chemistry analysis of primary coolant: 256 uCi/gm I Electrical: nomal Power: 100% All Ts: nomal Pzr level: -5t and decreasing oz oressure: 2200 osia and oecreasing Containment pressure: 1 osic Contaiment radiation moni to's: r id mR/hr Electrical : nomal Power: 100% All Ts: nomal Dzr level: -2% and increasing Pzr pressure: 2220 and decreasing Blowdown process monitors: increasing g, j Condenser air ejector monitors: 5x10-3yi/cc I 1 l Electrical : 6.9 kV and 4.16 kV buses are lost
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.- 6 Points
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Avail able QUESTION 8.05 - The Tech-Specs specify a' Limiting Condition for Operation (LCO) with respect to the Auxiliary Feedustar systa ..
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; What is the LC0 for Ibdes 1-3? (1.51 I For what does the OPERA 8ILITY of be Auxiliary Feedwater Systes provide assurance? (1.0) With one (1) auxiliary feedwater pap inoperable, what action is required? (1.0)
00ESTION 8.06 List the letter designations of those statements chosen from the following statements which are correct. The statements are in I response to, " Temporary changes to procedures may be made provided:" (2.0)
(a.) Critical operation of the unit shall not be resumed until authorized by the Comission (U.S. NRC).
(b.) The intent of the original procedure is 'not altere (c.) The change is approved by two (2) members of the plant management staff, at least one of w)om holds a SRO License in the affected uni (c.) The enange is cocuentec, reviewed by the POSRC anc acor oved oy the Dian: Manage- wi nin 21 days of impl emen tati on.
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Points Avail able QUESTION 8.07 18ut pemission (from whom ad at dat check points) does a radiation worker need to complete a task utrict is espected to tacrease Ws exposure Ipy ZIX! ww trs W h erker j is E years old, has a cryletma E Fwe 4 ame nas a radiation *
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history of 13000 mren lifetime, 2I00 mrum for. the year, and 600 mram for the quarte (1.5) 1 QUESTION 8.08 Technical Specification 3/4.1.1.5 states the lowest loop operating tamperature for the RCS T a ve shall be 3515'F when the reactor is cri tical . Explain dat four (4) things this specification ensure (2.0) How often must this be determined when Tave is less than 525*F with the reactor critical? (0.5) OUESTION 8.09 What is required of personnel before they can be designated as
" escorts" in Radiation Wort Areas? (1.0)
00ESTION 8.10 When a shift sucervisor places his/ hee signature on an RWP, ne/she is verifying that certain conditions nave and will List I cxist, and that certain concitments will be kep two (2) of tnese conditions or commitwent (2.0: I
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5.0 THEORY OF NUCLEAR POWER PLANT OPERATION, FLUIDS AND (25) THERMODYNAMICS ' Points Available The following statements apply to QUESTIONS 5.01, 5.02, 5.03, 5.04 and 5.0 The C-E Training Facility " power plant" has been operating con-tinuously at a steady 100*. of full power for 10 days. All control rods (CEAs) are fully withdrawn from the nuclear. reactor core (ARO). All controlled parameters are equal to their respective programed values. The fuel-burnup status is that the core has
, reached 12K MWD /T in cycle 6. The present boron concentration is 200 ppm. The generator is operating with a pf of 0.95 laggin Use any of the provided figures and tables. Show your work and your procedures for arriving at your answer .
00ESTION 5.01 What is the expected value for the AT across the core? (0.5) What is the expected axial neutron-flux shape; i.e., sketch (cualitatively) the themal neutron flux as a function of
axial distance. Explain your sketch; i .e. , provide the rationale for the shape of your sketc (2.5) If the flowrate of the primary coolant through the core was reduced, explain the effect this would have on the AT across the cor (1.0) If the flowrate of the primary coolant through the core was reduced, explain the effect this would have on the axial
- neutron-flux shap (1.0)
ANSWER 5.01 If, for 100*. of full power TH = 599'F TC = 544*F, l then aT = 55' (+0.5) 52 7, Wf
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Points Available i ANSWER 5.01 (contd) core height h neutron core midplane flux (+1.0' for shape) With a fuel-burnup status of 12X MWD /T, the core is at End-of-Li fe (EOL) (+0.5). At EOL, the U-235 fuel has been dis-proportionally burned-up in the bottom half of the cor This shift in the relative fuel concentration to the top half of the core is responsible for the peak in the top half of the core. At full power the moderator density , decreases with core height. The higher moderator density in the bottom half of the core is responsible for the thermal-neutron peak in the bottom half of the core. (+1.0) c. If the flowrate is reduced, more heat will be added to the coolant as it passes through the core. This will raise the core AT (+1.0). An increase in the temperatures of the core will cause the power produced in the core to try to decrease. However, if the electrical output is to remain the same, the secondary system will adjust and the increase in the core aT will remain to offset the decrease in the flowrate of the primary coolan l If the flowrate of the primary coolant is reduced, the core temperature will rise. The increase in temperature will be greater in the top half of the core than in the bottom hal This will cause the neutron flux to shift to the bottom of the cor (+1.0)
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Points Available
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ANSWER 5.01 (contd) Reference ( s) C-E Training Center: " Reactor Theory," Flux Distribution, Section 6.3, Figures 6-24 through 6-29 and Figure 6-3 QUESTION 5.02 What is the temperature in the Pressurizer? (0.5) What is the margin of subcooling in *F for the coolant in the hot leg and for the coolant in the cold leg? (1.0) Plot the point on the figure, " Reactor Coolant System Pres-sure Temperature Limitations," on page 2 of the Technical Data Book that corresponds to the present operating condi-tions, and detemine the margin in psi to the 50*F Subcooled margin limi (1.0) If the power level was reduced from 100". to 0*., does the mass of the primary coolant increase or decrease? Explain your answe (1.5)
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b ANSWER 5.02
. The temperature in the Pressurizer is the saturation tempera-ture for 2250 psia and is 653*F (+0.5 for 651*F to 655*F). Margin of subcooling = Tsat (2250 psia) - TH for the hot leg -
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= 653 - 599 = 54*F (+0.5 for 52*F to 56*F)
for. the cold leg = 653 - 544
= 109'F (+0.5 for 107*F to 111*F)
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. Points Availabl e ANSWER 5.02 (contd) c. See the attached figure. (+0.5) The psia margin to the 50*F subcooled curve is 2250 - 1500 = 750 psia (+0.5) d. Considering the dashed curve on the figure on page 6 of the Technical Data Book, the actual operating point is at a higher volume than that corresponding to a constant mass. Hence, in going from 100% to 0% the mass,has increased (+0.5). The volume' has decreased; the temperature'has also decreased , which increases the density of the coolant. The increase in density produces a greater effect than that of the decrease in volume, so the mass has increased (+1.0).
