ML20197A556
| ML20197A556 | |
| Person / Time | |
|---|---|
| Site: | Columbia  |
| Issue date: | 04/22/1986 |
| From: | Elin J, Pate R NRC OFFICE OF INSPECTION & ENFORCEMENT (IE REGION V) |
| To: | |
| Shared Package | |
| ML20197A541 | List: |
| References | |
| 50-397-OL-86-01, 50-397-OL-86-1, NUDOCS 8605120344 | |
| Download: ML20197A556 (100) | |
Text
. _ _ _
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U. S. NUCLEAR REGULATORY. COMMISSION REGION V
-Examination Report No. 50-397/0L-86-01 Facility: Washington Nuclear Plant No. 2 Docket No. 50-397 Examinations administered at Washington Nuclear Plant No. 2, Richland, Washington from p 4c G G ~1985.
a/4 -tal 76 - /
Chief Examiner: If ./
- t Y 7[f[
R( J. Pate, Ckief Date Signed Reactor Safety Branch
, Approved: .
i &
- g. O. Elih, Chief Ddte Signed
@perations Section
"*"*'Y' (o %(o Examinations on No.e Los 6-6, ivo4 Written and operating (oral and simulator) examinations were administered to four SRO and seven RO candidates. All R0 and.SRO candidates passed the examinations.
8605120344 860424 7 PDR ADOCK 0500 V
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- a _ . _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ . _ _ _ _ , _ _ __
REPORT DETAILS
- 1. Examiners
- Lenord Wiens, NRC L. Miller, NRC R. Pate, NRC M. King, EG&G J. Sherman, EG6G
- Chief Examiner
- 2. Examination Review Meeting An exam review meeting was held after the written exam was administered on February 6, 1986. The facility comments and subsequent Region V responses are attached.
- 3. Exit Meeting At the conclusion of the site visit February 6, 1986, the examiners met with representatives of the plant staff to discuss the results of the examinations. Those individuals who clearly passed the oral and simulator examination were identified in this meeting.
The current status of the plant simulator was discussed. The simulator was found to be very limited in the number of malfunctions that could be simulated. The examiners noced that the WNP-2 simulator was marginally acceptable and the problem appeared to need prompt and effective management attention.
- a. Attendees were:
NRC Robert Pate, Chief, Reactor Safety Branch Lenord Wiens, Senior Reactor Engineer, OLB HQ Lee Miller, Training and Assessment Specialist Mike King, Examiner, INEL Jeff Sherman, Examiner, INEL Utility John Wyrick, Licensed Training Manager Jack Baker, Assistant Plant Manager, WNP-2 Lou Frank, Principle Training Specialist, WNP-2 Bob Beardsly, Assistant Operations Manager, WNP-2
- b. The examiner reported that there were four candidates that were a clear pass on the Operating Examination (Oral). The criteria used for determining whether a candidate passed the oral examination was discussed.
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WNP-2 FACILITY C0f00ENTS AND RESOLUTION REACTOR OPERATOR EXAMINATION GIVEN ON 2/4/86
- 1. Facility Comment on Question 1.02 Also give credit for using doubling count rate the new Keff is half the distance to one f.e. 100 200 cps Keff .95 Keff .975 200 250 cps Keff .975 Keff .981 Examiner Resolution .
Comment rejected, because method only works for a one-step doubling, not for this situation.
- 2. Facility Comment on Question 1.03 _
Stating half life of the longest lived precursor was not asked in the question, and should not be required for full credit.
Examiner Resolution Agree with comment and answer key changed accordingly.
- 3. Facility Comment on Question 1.04 Section d. Also accept increases - due to less flow losses see attached G.E. HTTFF pages 7-94, 7-95.
Examiner Resolution '
Correct answer is changed on the answer key to INCREASES for part 'd'.
- 4. Facility Comment on Question 1.05 Also accept the following answer from G.E. THTFF. (See attached page 9-51)
Examiner Resolution Comment rejected, because the question asks WHY is flow orificing necessary, not HOW it is accomplished. For full credit, answer must include the flow starvation effect of increased voiding on higher power fuel bundles.
- 5. Facility Comment on Question 1.07(b)
Also accept " resonance absorber build-up causing more resonance capture."
s Question did not state " list the isotope."
Examiner Resolution The facility comment is correct, but the question asks for the primary effect. PU-240 build-up is the correct answer.
- 6. Facility Comment on Question 1.08 Accept for part b "any number less than 2%". To memorize values for a table to 1/10 of a percent is unrealistic. Also there is no direct
" decay heat" meter or indicator in the control room.
Examiner Resolution The answer key was changed to accept 0.5% to 2.0% for part b; this -
increases the range of acceptable answers.
- 7. Facility Comment on Question 1.10 Accept 57 F 1 F question did not ask to determine cooldown rate to -
nearest 1/00 F. Also in changing PSIG to PSIA accept use of 15#
vice 14.7#.
Examiner Resolution Answer key changed to accept a wider more realistic range and 15 psi.
- 8. Facility Comment on Question 1.12
- a. The "why" section of the question does not ask the student to state two reasons why. " Tripping off line" should not be required for full credit. Should accept any one of the three.
- w. .
Ref. Examiners Stand ES 202 #18 open ended question's should be avoided.
Examiner Resolution Will accept over-heating, electrical damage, or tripping off-line for full credit.
- 9. Facility Comment on Question 1.13 Comment - delete the question. Question not covered by learning objectives and can not find the answer in the stated reference.
Examiner Resolution Comment accepted, question and point value (2.00) deleted. Answer can't be referenced in WNP-2 documents. Section 1 becomes 22.00.
2
w SAMPLE PROBLEM: .
(Continued)
Solution:
/'
/
/ 4 The total NPSH on the recirculation pump is calculated by b ,
first determining the inlet pressure P; in Equation 7 43.
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i V I [
i=Pdome + PHO- 2 losses where:
P dome = pressure as measured in the steam dome (lbf/f t )
P H O = pressure due to the water column (Ibf/f t2) 2 P
gg33,3 = pressure loss due to irreversible flow losses (Ibf/f t )
The dome pressure is 1000 psia. The pressure due to the height of the water column is the density of water times the height of the column (plus o change in units to Ibf/ft2). The
. density of the water.is taken as the saturation density of -
20-Btu /lb-subcooled water (47.3 lb/ft3) which con be found from a table of subcooled water properties. The irreversible' losses are o function of the square of the fluid velocity'uncthe effects of elbows, pipe fittings, and suction valve in the" recirculation pump suction line. It is normally 20 psia of rated conditions and decreases as the square of the flow rate.
Equation 7-43 beccmes:
7 , ,
- )
, U00 tbf/in < 144 in"/f t' -
m1 f' x 47.1 thm/fi 1 x 32.2 i t / wc "
'q,; !! >t. ."cD ~ Un'a
- 29 ibf /;n' /. I W m">/fi2 t
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?.t : 144,000 lbf/f t ~
7 94 L
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.. .. t SAMPLE PROBLEM:
[( (Continued)
The saturation pressure is found in the folicwing way. First, find the saturated liquid enthalpy at 1000 psia. This turns out to be 542.6 Btu /lbm. But since the water in the downcomer is subcooled 20 Btu /lbm by the feedwater, the actual enthalpy at the eye of the pump is 522.6 Btu /lbm. The saturation pump which corresponds to this liquid enthalpy if approximately 875 psia.
P = 875 E x 1441"2 = 126,000 M in 2 f,2 f,2 The NPSH is then:
(7-43) NPSH = (P; - P3) x 9 c f 0 9 (k
NPSH = (144,000 - 126,000)M x 32.2 lbm-f t 2
ft lbf-sec 2
- 4. lbm 32.2sec U.- 2 x 62.4 ff NPSH = 18.000 lbf/ff 62.4 lbf/f t 3 NPSH 288 f t of H2O The cnn -ibution of the water column above the pump can be calcuict separately by first assurning no subcooiing and neglec;iri ; tha suction line head toss. ,
- 7-95 L
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% 'd' s CORE ORIFICING We mentioned that for single or two-phase flow, the constant term k represented a resistance due the inlet orifices which are placed in the fuel support pieces in the core support plate.
One might ask why would we artificially and intentionally create a flow /estriction?
To obtain a qualitative picture of the ef fect of core inlet
/ orifices, first consider the BWR core without inlet orifices.
,/ The pressure drop across all the fuel assemblies is the same -
' since, as we said, they share a common inlet and outlet plenum. Assume further that all the bundles have the same
\ flow resistance characteristics so that, at zero power and minimum recirculation system flow, all the fuel bundles have the same flow. -
Now increase core power as in a normal startup where there are some high powered bundles and some low powered bundles.
As bundle power reaches the point of increasing water temerature in the channel, the bundle flow will increase. This occurs because the hotter water in the channel is.less dense than the water in the downcomer region and gravity will cause an increase in flow in the warmer bundles. In addition, as boiling begins, the buoyant force of the steam bubbles will cause a further increase in bundle flow.
As power continues to increase, however, the channel quality in the highest powered bundle increases as does the two-phase flow friction multiplier 2 2o (See Figure 9-20). The result is' a forge increase in flow resistance as quality increases. Since the channel pressure drnp is controlled by the inlet and outlet plenums (i.e. for constant a P), equation 9-20 indicates that the flow through the_ fuel bundle will decrease as R increases.
The result is that flow which should go to the highest powered bundle is being diverted to lower powered bundles. That is, the flow seeks the path of least resistance. This is, of course, undesireable.
Flow orifices are provided at the bundle inlet to minimize the undesireanle ef fect of a quality increase on bundle flow. The iniet orifice has the effect of proviaing a larger resistence to
!!o.v co that any additionc! flow resistance caused by two-pnose f' w is acc"ptnbly small in comparison.
Tnere e,
- riassic 2nciogy se this effect which can Se iii stretee w th~ simple electr;cai exo:nple snown in Fiqure
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' 3. The examination should include questions to determine a candidate's understanding of his responsibilities related to the administrative procedures, precautions, environmental and radiation release require-men *s, and pressure / temperature limits.
4 Questions on health physics and chemistry procedures should be determined on the casis of the facilities' type of health physi s coverage.
,,5. Extended definitions questions (e.g., 6-factor formula) should be avoided.
- 6. Questions on detailed system characteristics or instrumentation, su.n as annunciator logic or setooints, should be avoided unless required for safety system operations.
- 7. Questions should be based on
- a. a review by the examiner of material provided by the facility
- b. a review of past examinations given at the facility
- c. content validity study results, when available
- 8. Other sources of questions are g
- a. standard questions and answers
- b. Examination Question Bank
- c. examinations on similar facilities
- d. personal file of questions and answers
- a. ..
- 9. A rule of thumb is
- a. approximately 55 to 70 responses for a 6-hour examination
- b. a response that requires about 3 to 4 minutos to write
- 10. Examinations shall be 6 hours6.944444e-5 days <br />0.00167 hours <br />9.920635e-6 weeks <br />2.283e-6 months <br /> long.
__, 11. Examination questions should consist of short word sentences using the terminology of the facility as :uch as practicable.
- 12. " Discuss"-type questions should te avoided; questicns should be specific to elic; short precise answers.
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[-'13. Practical realistic questians tnat relate to aperator kno ledge and required coerating practice should be used.
p 14 Multipart cues *' ens should be troken down irto logical sequential parts. The answer eneet should 3F.os points assignea for subparts of answers.
4 Examiner Standards 4 of 6
- 10. Facility Comment on Question 2.03
- b. This requires the operator to memorize switch positions - which are labeled and would definitely be referred to where any switch man-ipulation is required. This switch is covered in our training material but memorization of each switch position is not reauired by our learning objectives.
Examiner Resolution Agree with comment and the switch position portion of answer is deleted for full credit. Interpretation of meter reading is still required.
4
- 11. Facility Comment on Question 2.05
- a. Should also accept - prevent exceeding design external to internal containment P (2 psid) (for any reason - wouldn't have to be re-stricted to " condensing steam").
Examiner Resolution Disagree with comment. The design purpose, as stated in the reference, is to prevent a vacuum in the primary containment which would occur while condensing steam.
- 12. Facility Comment on Question 2.07
- b. RCIC should also be accepted as a system redundant to HPCS see attached T.S. page B 3/4 5-2. Operators are trained to utilize RCIC as a backup to HPCS. Also recognized in T.S. 3.5.1 in Div.
3 ECCS.
Examiner Resolution Agree with comment and the answer key is changed to accept either ADS oj; RCIC.
- 13. Facility Connent on Question 2.08 Answer #1 "RCIC equipment area and/or pipe routing area high temp" should be accepted as 2 separate signals if so listed.
Examiner Resolution Agree with comment and will give credit for two separate signals, if so listed.
e 14. Facility Comment on Question 2.09
- b. The stop control for HPCS in the control room is a switch, not a push button.
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. . 6 EMERGENCY CORE COOLING SYSTEM C BASES ECCS - OPERATING and SHUTOOWN (Continued)
The capacity of the system is selected to provide the required core cooling.
The HPCS pump is designed to deliver greater than or equal to 516/1550/6350 gpm at differential pressures of 1160/1130/200 psig. Initially, water from the condensate storage tank is used instead of injecting water from the suopression pool into the reactor, but no credit is taken in the safety analyses for the
/
condensate storage tank water.
'/ : With the HPCS system inoperable, adequate core cooling is assured by the O.7 j0PERABILITYofthe.redundantanddiversifiedautomaticdepressurizationsystem .,
and both the LPCS and LPCI systems. In addition, the reactor core isolation v / cooling (RCIC) system, a system for which no credit is taken in the safety analysis, will automatically provide makeup at reactor operating pressures on a reactor low water level condition. The HPCS out-of-service period of 14 days is based on the demonstrated OPERASILITY of recundant and diversified low pressure core cooling systems.
The surveillance requirements provide adequate assurance that the HPCS system will be OPERABLE wnen required. Although all active components are testable and full flow can be demonstrated by recirculation through a test loop during reactor operation, a complete functional test with reactor vessel
.njection requires reactor shutdown. The pump discharge piping is maintained full to prevent water hammer damage and to provide cooling at the earliest moment.
.NM "E.)
Upon f ailure of the HPCS system to function properly after a small break loss-of-coolant accident, the automatic depressurization system (ADS) automa- .
tically causes selected safety / relief valves to open, depressurizing the reactor so that flow from the low pressure core cooling systems can enter the core in time to limit fuel cladding temperature to less than 2200*F. ADS is cons'erva-tively required to be OPERABLE whenever reactor vessel pressure exceeds i 100 psig. This pressure is substantially below that for which the low pres-sure core cooling systems can provide adequate core cooling for events requiring ADS.
ADS automatically controls seven selected safety / relief valves although the safety analysis only takes credit for six valves. It is therefore appropriate to permit one valve to be cut-of-service for up to 14 days without materially reducing system reliability.
3/4.5.3 500:RESSION CWMSER The suosressier. cn2mber is recuired to be OPERABLE as part of the ECCS to ensure that a suffi: tent supply of water is available to the HPCS, LPCS, and LPCI systems in t:e went of a LOCA. This limit on ser;:ression cha:ber minimum _
water volu e ensures tnat sufficient water is available to permit recirculation '
cooling fios to tne core.
The C?IMEILITY of the su::oression enameer in '
OPERATICHAL CONDIT!"N 1. 2, or 3 is required by Specification 3.6.2.1.
WASHINGTON NUCLEAR - UN!T 2 5 3/4 5-2 D 1
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3/4.5 EMERGENCY CCRE CCOLING SYSTEMS 3/4.5.1 ECCS - OPERATING LIMITING CONDITION FOR OPERATICN 3.5.1 ECCS divisions 1, 2, and 3 shall be OPERABLE with: -
U a. ECCS division 1 consisting of: -
- 1. The OPERABLE low pressure core spray (LPCS) system with a flow path capable of trking suction from the suppression chamber and transferring the water through the spray sparger to tne reactor vessel.
- 2. The OPERAELE low pressure' coolant injection (LPCI) subsystem "A" of the RHR system with a flow patn capaole of taking suction from the suppression enameer and transferring the water to the reactor vessel.
- 3. Seven OPERABLE A05 valves,
- b. ECCS division 2 consisting of:
/ 1. The nPERABLE low pressure coolant injection (LPCI) subsystems "B" and "C" of the RHR system, each with a flow path capable of taking suction from the suppression chamber and transferring the water to the reactor vessel.
\; 2. Seven GPERABLE ADS valves. -
- c. ECC5 division 3 consisting of the OPERABLE nigh pressure c::re spray (HPCS) system with a flow path capable of taking suction from the suppression chamber and transferring the water through the spray sparger to the reactor vessel.
APPLICA3ILITY: CPERATIONAL CONDITICNS 1, 2"#, and 3". N" "ine A05 is not recuired to be CPERA3LE aren reactor steam Ocme pressure is less than or equal to 123 psig.
! #See Special Test Exce:: tion 3.10.6. T WASHINGTON NUCLEAR - UNIT 2 3/4 5-1 s .
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.. . 8 EMERGENCY CCRE CCOLING SYSTEMS '
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LIMITING CCNOITION FOR OPERATION (Centinued) w.-
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i N a. For ECCS division 1, proviced that ECCS divisions 2 and 3 are OPERABLE:
I E g -> ". . s , 1. With the LPCS system inocerabla, restore the inoperable LPCS
.,-Fi system to CPERABLE status within 7 days.