Reference ( s) C-E Training Center: Technical Data Book, p. . Generic: Academic Program for Nuclear Power Plant Personnel," Volume III, pp. 2-45 through 2-56, General Physics Corporatio <
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REACTOR COOLANT SYSTEM PRESSURE TEMPERATURE LIMIT ATION S i e 4 . 9 J 2@ - LOWEST
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SERVICE TEM , (g, f} ,gfd l NEATUP
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2000
- COOLDOWN g E
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g - ,a
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E 90*F 5 SU8 COOLED g MARGIN
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MAXIMUM PMESSURE PCR SDC OPERATION 270
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O 100 200 300 ADC 900 . 800 INDICATED REACTOR COOLANT TEMPERATURE T,'F
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FIGURE 5.0 (ANSWER)
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Points Available 00ESTION 5.03 If the control rods (CEAs) were inserted while in the manual sequential (MS) mode until Bank 6 is at 54 inches inse<4ed,"'4, and if the boron concentration was adjusted to maintain the present power level of 100%, what would be the new concen-tration of boron required for steady operation? Neglect any effect from changes in the zenon or samaritan concentration (1.5) How many gallons of boric acid or makeup water would have to be added? (1.0) According to the CEA Insertion 1.imits criteria, is this operation allowable? If not, what insertion would be allowable? (1.0) ANSWER 5.03 Using the figure on page 11, ao rods = 0.5% ak/ (+0.5) Using the figure on page 17, 0.5% ak/k = 50 appm-boro Or using the data on page 7, (0.5% ak/k) (87 pps/% ak/k) = 43.5 appm-boro (+0.5 for either appe procedure) Hence, the new baron concentration is 200 - 50 = 150 pp (+0.5 for 200 - appm) Using the equation on page 16, MWV = 63000 in 200/150 = 18,124 ga Or using page 19, the curve shows 2 x 10# gal .
(+1.0 for either result) No, at 100% power the limit is 80 inches on Group See page 1 of the Technical Data Boo (+1.0)
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Points Available ANSWER 5.03 (contd) Reference ( s) C-E Training Center: Technical Data Book, pp.1,11,16, 17 and 1 QUESTION 5.04 The " dispatcher" calls and inforsns you that your power plant is going to have to carry a lagging pf of 0.85 due to changes in the other generators feeding the large (infinite) gri (The initial plant conditions are stated prior to QUESTION 5.01.) List the adjustments that must be made in the control room for the power plant to meet this new requiremen (1.5) After the adjustments have been made, what will be the real or true output power, the reactive output power and the apparent output power? (1.5) ANSWER 5.04 a. e increase the generator field current (+0.5) e decrease the power level of the plant (MfeMMf
- throttle the turbine control valves (+0.5) H, pgessu#Ej -
borate the match Tay, to Tref (+0.5) Using the figure on page 5 of the Technical Data Boot, o 800 Mi e 500 MVARS e 943,000 XYA (+0.5 each) Reference ( s) C-E Training Center: Technical Data Book, p. . C-E Training Center: NSS Program, Turbine Generator, pp. 1, 31 and 3 Section 5 continued on next page -
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Points Available QUESTION 5.05 The power plant is to be taken from 100% to 80% of full powe Assume Tay,= Tref at both 100% and at 80%. What would be the magnitude of the change in reactivity (in
% ap) due to the change in the temperature in the power plant? Specify whether this change would add or take away reactivity from the cor (1.0) What is the change in reactivity (in 1 ap) due to the change in the moderator temperature? Specify magnitude and directio (1.5) If the core were a fresh core; i.e., zero MWD /T on cycle 6, how would your answer to "b." above change? Expl ai (1.5) Assuming that this maneuver takes 20 minutes, make a sketch on the provided graph (Figure 5.05) of the xenon worth as a function of time. Show time from the point in time of decreasing power and for the next 50 hour (1.5)
ANSWER 5.05 a. 'Using the EOL curve from the figure on page 21 of the Tech-nical Data Book, 100% -1.5% 80% -1.23% 0.27% has been adde (+0.5 for 0.24% to 0.30% and +0.5 for "added")
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June 18,1985 ' . . Points Avail able ANSWER 5.05 (contd) T,,,(100%) = 572*F j T,,,( 0%) = 532*F 40'F ATay, (100% to 80%) = (0.2)(40) = 8*F (+0.5) , Using og from page 8 of the Technical Data Book, ap = (-1.96 x 10-4) (-8)
= 15.7 x 10-4 %so = 0.157 (+0.5)
This reactivity is added (+0.5) to the core.
1 At BOL (see page 8), og = -0.32 x 10-4 tap = 0.0256.
i Qualitatively, in going from EOL to BOL, the 2g becomes smaller in magnitude, hence the reactivity change due to the moderator-temperature change would be less in magni tud (+0.5) If the temperature of the moderator were reduced, then the density of the coolant would increase which would increase the amount of boron in the core which in itself would cause a decrease in reactivity. This decrease offsets the increase in reactivity due to the more effective " slowing-down" of the more-dense moderator. Hence, the greater the initial concentration of boron, the smaller in magnitude is the moderator temperature coefficient. There is a higher concen-tration of boron in the core of BO (+1.0) See the attached figur Reference ( s) i C-E Training Center: Technical Data Book, p. . CE Training Center: " Reactor Theory," pp. 500 (82 J81/ds-23, Figures 6-34 through 6-37.
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** -3. 0 . - . 0 - L5 * -LO 10 E 20 m g g O *' . HCMt3 FIGURE 5.05 (QUESTION) - Section 5 continued on next page - .
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Points Available 00ESTION 5.06 l For QUESTION 5.06, refer to Figure 5.06 which shows a condenser hotwell and condensate pump. Recall that 1 ata = 29.92 i f Hg = 33.9 ft H2 Select the letter designation below that most accurately gives the absolute pressure in psia in the condense (a.) 1.93 psia (b.) 2.31 psia (c.) 12.77 psia ( d.) 16.91 psia (1.0) ANSWER 5.06 26 in. Hg vacuum = 29.92 - 26
= 3.92 in. Hg pressure (+0.5)
3.92 , x 29.92 1 x = 1.93 psia The answer is (a.),1.93 psi (+0.5) Reference (s)
. Generic: " Academic Program for Nuclear Power Plant Personnel," Vol . III, " Nuclear Power Plant Technology." St. Lucie 1&2: " Power Plant Thermodynamics," pp. 7-10.
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C OHp fH S K A i HeTWELL A 4 in. Hy va.a uam
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f ,w JC - - _ _ _ Aj coH9CN.sn Yz PUMP FIGURE 5.06 (00ESTION)
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Points Available QUESTION 5.07 Why is "NPSH" important; i.e., what concern does it address? (1.0) Why is the following statement true: "If the flowrate through a centrifugal pump was increased, then the ' required NPSH' would be increased."? (1.0) ANSWER 5.07 NPSH is important in considering the operation of centrifugal pump Sufficient pressure is required at the inlet to the pump in order to prevent the water from flashing to steam in the " eye" of the pump and thereby causing pump cavitation -
(+1.0). Cavitation causes erratic flowrates and can cause damage to the pump. The damage to the pianp comes from the collapse of the vapor bubbles along the impeller vanes (where the pressure is higher than in the " eye") (+0.5).
(+1.0 max) If the flowrate through a centrifugal pump was increased, then the pressure drop from the suction to the " eye" would increas Hence, to avoid cavitation the suction pressure must be higher (+1.0).
Refert sce( s) Gr.neric: " Academic Program for Nuclear Power Plant Person-r.al," Vol. III, " Nuclear Power Plant Technology," pp. 2-232 to 2-236, General Physics Corporatio . C-E Training Center: " Thermodynamics Review," p. 2 . I i
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Points Available QUESTION 5.08 ' While cooling the power plant on natural circulation, how can the cooldown rate be controlled? How can the cooldown rate be increased? (1.0) ANSWER 5.08 The cooldown rate can be controlled by controlling the steam and feed flowrate in the Steam Generators (+0.5). If the steam and feed flowrate were increased, then the cooldown rate would
; be increased (+0.5).
, Reference ( s)
, C-E Training Center: Natural Circulation-Loss of Forced Coolant Flow," p. End of Section 5 -
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6.0 PLANT SYSTEM DESIGN. CONTROL AND INSTRUMENTATION (25) Points Available OUESTION 6.01 Figure 6.01 shows part of the Feedwater Control System (FWCS).
Figure 6.01 shows the " FLOW CEMAND" signal being routed to die Ma feeb I" v^ C ";;;;;c :r;" M3h - . 'i ' "W ." Describe how this FLOW DEMAND signal is generated. (Assume that the master controller is in AUTO.) In particular, list the sensor signals that are used as inputs, the inputs that are manually set, and specify how these inputs are processed and combine Do not describe test or calibration functions and operation (3.0) ANSWER 6.01 See the attached figure which shows the input signals: e 1 eve 1 (2) e TMC L EvCL SJtrk B PAGsED e feedwater flowrate 7##t e d A 4E60/t.A(- CIhG Act ct= A*CfD T: L F/f. Sf7A e steam flowrate ietel setpoint
*THE FEEDWArFR Flew'R A rE e
IS PASSCC THRC.Ur1 A L &fk and how they are processed and contined: gg4 cyfg g y 7 ggig gja g9g gg N 'T e ' eve! ;f; :1 p::::d tr. cer : 10:1" ; -i -cu i t Tl T hf
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- Section 6 continued on next page - .
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17 C-E TRAINING CENTER
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Points Available ANSWER 6.01 (contd) Reference (s) C-E Training Center: " System Description, Feedwater Control System, San Onofre Nuclear Generating Station Units 2 and 3," Description No.1370-ICE-6425, Revi-sion 00, pp. 6-13 of 85 and 56-60 of 85, Combustion Engineering, Inc. , Windsor, C ,
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June 18,1985
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- Section 6 continued on next page - ;
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. - Section 6 continued on next page - .