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- 2. With LPCI subsystem "A" inoperable, restore the inoperable LPCI subsystem *A" to CPERABLE status within 7 days.
- 3. With the LPCS system inocerable and LPCI subsystem "A" inoperacle, restore at least.the inocerable LPCI subsystem "A" or the inoperable LPCS systam to OPERABLE status within 72 hours8.333333e-4 days <br />0.02 hours <br />1.190476e-4 weeks <br />2.7396e-5 months <br />. '-
- 4. Otherwise, be in at least HOT SHUTCCWN within the next 12 hours1.388889e-4 days <br />0.00333 hours <br />1.984127e-5 weeks <br />4.566e-6 months <br /> and in CCLO SHUTCC'aN within the following 24 hours2.777778e-4 days <br />0.00667 hours <br />3.968254e-5 weeks <br />9.132e-6 months <br />.
- 1. With either LPCI su:: system "B" or "C" inocerable, restore the inceerable LPCI subsystem "B" or "C" to CPERABLE status within 7 cays. ,
- 2. With both LPCI subsystems "B" and "C" incperacle, restore at least the inocerable LPCI subsystem "B" or "C" to CPERABLE status within 72 hours8.333333e-4 days <br />0.02 hours <br />1.190476e-4 weeks <br />2.7396e-5 months <br />.
L 3. Othemise, be in at least HOT SHUTCCWN within the next 2 hours2.314815e-5 days <br />5.555556e-4 hours <br />3.306878e-6 weeks <br />7.61e-7 months <br /> -.
and in COLD SHUTCCWN within the following 24 hours2.777778e-4 days <br />0.00667 hours <br />3.968254e-5 weeks <br />9.132e-6 months <br />".
N,/
/r\ ' RCIC system are OPERABLE:
r.
- 1) With ECCS division 3 incperable, restore the incoerable division to OPERABLE status within 14 days.
- 2) Othemise, be in at least HOT 3HUTCC%N within the next 12 hours1.388889e-4 days <br />0.00333 hours <br />1.984127e-5 weeks <br />4.566e-6 months <br /> anc in C':LD 3HUTCC*nN within the folicwing 24 hcurs.
- c. For ECC3 civisiens 1 and 2, previcac that ECCS civision 3 is CPERABLE:
1
- 1) Witn .. CI suosystem "A" acc either LPCI suosystem "S" or "C" inc:e-sola, restore at least the ico::e-acle L?CI subsystem "A" 7c cr .ne metectole L?CI suosys:.am "S" er "C" to CPERAELE stans wi nin 72 nears. ~~
%neneur wo or =cre RER suosystems are ine::erable, if unacle to attain CCLD l SHUTCC%N as reOuireo ::y this ACTICN, maintain reactor 00cian tem erature
( as low as practical oy use of alternate neat removal metnocs. -=
WASHINGTCN NUCLEAR - UNIT 2 3/4 5-2
Examiner Resolution -
Comment noted, but this does not change the answer to the question.
- 15. Facility Comment on Question 2.10
- a. Answer #1 also accept "undervoltage" on associated bus.
Examiner Resolution Facility comment is correct and answer annotated to accept undervoltage.
- 16. Facility Comment on Question 2.11 RWCU pump no longer trip on high RCC temp., they would trip on V-4 not being full open. Also FCV-33 auto closes when V-4 (or V-1) goes closed.
This response should also be accepted.
Examiner Resolution Comment was verified to be correct. The RWCU pumps will not trip directly on high RCCW temperature, but they will trip indirectly as a result of V-4 closing due to high temperature at NRHX outlet.
FCV-33 auto closes when V-4 closes.
Answer is changed to:
- 1. Affected components will be non-regenerative heat exchangers [0.25],
V-4 [0.25], reactor water cleanup pumps [0.25], and FCV-33 [0.25].
- 2. High temperature at NRHX outlet will cause isolation valve V-4 to close. [0.5] V-4 closure causes RWCU pump trip [0.25] and FCV-33 A closure [0.25]. #
Due to facility comments, reference is changed to: WNP-2 Systems, Volume I, Tab. 9, pp. 7-8.
- 17. Facility Comment on Question 2.12 Answer #1 "Feedflow < 30% w/ 15 sec. T.D." Time delay should not be required for full credit. Question asks for setpoints only.
Answer #4 ">142 # turbine press." is when the trip is, available, not the setpoint at which the trip occurs Setpoint is: when the throttle valves are not full open or upon low EH fluid pressure <1250#.
Ref. T.S. 3/4 3-44.
Examiner Resolution Answer #1: The time delay is part of the condition and is required for full credit.
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' .f If a high pressure develops in the piping downstream of the ..
2h.~,i" g %,-. :%:. FCV, a high pressure setting on PS-14 will close the FCV at -2
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- 3. xClosure of either V-1 or V-4 will cause FCV-33 to close.
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.- FCV-33 closes to prevent system depressurization which causes the hot water in the system to flash to steam (the water flashing to steam and resulting water hammer when ficw is -
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-'. re-established could possibly damage the system piping or heat
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. exchangers). -
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When V-1 and V-4 are both open, FCV-33 will autcmatically reopen to the position determined by its manual controller on -
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" M . , C. The following will cause a pump trip:
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Inlet isolation valves (V-1 or V-4) not full open '
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'-? d2. ' Low system flow as sensed by FE 70 gpm *Me >
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,..f.r ' - operated) which senses effluent flow and maintains a constant ,
., -,f2- ' flow rate through the vessel for varying pressure drops across
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t e. the fil ter. The vessel dP should range from a low of.1 psid r' 'i-
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- 2. The ficw rec e mr control stations 74A and 748 start and stop
,., their respec:' in vessel hol d pump. The hold pump will start
' r."
at less tnaa mal to 100 gpm and stoo at greater than 100
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i nc S are located on Panel 26 in the r..adwaste IV control rocm). ::y 1
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Agree with comment concerning answer #4. Answer #4 deleted and replaced with the following:
- 4. Turbine throttle valve-closure 55% closed.
- 5. Turbine Governor valve - fast closure 21250 psig.
Add to existing reference:
WNP-2 Technical Specifications 3/4 3-44.
- 18. Facility Comment on Question 2.13
flow. Also, the minimum flow valve does not receive its flow signal from the same F transmitter and therefore is not affected by this failure.
Examiner Resolution Some of the facility's comment is correct, therefore, will accept possible RCIC overspeed due to attempting maximum flow. However, the RCIC flow control transmitter (FT-3) and RCIC min flow valve flow switch (FIS-2) g are separate, but are arranged in parallel such that a break in the D/P j( cell on FT-3 will cause a zero D/P to be seen on both instruments. The min flow valve will remain open during this failure. Additional reference:
WNP-2 Drawing M519 (RCIC System)
General Coment: This section (03) was very well written. Questions were clear and concise, answers were brief but complete. Every ., ques tion
( was something an operator should know! ,
- 19. Facility Coment on Question 3.04
- b. Placing the master controllers to manual does not reset the "setpoint setdown", it would take manual control of the feed pump, but taking manual control of a feed pump controller (601A and/or B) would also give you manual control of the feed pump speed. I would not expect 2 actions or for the second action I would accept takng manual control of any of the feed pump controls.
Examiner Resolution The facility comment would achieve the desired effect. Will accept manual control of individual feed pump in lieu of master controller placed in manual in part b.
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ALLOWABLE IRIP IllitCT10!I 1 RIP SEIP0lNT VALUE E
p 1. luibine llirottle Valve-Closure 1 5% closed 5 7% closed
'2'1 2> 2. luitiine covernor valve-East Closure > 1250 psig > 1000 psig E .
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- 20. Facility Comment on Question 3.05
- c. When level error matches steam-feed flow error part of answer was not asked, nor should be required for full credit.
Examiner Resolution The comment is noted, however, this part of answer required to complete the description as to why level stops decreasing.
- 21. Facility Comment on Question 3.08 Also accept word description instead of valve numbers. .
V-123 = under piston or insert drive wtr valve V-121 = over piston or insert exhaust wtr valve V-122 = over piston or withdraw drive wtr valve V-120 = under piston or withdraw exhaust wtr valve .
Examiner Resolution Will accept either word description or valve numbers.
g- 22. Facility Comment on Question 3.10 I(, b. We don't have a "t:1umbwheel mode selector switch" for bypassing LPRM's. We have a small toggle switch inside the associated APRM cabinet. Also, for answer number 3 in this part you should accept
" bypass lite indication of the full core display (P603)". This is the correct terminology.)
Examiner Resolution Comment on bypass switch noted. Comment on bypass light indication rejected. The light indication on P603 is the four rod display (same as answer 2).
- 23. Facility Comment on Question 4.03
- b. Correct answer = None RCIC auto shif t on low CST level (Ref. Volume III, Tab 3 P 19) however, this is a misleadino question.
Examiner Resolution Agree with facility comment. The examiner was in error. Part b of question was deleted and section 4 point value reduced to 25.0.
. 24. Facility Connent on Question 4.04 f
\~- Should accept any 5 of the 11 steps of this PPM. These 11 steps are generally treated as immediate actions - however none of these steps are defined as Immediate Actions since this is not an abnormal procedure.
3 6
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Examiner Resolution 1(, Agree with facility comment and 5 of the 11 steps required for full credit.
- 25. Facility Comment on Question 4.05 Should give full credit even if... " including the public.." is not included, since injury to personnel is all 1.clusive - includes any persons public or employees.
Examiner Resolution Agree with comment. The words " including the public" not required for full credit.
- 26. Facility Comment on Question 4.06
- a. Also accept (P&RTS bases, 3/4 4.1), " ensures adequate core flow coastdown following a LOCA."
- b. Any answer that refers to positive reactivity addition should be accepted since no reason is given in the precaution section.
f' Examine Resolution P
Agree with facility comments.
- a. Full credit will be given for " ensures adequate core flow coast-down following a LOCA".
s b. Comment Accepted. __
~
- 27. Facility Comment on Question 4.07 This answer requires memorization of a normal operating procedure - which is not required for ES202 page B.4. Also, if the " Water Leg Pump Discharge Press. Low" alarm is lit - the HPCS pump should not be started.
Ref.: PPM 2.4.4 Pre req. F and Limit. C (rev. 2)
Examiner Resolution ES 202 part B.4 stated "The candidate is not expected to have normal procedures committed to memory but should be able to explain reasons, cautions, and limitations of normal operating procedures." The question refers to a limitation on operation of the HPCS.
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WASHINGTON PUBUC POWER SUPPLY SYSTEM PLANT PROCEDURES MANUAL WNP. 2 PROCEOu AE puusEA APF4oVE
' QATE
' +4.601.Al-6.7 g -
8/22/84 VOLum4MAME '
f 4 ABNORMAL C SECTIC M ITION PROCEDURES 4.601.Al ANNUNCIATOR RESPONSE. P601 ANNUNCIATOR Al TIT b5
- 4.601.Al-6.7 HPCS WATER LEG PUMP OISCHARGE PRESSURE LOW l i 1 l 6-7 WINDOW l l l SOURCE l AUTCMATIC ACTIONS l l. I I
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s ' HPCS l HPCS-PIS-13 ( 6 53 PSIG) l NCNE I
, I WATER LEG l l l l PUMP OISCH I l l l PRESS LOW l l l I I I I l i 1 1
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kh 1. Verify HPCS Water Leg Pmp (HPCS-P-3) discharge pressure less than or equal to 53 PSIG as read on HPCS-PI-13(P601).
- 2. Verify HPCS Water Leg Pmp (HPCS-P-3) running (P601).
'- NOTE: The following step is designed to prevent pump run-out and possi-ble loss if LOCA shculd occur when HPCS-P-1 is running in tre test mode.
- 3. If tre Water Leg Pmp fails wten HPCS is required to be operable, start and run HPCS-P-1 in a test mode while maintaining approximately 1,200 psig disenarge pressure; maintain these conditions until the Water Leg Pmp is made operational.
4 If during a power interruption, the HPCS Water Leg Pmp Discharge Pres-sure Lcw ala:m is received and adecuate core cooling is assured, told HPCS-P-1 centrol switch in the STCP position or otherwise prevent pmc start until tre system can be filled and vented.
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SAFETY RELATED CONTROLLED COPY C. Vencers Manual
- GEX-71334
O. FSAR
, s. Secticns 6.3, 7.3, 6.3 2.4.4.3 P erecuisites A.
The Reactor Building Heating, Ventilaticn and Air Ccnditiening System in operatim to succort HPCS System Cperation. '
B. .'
IPCS service water system available to support HPCS diesel and WCS system ccerations. .
C. The cencensate stcrage tanks have the repuired amcunt of water to l
{ .t sucport HPCS cperaticn (7'7" each tank cr 13'3" in ene tank minimum per Technical Scecificatien 3.5.3). -
O. Must have at least minimun fuel supply (30,CCO gallens) en site for FPCS diesel.
E. The suppressicn pcol level normal (31 ft. 2 in. to 30 ft.10 in.).
F". The HFCS pu:c shculd not be started when the "HPCS WATER LEG PU4P
/
DISm RESS LCW" alarm is lit. The water leg pwp is cesigned to
/ remain in service thrcughcut system cceratim and utring stancby
/
status. Refer to Technical Specificaticn 3.5.1. 1
/
2.4.4.4 Limitaticns A. Cbserve RWP require:ents per PPM 11.2.8.1. .
~..-
- 8. The HPCS System snall not be remcVed frcm service anytime it is required to be cperable by Technical Scecifications. See Tectnical hg' Specificaticn 3.5.1.
a
< C. If during or follcwing a pcwer interrection the HPCS WATER LEG FWP f
/
/ OISm RESS LCW alarm is received and acecuate core cooling is assured, hold the p sto centrol switch in the STOP or otherwise t
i y
1 f ;
prevent puro start ntil.the system ~- '
can be filled and vented.
^
D. The FPCS System sha'l be maintained full anytime the system is re-
/[, .7 quired to te ccerable. If water leg amp FFCS-P-3 fails, start and
/
vt s ccerate -FCS o-1 cn recirculation to the CST until corrective action C _. dS C7012
- E. The concensate stcrage sna succly system may cnly be used to flush the EPCS Zystei. It snall never te used to keep the system charged or li.nec to to an trattencec system.
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- MOCEQuME NUMgER e4 h # 114C N N UM S E R
- AGE NYMMEH 2.4.4 2
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2.4.4-2 of 17
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Since the question did not state the cause of the water leg pump failure,
( if the candidate assumes (and states) the failure is due to power loss, half credit will be given for answering; the HPCS pump switch should be held in STOP or otherwise prevent pump start until the system can be filled and vented.
- 28. Facility Connent on Question 4.08 This response based upon PPM 4.2.1.2 Rev. 2 should be:
A Notify CRS B Verify Auto Actions C Take Manual Control of FWLC and reduce RPV level This is the current PPM in use and given to the licensee candidates :
prior to the exam.
Examiner Resolution Comment response will be accepted for full credit and the answer key :
changed, per reference stated by facility.
- 29. Facility Comment on Question 4.10 4.10 a. No reason is given in PPM 2.4.4 (HPCS) for this limitation -
Also our Supp. Pool Chem, results show that it is well within the Chem. Specs. for the RPV.
- b. Again - no reason given, however, to prevent overheating of motor windings should be given full credit.
Examiner Resolution A .
Per ES 202 candidates are expected to be able to explain reasons for limitations in normal operating procedures, whether they are stated in the procedure or not. Facility offers no alternative answers, and therefore the answer key was not changed.
- 30. Facility Comment on Question 4.12 The operators are not required to memorize normal operating procedures.
(ES 202 B.9.) These checklists are referred to during each turnover by the R0 and memory does not nor should not be relied upon to complete these checklists.
Also this is not required per our Volume 1 PM Learning Ob.fectives.
Invalid question. Also the point value is excessive on this question -
3.5 pts 14%. If an operator didn't memorize the turnover checklist he is down to an 86% on Section 4.
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Examiner Resolution Comments noted. ES 202 part B.4. states that administrative procedures may be included to the extent they are directly applicable to an operator.
The R0 initials the Shift Turnover Checklist each time he/she takes the watch, so it is reasonable to expect the candidate to know what is on tr.a list. Point value is 20% of section, thus meeting requirements o' ES-107.
Will accept other implicit items included as part of the items listed.
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WNP-2-FACILITY-COMMENTS-5.03 Point total for the answer does not add up to the point value for the question.
5.04 Question did not ask for a discussion of boiling boundary only voids, and core net reactivity. Discussion of boiling boundary should not be required for full credit.
5.06 Tech Specs and our procedure (7.4.1.1.2) define shutdown margin as the amount sub critical with a " cold" " clean" core with highest reactivity rod withdrawn. If the candidate assumes that the stated shutdown margin is cold and clean then he should be given credit for answering that the SOM is acceptable and that there would be no consequences over -
the subsequent 20 hours2.314815e-4 days <br />0.00556 hours <br />3.306878e-5 weeks <br />7.61e-6 months <br />. (Note: per PPM 7.4.1.1.2 we do not measure SDM until shutdown for 50 hours5.787037e-4 days <br />0.0139 hours <br />8.267196e-5 weeks <br />1.9025e-5 months <br /> and it is corrected for temperature).