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Points Available ()UESTION 6.02 What alternate sets of instrumentation can be used to verify l each of the following indications? ! nuclear excore instrumentation (TWO sets required) (1.0) subcooled-margin meter (ONE set required) (0.5) Tavg (TWO sets required) (1.0) ANSWER 6.02 . AT power Feed Gowrate, feed temperature, SG press (secondary calorimetric) Incores TH , T e, primary flowrate (primary calorimetric) 5, USE orHER EJ MtGS (+0.5 each, +1.0 max) PZR pressure, Tg (+0.5) H . Tg, Te 2. SG pressure 3. Incore TCs (+0.5 each, +1.0 anx) 4 , CTH ER JP RS c hap #rL S ! !
- Section 6 continued on next page - .
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21 C-E TRAINING CENTER
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June 18,1985 . Points Available OUESTION 6.03 Sketch the flow paths associated with Safety Injection Tank (SIT) , 11A as directed below. Indicate valve types but not valve number designation The sketch should show the SIT and the flow path (s) to the Loop 11A cold leg, including all valve (1.0) Include on the sketch the flow path (s) that connect the AUX HPSI, MAIN HPSI and LPSI to the Loop 11A cold le Show all valves and flow sensors associated with this portio (1.0) Include on the sketch the flow path (s) that allow for a measurement of leakage through the loop-entry check valve Show all valves and flow sensor (1.0) If the level indicator for the SIT initiates a LO alam, specify two (2) additional Control Room indications that you would check as verification that the SIT level is indeed lo (1.0) ANSWER 6.03 a- See the attached sketch which is Figure 1 of the section " Engineered Safety Features." (+1.0 each) d. Check the redundant LO alam which is set for 187 inches, the SIT pressure which would alam LO at 205 psig and verify the positions of valves 611 and 618. (+0.5 each, +1.0 max) Ref erence( s) C-E Training Center: NSSS Program, Simulator Training Manual, " Engineered Safety Features," pp.1- l l
- Section 6 continued on next page -
22 C-E TRAINING CENTER ,
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June 18, 1985
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- Section 6 continued on next page -
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23 C-E TRAINING CENTER
- ' June 18, 1985 .
Points Available OVESTION 6.04 Answer TRUE or FALSE to each of the following statements concern-ing the Containment Iodine Removal Syste Each filter unit consists of a moisture seperator, a high-efficiency particulate air filter and a fa (0.5) The filter-unit fan starts on either a CIS or a SIA (0.5) c. With two (2) filter units operating and with the other postulated, conservative assumptions associated with a LOCA, the two (2) hour dose at the site boundary and the total dose at the outer perimeter of the low-population zone are within
the limits of 25 rem to the whole body and 300 rem to the thyroi (0.5) Number 12 and 13 filter units can be lined up to receive electric power from either vital bus through the use of remote-operated disconnect (0.5) AN54ER 6.04 True Fal se True Fal se (+0.5 each) e Ref erence( s) C-E Training Center: Simulator Training Manual,
" Engineered Safety Features, pp.16-1 l l
l I , i
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24 C-E TRAINING CENTER
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Points Available QUESTION 6.05
- Specify the letter designation of the most correct statement from those given belo (1.0)
(A.) The purpose of the Auxiliary Feedwater System is to pro.-
vide feedwater for the removal of heat from the primary coolant af ter a Steam-Generator tube-rupture event has been diagnose (B.) The auxiliary feedwater pump has a capacity of 57, of rated feedwater flow at rated hea (C.) At a level of -40 inches or less in a Steam Generator, the steam inlet valve automatically opens providing steam to the Auxiliary Feedwater Pump Turbine and automatically opens the auxiliary feedwater inlet valv (D.) During plant cooldown, the preset (for the Auxiliary Feed-water Pump) turbine speed is greater than that required to supply feedwater for decay-heat rsmoval and cooldown and the speed must be lowered by means of the local manual speed control.
l ANSWER 6.05
- (B.) (+1.0)
' Reference (s) C-E Training Center: Simulator Training Manual, " Steam, - Feed and Condensate Systems," pp. 44 and 45.
- Section 6 continued on next page -
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25 C-E TRAINING CENTER
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Points Available QUESTION 6.06 . Describe the construction and operation of the letdown process radiation monitor. In particular: How and from where is the sample 'obtained that is to be measured for its radioactive level? (1.0) What type of detectors are used and what type of circuitry is used? What infomation or output is the result of the design of these detectors and associated circuitry? (1.0) Describe the impact that ABNORMAI. flowrates of the Component Cooling Water would have on the operation of the process radiation monitor. Give reasons fur any stated impact (1.0) How would you use the process radiation monitor infomation available in the Control Room to determine that you have had a sudden " crud-burst?" (1.0) ANSWER 6.06 a. A portion of the letdown flow is taken (via a 1/2-inch line) downstream of the letdown flow-control valves and upstream
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of the flowrate sensor and the purification filters. The pressure drop across the purification filters is used as the driving force for the flow through the process radiation
[ moni tor. In the 1/2-inch line is a sample volse which is ! " viewed * by the detecto (+1.0) The detector is a -scintillation detector whose output feeds a log ratameter channel and a linear ratameter channel. The signal is processed to measure gross activity and also 1-135 activi ty. The latter is obtained by the use of a discrim-inator circuit that is set to count only those pulses that would come from a certain energy - associated vith 1-13 A two-pen recorder for gross and I activity, selected range and H1 alam are available on IC0 (+1.0)
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June 18,1985
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Points Available ANSWER 6.06 (contd) If the flowrate of the CCW decreases sufficiently that the water temperature of the primary coolant exciting the let-down heat-exchanger equals or exceeds 145'F, then the pro-cess radiation monitor would be isolated. This location is designed to protect the process radiation monitor and the boronometer from damage from high temperature (+1.0) In order to detemine " sudden" versus " gradual," the trending inforination of the strip-chart recorder would be used (+0.5). A crud-burst would probably not cause an increase in the I-135 indication, but would cause an increase in the gross radiation signal (+0.5).
Reference ( s) C-E Training Center: Simulator Training Manual,
" Chemical and Volume Control System," pp. 9 and 1 ESTION 6.07 If, during reactor plant operations at 95t power, a feedline rupture were to occur inside the contairment, what are ALL the Engineering Safety Features (ESFs) that could possibly be actuated and what signals will cause these actuations?
Include setpoints and logi (3.1) ANSWER 6.07 SIAS (+0.4) and CIAS (+0.4) - High containment pressure (+0.2), 4 psia (+0.2) 2/4 (+0.1).
CSAS (+0.4) - High contaiment pressure (+0.2) 4 psig (+0.2) 2/4 (+0.1) SGIS (+0.4) - Low S/G pressure (+0.2) 500 psia (+0.2) 2/4 (0.1).
(3.1 max) Reforence( s) ESF Systm Description, pp. 22-2 . Steam, Feed and Cond. System Description, p. 4 .
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Points Avail able QUESTION 6.08 The power plant is inn equilibria (steady) operation at 100?, of full power. RRS Chaannel-Y is selected to calculate T ave from Loop- RRS Channel--X is selected to calculate Tave from Loop- The Pressurizer lever control is selected to Channel-X. Tem-perature elment TE-ill1X (Loop-1 hot-leg temperature element) fails and reads 150*'t low, Explain how and ey the charging pumps and letdown valves should respond t= the failed TE-111 (2.0) What actions shonald the operator take to compensate for the failed tempeerature indication? (1.0) ANSWER 6.08 , The controlling 7 ave will drop (+0.25) that will cause the RRS to produce al ministan Pressurizer level setpoint (+0.75) that will cause ninism charging (+0.5) and maxistan let-down (+0.5) . Change RRS Channeel-X to calculate Tave from Loop-2 (+1.0) or change Pressurizzer level control to Channel-Y (+0.5).
(+1.0 an) or 3r.r;r orn rpPRS w ca sterc AmA r A kw L k4 u;c- c. c: ret PM tr/r. ser Angr Reference ( s) Millstone 12- Book 9, Systm Description #5, "RRS." Millstone 12: Boot 9, Systas Description 96, " Pres-surizer Co:sterol," System 5, Figure 1.5, System 6, pp. 1- . C-E Training Center: NSSS Program, " Reactor Regulat-ing Systen.*
- C-E Training Center: NSSS Program, " Pressurizer Level
> and Pressurs Control Systans," pp. 6-14.
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. - Section 6 continued on next page - : .
. - 28 C-E TRAINING CENTER ' - June 18,1985 O
L Points Avail abl e QUESTION 6.09 Will the plant trip as a result of the following situation Explain your answer Consider each situation separatel volt vital bus A is de-energized and channel B pressurizer pressure indication fails hig (0.8) The Flow Dependent Setpoint Selector Switch is placed in the 3 pump position while at 705 power and Axial Shape Index is-0 . 3 . (0.8) SG 1 pressure channel A fails to 400 psia and SG 2 pressure channel B vails to 450 psia meile at 60% powe (0.8) ANSWOR 6.09 Yes (+0.3) de-energizing channels provide a trip signal (+0.5).
. No (+0.3) flow, power, TM/LP, and APD values are all less tnan the reduced setpoints (+0.5). Yes (+0.3) each SG pressure channel auctioneers lowest pressure from the SGs (+0.5).
Reference ( s) SD: RPS , pp . 5, 13, 14-16, 36-3 End of Section 6 -
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79 C-E TRA!N2NG CENTER ) June 18,1985
'[* l 7.0 PROCEDURES - NORMAL, ABNORMAL, EMERGENCY AND RADIOLOGICAL CONTROL (25) '
Points , Available l
- QUESTION 7.01 j l
Assee that the scenario described below has occurred for each-of the four (4) conditions (a, b, c, and d). Designate the most probable cause by choosing from the list of possible causes.
j Consider each of the four conditions separately.
< j Scenario: The power plant has been operating at 100t of full ! power for two (2) months. All four (4) RCPs trip, and after ] 15 minutes your operators inform you that natural circulation
! flow has NOT developed.
i ! The wide-range levels for both Steam Generators are indicating - less than 0% and T C is increasing above the T sat for the l Steam Generator (0.75)
i Both TC and the pressures in the Steas Generacors are
- increasin (0.75)
j The RC5 ioop LT is increasing, and the Pressurizer level , is errati (0.75) i I The Pressurizer level is low, and there is a mismatch between the loop RTDs and the core exit therinocouple (0.75) , Possible causes: i (1.) Condensible voids developing in RCS flow path.
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(2.) Failure of both Main Feedwater Pap (3.) The Auxiliary Feedwater Pumps are operating at maximum flowrat t
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; (4.) Inadequate secondary steam flowrat i i
! - Section 7 continued on next page .
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June 18,1985 Points QUESTION 7.01 (contd) Available (S.) Pressurizer relief valves stuck close . ,
(6.) Inadequate RCS inventor (7.) More than one CEA stuck in the out position.