5.07 Unable to find answer to part "b" in the stated reference. The statement "the increase frictional resistance lowers the total flow to less than twice the original flow" should not be required for full credit.
Part B - delete [0.5] following "less than double the original flow",
otherwise point value is 2.0 for total question.
5.08 Also accept per the abnormal procedure 4.4.4.2 (see attached copies) reactor power and pressure perturbation, reactor vessel level perturbation or the following explanation:
Turbine load - may initially decrease due to spraying cold water on steam exiting the core. If reactor power then increased (without causing a scram due to APRMs) because of colder water being returned to
,6- the down comer and flowing into the bottom of the core pressure-will increase and turbine load will -increase. -
Reactor water level indication - may initially decrease due HPCS spray causing a pressure drop (steam condensing, void collapse) and water from the downcomer will flow into the core because of less back pressure. As HPCS continues to inject this effect is overcome by the amount of water injected and reactor water then increases until it is compensate by the FKLC system.
Feedwater flow - may initially increase a small amount due to the indicate water level decrease but as HPCS continues to inject and reactor water level increases the FWLC will then act to lower the feed rate from the feedwater pumps.
As of date, we do not have data for this transcient so it is difficult to predict actual plant response.
In Section 6.0 question #5 is the same as 5.8b, this double ,ieopardy is not allowed by the examiners standard, b
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5.11 Unable to find answer for shape of curves drawn on Figure 1 in stated reference for the answer key. Also, this curve is not a standard curve used by the operating unit. And the discuss answer given in the key requires that students have the bases for Tech Spec memorized verbatim which is unrealistic.
Give full credit for either drawing the curves or. discussion of effects.
5.12 Section "b" also accept for full credit: " Increasing the flow rate -
increased the heat removal capability and places the bundle farther from OTB."
6.02 No comment on answer, however - question stipulates a mode (flex auto) of RFC that we do not use at present.
6.03 a. should give full credit for 1) loading in U234 " breeder" material
- 2) reduction of the " sputtering" effect also this was not required knowledge item as part of the LPRM learning objectives
- b. no comment 6.04 "The drive piston"... should not be required for full credit. A similar response such as the " control rod" moves past the "00" position should deserve full credit as well.
6.05 Part ii asks for FWLC response with a HPCS initial at 90% power - this item is not covered in our " Systems training mat'l and the plant response would be dependant upon how fast the FWLC system & RFPs could b-respond to the increase in level part d could also be acceptable (RPV level until turbir.e trip). Also this question is also asked in -
question 5.8c (FW response due to HPCS initial) Couble Jeopardy 6.06 Answered Part b the flow elements are located in the pump suction rather than the discharge ref systems manual Vol.1 Tab 6 Fig. 2.
6.07 Answer Part d should also accept RHR pump 2b will not start
- due to suction valve interclock - no suction path, no injectinn because the pump did not start.
- won't start due to bkr cycle close then open 6.08 a. should also accept - provide a more stable flux signal to minimize FCV ball valve wear due to hunting 6.09 a. also accept - increase to 100% recire flow general comment - again- RFC not used in Loop Auto
. ~
6.10 " Anticipatory" scrams is very vague all scrams are due " anticipating" further plant problems degraded conditions. We do not classify scrams under this type of category. This cuestions assumes all other scrams de not anticipate other problems. Should accept any scram signal -
with justification.
Question Assumes only 3, but gives 4 as answer 6.11 General comment question #5: 6.02 (2 pts) 6.08 (3 pts) and 6.09 (3 pts) all referred to the Recirc Flow Control System and its components or operation in auto modes we currently do not use. Also, the total point value of these questions 8 pts accurate for 32% of this section of the exam on the topic of the Recirc Flow Control system which is not in .
agreement with ES-107 C.5 "no topic is worth more than 20% of that category.
7.01 A. The answer does not match the question. The question asks "What do Tech Spec's say?" Per Tech Spec, the required action is "to reduce suppression pool temp to 90 F within 24 hours2.777778e-4 days <br />0.00667 hours <br />3.968254e-5 weeks <br />9.132e-6 months <br />," (3.6.2.1, action 6). In addition, this question requires memorization of a 24 hour2.777778e-4 days <br />0.00667 hours <br />3.968254e-5 weeks <br />9.132e-6 months <br /> Tech Spec Action Statement.
B. It requires memorization of a lor.g term (24 hcur) action statEmJnt.
C. Same comment as "B" above.
.*- 7.04 General Comment Just because both loops of S/D cooling are IN0P., per 5.3.5. Tech Spec 3.4.9.1 or 314.9.2 does not require that alt. S/D -
Cooling be performed.
- c. Per PPM 5.3.5, Alternate Shutdown Cooling, Alt. Shutdown cooling can be accomplished by any low pressure ECCS pump, thus LPCS, RHR "C", or either RHR "B" or "C" may be used to inject into the Reactor Vessel. Should accept any of these as correct or should not " count-off" for not listing PHR A or B as injecting into RPV.
- d. PPM 5.3.5 is used in accident conditions where cooldown is required and normal S/D cooling cannot be accomplished. It is not reasonable to expect an operator to " memorize" this procedure or the numbers in it. This RPV pressure response is only a " rough estimate" to provide indirect indication of core flow. Step 6.3 is used to supercede this to ensure a cooldown rate of less than 100 F/hr. Should accept "less than 120psig (to allow low press.
ECCS pumps to inject)."
7.06 a. This question requires memorizatico of a limitation not an
( immediate action on a procedure which is not an abnormal or an L emercency procedure.
6 lf( '
7.08 The question states that a reactor shutdown was performed. This would make part 1 of the answer "not applicable." Should accept " items performed during the outage were easily tracked and controlled by shift staff." (PPM 3.1.4, discussion A, paragraph 1).
7.09 Answer key does not give all possible answers per PPM 4.4.2.1. Should also accept the other answers shown included in Section 4.4. 2.1.4.C (RWCU system heat exchangers or CRD & reject vra RWCU). Note: the answer key does not refer to the latest revision of this. PPM. we have ,
attached a copy of the latest revision. '-
7.10 B.
- 1. You cannot make the assumption that because a control rod is "untrippable" it is immovable. Thus, part B does not " jive" with part A. ,
- 2. Should also accept step 7 of PPM 5.1.3 as a full credit answer."
If the reactor cannot be shutdown before suppression pool temperature reaches 110 F."
8.01 This question requires memorization of a normal operating procedure, does not agree with our learning objectives.
8.03 A. Should accept any three of the six listed on PPM 1.11.3, Health Physics Program under (radiological) " conditions that require an RWP" (page 9 of 25) 4-8.04 This answer requires memorization of an administrative procedure. This also not included in our learning objectives. -
8.05 b. should accept either safety related or 8.06 This question is very vague! It assumes that the shift manager either does not or cannot (for whatever reason) remove the clearance order, then perfcrm his test. If you want the SR0 to know the restrictions placed on " Temporarily lifting danger tags," then ask it that way! The PPM 1.3.8 states the S. M. authorizes the temporary lifting at taos, not system checkout.
8.08 Answer #7 contains 3 distinct guidelines to be accomplished - should consider these separately.
8.09 PPM 1.3.1, Standing orders has been updated to be consistent with our Emergency Operating Procedures. Please see latest rev. (attached) of PPM 1.3.1 for correct answer.
8.11 This question is not valid because the justification for use or non-use has not been documented or clarified in normally " testable" information. In addition, the lates revision of PPM 1.3.1, no longer includes this instrument on the " unqualified Instrument List."
P.ecommend deletion from exam. (see attached PPM 1.3.1).
'I l 4 ,
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I' d. U. S. NUCLEAR REGULATORY COMMISSION REACTOR OPERATOR LICENSE EXAMINATION i
FACILITY: WNP-2 S REACTOR TYPE: BWR-GES DATE ADMINISTERED: 86/02/04 EXAMINER: SHERMAN. J.
{
APPLICANT:
l '
1 INSTRUCTIONS TO APPLICANT:
{
i U o separate paper for the answers. Write answers on one side only.
7 Stcple question sheet on top of the answer sheets. Points for each l qu stion are indicated in parentheses after the question. The passing Crzde requires at least 70% in each category and a final grade of at F'
least 80%. Examination papers will be picked up six (6) hours after g the examination starts.
b % OF
, CATEGORY % OF APPLICANT'S CATEGORY r VALUE TOTAL SCORE VALUE CATEGORY
[ J.2. 0 o LE7N
- 2 '_ . ?? E0.04 1. PRINCIPLES OF NUCLEAR POWER PLANT OPERATION, THERMODYNAMICS,
.,o HEAT TRANSFER AND FLUID FLOW gq S C $2.++
, t's-25 2"J-te 2. PLANT DESIGN INCLUDING SAFETY AND EMERGENCY SYSTEMS 1 S.I' *l
- 25.00 E4,94 3. INSTRUMENTS AND CONTROLS Y bai. C 0 g r. Y\
2" "J 36-9+ 4. PROCEDURES - NORMAL, ABNORMAL, EMERGENCY AND RADIOLOGICAL
'E CONTROL
[ 9 4 . ') $'
100.25 100.00 TOTALS r
k, 1 FINAL GRADE %
,( All work done on this examination is my own. I have neither lf Civsn nor received aid.
I j{ APPLICANT'S SIGNATURE It i!
i 6
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'V. .
- 1. PRINCIPrE OF NUCrRAR POWER PLANT OPERATION. PAGE 2 TWRMODYNAMICS. MAT TRANSFER AND FLUID YLQW
. s I;-
QUESTION 1.01 (1.50) -~
'), ; $ +- -
j- INCREASING recirculation flow causes movement of the boiling t boundary and a power increase. Name the DOMINANT reactivity K coefficient (s) which are acting during the power increase when:
- a. The boiling boundary moves upward (0.5)
- b. The boiling boundary moves downward (1.0) s o[
[
QUESTION 1.02 (1.50) f During a reactor startup, Keff is .95 when the SRM channels p read 100 cps. What will the new Keff be when SRM channels trad 250 cps?
STATE ANY ASSUMPTIONS YOU MAKE.
- k. SHOW ALL YOUR WORK.
F E _
QUESTION 1.03 (2.00)
Saveral minutes after a reactor scram you notice indicated power dccreasing on an 80 second period. What causes this behavior and why is the period 80 seconds?
)(
L
- QUESTION 1.04 (2.00) p t State whether the following changes will directly affect L AVAILABLE recirculation pump net positive suction head 6 (NPSH): (Limit answer to: INCREASE, DECREASE, or NO AFFECT)
I
- a. Feedwater temperature increases (0.5)
{ b. Reactor pressure decreases (0.5)
[ c. Reactor water level increases (0.5)
L j d. Recirculation pump speed decreases (0.5) i I
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(***** CATEGORY 01 CONTINUED ON NEXT PAGE *****)
i 4
< 1. PRINCIPrRR OF NUCrRAR POWER PLANT OPERATION. PAGE 3 TmL1tMODYNAMICS . mtAT '"RANSFER AND FLUID FLOW
' s
! QUESTION 1.05 (2.00) ,
- Flow orifices serve to provide the majority,of pressure drop across the core. WHY is flow orificing necessary in a BWR?
- QUESTION 1.06 (2.00) p When called upon for an emergency shutdown, the Standby Liquid Control System must inject at a rate neither too slow or too I' fast. DESCRIBE why this is the case.
r r
g QUESTION 1.07 (1.50) h The fuel temperature (Doppler) coefficient of reactivity changes over core life, y
i a. Does it become HORE NEGATIVE OR LESS NEGATIVE 7 (0.5)
I b. DESCRIPE the primary effect that causes it to change over core life. (1.0) a r
QUESTION 1.98 (1.00) c WHAT percentage of full power is produced by decay heat at the following times after a reactor shutdown from 100 % power?
G (Assume equilibrium reached prior to shutdown.)
r p a. One second after shutdown.
" b. One hour after shutdown.
l QUESTION 1.09 (1.50)
During a reactor startup, the IRM readings go from 30 to 65 on the L. came range in 2 minutes with no operator actions. What PERIOD is the reactor on? SHOW YOUR WORK.
h.
.I c (***** CATEGORY 01 CONTINUED ON NEXT PAGE *****)
1:
I- - .- . __ __. _ .-- - . . - . . - - --. _. - _. . .- -.
- 1. PRINCIP m OF NUCLEAR POWER PLANT OPERATION. PAGE 4 f[. THERMODYNAMICS. HEAT TRANSFER AND FLUID FLOW i
- j. -
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QUESTION 1.10 (2.00)
A reactor vessel is depressurized from 1000 psig to 600 psig in a period of one hour. WHAT is the cooldown rate? Assume saturated b conditions. SHOW YOUR WORK.
(
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- QUESTION 1.11 (3.00) b-Following a normal reduction in power from 90% to 70% with fi racirculation flow, HOW will the following change (increase, dscrease, or remain the same) AND WHY:
- o. The pressure difference between the reactor and the turbine
- steam chest.
3 L b. Condensate dep'ression at the exit of the condenser.
E c. Final Feedwater temperature.
e QUESTION 1.12 (2.00)
C o. WHAT is " pump runout" and WHY is it an undesirable condition? (1.0)
- b. Define the term cavitation, and GIVE TWO (2) examples of
~~
detrimental effects. (1.0) k.
?'
l QUESTION 1.13 (2.00)
L What are the two considerations that determine the maximum and cnd minimum control rod speed limit? (Consider normal movement, not scram)
L i m k
x a
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(***** END OF CATEGORY 01 *****)
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- 2. PLANT DESIGN INCLUDING SAFETY AND EMERGENCY SYSTEMS PAGE 5 7
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4.
QUESTION 2.01 (2.00) i L STATE TWO purposes for the RHR water leg pumps. ,
l I
f QUESTION 2.02 (1.50) k An "IRM Inoperative" annunciator is received. What THREE conditions
[ could be the cause of this alarm?
E h QUESTION 2.03 (2.00)
Regarding the APRM system:
- b. STATE the meter switch position you would use to determine
? the number of LPRMS in service on an APRM channel. Include required interpretation of the meter reading. (1.5)
L L
f
,l' QUESTION 2.04 (2.50) p.
NAME five of the six physical barriers which will limit the release cf fission products from the fuel to the environment.
,\-i QUESTION 2.05 (2.00)
Ccncerning the reactor building-to-wetwell relief lines:
t I c. WHAT purpose do they serve? (1.0) i b. Under WHAT condition will they relieve (setpoint required)? (1.0)
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(***** CATEGORY 02 CONTINUED ON NEXT PAGE *****)
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.2. PLANT DESIGN INCLUDING SAFETY AND EMERGENCY SYSTEMS PAGE 6
[- .
E L QUESTION 2.06 (1.00) h b " Souping", a light load phenomenon in a diesel generator, occurs when:
J (CHOOSE ONE)
- c. Excessive oil enters the combustion chamber and accumulates in the p exhaust system.
J b. Electrical loads are added too quickly immediately after the b engine reaches rated speed.
[ c. Diesel fuel excessively contaminates the lubricating oil r system
- d. The water jacket cooling system becomes fouled with biological contaminants, b
QUESTION 2.07 (1.50)
E Regarding the High Pressure Core Spray System:
L b o. List the two auto initiation signals (setpoints not required). (1.0)
- b. STATE the system which is redundant to HPCS for small break L
protection. (0.5) e 1
f QUESTION 2.08 (2.50)
I LIST five of the six conditions that will cause automatic isolation
. of-the RCIC system (setpoints not required). (2.5)
F i
b QUESTION 2.09 (1.00)
Are the statements below TRUE or FALSE concerning the High Pressure Core Spray Diesel Generator System?
4 jf 1. The bus SM-4 is powered by the diesel generator due to loss of preferred power. After the preferred power source is made available, the operable preferred power source breaker is closed. The HPCS-DG output breaker
' I automatically OPENS. (0.5) f 2. There are two NORMAL engine stop pushbuttons: 1. in the Control
!j Room. 2. on the local panel. BOTH pushbuttons are lt
,1 ALWAYS active. (0,5)
(***** CATEGORY 02 CONTINUED ON NEXT PAGE *****)
I 2. PLANT DESIGN INCLUDING SAFETY AND EMERGENCY SYSTEMS PAGE 7 I .
c s
P t QUESTION 2.10 (3.00)
E
, c. Regarding the Plant Electrical System, the standby diesel i generators start automatically upon receiving what g
signals? (three required) (1.5)
When a LOCA signal is present to the Standby Diesel Generator fe b.
System, what three auto shutdown signals or functions are NOT I; bypassed? (1.5) 1 b
E-E QUESTION 2.11 (2.00) e WHAT will happen to the Reactor Water Cleanup System if Reactor Closed Cooling Water is lost to all components RCCW serves in the p RWCU system?
- Ycur answer should include the affected components, and any trips er interlocks associated with this condition. (Setpoints not required)
L QUESTION 2.12 (2.25)
E Regarding the recirculation pumps, WHAT are THREE conditions that will cause a fast speed trip AND an LFMG auto start?
p (Setpoints ARE Required.)
h QUESTION 2.13 (2.00) r A2sume the RCIC system receives an initiation signal, all system 4
components function properly, except the items listed below.