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(8.) Steam-line ruptu're on Steam Generator ANSWER 7.01 (2.) (4.) (1.) (6.)n(8.)(+0.75each)'
Reference (s) C-E Training Center: AOP- , s
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Points Available
. QUESTION 7.02 "
At the moment a bubble is formed in the Pressurizer, how should each of the following parameters respond? the letdown flowrate (0.5) the Pressurizer pressure (0.5) the Pressurizer temperature (0.5) the Pressurizer level (0,5) ANSWER 7.02 Ouy: :1 ' I *'.trAs t at;,th = r n M M C' stable decreases (+0.5 each) Reference ( s) , C-E Training Center: SOP-C5, g. 3, Rev. 1, p. 4.
e
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Points Available 00ESTION 7.03 Specify the letter designation of the most correct phrase with which to complete the following sentence concerning the Pres-3urizer Pressure Control Syste To increase the heat supplied by the proportional heaters with the manual / automatic station for the Pressurizer heaters and spray in the Manual Mode, you must . . . (0.65)
(A.) increase the controller outpu (B.) adjust the pressure setpoint highe (C.) decrease the centroller outpu (D.) adjust the pressure setpoint lowe ANSWER 7.03 (C.) (+0.65)
Reference (s) C-E Training Center: 50P-C6, a. (4 ) , p. .
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s 33 C-E TRA8NING CENTER l
, , June 18, 1985 l Points :
Available QUESTION 7.04 In accordance with E0P 7A, Reactivity Control, what are the THREE (3) reactivity control success paths and WHEN would each be used? (3.0) ANSWER 7.04 CEA insertion (+0.7) used when greater than 2 rod bottom lights not lighted (+0.2) and rods indicate out on
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metrascope (+0.1). Borate via CVCS (+0.7) when greater thar. 2 CEA not inserted reactor power >1% or increasing (+0.1) RCS pressure >1250 psia (+0.2). Borate via HPSI (+0.7) when greater than 2 CEA not inserted reactor power >1% or decreasing (+0.1) and RCS pressure
<1250 psia (+0.2). .
4, CC A Oti r'E 00eN IF Pit T rt 3 f Azas [ (.0] Reference ( s) C-E Training Center: E0P 7A, Figure 5.5.
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34 C-E TRAINING CENTER
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Points QUESTION 7.05 In preparation for reactor plant heat up, RCP 11-A has been started. If both RCP 11-B and RCP 12-A had tripped on their initial start, what should be the pump starting sequence to establish three running pumps in the shortest time if all subsequent RCP starts are successful? Explain your reasonin (2.0) ANSWER 7.05 Start RCP 12-A then RCP 12- (+0.6) A cold RCP can be started three consecutive times if no other pump is operating in the loop. (+0.7) An RCP must be idled for 60 minutes between starts if there is , a pump operating in the loop. (+0.7)
(+2.0 max)
Reference ( s) C-E Training Center: SOP-C3, p. 2.
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Points Available QUESTION 7.06 In each of the following situations, specify the conditions which would require that the power plant be trippe _ decreasing condenser vacuum, while at 80% power (0.7) loss of an operating Component Cooling Water pump while at 50% power (1.1) a reactor coolant system leak which is slowly increasing while at 50% power (0.7) ANSWER 7.06 low Vacuum Alarm (20" Hg) (+0.7)
; loss exceeding 10 min (+0.5) or RCP seal high tamperature alarm (250*F) (+0.3) or RCP-bearing high temperature alarm (195'F) (+0.3) leak greater than charging pump capacity (132 gpm) (+0.7)
- Reference ( s) C-E Training Center: AOP 1, p. ' C-E Training Center: AOP 8, p. . C-E Training Center: ADP 3, p. 2.
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June 19,1985 Points Available OUESTION 7.07 Following a reactor trip, what FOUR actions must be taken besides verifying proper automatic functions if all systems operate nomally? (2.5)
ANSWER 7.07 Depress both reactor trip pushbutton (+0.7) Manually trip turbine. (+0.7) Open generator exciter breaker. (+0.6 ) 4 Announce Reactor Trip over the public address syste (+0.5) Re ference( s) E0P 1, pp. 2- QUESTION 7.08 In general, what actions need to be taken if Component Cooling j Water is lost because of a high-level in the head tank? (1.5)
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ANSWER 7.08 Locate and isolate leak from an RCS canponen (+1.5) n /Ehsr.# rar Srnr as o r t e r v n- v se ro re r ncA C rn r< nec Referencefs) RESTORE Ch.W. f-tfECx fot :FC:~.41:orG of R(,5 L EdN4fl Z/ r0 [' W* $ 'I*I AOP-3, p. 1.
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37 C-E TRAINING CENTER June 18,1985
- . ' O Points '
Available QUESTION 7.09 l
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Explain why each of the following events would or would not require suspension of movement of irradiated fuel inside the contai nmen ,
! A maintenance person opens the inner door of the air lock l to exit the contaimnent. The outer door of the air lock is shu (0.75) Securing of the operating shutdown cooling loop which leaves both loops secured but operabl (0.75) The damper of a spent fuel ventilation exhaust fan fails shut due to a loss of instrument air to the dampe (0.75)
ANSWER 7.09 Would not (+0.35) only one door of the air lock is required to be closed (+0.4). (+0.75) Would not (+0.35) i hr without SDC is allowed by Technical Speci fications (+0.4). (+0.75) Would not (+0.35) one exhaust fan would still be operable and that is all that is required (+0.4). (+0.75) Reference ( s) T.S., 3/4 9-12, 3/4 9-13 SD: Containment System, p. 30.
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June 18,1985 Points
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Available QUESTIDW 7.10 , Under what conditions would a formal ALARA review be required? (1.0) l l ANSWER 7.10 All jobs ard tasks that would involve radiation exposure greater than 1 person-re (+1.0) Reference (s) C-E Training Center: Radiation Protection Training Manual, p. ds-3 QUESTION 7.11 Indicate by title the person who: l will act as the Manager of Control Room Operations during an emergency (0.6) is responsibile for the assessment, classification, and declaration of energencies (0.6) initially asstanes the responsibilities of the Director of Station Emergency Operation (0.6) ANSWER 7.11 shift supervisor shift supervisor or. duty officer snift supervisor (+0.6 each) Reference ( s)
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QUESTION 7.12 Assume that a Steam Generator tube rupture has been verified at a leak rate of 2 gp Emergency Operating Procedure E0P-4 cautions the operator to reduce the RCS T h to less than 525'F before isolating the affected Steam Generator. What is the reason for this caution? (1.0) ANSWER 7.12 To minimize the possible lif ting of the Steam Generator safety valves. (+1.0) Reference (s) C-E Training Center: E0P-4, p. QUESTION 7.13 Following a Steam Generator tube rupture, the operator is instructed to control the RCS pressure, maintaining it below 1000 psia. What three (3) systems are available to the operator to effect this control of the pressure? (1.8)
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AMSWER 7.13 e Main spray depressurization e Auxiliary spray depressurization e Throttling of the HPSI pumps (+0.6 each) Reference ( s) C-E Training Center: E0P-4, p. End of Section 7
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8.0 ADMINISTRATIVE PROCEDURES, CONDITIONS AND LIMITATIONS (25) Points Available OUESTION 8.01 For each of the following situations indicate what REQUIREMENT, if any, applies and what ACTION, if any, should be taken. Con-sider each situation separatel Diesel Generator A's operability load test, which is required every 31 days, is scheduled for today. The last three tests were completed 36, 68, and 102 days ago, respectively. The plant is at 100% powe (1.5) The plant is at 295'F and heating up at l'F per minute, when an HPSI pump is found to be inoperabl (1.5) The plant is at 100". power when it is determined that the heat tracing circuits for both boric acid storage tanks are inoperable and cannot be repaired for 4 day (1.5) ANSWER 8.01 Each test should be conducted within 25% of the required time (+0.35) and each three consecutive time intervals should be within 3.25 of the required time interval (+0.4).
Declare DG A inoperabl (+0.25) Prove operability of DG B within 1 b (+0.3) Conduct load test on DG A. (+0.2) t , I.0) Two HPSI pumps are required to be operabl ( 1.7 0 Ensure I operable pumps off separate power supplie 'M_?5M l ( , c. C) l Unable to comply with LCO of Action Statemen (+0.75)
(General Statement T.S. 3.03.)
l Start shutdown within 1 hour. (+0.25) Hot standby within next ' 6 hr. (+0.25) Hot Shutdown within following 6 hours. (+0.25) c n e Sen DCh' c " w 2 Y k ( ni ST) l Reference ( s) h*8^Ml , j C-E Training Center: Safety Technical Specifications, l pp. 3/4 0-1, 3/4 0-2, 3/4 5-3, 3/4 1-16, 3/4 8-1.
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June 18,1985 Points , Available 00ESTION 8.02 Complete the following table to indicate the minimum shift crew composition in the applicable mode (2.0) MINIMUM SHIFT CREW COMPOSITION # LICENSE CATEGORY APPLICABLE MODES 1,2,3,&4 5&6 50L OL Non-Li censed Shift Technical Advisor If the sinimum shif t crew composition of Question 8.02 above ' canect be met, what is the maximum time allowable to restore the shift crew composition to within the minimum requirements? (1.0)
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June 18,1985 Points Avail able ANSWER 8.02 MINIMUM SHIFT CREW COMPOSITION # LICENSE CATEGORY APPLICABLE MODES 1, 2, 3, a 4 5&6 SOL 2 1$ OL 2 1
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Non-Licensed 3 3 Shift Technical Advisor 1 0
*Does not include the licensed Senior Reactor Operator or Senior Reactor Operator Limited to Fuel Handling, supervising CORE ALTERATIONS during fuel reloadin (+0.25 each)
b. Number of shift crew composition may be less than the minima requirements for a period of time not to exceed 2 hours M*)h4) in order to acconunodate unexpected absence of on duty shif t crew menbers provided iimmediate action is taken to restore the shift crew M? composition to within the minims requirements of Table 6. Reference (s) C-E Training Center: Technical Specifications, Table 6.2.1, p. 6- .
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June 18,1985 Points Available QUESTION 8.03 a. If a bypass device is to be used on a system and it is determined that using the bypass device WILL cause adverse environmental impact, what form of approval is required to use the bypass device? (1.0) b. Can safety tags be lifted for any reason other than clearing the tags? EXPLAI (1.0 ) ANSWER 8.03 a. POSRC approved procedur (+1.0) b. Yes (+0.5), to allow testing when it is expected the tags may be re-hung (+0.5). (+1.0) Reference (s) AP-6, pp. 9-10.