Each failure is present prior to the initiation signal being rcceived.
Describe the RCIC systems response for each of the following
, end justify your answer. Consider each item seperately.
t
- c. The turbine exhaust valve (RCIC-V-68) is stuck shut.
- b. The D/P cell, for the RCIC flow control element, has a perforated diaphram.
I i
t i
(***** END OF CATEGORY 02 *****)
t
- 3. INSTRUMENTS AND CONTROLS PAGE 8 1.
( '
L- QUESTION 3.01 (2.50)
E STATE five signals which will automatically close the main J cteam isolation valves (MSIV's). (Setpoints not required.)
i E
i j QUESTION 3.02 (1.00)
L ALL 125 VDC power has been lost.
I Which of the following is correct:
I' a. SRV's will be operable in the pressure relief, safety relief, and AUS modes.
[ b. SRV's will be operable in the pressure relief -and 4DO-modeaM.
g d. SRV's will be operable in the ADS mode 44MAF.
i f QUESTION 3.03 (2.00)
- E ch condensate filter demineralizer (demin) has a hold pump I
casociated with it. Regarding hold pumps:
I c. WHAT is their purpose? (0.5)
P b. STATE the conditions under which the pump will start or stop
- if in the auto mode. (1.0)
- c. WHAT problem may exist with the demin if the hold pump failed to start in auto when it is required to? (0.5) i j CUESTION 3.04 (2.00)
R:garding the setpoint setdown feature of the feedwater level l control system:
l '
- c. WHAT TWO changes take place automatically in the control 7
l! circuitry following a reactor scram?
lI b. The feedwater level control system will remain in the setpoint i, setdown condition following a scram until one of two actions l; is performed by the operator. STATE these TWO actions.
ll 1f
[ (***** CATEGORY 03 CONTINUED ON NEXT PAGE *****)
{
i
, . ~ . . _ _ - - _ ._ __ _
i:
- 3. INSTRUMENTS AND CONTROLS PAGE 9 4
14 g
I
[i QUESTION 3.05 (3.00)
V I- DESCRIBE the changes in the following parameters in the FIRST F FEW MINUTES following the low failure of one steam flow transmitter.
I (Assume 100% reactor power, three element FWLC, and no operator
! cetion) k
[ c. Reactor power t- b. Feedwater flowrate
{ c. Reactor water level g
[
- f. QUESTION 3.06 (2.00)
$ Fcr each of the following state whether a ROD BLOCK, HALF-SCRAM, p FULL-SCRAM, or NO PROTECTIVE ACTION is generated for that k
c ndition.
I NOTE: If two or more actions are generated, i.e. rod block and a I
half-scram, state the most severe, i.e. half-scram.
I' Assume NO operator actions.
g c. APRM downscale, Mode Switch in Run.
!' c. Both Flow Conversion Units Upscale (>108% flow),
Mode Switch in Run.
I d. APRM A and D >20%, Mode Switch in STARTUP. (4 0 0.5 ea.)
F K '
?
If
[- QUESTION 3.07 (2.00)
S For each of the IRM (Intermediate Range Monitoring) range changes below, provide the following (Mode Switch in STARTUP):
- 1. The indicated level on the NEW RANGE.
- 2. All automatic actions initiated as a result of the indicated L level on the NEW RANGE.
L
- c. Switch from Range 5, Reading 25, to Range 7.
f 5 b. Switch from Range 6, Reading 39, to Range 5.
h,
, QUESTION 3.08 (2.50) .
s LIST the sequence of valve movement that occurs when a single notch rod withdrawal is demanded from the Reactor Manual Control System.
}
e (Specific times are not required; correct sequence IS required; I FIVE events for full credit.)
h
(***** CATEGORY 03 CONTINUED ON NEXT PAGE *****)
u ..-..-..a----. . . . . . . ..
I c.- -
i 3. INSTRUMENTS AND CONTROLS PAGE 10 i
c.
QUESTION 3.09 (2,50) l Both the SRM and IRM compensate their detector signals with a unique
[ type of discrimination process.
Il (0.5)
- c. WHAT type of radiation does the discriminator eliminate?
(
} accomplish this task. (1.0) h c. WHY is there a difference between the two discrimination g processes? (1.0)
[
QUESTION 3.10 (2.50)
Answer the following questions concerning the Local Power Range Monitors.
[
b c. What is used in the detector to extend the neutronic lifetime? (0.5) h b. Where are the three (3) indications that the thumbwheel mode selector switch is in the bypass position? (1.0) t-
- c. Where is the signal of the flux amplifier fed, when the mode (1.0) selector switch is in operate? (Four required for full credit.)
h QUESTION 3.11 (3.00) b C:ncerning the Area Radiation Monitors (ARMS):
c 1. What PURPOSE do these instruments serve? (1.0)
- 2. What is INDICATED by the following lights on the front face r of the indicator and trip unit-d_ a. WHITE light [0.5]
. b. AMBER light [0.5] (1.0)
' I:
L 3. Assume several ARM's in the reactor building area are alarming.
l HOW would the control room operator determine the specific locations of the alarming ARM's? (1.0) jL l
1 i
l'
(***** END OF CATEGORY 03 *****)
l
'ta
h
- 4. PROCIDURES - NORMAL. ABNORMAL. EMERGENCY AND PAGE 11 k RADIOLOGICAL CONTROL QUESTION 4.01 (1.00) 7;
- According to Emergency Procedure 5.1.3 (Reactor Power Control), two cystems OTHER THAN Standby Liquid Control could be used"to inject boron into the reactor vessel. Name the two systems.
i P
QUESTION 4.02 (2.00)
WHAT are the entry conditions for Emergency Procedure 5.4.1 --
p' Station Blackout? (FOUR itemr) p c:
fQUESTION4.03 'J . 00 ) " %.oo)
I Emergency Procedure General Precautions, Caution #7, requires Cpecific operator action if "HPCS Suction Switchover Suppression Pool Level High or HPCS/RCIC Suction Switchover CST Level t
Low alarms occur".
- c. WRAT operator action is required?
O t. ". . 'c s u i. L o s,wltchcd eenmell, in e OCT ivw level t V
F k QUESTION 4.04 (2.50) t r According to General Operating Procedure 3.3.1 (Reactor Scram),
WHAT are the first FIVE steps required to be performed? (Each STEP
- 2;y contain more than one item.)
f QUESTION 4.05 (1.00) i According to the Standing Operating Orders (Procedure 1.3.1),
under what conditions may operating personnel depart from opproved operating procedures?
I I
r
(
(***** CATEGORY 04 CONTINUED ON NEXT PAGE *****)
- 4. PROCEDURES - NORMAL. ABNORMAL. EMERGENCY AND PAGE 12 F RADIOLOGICAL CONTROL h
7 w -
r QUESTION 4.06 (3.00)
.c
! Procedure 2.2.1 (Reactor Recirculation System), contains i o number of restrictions for operation of the recirculation
$ cystem. STATE the reason (s) for the following limitations:
4
[ c. For two loop operation, maintain loop-to-loop flow mismatch less than 10% when less than 70% rated core flow and
- i. maintain loop-to-loop flow mismatch less than 5% when greater i than 70% rated core flow.
L E b. Do not start an idle reactor recirculation pump during the
[ approach to criticality or when critical below the power range.
[.
E c. Following a recirculation pump trip, close the discharge
{ valve. After 5 minutes (maximum), open the discharge valve.
y
[ QUESTION 4.07 (1.00) b WHAT action must you take if the water leg pump for High Pressure
[ Core Spray fails when HPCS is required to be operable, according to p Procedure 2.4.4 (High Pressure Core Spray System)?
b:
i V
r QUESTION 4.08 (3.00) i LIST THREE IMMEDIATE ACTIONS you would take upon a reactor HIGH L water level condition according to Abnormal Condition Procedure t
4.2.1.2 (Reactor Vessel High Water Level).
r L
QUESTION 4.09 (2.00)
DESCRIBE TWO problems, other than CRD accumulator problems,
[ which will occur upon a complete loss of
{
CRD flow, according to Abnormal Conditions Procedure
- f, 4.1.1.2 (Complete Loss of CRD Drive Flow).
I n .
i
( (***** CATEGORY 04 CONTINUED ON NEXT PAGE *****)
I l
l
- 4. PROCEDURES - NORMAL. ABNORMAL. EMERGENCY AND PAGE 13
[ RADIOLOGICAL CONTROL
[
[. -
E E
QUESTION 4.10 (2.00)
- C STATE the reason (s) for each of the following limitations in the b High Pressure Core Spray System Operating Precedure. ;
- c. Whenever the HPCS system is discharging into the vessel under
}- any conditions other than an actual loss of coolant accident,
- make sure that pump suction is being taken from the condensate e storage tank.
h
- b. Pump Start Limitation - Two starts in succession from ambient 1 temperature or one start from rated temperature.
b
/ QUESTION 4.11 (3.00) b What are SIX of the NINE immediate operator actions you r; would perform if a control room evacuation were ordered by the Shift Manager? (Assuming you have time.). (3.0)
{
h l- QUESTION 4.12 (3.50)
The Control Room Operator Shift Turnover Checklist requires the i cncoming CRO to take certain actions.
I c. What are FOUR items to be reviewed DURING turnover? (2.0)
- b. What are THREE items to be reviewed SHORTLY AFTER v
r shift turnover? (1.5) i t
L
[
t t
i t
I
(***** END OF CATEGORY 04 *****)
(************* END OF EXAMINATION ***************)
b
- 1. PRINCIPLES OF NUCLEAR POWER PLANT OPERATION. PAGE 14 I THERMODYNAMICS. HEAT TRANSFER AND FLUID FLOW E
ANSWERS -- WNP-2 -86/02/04-SHERMAN, J. ]
f ANSWER 1.01 (1.50) *f ~
il (0.5)
F c. Void coefficient [0.5]
f b. Void coefficient [0.5], fuel temperature (Doppler)
) coefficient [0.5] (1.0)
$ REFERENCE l WNP-2 Reactor Theory Student Text, Pg. 21 r
ANSWER 1.02 (1.50)
CR1(1-Keff1) = CR2(1-Keff2) [1.0)
F CR1/CR2(1-Keff1) = (1-Keff2) f l 100/250(1 .95) = (1-Keff2)
L .C2 = (1-Keff2)
- h. K ff2 = .98 [0.5]
, REFERENCE
' WNP-2 Reactor Theory Student Text, Pg. 8 ANSWER 1.03 (2.00)
Delayed neutron precursors rele ing delayed neutrons.[1.0] Longest liveddelayedneutronprecursorhashalflifeof56 seconds) l, produces period of -80 seconds. [1.0]
't
. REFERENCE
! WNP-2 Reactor Theory Student Text, Pg. 24 l
ij-
,5 ANSWER 1.04 (2.00) l o. DECREASES (0.5) i t, b. DECREASES (0.5) l[ c. INCREASES (0.5)
'I: d. POT AITECT INCREAf ES (0.5) i l) u
L b 1. PRINCIPLER OF NUCLEAR POWER PLANT OPERATION. PAGE 15
) TW RMODYNAMICS. HEAT TRANSFER AND FLUID FLOW
?
. ANSWERS -- WNP-2 -86/02/04-SHERMAN, J.
e REFERENCE f .
G.E. Thermodynamic, Heat Transfer, and Fluid Flow, Pg. 7-92 +
i
).
I ANSWER 1.05 (2.00) ;
? Higher bundle power causes increased voiding and therefore increased j' rssistance to coolant flow [1.0]. If no orificing, high power (central) bundles would be starved of cooling, while more coolant b flow would be diverted to lower power (peripheral) bundles [1.0]. (2.0)
I r
REFERENCE L WNP-2 Systems Vol. I, Tab I, Pg. 14 r
ANSWER 1.06 (2.00)
- 1. If injection is too fast, uneven mixing could occur and g
power chugging could result. (1.0)
I' 2. If injection is too slow, the initial shutdown margin would be reduced due to lower concentration of fission product poisons. (1.0)
REFERENCE WNP-2 Systems Vol. III, Tab 8, Pg. 8 ANSWER 1.07 (1.50)
- n. MORE NEGATIVE (0.5)
- b. Pu-240 builds up with exposure. (1.0)
,' (Pu-240 resonance absorption is stronger than V-238)
I REFERENCE
{ WNP-2 Reactor Theory Student Text, Pg. 19 ANSWER 1.08 (1.00) i (0.5) i n. 6.0% (Range 5.0 - 7.0%)
{ b. 1.6% (Range 4.t 1.0%) (0.5)
! o r -- z,o%)
i i
1 l
I
PAGE 16
' k' 1. PRINCIPMR OF NUC MAR POWER PLANT OPERATION.
TERMODYNAMICS. MAT TRANSFER AND FLUID FLOW
. ANSWERS -- WNP-2 -86/02/04-SHERMAN, J.
REFERENCE L
WNP-2 Reactor Theory Student Text, Pg. 9 .
L ANSWER 1.09 (1.50) i P = Po e t/T P/Po = e t/T
[
F In(P/Po) = t/T Y T = t/In(P/Po) (1.0)
= 120 sec./ln(65/30) e = 155 sec. (0.5)
REFERENCE WNP-2 Reactor Theory Student Text, Section 6, p. 14 ANSWER 1.10 (2.00) i From Steam Tables 1000 psig + 14.7 = 1014.7 psia [ dam *O @ WJd IN 600 psig + 14.7 = 614.7 psia (3; t,,J . 4- f 4,7 p) (0.5)
C Taatl = [(14.7/50)(550.53 - 544.58)] + 544.58
= 546.33 F (0.5)
Tcat2 = [(14.7/50)(494.89 - 486.20)] + 486.20
= 488.75 F (0.5)
[
' l Cooldown rate = 546.33 - 488.75 = 57.57 F (0,5)
( Auf t.e f kle ranje 57 l#F)
REFERENCE Steam Tables s 1
I.-.,_...._..._.-.. , _ e e _- --
E.
t . .
K' 1. PRINCIPLES OF NUCr EAR POWER PLANT OPERATION 1 PAGE 17
[
THFRMODYNAMICS. HEAT TRANSFER AND FLUID FLOW ANSWERS -- WNP-2 -86/02/04-SHERMAN, J.
s,
'f L 4 E ANSWER 1.11 (3.00) *
[ c. Decreases [0.25]. There is less steam flow, therefore, less (1.0)
{
w pressure drop through the main steam lines [0.75].
F b. Increases [0.25]. With the same amount of cooling water through t
i-the condenser [0.25] and less of a heat load [0.5]. (1.0)
I c. Decreases [0.25]. Less extraction steam from the turbine to f heat the feedwater [0.75]. (1.0) k e REFERENCE
[ EIH Heat Transfer Lesson Plan, pp. 75 & 78, and EIH Nuclear Training, p. 10.4-11.
p i WNP2 G.E. Thermodynamics, Heat Transfer, and Fluid Flow, pg 6-68 p
L
( ANSWER 1.12 (2.00)
[ c. Running a centrifugal pump at minimun ad and L maximum pacity 'O.5). Runout causes ctrical over heating ossible electrical damage, an ikely I tripping off line 'O.5). (1.0) esly ont repJs al he 4. ll credit)
- b. Cavitation is the flashing to vapor of liquid in the pump suction or a low pressure area.[0.5] Cavitation results in any or all of the following
- excess vibration and noise, reduced i Pump efficincy, pitting and corrosion of pump impeller.
F (not limited to these 5) (2 required at 0.25 each) (1.0)
REFERENCE WNP2 G.E. Thermodynamics, Heat Transfer and Fluid Flow, pg 7-124 & 7-91 i
l r
a l
r i
i
)
r f
18
- 1. PRINCIPLES OF NUCLEAR POWER PLANT OPERATION. PAGE THERMODYNAMICS. HEAT TRANSFER AND FLUID FLOW ANSWERS -- WNP-2 -86/02/04-SHERMAN, J.
Y .
ANSWER 1.13 (2.00) Q uerh m ad f** **$4 (2.0) delt ted 4
f 1. Maximum control rod speed is intended to limit the rate at which control rods can be withdrawn during reactor f
- startup. Fast reactivity insertion rate results in L short periods and could cause reactor core to rapidly 4
overheat and become damaged. (1.0) b 2. The rate of control rod speed must be sufficient to fi r~ overcome xenon reactivity decrease during burnout. (1.0)
.E REFERENCE r WNP2 Systems Vol II, Tab 8 r
F b.
p I-r L
i 1
I 1
l E
6 k,
t I
i t I
f
~
(-
F.
PLANT DESIGN INCLUDING SAFETY AND EMERGKNCY SYSTEMS PAGE 19 p 2.
ANSWERS -- WNP-2 -86/02/04-SHERMAN, J.
h g.. . -
f.
?
4 h ANSWER 2.01 (2.00) ,
- 1. Time required for water to reach reactor vessel upon
? system initiation is minimized. (1.0) 1 2. Possibility of water-hammer damage is minimized. (1.0)
?
REFERENCE I WNP-2 Systems Vol. III, Tab 6, Pg. 7 i
J ANSWER 2.02 (1.50)
- 1. Detector high voltage low (0.5)
- 2. Module unplugged (0.5)
- 3. IRM mode switch not in " operate" (0.5)
REFERENCE WNP-2 Systems Vol. II, Tab 2, Pg. 23
}.
i i ANSWER 2.03 M ( l.TO 0)
- a. 14 (0.5) ,
- b. 6 ount" switch positio d 0.5 7 Meter reading (in percent) divided by five yields number f of LPRM inputs. (1.0) s
~
REFERENCE 1
WNP-2 Systems, Vol. II, Tab 4, Pg. 12 r.
ANSWER 2.04 (2.50)
- 1. Fuel pellet itself
- 6. Secondary containment [5 required, 0.5 each] (2.5)
[
I e
t i l
'I
r.
[- 2. PLANT DESIGN INCLUDING SAFETY AND EMERGENCY SYSTEMS PAGE 20 ANSWERS -- WNP-2 -86/02/04-SHERMAN, J.
REFERENCE I WNP-2 Systems, Vol. III, Tab. 11, Pg. 4 f W.
l ANSWER 2.05 (2.00)
{ b1 1; c. Equalize pressure differential that may occur between secondary L containment and suppression chambe thus preventing a vacuum f from develo 3 (1.0) y steam.(o.ss) ping in the primary containment g,g3due to condensing L
F b. When pressure in secondary containment exceeds pressure in
}
suppression chamber [0.5] by 0.5 psig [0.5]. (1.0)
L I
REFERENCE I WNP-2 Systems, Vol. II, Tab 11, Pg. 11 k
f ANSWER 2.06 (1.00) a (1.0) fL*
REFERENCE WNP-2 Systems, Vol. IV, Tab 19, Pg. 70 E
ANSWER 2.07 (1.50)
- c. Low reactor water level (0.5)
High drywell pressure *
(0.5)
-f, (0,5)
- b. ADS o v- K cics (oM regArc/ A ll creJit) i REFERENCE f WNP-2 Systems, Vol. III, Tab 1, Pg. 5 F
I i
r e
f T -
I
i F- 2. PLANT DESIGN INCLUDING SAFETY AND EMERGENCY SYSTEMS PAGE 21
( .
E ANSWERS -- WNP-2 -86/02/04-SHERMAN, J.
c .
u I
t
- ANSWER 2.08 (2.50)
- 1. RCIC equipment area :nd/er ipe routing area high
! temperature (cred si 4.c 2 44s we ry if J .'s ted uparaielf )
i 2. RCIC equipment area high differential temperature
[f 3. Low steam supply pressure
I REFERENCE r WNP-2 Systems, Vol. III, Tab 3, Pg. 14
'r c ANSWER 2.09 (1.00) i 1. FALSE (0.5) 0
- 2. False { The LOCAL / REMOTE switch will select the active button } (0.5)
[ REFERENCE F WNP2 - Systems, Vol IV, Tab 9, pg 43 I
ANSWER 2.10 (3.00)
/
loss of voltageo(A[nW derv d e.08 associated) class 1E bus j i c. 1.
- f. 2. high drywell pressure l 3. low reactor vessel water level (3 @ 0.5 ea)
- b. 1. Engine overspeed
- 2. Generator differential current 1 3. Incomplete sequence (3 @ 0.5 ea) il. REFERENCE
!!- WNP2 - Systems, Vol IV, Tab 9, pg 43 l;
f 1
l 1
h i
k b' 2. PLANT DESIGN INCLUDING SAFETY AND EMERGENCY SYSTEMS PAGE 22 i I
[.
l 1-ANSWERS -- IOP-0 -86/02/04-SHERMAN, J.
1.
Affected components will be non-regenerative heat exchangers [0.25]
[0.25), reactor water cleanup pumps [0.25), and FCV-33 [0.25].
- 2. High temperature at NRHX outlet will cause isolation valve V-4 to ANSWER 2.11 \ L , 3,V-4 ci sure causes RWCU pump trip [0.25] and FCV-33
{g ,, - v - - -
i
~
f1. Affected components will be RWCU pumps [0.5] anc
} non-regenerativ exchangers [0.5]. (1.0)
A 2. Operating RWCU will +2ip vi. ng temperature (0.5]. Hi h temperatur NRHX outlet will cause outboard isolation val e
((
e
-(V-4, o close [0.5]. (1.0)
REFERENCE f WNP2 - Systems, Vol I, Tab 9, pg# pp 7 ed L
i.
ANSWER 2.12 (2.25) [%ua,@#4)Aw.% h q E , G '" "3
- 1. Feedflow <30% with 15 sec. T.D. (cavitation Limit)
- 2. Main Steamline/ Pump suction temperature differential (9.9F L 3. Reactor vessel low level 3
- 1. RET T TErbine T,rw 0;7es! re;;;t s14? pa a turbine r5-f -tage pwos-(Tripe G G.5 ca) (Sci,yviu s G G.25 eg]-
[- g A>
-. - c L
_. . - . '( %
WNP2 - Systems, Vol I, Tab 6, pg 38, w u e sJr.s, 3/q ,3-44 ANSWER 2.13 (2.00)
- c. RCIC will not initiate, [0.25] the RCIC steam stop valve (RCIC-V-45) will not open if the exhaust valve (RCIC-V-68)
Jt is not full open [0.75]. an4 m,th( min ove e (1.0)
- b. RCIC will inject at maximum rat Aand Now,d) va lve wi'.1 not respond, (will remain open) [0.25] the flow signal is at minimum due to the zero d/p sensed therefore demanding Max.
flow from the RCIC system [0.75]. (1.0) i lf REFERENCE
} FNP-2 System & Procedures Vol III RCIC L.P. pg 11, 12 & 13 I
- 4. Turbine Throttle valve-closure < 5% closed.
[ ;
I 5. Turbine Governome valve - fast closure 21250 psig.
5 O
lfi
p . . _ _. .- . _ _ _ _ . _ _ _ _ . . . - _ . _ . _ . .
Y T. -
0 3.. INSTRUMENTS AND CONTROLS PAGE 23 F
I, ANSWERS -- WNP-2 -86/02/04-SHERMAN, J.
i ANSWER 3.01 (2.50) .
- 1. Reactor water low level (Level 2) 2.~ Main steam line high radiation
- 3. Main steam line high steam flew
- 4. Main steam line low pressure with mode switch in run
? 5. Main steam line tunnel high temperature, or high ventilation system differential temperature f
- 6. Main condenser low vacuum.
1 (5 @ 0.5 ea.)
k
$ REFERENCE WNP-2 Systems Vol. V, Tab 1, Pg. 16 r
h L
ANSWER 3.02 (1.00) e L REFERENCE WNP-2 Systems, Vol. V, Tab. 1, Pg. 33 c
ANSWER 3.03 (2.00)
- c. Hold pumps maintain enough flow to maintain precoat on the filter elements. (0.5)
, b. Starts in auto upon low flow (<1,600 rpm) through associated i derin unit. Stops when condensate flow re-established. (1.0)
F c. Precoat may have been lost. (0.5) i t
REFERENCE WNP-2 Systems, Vol. Y, Tab 11, Pg. 7
[
L t
p
[ ANSWER 3.04 (2.00) e.
[S- a. 1. Transfer from three element to single element control. (0.5)
- 2. Reactor vessel level setpoint transferred to +18 inches. (0.5)
I or in&. del es.k.IIer ,
- b. 1. Master controllerAis placed in manual. (0.5)
- 2. Setpoint setdown reset pushbutton is depressed. (0.5) l?
!I I,
i!
r L_
m
'P
- g. .
- 3. INSTRUMENTS AND CONTROLS PAGE 24 I[
P ANSWERS -- WNP-2 -86/02/04-SHERMAN, J.
t, i
7 REFERENCE ,
WNP-2 Systems, Vol. V, Tab. 14, Pg. 6 ANSWER 3.05 (3.00) e 7 c. No change (1.0) ic b. Decreases initially due to false low steam flow indication, 6
then returns to the same as initially when level equilibrates e
at lower point. (1.0)
- c. Lowers due to decreased feed flow initially, then stabilizes at some new lower point when level error matches steam-feed t flow error. (1.0)
L REFERENCE b WNP-2 Systems, Vol. V, Tab 14, Pg. 15 V
r ANSWER 3.06 (2.00)
- 2. Rod block
- b. Half-scram
- c. Rod block
[. d. Full scram F REFERENCE WNP-2 Systems, Vol. II,. Tab 4, Pg. 20 ANSWER 3.07 (2.00)
- c. New reading on Range 7 is 2.5. No auto actions. (1.0)
. b. New reading on Range 5 is 39. IRM high rod block and HI-HI half scram will be in. (1.0)
REFERENCE L WNP-2 Systems, Vol. II, Tab. 2, Pg. 14 l
-t
.t
- t. .
U 3. INSTRUMENTS AND CONTROLS PAGE 25 s
ANSWERS -- WNP-2 -86/02/04-SHERMAN, J.
v
'I F ANSWER 3.08 (2.50)
I~
- 1. Insert valves (121 and 123) activate to lift rod off collect f fingers.
T 2. Insert valves (121 and 123) close.
- 3. Withdraw valves (120 and 122) activate to withdraw rod.
} 4. Valve 122 (drive pressure) shuts.
W 5. Valve 120 (settle valve) shuts. (5 9 0.5 ea.)
desw.fr,on ide*l a f v4<c ne d -es AI.I s aa.ep Y word e REFERENCE WNP-2 Systems, Vol. II, Tab 6, Pg. 13 ANSWER 3.09 (2.50)
- o. Gamma (0.5) t b. SRM - pulse height discrimination.
IRM - Cambelling OR mean square voltage. (1.0) g c. Due to the low number of events and greater sensitivity, the t SRM deals with individual counts (pulses), while the IRM signal is a voltage level due to pulse overlap. (1.0) r REFERENCE WNP-2 Systems, Vol. II, Tab 1, Pg. 14 Tab 2, Pg. 7 e
[
E ANSWER 3.10 (2.50)
I c. U234 is added to the coating (0.5)
' b. 1. Bypass light on panel 608 r 2. The four-rod group display ( P 603)
- c. 1. LPRM upscale trip circuit.
I 2. LPRM downscal trip scircuit.
- 3. The associated APRM channel.
- 4. Plant Process computer (analog input).
II 5. Rod Block Monitor.
- 6. Four Rod display on C05 [4 @ 0.25 ea) (1.0)
}
l
, REFERENCE l' WNP2 - Systems, Vol II, tab. 3, pg 7,13,15 l
l?
E
- l
- 3. INSTRUMENTS AND CONTROLS PAGE 26 ANSWERS -- WNP-2 -86/02/04-SHERMAN, J.
E. ,
k p
in b
l ANSWER 3.11 (3.00) ,
- 1. Warn of abnormal radiation levels in areas where radioactive h" material may be present, stored, handled, or inadvertently
[ introduced. (1.0)
L I 2. a. Downscale trip [0.5]
[ b. Upscale trip [0.5] (1.0)
.f
? 3. Only one annunciator window is provided for reactor building F- area, so operator would have to check back panels to see which ARM's are alarming [0.5]. Specific locations of those ARM's would be determined from plant drawings (labels accepted if f
actually installed in plant)[0.5]. (1.0)
L REFERENCE WNP-2 Systems, Volume III Tab. 10, pp.1,6.
1 N
e L
i i
lt
- s
- l it l-I il';
s l?
if t:
- 4. PROCEDURES - NORMAL. ABNORMAL. EMERGENCY AND PAGE 27
. RA.DIOLOGICAL CONTROL
. ANSWERS -- WNP-2 -86/02/04-SHERMAN, J.
h;. ANSWER 4.01 (1.00)
- 1. RCIC (0.5)
! 2. RWCU (0.5)
REFERENCE WNP-2 Procedure 5.1.3, Pg. 5
[
I ANSWER 4.02 (2.00) i Loss of ALL of the following:
- 1. Offsite 230 KV startup power
- 2. Offsite 115 KV backup power s 3. Diesel Generator #1
, 4. Diesel Generator #2 (4 9 0.5 ea.)
e REFERENCE WNP-2 Procedure 5.4.1, Pg. 1 b
c; ANSWER 4.03 g (/,O0)
O. Confirm automatic transfer of or manually transfer HPCS (1.
and RCIC suction from the CST to the suppression pool.
@. RCICQ e
L REFERENCE WNP-2 Procedure 5.0.0., Pg. 2 I
r ANSWER 4.04 (2.50)
., 1. Initiate a manual scram H 2. Place mode switch in SHUTDOWN
- 3. Verify reactor power is decreasing (check APRM's) .
- 4. Verify all control rods have been inserted (various methods)
- 5. Verify reactor water level is being restored by one or more of Feedwater, RCIC, or HPCS (5 9 0.5 ea.)
AIJO accepV 5 fef f G- l} on 613 fed,4 i
I
3 t '
40% $. w Cenbud I.
1:
-[ 6 Wrify .90CXY Breakers 4885 and 4888 have opened. If y- not, depress either emergency tri:: push button en Panel C.
4, . ,
i 7 Verify the Main Turbine has tripoed. If net, depress both emergency trip pushbuttons en Panel B. ,
f 3. Press the red " RESET" pushbutten en the reheat te."pera-s tare controller until it backlights red, -
- 9. Verify the 4160V and 690GV buses have suto transferred g*
to the startt.p transformer.
l' "
80 Verify both Fesctor recirculatien ptrts have auto transferred to 15 hr.
t '
Verify scrsn discharge volume vent and crain valves indicate closed at Panel P603.
L i
t' '
i-f'
- i i
^f
'i lE r
,i Y
^
_ _ ~ - - _ _ _ _ _ ,
[- 4. PROCEDURES - NORMAL. ABNORMAL. EMERGENCY AND PAGE 28 y RADIOLOGICAL CONTROL 7
- f. r ANSWERS -- WNP-2 -86/02/04-SHERMAN, J.
.0 I
E REFERENCE WNP-2 Procedure 3.3.1, Pg. 3 ,
[f 7
i
$ ANSWER 4.05 (l.00) s k Wherenecessrytopreventinjurytopersonnel,fincludingthe
[ public) or damage to the facility. (1.0)
L REFERENCE re
$ WNP-2 Procedure 1.3.1, Pg. 3 ANSWER 4.06 (3.00) ras 3
- o. Prevent possible vibration of jet pumps,and riser braces.to.s') (1.0)
O_g Ens << ad < y ..r. co a p <a art Jo Eoos. : L OL A ,
P b. On ection of colder water from idle recire loop co,uld add
[ enough positive reactivity to cause criticality and/or power excursion. (1.0)
- c. Close discharge valve to prevent reverse rotation of pump.
Open discharge valve to maintain temperature of idle loop. (1.0) p p
REFERENCE
- WNP-2 Procedure 2.2.1, Pg. 3 i
j ANSWER 4.07 (1.00)
.1 Start and operate HPCS on recirculation to the CST until corrective I cction is completed. ( All e e ,J I F) , (lf 4 ,,g e, j, g s, f, ,,, (l.0)
,% ,a Syor lost , H PCS ge n 4 i k.tk to < ;A ,, ,4 p,,. , ;, ,
{
p re unt( guy inq , at.1 f t,e 7,a b f.h d REFERENCE < 7s y,,,.
-l WNP-2 Procedure 2.4.4, Pg. 2 g j,g,yg,,rg 4
- 4. PROCEDURES - NORMAL. ABNORMAL. EMERGENCY AND PAGE 29 E4DIOLOGICAL CONTROL
- . ANSWERS -- WNP-2 -86/02/04-SHERMAN, J.
/ l.#7.Notify CRS Verify Au w Actions
, 7. Take Muwal Control of FWL.C and reduce ItPV level ANSWER 4.98 (3.00) f fm
~ / _J ~
(1
/. e manual control of the controlling feedwater controller and
/ reduc feed rate to decrease reactor water level to i I normal . !
t E' 2. If level can't be ret to normal with c rolling controller, l r% '
adjust applicable RFP go controll FW/RPV DP to lower j level to normal. I 1 ,
/
- 3. If the high level condi occurred due to failure of level
! i instrumentation, to the alternate level in ment. ,/
a 4. Verif no ECCS system was inadvertently initiated .
/
L ure them if they are not needed .
g _/~~( 3 .ed- 9 1. 0 e a . )
REFERENCE WNP-2 Procedure 4.2.1.2, Pg. 2 w
ANSWER 4.09 (2.00)
- 1. Limits operator ability to control reactor
( 2. Loss of CRD cooling flow will cause control rod drives to be subjected to high temperatures, resulting in reduced seal life.
/ 3. Loss of flow to recirculation pump seals.
(2 9 1.0 ea.)
REFERENCE WNP-2 Procedure 4.1.1.2, Pg. 2 i
ANSWER 4.10 (2.00)
- c. CST water is cleaner than suppression pool. Want to minimize unnecessary contaminants in the reactor vessel. (1.0)
- b. High current through pump motor windings during start attempt heats them up. Excessive starts could damage windings. (1.0)
L
p 4. PROCEDURES - NORMAL. ABNORMAL. EMERGENCY AND PAGE 30
[e RADIOLOGICAL CONTROL p ANSWERS -- WNP-2 -86/02/04-SHERMAN, J.
p E REFERENCE
[ WNP-2 Procedure 2.4.4, Pg. 3 Y
L h
a j
ANSWER 4.11 (3.00)
- 1. Manually scram reactor.
f
. 2. Place mode switch to shutdown.