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Points Available i I 00ESTION 8.04 Classify the following conditions according to the ' Emergency Plan" in E0P-9. Consider each part of the question, each event, as separate and unrelated to the other events. Specify each as Unclassified, Unusual Event, Alert, Site Area Emergency or General Emergenc (3.0) Power: 100% All Ts: (Te , T h, Tave.-Tfeed, Tsteam, ...): nomal Pzr level: nomal Pzr pressure: nomal - Letdown process monitor: alams Chemistry analysis of primary coolant: 256fi/gm .I ' ' ' Electrical: nomal Power: 100% All Ts: normal Pzr level: -Si and decreasing Pzr pressure: 2200 psia and decreasing Contairment pressure: 1 psig Containment radiation monitors: 10# mR/hr El ectrical : nomal Power: 100% All Ts: nomal Pzr level: -2% and increasing Pzr pressure: 2220 and decreasing
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Blowdown process monitors: increasi ng
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Condenser air ejector monitors: 5 x 10-3 Ci/cc I Electrical: 6.9 kV and 4.16 kV buses are lost-ANSWER 8.04 Unusual Event (+1.0) fuel element failure Site Area Emergency (+1.0) LOCA Alert (+1.0) S/G TR Reference ( s) , C-E Training Center: E0P-9, Table '
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Points Available QUESTION 8.05 The Tech-Specs specify a Limiting Condition for Operation (LCO) I with respect to the Auxiliary Feedwater syste j What is the LCO for Modes 1-3? (1.5) For what does the OPERABILITY of the Auxiliary Feedwater ; System provide assurance?
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' c. With one (1) auxiliary feedwater pump inoperable, what action is required? (1.0) l ANSWER 8.05 .7.1.2. At least two (2) steam turbine-driven steam gener-ator auxiliary feedwater pumps and associated flow paths shall be operable (+1.0) and capable of automatically
- initiating flow, within the limits of acceptable operation to each steam generator (+0.5). The operability of the auxiliary feedwater system ensures that the reactor coolant system can be cooleo down to less than 300*F from normal operating conditions in the event of a total loss of offsite powe (+1.0) With one (1) auxiliary feedwater pump inoperable, restore at least two (2) auxiliary feedwater pumos to operable status within 72 hours or be in hot shutdown within the next 12 hours. (+1.0)
Reference ( s) i C-E Training Center: Technical Specifications, ! " Auxiliary Feedwater System," 3.7.1.2, pp. 3/4 7-4.
l C-E Training Center: Tecnnical Specifications, l " Basi s," 3/4.7.1.2, pp. B 3/4 7-2.
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Points Available 00ESTION 8.06 List the letter designations of those statements chosen from the i following statements which are correct. The statements are in j response to, " Temporary changes to procedures may be made provided;" (2.0)
(a.) Critical operation of the unit shall not be resumed until authorized by the Commission (U.S. NRC).
(b.) The intent of the original procedure is not altere (c.) The change is approved by two (2) members of the plant management staff, at least one of whom holds a SRO License in the affected uni (d.) The change is documented, reviewed by the POSRC and approved by the Plant Manager within 21 days of implementatio ANSWER 8.06 (b.) and (c.) - correct o (a.) and (d.) - incorrect (+0.5 each) Re'erence( s) C-E Training Center: AP-10, Rev. O, p. 6 of 7, Appendix 10- . !
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47 C-E TRAINING CENTER June 18, 1985
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QUESTION 8.07 l What permission (from whom and at what check points) does a radiation worker need to complete a task which is expected to increase his exposure by 2000 mrem this quarter? The worker is 30 years old, has a completed NRC Fom 4 and has a radiation history of 13000 arem lifetime, 3000 mrem for the year, and 600 mrem for the quarte . (1.5)
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ANSWER 8.07 HP Supervisor's approval for exceeding 1000 mr/qt. (+0.5)
, . :.mc. 4tP Department Head approval for exceeding 2000 mr/qt. (+0.5)
HP Supervisor and Station Superintendent approval for exceeding 5000 nr/yr. (+0.5) j Reference ( s) !
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June 18,1985 s Points Available QUESTION 8.08 Tec'hnical Specification 3/4.1.1.5 states the lowest loop operating temperature for the RCS Tave shall be 1515'F when the reactor is , cri tical . l Explain what four (4) things this specification ensure (2.0) How often must this be determined when Tave is less than 525'F with the reactor critical? (0.5) ANSWER 8.08 /4.1.1.5 MINIMUM TEMPERATURE FOR CRITICALITY This limitation is required to ensure (1) the moderator temperature coefficient is within its analyzed temperature range (+0.5), (2) the protective instrumentation is within its normal operating range (+0.5), (3) the Pressurizer is capable of being in an OPERABLE status with steam bubble (+0.5), and (4) the reactor pressure vessel is above its minimum RTND7 temperature (+0.5). , At least once per 30 minute (+0.5) Reference ( s) C-E Training Center: Technical Specifications,
" Basi s ," 3 /4.1.1.5, p. B3/4 1-2.
' 2. C-E Training Center: Technical Specifications, 3/4.1.1.5, Minimum Temperature for Criticality, p. 3/4 1- .
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Points Available QUESTION 8.09 What is required of personnel before they can be designated as
" escorts" in Radiation Work Areas? (1.0)
ANSWER 8.09 An escort must have been trained as a radiation worke (+1.0) Reference ( s) C-E Training Center: Radiation Protection Training Manual, p. ds-1 l
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QUESTION 8.10 When a shift supervisor places his/her signature on an RWP, he/she is verifying that certain conditions have and will exist, and that certain comitments will be kept. List two (2) of these conditions or comitment (2.0) ANSWER 8.10 e No plant evolutions are planned which could change the radiological conditions stated in the RW e Operations (Supervisor) will notify HP whenever any plant evolution has taken place or is to take place that would change the radiological conditions in the area listed in the RW e The plant is not and would not be jeopardized by the work indicated on the RW (+1.0 each bullet, +2.0 max) Reference ( s) C-E Training Center: Radiation Protection Training Manual , p. ds-4 End of Section 8 -
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