3
- 3. Manually close all MSIV's, MS-V-16, MS-V-19, and t~ RWCV-FCV-33.
{ 4. Verify APRM downscale lights illuminated.
- 5. Trip main generator and ensure that auxiliary power is being J~ supplied from.TR-S.
- 6. If recirculation pumps not on LFMG, transfer them to LFMG.
[ 7. Ensure only two condensate booster and two condensate pumps h are operating.
- 8. Ensure not more than two circulating water pumps are running.
- 9. Verify RFW valve positions.
[6 required 9 0.5 each] (3.0) w i
[ REFERENCE WNP-2 Procedure 4.12.1.1 L.
L ANSWER 4.12 (3.50)
. c. 1. Control Board Walkdown ,
(0.5)
! 2. Active Surveillances/Ta % (0.5)
- 3. Control Room Operators Log & W 4. e-=- m *M (0.5)
- 4. Offgoing CRO Summary Sheet a % A r eu.) (0.5)
- 5. % J o w- .
S b. 1. SRV indications 4. %dd #4 % i9-y* *%M
" *Pa*
^W .S (0.5)
I 2. Annunciator test ' N H5 (0.5) i 3. Indicating lamp survey Y. C C % , _ ,3 _..... (0.5)
~ '
~;
REFERENCE - .
f WNP-2 Procedure 1.3.6 .,
- r I
l i
1i -
L
=
U. O. NUCLCAR GCCULATORY COMMIE 5 ION SENIOR REACTOR OPERATOR LICENSE EXAMINATION s
FACILITY: W NP-2___________________
f jJ _
RCACTOR TYPE: _BWR-g[5_________________
L' DATE ADMINISTERED: _86/9gf93________________
EXAMINER: _ MILLER 2_L.______________
APPLICANT: _________________________
INSIBUCIIQUS_IQ_9EELIC6UIl Use separate paper for the answers. Write answers on one side only.
Staple question sheet on top of the answer sheets. Points for each question are indicated in parentheses after the question. The passing grade recuires at least 70% in each category and a final grade of at least 80%. Examination papers will be picted up s 1 >: (6) hours after the examination starts.
% OF CATEGORY % OF APPLICANT'S CATEGORY
__Y66Ug_ _IDIe6 ___SGOBE___ _YeLug__ ______________geIgggsy_____________
_2D 99__ _25t99 ___________ ________ 5. THEORY OF NUCLEAR POWER PLANT OPERATION, FLUIDS, AND THERMODYNAMICS
_2Ez99__ _25z99 ___________ ________ 6. PLANT SYSTEMS DESIGN, CONTROL, AND INSTRUMENTATION
_2E 99__ _29z99 ___________ ________
- 7. PROCEDURES - NORMAL, ABNORMAL, EMERGENCY AND RADIOLOGICAL CONTROL
_2Ez99__ _25199 ___________ ________ 8. ADMINISTRATIVE PROCEDURES, CONDITIONS, AND LIMITATIONS 199199__ 199199 ___________ ________ TOTALS FINAL GRADE _________________%
All work done on this examination is my own. I have neither given nor received aid.
kPPLICAsi s siss5iURs-----~~~~~~~-~
i, @e -- THE96Y_9E_NyC(g68_EQWES_EL@NI_QEEE611gN2 _ELVIDHi_eNQ PAGE 2 ISEBdQ9YNed1GE E
o 4
QUESTION 5.01 (1.50) i For each of the following events, STATE WHICH coefficient of i reactivity (Power coefficient, void coefficient, moderator coefficient or doppler coefficient) would act FIRST to change reactivity.
I a. Control rod drop at power (0,5) i b. SRV opening at power (0.5)
- c. One recirc pump trips while at 50% power (0.5) i DUESTION 5.02 (2.00)
Three (3) minutes following a reactor scram from high power, indicated reactor power is 75 on range 4 and decreasing.
- a. WHAT will INDICATED power be one (1) minute later?
0 (Show calculations) ($.0)
- b. Explain WHY power decreased at this rate. (1.0) 1 4
l QUESTION 5.03 (1.50) f'
.Concerning control rod worth, compare withdrawing a center control rod at 90% rod density to withdrawing a center control rod at 40% rod density. Which situation is control rod worth l greater for the withdrawn control rod? Explain your answer.
J
, QUESTION 5.04 (2.00) i The reactor in operating at 75% power. Recirculation flow is subsequent., .ncreased to provide 100% power and 100% flow.
Dascribe HOW the reactor power is increased with the increased recirc flowrate. Continue your description until stnady state conditions are reached (include CORE VOID CONTENT AND CORE NET REACTIVITY in your discussion).
(*****- CATEGORY 05 CONTINUED ON NEXT PAGE *****)
5:__IBE96Y_9E_NuCLges_EgWEB_E69NI_gEEBBIlgN3 _E(QlDS 3_8ND PAGE 3 ISEBMgpyNOMICS QUESTION 5.05 (1.00)
What is the ' REVERSE POWER EFFECT' which can occur in the core during control rod movements.(Include how it occurs in your explanation)
OUESTION 5.06 (2.00)
A reactor has just scrammed from entended full power operation.
Ten (10) hours later cooldown is complete, and the SDM is measured at that time to be 1% dk/k.
A. Is the measured SDM acceptable? EXPLAIN (1.0)
B. What are the consequences, if any, to changes in the SDM for the next 20 hours2.314815e-4 days <br />0.00556 hours <br />3.306878e-5 weeks <br />7.61e-6 months <br />? (1.0) l QUESTION 5.07 (1.50)
Consider a real plant system (NON-IDEAL) with two centrifugal pumps in parallel, one of which is running. The second pump is started.
A. System flow will be: (Choose the correct answer.)
((NOTE BOTH PUMPS OPERATING G 1800 RPM)) (0.5)
- a. double the original flow
- b. less than double the original flow
- c. greater than double the original flow
- d. the same only the discharge head changes B. EXPLAIN YOUR CHOICE of system flow response in part A. (1.0)
(***** CATEGORY 05 CONTINUED ON NEXT PAGE
- )
l
5,__ISEQBy_QE_NQGLEGE_EQWE8_E(@NI_QEEB8IlgN1 _ELUlpS,_GNQ PAGE 4 IHE8MQQyN6dlCS QUESTION 5.08 (3.00)
The reactor is operating at 100% power when HPCS inadvertantly initiates. Describe the response of the following parameters during the transient, including why the parameter changes as it does. Assume NO SCRAM occurs. Continue your description until steady state conditions are reached.
- a. Turbine Load (1.0)
- b. Reactor Water Level (1.0)
- c. Feedwater Flow (1.0)
QUESTION 5.09 (2.50)
With regard to the MAPLHGR thermal limits
- a. Briefly, WHAT is the reason, or bases for having a MAPLHGR thermal limit? (1.0)
- b. WHICH TWO of the following four parameters affect the the MAPLHGR LIMIT? (0.5)
- 1. Moderator Temperature
- 2. Type of fuel
- 3. Fuel exposure
- 4. Reactor pressure
- c. If a P-1 isselected on the Process Computer, the program provides, among other things, MAPRAT. What is the relationship between MAPRAT and MAPLHGR? (1.0)
(***** CATEGORY 05 CONTINUED ON NEXT PAGE **+**)
Dz__IHEQBy_QE_NUQ(E@B_EQWEB_E6@NI_GPER@llQNi _ELQIQSx_AND PAGE 5 IHEEdQQyN@dlQs QUESTION 5.10 (3.00)
Following a normal reduction in power f rom 90% ' a 70% wi th rccirculation flow, HOW will the following change (increase, decrease, or remain the same) AND WHY:
- a. The pressure difference between the reactor and the turbine steam chest. (1.0)
- b. Condensate depression at the exit of the condenser. (1.0)
- c. Final Feedwater temper ature. (1.0)
OUESTION 5.11 (2.00)
Indicate on the attached figure (FIGURE 1) or discuss, HOW the following stresses vary across the reactor vessel wall during a reactor heatup. (2.0)
- a. Thermal stress
- b. Pressure stress
- c. Combined thermal / pressure stress QUESTION 5.12 (3.00)
Since the parameters which result in fuel damage are not directly obgervable during reactor operation, CRITICAL POWER is adopted as e convenient limit. For each of the following factors, Otete HOW critical power will change (increase, decrease or rcmains the same). EXPLAIN your answer.
- c. Inlet subcooling increases (1.0)
- b. Core flowrate increases (1.0)
- c. Reactor pressure increases (1.0)
(***** END OF CATEGORY 05 *****)
6 2__ELONI_@y@lEM@_ DESIGN 3_CQNIBQL2_6ND_INSISQMENI@llgN PAGE 6 QUESTION 6.01 (2.00)
With regard to the Intermediate Range Monitoring System (IRM), answer the following questions:
- a. If the physical hardware that provides gimma discrimination for the Intermediate range is not considered, is gamma discrimination necessary over the entire Intermediate range of the neutron instrumentation? EXPLAIN YOUR ANSWER. (1.0)
- b. What three conditions will result in an IRM inoperative Rod Block? (1.0)
OUESTION 6.02 (2.00)
Unit 2 is operating at 100% rated thermal power, with recirc in Master Manual (flux automatic). An operator inadvertently INCREASES the " Pressure Set" on EHC by 5 psig.
ASSUME: 1. No Further Operator Actions
- 2. All other DEH control settings are normal
0% Steam Flow Position o Power - 100% Rated Thermal Power o Pressure - 1005 psig NOTE: FIGURE # 2 IS ATTACHED FOR REFERENCE Which of the following most accurately describes both the INITIAL RESPONSE and FINAL STATUS of the different parameters and components.
a b c d INITIAL RESPONSE o TCV's lCLOSE ('83%) lCLOSE ('83%) ICLOSE ('83%) i NO CHANGE o BPV's i NO CHANGE ! OPEN (*17%) ! NO CHANGE : OPEN (*25%)
o Power i INCREASE : NO CHANGE : INCREASE l DECREASE o Pressure : INCREASE l NO CHANGE I INCREASE I DECREASE FINAL STATUS ! ! ! 1 l I 1 !
o TCV's : *100 %
- 83 %
- B3 % l 100 %
o BPV's : O% !
17 % :
17 % ! O%
o Power *100 % : 100 % ! > 100 % ! < 100 %
! o Pressure >1005 psig i 1005 psig! >1005 psigt <1005psig i
j (***** CATEGORY 06 CONTINUED ON NEXT PAGE *****)
l
, ez__ELONI_gy@lgdQ_QEglGN i_QQN18Q(1_@ND_idgIBydENI@llgN PAGE 7 QUESTION 6.03 (1.50)
- e. The new LPRM detectors (the NA-2OO detector) have extended lifetimes, at least 3 times that of the NA-100 detectors for the same end of life output currents.
What are the two features of the new LPRM detectors which allow the extended life ? (1.0)
- 6. A Core Thermal Power and APRM Calibration program (DD-3) is performed and shows APRM A with a Gain Adjustment Factor of 1.03. What does this tell the operator about the relationship between actual and indicated power on APRM channel A? (0.5)
QUESTION 6.04 (1.50)
Following a reactor scram, the four rod display position go2s blank, but the green full-in light on the full core display for that control rod is lighted. Is this normal? If so, explain why it occurs. If not, describe the probable cause. (1.5)
QUESTION 6.05 (2.00)
For each of the following situations (1 and 11) select the correct Feedwater Control System / plant response from the list (a through e) which follows.(An answer may be used more than once, cnd no operator actions are taken.)
- 1. The plant is operating at 70% power, in 3-element control, when one of the Steam Flow Detectors FAILS DOWNSCALE.
ii. The plant is operating at 90% power, in 3-element control, when HPCS inadvertently initiater, and injects FEEDWATER CONTROL SYSTEM / PLANT RESPONSE
- a. Reactor water level decreases and stabilizes at a lower level.
- b. Reactor water level decreases and initiates a reactor scram.
- c. Reactor water level increases and stabilizes at a higher level.
- d. Reactor water level increases and initiates a turbine trip.
- e. None of the above.
(***** CATEGORY 06 CONTINUED ON NEXT PAGE *****)
I.
, 6i__ELhNI_@ygIEd@_QE@lGN,_GQNIBQL2_QNQ_INSIBQUENI@IlQN PAGE. 8 l
J QUESTION 6.06 (2.00)
Ths APRM scram function actually consists of two separate satpoints:
, o. - FILL IN THE BLAJKS: .
. Flow Biased Scr am *w + __ 2)( __%
(1) __
(0.25)
Fixed Scram -
__ (3) __% (0.25) l
- b. LIST the specific location (s) of the sensor (s) which measure the variable "w". (0.5)
- c. While operating at power, one MSIV fails shut resulting in
- a brief (* 1 second) flux spike to 121% power. STATE which of the two scram setpoints mentioned above (one or both) should initiate a reactor scram. JUSTIFY your choice. (1.0) j i QUESTION .6.07
< (4.00)
. D2 scribe the operation of the following systems for the l following given conditions. Consider each situation -
! separately. State if the system pump will start (and run)
]' and if the system will inject into the vessel and WHV.
GIVEN: The system receives a valid initiation signal.
I A. -The RCIC exhaust valve (RCIC-V-60) is stuck shut. (1.0) 3
} B. The LPCS suction valve is shut. (1.0)
C. The RHR pump 2C suction valve is shut. (1.0)
D. The RHR pump 2D suction valve is shut. (1.0)
(***** CATEGORY 06 CONTINUED ON NEXT PAGE *****)
6z__ELONI_gy@Igdg_QgglgN _CQUIBQL2_@NQ_INgIBUdENI@I19N 2 PAGE 9 l
QUESTION 6.08 (3.00)
With regard to the Recirculation Flow control system, answer the .following questions.
- a. What is the purpose of the FLUX ESTIMATOR 7 (1.0)
- b. What two conditions will cause the FLUX ESTIMATOR to l
" automatically" shift from estimated flux signal to l
neutron flux as the flux feedback signal to the flux i estimator. (1.0)
- c. Give 5 of 10 conditions that will cause a FLOW CONTROL VALVE to " LOCK-UP" (1.0)
QUESTION 6.09 (3.00)
Concerning the RECIRCULATION FLOW CONTROL SYSTEM NETWORK, with the network in automatic flow control (flux manual loop automatic) and 90% core flow:
A. what would be the results on individual loop flows (GIVE APPROMIXATE VALUES) if the Recirculation loop f
flow A feedback signal DECREASED TO ZERO 7 EXPLAIN YOUR ANSWER (RECIRC FLOW CONTROL NETWORK Figure 3 attached) (2.0) l B. What would stop the flow control valve movement with the conditions as stated in part A 7 (1.0)
I QUESTION 6.10 (3.00)
What are three (3) anticipatory scrams, how is each sensed, end when is each bypassed ? (INCLUDE SETPOINTS) l l
l l
l
(***** CATEGORY 06 CONTINUED ON NEXT PAGE *****)
6i__EL8NI_EYSIE55_ DESIGN 2_G9 NIB 962_eND_INSIBydgNIGI196 PAGE 10
-QUESTION 6.11 (1.00)
Thz Main Generator is on line at 810 megawatts when a hydrogen lock in the generator reduces hydrogen pressure to 45 psig.
Using the attached figure #4 (Estimated Capability Curve),
which of the following is the ma>:imum lagging Reactive load allowed on the generator, if a power factor of 0.975 is to be maintained.
A. 890 MVAR B. 200 MVAR C. 280'MVAR D. 430 MVAR
(***** END OF CATEGORY 06 *****)
Zz__ES99EDuggg_:_NgBMeb,_egggede62_gegsggNgy_eNg PAGE 11 89DI96991Ge6_G9NIS96 QUESTION 7.01 (2.00)
Briefly explain-WHY each of the following RECIRCULATION PUMP STARTING LIMITATIONS are necessary. De specific.
A. The pump in an idle recirculation loop shall not be started unless the temperatures of the coolant within the idle and the operating recirc loop are within 50 degrees F of each other. (1.0)
B. If the temperature of the water in the lower head is more than 145 degrees F below vessel saturation temperature, the recirc pump shall not be started (1.0)
QUESTION 7.02 (2.50)
For the suppression chamber water temperatures listed below, WHAT ACTIONS are required by the Technical Specifications with the unit in operational condition 1 or 27 A. 97 degrees F (0.5)
B. 108 degrees F during RCIC testing (1.0)
C. 121 degrees F following a scram with the MSIV's SHUT. (1.0)
QUESTION 7.03 (2.00)
According to the TECHNICAL SPECIFICATIONS for the REACTIVITY CONTROL SYSTEMS:
A. When must the Rod Worth Minimizer be operable and what actions are necessary to continue operation with the RWM inoperable? (1.0)
B. Define the term " Limiting Control Rod Pattern". (1.0)
(***** CATEGORY 07 CONTINUED ON NEXT PAGC *****)
e
. Zz__BBgCggyBgS_:_NQBUS62_8pNQBd@(2_EdEBGENCy_GNQ PAGE 12 6091969GIC8(_GQNIBQL QUESTION 7.04 (2.50)
A reactor Cooldown is in progress and NEITHER loop of RHR ccn be placed in the Shutdown Cooling Mode. In accordance with Emergency procedure 5.3.5, Alternate Shutdown Cooling (contingency) has been established to remove decay hOat and continue the cooldown. STATE the condition / status of the following components / parameters when operating in this mode of shutdown cooling.
- c. MSIV's (0.5)
- b. SRV's (0.5)
- c. RHR Loops A & B (0.5)
- d. Reactor Pressure (Compared to Suppression Chamber Pressure) (0.5)
O. Reactor Level (Provide numerical value or component reference) (0.5)
QUESTION 7.05 (1.50)
A LOCA has occurred and a high temperature steam environment Cxists in the drywell. EXPLAIN why the drywell sprays must NOT be initiated in the " Unsafe" region of attached Figure #5 "Drywell Spray Initiation Pressure Limit".
(***** CATEGORY 07 CONTINUED ON NEXT PAGE *****)
l
PAGE 13
,-Zz__EB9GEDUEES_:_NQ8d6b4_@@N96d@b2_EdE6@gNQy_@NQ BOD 196991Geb_G9NIS96 l
1 I
QUESTION 7.06 (3.00)
A. Procedure 3.3.1 REACTOR SCRAM gives operating requirements for 2 situations for the RWCU system and states that the operator should observe these operating requirements. (CONSIDER each situation seperately and occurring after a reactor scram).
SITUATION #1 - The RFW CAPABILITY is maintained (RFWDT OPERATING) and the RWCU isolates.
SITUATION #2 - The RFW CAPABILITY is lost (RFWDT TRIP OR MSIV CLOSURE) and the RWCU does not isolate.
- 1. What are the operating requirements for situation #17 (0.75)
- 2. What are the operating requirements for situation #27 (1.25)
- 3. What is the purpose of these operating requirements? (0.5)
B. After a reactor scram with the loss of condensate booster pumps, WHY should the RWCU be lined up to return to the RPV before restarting a booster pump, per procedure 3.3.1, Reactor Scram. (0.5)
QUESTION 7.07 (1.50)
During a reactor shutdown with the Reactor power at 18% the Rod Sequence Controller becomes inoperative. The STA recommends that he will act as the second knowledgeable individual to allow continued rod insertion. As the Shift Manager would you follow the STA's recommendation ? EXPLAIN your answer.
(***** CATEGORY 07 CONTINUED ON NEXT PAGC *****)
Zz__BB9GEDUBES_:_NQBd863_6BNQBM863_EMEBGENQY_8ND PAGE 14 88D196991G96_G9NIBg6 QUESTION 7.08 (1.00)
As the Shift Manager you have determined that a Reactor chutdown has not been extensively disruptive to the normal alignment of systems, and therefore you can use the Minimum Startup Checklist to ensure preparations are made for a safe end orderly Reactor startup. The Minimum Startup Checklist nsed not be completed in full if two (2) conditions are met.
What are those two (2) conditions.
QUESTION 7.09 (3.00)
Tho plant is in HOT shutdown (condition 3) with loop 'A' of RHR unavailable. On a loss of the RHR 'B' HX, WHAT are 3 of the 5 proferred alternate methods provided in PPM 4.4.2.1 LOSS OF RHR SHUTDOWN COOLING MODE LOOPS to control reactor temperature.
(For each method, state the system used to provide the reactor cooling water, source of the water used, and the heat sink.)
(***** CATEGORY 07 CONTINUED ON NEXT PAGE *****)
Zz__EB99EDUBES_ _NQBM963_8pNQBd@L3_EMEBQENQy_8ND PAGE 15 88DI96991986_QQNIBQL QUESTION 7.10 (3.00)
A reactor scram has occurred. Four adjacent control rods have feiled to insert past position 06.
A. Match the following sets of indications with the appropriate potential problem type.
- 1. 3 RPS white lights are ON, a. Air problem 1 RPS white light is OFF, 4 blue lights on the full core b. Hydraulic problem display, are NOT ON
- c. Electrical problem
- 2. All RPS white lights are OFF, 4 blue lights, on the full core display, are NOT ON
- 3. All RPS white lights are OFF, all blue lights, on the full core display are ON (1.0)
B. With a number of control rods immovable, such as above what further criteria needs to be met, per PPM ",.1.3 Reactor Power Control to require initiating Standby Liquid Control? (1.0)
C. If boron injection is required and can not be injected with Standby Liquid Control, what systems per PPM 5.1.3 Reactor Power Control (RPV/Q) are to be used to inject boron into the vessel. (1.C)
QUESTION 7.11 (3.00)
PPM 4.12.1.1 CONTROL ROOM EVACUATION, lists NINE (9) immediate opcrator actions that should be performed prior to leaving the control room. What are six (6) of the immediate operator cetions performed in the control room 7
(***** END OF CATEGORY 07 *****)
Dz__0Dd1NISIEGI1YE_EB9GEDUBES2_G9991I19BS2_0Np_6]dlIGIlgNS PAGr 16 QUESTION 8.01 (2.50)
During plant operation a Drywell entry is desired to locate o possible leak. Give 5 of 8 Prerequisites which must be met prior to personnel entering the Drywell.
QUESTION O.02 (1.50)
A licensee may take reasonable action that departs from a license condition or a Technical Specifications in an cmergency when this action is immediately needed to protect the public health and safety and no action consistent with license conditions and Technical Specifications that can provide adequate or equivalent protection is immediately cpparent.
A. Such an action shall be approved, as a minimum,by WHOM7 (0.5)
B. What two (2) notifications should be made prior to the above action if at all possible, but always as soon as possible afterwards ? (1.0)
OUESTION 8.03 (3.00)
Ancwer the following questions with regard to the issuance of a Radiation Work Permit (RWP).
A. What are the radiological limits that require the use of a RWP7 (1.5)
B. What are the resposibilities of the Shift Manager, when reviewing a RWP for approval? (1.5)
OUESTION 8.04 (1.00)
Prior to accepting a clearance order, a maintenance coper vi sor determines additional clearance tegn are required for a safe working condition. How are the additional clearance tags documented as approved on the previously cuthorized clearance order ?
(***** CATEGORY 08 CONTINUED ON NEXT PAGE *****)
B___8Dd1NISIBBIIVE_EB9GEDUBES2_GQNQ111gNg3_@NQ_LJdlIGIlgNS PAGE 17 1
QUESTION 8.05 (2.50)
A. After completing the maintenance on a piece of equipment the responsible maintenance group left the site without obtaining a clearance release. Operations has a need for this equipment. Name 3 individuals by title who can authori:e the clearance release if the individual it is issued to can not be contacted. (1.5)
B. When is " Redundant Verification" required by the equipment clearance and tagging procedure 7 (1.0)
QUESTION 3.06 (1.00)
After maintenance on a piece of equipment an operational
.chackout of the equipment is desired. According to the Equipment Clearance and Tagging Procedure, Give 2 of 3 rostrictions that would prevent the Shift Manager from cuthorizing the equipment checkout .
QUESTION 8.07 (1.00)
When is the Plant Manager's permission required bef ore the reactor ccn be restarted after a reactor scram 7 OUESTION 8.08 (3.00)
During the performance of a surveillance on a safety-related cyntem the need f or a TEMPORARY FROCEDURE DEVI ATION arises.
A. Give 4 of 7 specific guidelines that must be complied with in the preparation of the temporary procedures deviation. (2.0)
B. Is the procedure deviation required to be documented prior to its implemention ? EXPLAIN YOUR ANSWER. (1.0) l
(***** CATEGORY 08 CONTINUED ON NEXT PAGE *****)
e z __eQd1NISIEeIlyE_EEQQEQUEEgi_QQNQlIlgN@i_QNQ_LidlI@IlgNS PAGE 18 QUESTION 8.09 (3.00)
What is the guideline or instruction given #or each of the following per the standing operating order s, 1.3.1 attachment 1 ?
A. Overriding the automatic action of an ECCS system. (0.75)
B. Placing a controller in the manual mode from the automatic mode. (0.75)
C. A safety related motor operated valve has been manually backseated. (0.75)
D. The instructions for aligning more than 2 valves or circuit breakers. (0.75)
OUESTION O.10 (3.00)
Whct is the minimum facility staffing required for WNP-2 by tha Technical Specifications when in condition 2 (mode 2) cnd indicate what type of license is required for each cteffing position, if any 7 OUESTION 8.11 (2.00)
Fual Zone Indicator MS-LI-610 is classified as " unqualified" in the event of a LOCA per STANDING ORDERS / NIGHT ORDER PROCEDURE 1.3.1.
A. If this instrument appears to be working during a LOCA, can the instrument be used? (1.0)
D. Explain why the instrument can or cannot be used.
(DO NOT JUST SAY BECAUSE OF THE PROCEDURE) (1.0)
(***** CATEGORY 00 CONTINUED ON NEXT PAGE *****)
r-I- .
.9___9951NISIBBI1YE_EBQCEDUBES2_ggNp111gNg3_9NQ_LINII8IlgNg PAGE 19 QUESTION 8.12' (1.50)
You as the' Shift Manager. are reviewing a procedure change '
to a Technical Specifications surveillance which changes
.tha setpoint of the Reactor Vessel Steam Dome Pressure High Scram from less than or equal to 1057 psig to 1065 psig plus or minus 5 psig.
Does this procedure , change involve an unreviewed saf ety question concern? Explain your answer. -(1.5) f
(***** END OF CATEGORY 08 *m***)
(************* END OF EXAMINATION ***************)
Ez__IMEQSY_QE_UUCLEBB_EQWEB_ELONI_QBEB811 Qui _ELU10Si_eUD PAGE 20 IBEBdQQyN6dlQS ANSWERS -- WNP-2 -86/02/03-MILLER, L.
ANSWER 5.01 (1.50)
- a. Doppler or fuel temperature
, b. Void
- c. VOID (EACH 0.5 pts)
REFERENCE WNP-2 REACTOR THEORY - STUDENT TEXT pg. 26-31 G.E. REACTOR PHYSICS, REACTOR FUNDAMENTALS TRAINING f
ANSWER 5.02 (2.00)
- a. Using P = Po e to the t/T then P = 75 e to 60/-80 P = 75 e to -0.75 = 35 on Range 4 [1.03
- b. On down-power transients, the rate of power change is limited by the rate of decay of the longest lived precursors,thus retarding the rate of power decrease.[1.03 REFERENCE WNP-2 REACTOR THEORY - STUDENT TEXT pg. 14,24 G.E. REACTOR PHYSICS, REACTOR FUNDAMENTALS TRAINING
~ ANSWER 5.03 (1.50)
Withdrawal of a center control rod at 90% densi ty has greater worth. (0.5)
Tha control rod worth is proportional to the (local neutron flux / the core avarage neutron flux) squared.(0.25)
With 90% rod density the core average neutron flux is very small.
Withdrawing a central control rod, increases the local flux in the area of withdrawn rod substantially. Because the rod causes the value of the term (local neutron flux / core average neutron flux) squared to be large its worth for this condition is quite high. Higher than withdrawing the rod at 40% rod density, when core average flux will be higher. (0.75)
REFERENCE WNP-2 REACTOR THEORY - STUDENT TEXT pg. 26 - 36 i G.E. REACTOR PHYSICS, REACTOR FUNDAMENTALS TRAINING i
l -
k
_ _ _ _ _ _ _ _ __ . La _ . _ w _
PAGE 21 Ut__ISEQBY_RE_ NUCLE 68_EQWEB_ELOUI_QEES@l1QNs_E(Q1QS3_6NQ
. IEEBdQDYNOblCS ANSWERS -- WNP-2 -86/02/v5-MILLER, L.
i ANSWER 5.04 (2.00)
Voids initially decrease (0.25) as the increased flow moves the boiling boundary higher in the core.(0.25) The decrease in void content initially causes a positive reactivity addition. (0.25) As power increases, the rate of boiling increases (0.25), the increased void formation adds negative reactivity (0.25), and increased fuel temperature adds negative rocctivity (0.25). The boiling boundary returns to near its original level (0.25). The not reactivity returns to 0 at steady state conditions (0.25).
REFERENCE WNP-2 SYSTEMS VOL.I RECIRCULATION FLOW CONTROL pg.26 WNP-2 REACTOR THEORY - STUDENT TEXT pg. 28 ANSWER 5.05 (1.00)
Reverse power effect is a decrease in power with a notch withdrawal of a shallow control rod. The notch withdrawal causes bundle power to increase where the rod is withdrawn. The local power increase will cause increase voici content in the bundle. The negative effect of the increase voiding is larger than the positive effect of the notch withdrawal of the shallow rod.
REFERENCE WNP-2 REACTOR THEORY - STUDENT TEXT pg. 40 G.E. REACTOR PHYSICS, REACTOR FUNDAMCNTALS TRAINING ANSWER 5.06 (2.00)
A. Shutdown margin assumes xenon free for its LCD limit.
The shutdown margin is not acceptable if the reactor is only shutdown by 1% dk/k as measured at peak xenon. (1.0)
B. Since the peak xenon reactivity is greater than the 1% SDM a reactor restart would occur. (1.0)
REFERENCE WNP-2 Reactor Physics Sec.V Fission Product Poisons,part 2 Xenon bchavior after Reactor Shutdown WNP-2 TECHNICAL SPECIFICATIONS 1.39 pg. 1-7
5 1__IMEQBY_QE_NQgLE88_EQWEB_E66NI_9EEB611gN,_E6919@3_9ND PAGE 22 ISEBMQDYN@blg@
ANSWERS -- WNP-2 -96/02/03-MILLER, L.
ANSWER 5.07 (1.50)
A. answer'b, less than double the original flow (0.5)
B. Less than double the original flow when deltvering water into a piping system that offers frictional resistance, 2 pumps operating in parallel will encounter greater resistance to flow. The increased frictional resistance lowers the total flow to less than twice t.1e original flow. 01.03 REFERENCE l Thermodynamics, Heat Transfer and Fluid Flow pg. 7-121 ANSWER 5.08 (3.00)
- a. Turbine load would decrease (0.5) due to the decrease in reactor pressure caused by the cool water spraying into the upper plenum,
- b. Reactor level will increase (0.5). A level error must be generated to reestablish steady state conditions in the FWLCS (0.5).
- c. Feedwater flow will decrease (0.5). The HPCS injection i s providing a portion of the required feed for the reactor and this is not sensed by the Feed Flow detectors (0,5).
REFERENCE WNP-2 FSAR CH.15.5.1 INADVERTANT HPCS STARTUP ANSWER 5.09 (2.50)
- c. Minimize fuel damage during a DBA LOCA by limiting the peak clad temperature (to < 2200 F) -OR- limiting bundle stored energy. (1.0)
- b. 2 and 3. (0.5)
- c. MAPRAT = APLHGR/LIMLHGR -or- = APLHGR/MAPLHGR limit (1.0)
-or- -= (APLHGR) actual / (MAPLHGR) LCO max REFERENCE WNP-2 G.E. Thermodynamics, Chapter 9, BWR Thermal Limits pg 9-68, 9-71 and 9-74 l
l l
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5 t__IME98Y_9E_NyG(g@B_EQWE8_E(@NI_QEEB811gN2_E(y1Q@i_@NQ PAGE 23 ISEBdQQYN@dlG@
ANSWERS -- WNP-2 -86/02/03-MILLER, L.
ANSWER 5.10 (3.00)
- e. Decreases EO.253. There is less steam flow, therefore', less pressure drop through the main steam lines CO.753. (1.0)
- b. Increases [0.253. With the same amount of cooling water through the condenser CO.25] and less of a heat load [0.52. (1.0)
- c. Decreases CO.25]. Less extraction steam from the turbine to heat the feedwater [0.753. (1.0)
REFERENCE WNP-2 SYSTEMS VOL.V DEH Fig. 13A WNP-2 SYSTEMS VOL.V EXTRACTION STEAM WNP-2 SYSTEMS VOL.V CIRCULATING WATER ANSWER 5.11 (2.00) ,
During heatup, the thermal gradients in the reactor vessel wall produce thermal stresses which va y from compressive at the inner wall to tensil at the outer wall. These thermal induced compressive stresses tend to alleviate the tensil stresses induced by the internal pressure.
(Drawing attached Figure 1)
REFERENCE WNP-2 TECHNICAL SPECIFICATIONS BASES 3/4.4.6 ANSWER 5.12 (3.00)
- a. Critical power increases (0.25) due to the increase in enthalpy rise that is required to bring the coolant to saturated conditions (0.75).
- b. Critical power increases (0.25) due to increased power required to bring the coolant to saturation conditions or increasing the flowrate increases the heat removal capability and places the ;
bundle farther from OTB. (0.75)
- c. Critical power decreases (0.25) due to it requiring a smaller enthalpy rise to change a liquid into a vapor at higher pressures than at lower pressures (0.75).
' 5___IHEQBy_QE_NUCLEBB_EQWEB_E60NI_9EE60I1982_E6ulps,_OND PAGE 24 IHEBdgDyNgd195 ANSWERS -- WNP-2 -86/02/03-MILLER, L.
REFERENCE WNP-2 THERMODYNAMICS, HEAT TRANSFER AND FLUID FLOW 9-85 TO 9-87
6:__E68NI_SygIEM@_ DESIGN3 _CQNISQ62_@ND_INSISyMgNI@Ilgd PAGE 25 ANSWERS -- WNP-2 -86/02/03-MILLER, L.
ANSWER 6.01 (2.00)
- a. NO (0.25), the fission gamma and neutron signal produced for the upper portion of the IRM range is greater than the signal that is contributed by the decay gammas, and fission gammas are proportional to the number of fissions tal:ing place.(0.7P)
- b. 1. Detector High Voltage Low [0.333
- 2. Module Unplugged E0.333
- 3. IRM Mode Switch not in operate CO.333 REFERENCE WNP-2 Systems Vol. II, IRM pg. 23 WNP-2 SYSTEMS VOL. 11, SRM figure 10 ANSWER 6.02 (2.00)
A REFERENCE WNP-2 SYSTEMS VOL V DEH pg. 62 ANSWER 6.03 (1.50)
- a. 1. The U234 loading (79%) allows the detector to regenerate the fissile coating.
- 2. The sputtering effect is reduced.
b.. The APRM reading is lower than actual power (0.5).
REFERENCE WNP-2 SYSTEMS VOL.II LPRM pg.7 WNP-2 SYSTEMS VOL.I PROCESS COMPUTER pg.14 ANSWER 6.04 (1.50)
Yes,11t is normal (0.5). The drive piston moves the RPIS magnet past the "OO" reed switch (or the control rod moves past "00" position) and actuates only the green full-in light," overtravel in" reed switch (1.0).
i
6:__EL8NI_SYSIEMS_QE@lgN3 _QQUIBQL2_9ND_INEIBydENI@IlgN PAGE 26 ANSWERS -- WNP-2 -86/02/03-MILLER, L.
REFERENCE WNP-2 SYSTEMS REACTOR MANUAL CONTROL, pg 19 ANSWER 6.05 (2.00)
- i. a ii. c REFERENCE WNP-2 SYSTEMS VOL V. FWLC pg 15 WNP-2 FSAR 15.5.1 INADVERTANT HPCS STARTUP ANSWER 6.06 (2.00)
- a. .66
- w + 51% (0.25) 118% (0.25)
- b. Recirc Loop flow elements (pump suction) (0.5)
- c. Only the 118% fixed scram (0.5) This is because the flow biased scram incorporates a time delay into its actuation (* 6 seconds, representative of the fuel thermal time constant) (0.5)
REFERENCE WNP-2 SYSTEMS VOL II APRM pg. 9,13 AND 20
62__EL8NI_Sy@IEDS_DEgigN3 _ggNIBQL2_@Np_INSIBQDENIOllgN PAGE 27 ANSWERS -- WNP-2 -86/02/03-MILLER, L.
ANSWER 6.07 (4.00)
A. The RCIC pump will not start (0.25) and will not inject. (0.25)
(RCIC-V-45) will not open if the exhaust valve (RCIC-V-68) is not full open (0.5)
B. The LPCS pump will start and run (0.25) but will not inject (0.25) because there is no flowpath available.(0.5)
C. The RHR pump 2C will start and run (0.25) but will not inject (0.25) because there is no flowpath available.(0.5)
D. The RHR pump 2B will start and trip (0.25) and will not inject (0.25) because of the short duration of the start. (0.5)
(also accept RHR 2B will not start due to the suction valve interlocks,the supply breaker will shut and then reopen, there will be no injection due to no suction path and no pump running.)
REFERENCE WNP-2 System Vol III RCIC pg. 21 WNP-2 System Vol III RHR pg 19 and 20 WNP-2 System Vol III LPCS pg. 8
. . . . , J
6t__EL6NI_@Y@lEMS_QE@l@N2 _GQNIBQ61_86Q_lN@lBQMENI@llQN PAGE 28 ANSWERS -- WNP-2 -86/02/03-MILLER, L.
ANSWER 6.08 (3.00)
A. The flux estimator is used to provide a more stable flux feedback signal to the recirc flow control system than the actual APRM can provide to minimite the valve hunting.(also accept - to provide a more stable flux signal to minimize FCV wear due to valve hunting) (1.0)
B. 1. APRM flux signal greater than 110% of rated neutron flux (0.5)
- 2. The difference between the APRM and the estimated flux signals exceed +/- 5% of rated neutron flux. (0.5)
C. 1. An actual High Drywell Pressure signal. (1.68)
- 2. A High Drywell / Valve Motion Inhibit Relay Logic test switch signal.
- 3. Loss of any DC power supply.
- 4. Loss of 120 VAC power
- 5. Loss of control signal.
- 6. Pump motor overload / undervoltage
- 7. Oil temperature hot (150 degrees F)
- 8. Reservoir oil level low-low (20 gallons per segment)
.9. Pump discharge pressure loss (less than 1300 psig).
- 10. Servo valve control power loss.
(Any 5 of 10 0.2 pts each)
REFERENCE WNP-2 SYSTEMS VOL. I RECIRC FLOW CONTROL pg. 16,18 ANSWER 6.09 (3.00)
A. Individual loop 'A' flow would increase 10% (0.5)
(also accept increase loop 'A' flow to 100%)
individual loop 'B' flow would remain the same (0.5)
The error signal from the flow reference signal and the flux controller signal summer would be limited to 10% by the error li mi ter. (1. 0)
B. The flow control valve motion would stop when the valve opened the initial 10% . The position feedback would null the signal from the loop flow controller. (1.0)
REFERENCE WNP-2 SYSTEMS VOL.I RECIRCULATION FLOW CONTROL pg. 25 AND FIG.9 e
2
.hz__ELONI_SYSIEM@_DEgl@N2_GQNIBQ62_@NQ_lNQISyMENI@IlQN PAGE 29 ANSWERS -- WNP-2 -86/02/03-MILLER, L. l l ANSWER 6.10 (3.00)
- 1. MISV CLOSURE VALVE POSITION (0.3), greater than 6%(0.1) closed as sensed by limit switches on the valve (0.3), bypassed when the MODE. switch is not in RUN and RPV pressure is less than 1037 psig.(0.3)
- 2. TURBINE THROTTLE VALVE CLOSURE (0.3), greater than 5% closed (0.1) as sensed by fast acting switch on any throttle valve (0.3),
bypassed when power is less than 30% (as sensed by first stage pressure).(0.3) 3 .' TURBINE GOVERNOR VALVE FAST CLOSURE (0.3), less than 1250 psig on the EH oil (0.1) as sensed by a pressure switch (0.3), bypassed when power is less than 30% (as sensed by first stage pressure) (0. 3) .
REFERENCE
'WNP-2 TECHNICAL SPECIFICATIONS LASES RPS LSSS
-WNP-2 SYSTEMS VOL II RPS pg. 15-25 ANSWER 6.11 (1.00)
B. 200 MVAR REFERENCE, WNP-2 SYSTEMS VOL.IV MAIN GENERATOR FIG 9.
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PAGE 30 B89196gGIC86_CgNISQL ANSWERS -- WNP-2 -86/02/03-MILLER, L.
3 ANSWER 7.01 (2.00)
A. Prevents an undue stress on the vessel no:zles and bottom head region. (1.0) 4 B. ' Limits undue thermal stress on vessel (1.0)
REFERENCE WNP-2 TECHNICAL SPECIFICATIONS B3/4.4.1 pg B3/4 4-1 2
ANSWER 7.02 (2.50)
. A. Initiate suppression pool cooling or reduce suppression pool temperature to less than 90 degrees F.(0.5) 4 B. Stop RCIC testing (0.5) and initiate and restore average temperature to less than 90 degrees F (0.5) (within 24 hrs or be in HOT SHUTDOWN in the next 12 hrs and COLD SHUTDOWN within the following 24 hrs. NOT REQUIRED FOR FULL CREDIT) 1C. Depressurize the reactor pressure vessel (0.5) to less than 200 psig (0. 5) (wi thi n 12 hrs. NOT REQUIRED FOR FULL CREDIT)
I REFERENCE WNP-2 TECHNICAL SPECIFICATIONS, 3.6.2.1 pg 3/4 6-13 ANSWER 7.03 (2.00)
A. Whenever the reactor is in the startup or run mode and less than or equal to 20% rated thermal power.(0.25)
Verify control rod movement and compliance with the i prescribed control rod pattern by a second licensed l operator or other technically qualified member of the unit technical staff who is present at the reactor console.(0.75)
B. A pattern which results in the core being on a thermal hydraulic limit (i .e. , operating on a limiting value for APLHGR,LHGR or MCPR). (1.0)
REFERENCE WNP-2 TECHNICAL SPECIFICATIONS 3.1.4.1 WNP-2 TECHNICAL SPECIFICATIONS 1.19 i
_ - - _ - .__ _ - ._ _-~ _ _ _ _ _ - M
-Zr__889CEEMBEE_ _NQSUG(i_@ENQBd@b2_EdEBGENQY_@NQ. PAGE 31 86D1969G1G86_GQNI6QL ANSWERS -- WNP-2 -86/02/03-MILLER, L.
ANSWER 7.04 (2.50)
- a. Closed (0.5)
- 6. 2 (or 3) Open (0,5)
(also acceptable RHR A and B in suppression pool cooling)
- d. 76 - 120 psig > Suppression Chamber Pressure (0.5)
(any value between 76 and 120 psig is acceptable)
- e. R actor level > +648 inches above vessel =ero (Main Steam Line elevation) (0.5)
REFERENCE WNP-2 EMERGENCY PROCEDURES VOL.V , 5.3.5 ALTERNATE SHUTDOWN COOLING (CONTINGENCY)
ANSWER- 7.05 (1.50)
Because spraying the drywell may result in a drpressurization rate in the containment (drywell and supression chamber) which is beyond the capacity of the Rasctor Building-to-Suppression Chamber Vacuum Breakers,(0.5) resulting in negative containment pressures in excess of dazign, leading to failure of the primary containment.(1.0)
REFERENCE WNP-2 EPGs SEC. 08 pg.B.4-8
___ j
Zz__EEgCgpUggg_:_NOBdeLi_eENOBdBL1_EME60EUQy_8Up PAGE 32 EBR1969GIGOL_GONIERL ANSWERS -- WNP-2 -86/02/03-MILLER, L.
ANSWER 7.06 (3.00)
A.1. RWCU should be diverted to the condenser (0.5) before the differential between the return to the RPV and feedwater exceeds 165 degrees F.(0.25)
- 2. Cooldown and depressurire the RPV (0.5) and do not divert RWCU flow away from the RPV (0.5) unti'l the differential temperature is 165 degrees F or less (0.25).
B. The flow is returned to the RPV to repressurire the feed water piping to prevent excessive water hammer. (0.5)
REFERENCE WNP-2 PPM 3.3.1 Reactor Scram Recovery, pg 2 ANSWER 7.07 (1.50)
NO (0.5), Control rod movement is not permitted below 20%
rated thermal poirer except by scram, if the RSCS is not operable (1.0).
REFERENCE WNNP-2 VOL III GENERAL OPERATING PROCEDURES 3.2.1 NORMAL SHUTDOWN TO COLD SHUTDOWN DEVIATION.85-466 ANSWER 7.08 (1.00) 1.- The Reactor Trip and Recovery Report requires no corrective action ' prior to restart. (0.5)
- 2. Verbal concurrence is obtained from the Operations Manager or his designee to omit specified portions of the checklist.
(0.5)
REFERENCE' WNP-2 VOL.III GENERAL OPERATING PROCEDURES MINIMUM STARTUP CHECKLIST 3.1.4, pg.1
. . -- .~ - . - . _ - - . .. - _ . . - -. .. . - _.
Z:__E89GEQUBES_;_NQBM@63_@BNQSd@61_EdEBGENQY_@NQ. PAGE 33 BOD 196991GOL_GQNIBQL f
' ANSWERS.-- WNP-2 ~86/02/03-MILLER, L.
A 1
ANSWER 7.09 (3.00)
- 1. RCIC injection with steam rejected to the main condenser or j suppression pool.
- 2. Condensate pump injection via feedwater. lines with steam rejected to the condenser (or suppression pool) and then to CST's via filter demins (F.D.'s).
- 3. HPCS (f eed and bleed) suction from CST's with RPV steam rejected to the condenser (or suppression pool) and then to CST's via filter demins (F.D.'s).
I LPCS (f eed and bleed) suction on the suppression pool; steam 4.
to condenser.
i 5. RHR-C (feed and bleed) suction on suppression pool; steam to condenser.
i 6. Increase cooling water flow to RWCU system heat exchangers.
(Any 3 of the 7 Each method 1.0 pts.)
1 REFERENCE j WNP-2.VOL IV ABNORMAL CONDITION PROCEDURES 4.4.2.1 LOSS OF RHR SHUTDOWN COOLING MODE LOOPS pg 2 4 ANSWER 7.10 (3.00)
A. 1. c
, 2. a
- 3. b (3 9 0.33ea) (1.0)
B. Reactor / power is > 5% (0.5) or can not be determined and Suppression pool temperature > 110 deg F (0.5) (1.0)
C. 1. RCIC (0.5)
- 2. RWCU (0.5)
REFERENCE WNP-2 SYSTEMS VOL II RPS .
WNP-2 VOL V EMERGENCY PROCEDURES REACTOR POWER CONTROL 5.1.3 pg 2 WNP-2 SYSTEMS VOL I CRDH
Zz__P8QQEQQ8ES_ _UQBM861_8@UQBM863_Edg8QEUGy_8dp PAGE 34 BeQ196991GOL_GONIBQL ANSWERS -- WNP-2 -86/02/03-MILLER, L.
ANSWER 7.11 (3.00)
- 1. Manually scram the reactor.
- 2. Place the Reactor Mode Switch to Shutdown.
- 4. Verify that the APRM downscale lights are illuminated.
- 5. Trip the main generator and ensure that auxiliary power is being supplied from TR-S.
- 6. If the reactor recirc pumps are not being supplied from the LFMG, then transfer both recirc pumps to the LFMG.
'7. Ensure only two condensate booster and two condensate pumps are operating.
B. Ensure not more than two circulating water pumps are running.
- 9. Verify valve positions RFW-10 CONTROLLER IN AUTO,RFW-V-118, RFW-V-117A,RFW-V-117B OPEN. RFW-V-112A,AND RFW-V-112B CLOSED.
(ANY 6 OF 9 0.5 pts.EACH)
REFERENCE WNP-2 VOL.IV ABNORMAL CONDITIONS PROCEDURES PPM.4.12.1.1
a __0DMINISIB9IIVE_BBgCEgyBE@2_CQUp111gN32_8ND_LIMlI911985 PAGE 35 ANSWERS -- WNP-2 -86/02/03-MILLER, L.
ANSWER 8.01 (2.50)
A. Authorization for personnel to enter the drywell must be obtained from the Shift Manager and the Health Physics / Chemistry Manager prior to initiating this procedure. (Plant Manager's permission is required if the reactor is critical when personnel enter the drywell.)
B. The drywell must be deinerted prior to entry. (ensure oxygen readings > 19.5% on containment oxygen monitors)
C. Prior to entry into the Primary Containment with the reactor critical, reactor power level must be less than or equal to 5% and stable.
D.. The Transverse in Core Probes (TIPS) must be withdrawn from the core and tagged out of service.
E. The nitrogen inerting system must be tagged out of service.
F. An RWP must be initiated (by the Shift Manager).
G. The CAS or SAS shall be notified immediately prior to opening any hatch into the drywell and a security officer posted.
H. The Plant Manager, or his designee, shall provide a list of personnel authorized access to the Primary Containment.
(ANY 5 OF THE B O.5 PTS. EACH)
REFERENCE WNP-2 ADMINISTRATIVE PROCEDURES 1.9.3 PERSONNEL ENTRY TO PRIMARY CONTAINMENT PG. 1 AND 2 ANSWER 8.02 (1.50)
A. The action shall be approved by a licensed Senior Operator as a minimum. (0.5)
B. Operations Manager (0.5)
NRC (0.5) Operations Center 7
. B___eedlNISIEGIIVE_EEQCEQQEEgi_CQNQlliQNSt_@NQ_LidlI@IlgNS PAGE 36
' ANSWERS -- WNP-2 -86/02/03-MILLER, L.
REFERENCE 10 CFR 50.54 X AND Y , ADMINISTRATIVE PROCEDURES 1.2.3 pg.2;1.1.3 pg.3; 1.3.1 pg.2 ANSWER 8.03 (3.00)
A. DIRECT RADIATION >/= 2.5 MREM /HR (0.5)
AIRBORNE RADIOACTIVITY >/= 25% OF MPC (0,5)
CONTAMINATION >/= 1000 DPM/100CM2 BETA GAMMA OR
>/= 100 DPM/100CM2 ALPHA (0.5)
B. To determine the effects of the task on the overall i plant and the effect of other plant parameters on the task to be performed. (1.5)
REFERENCE WNP-2 HEALTH PHYSICS PROGRAM PPM.1.11.3 pg.9 4
ANSWER 8.04 (1.00)
The changes need to be initialed by the shift manager to
- signify his approval.
}
REFERENCE WNP-2 ADMINISTRATION PROCEDURE 1.3.8 PROCEDURE DEVIATION 84-1058 i
ANSWER' 8.05 (2.50) i A. 1. The worker's supervisor
- 2. The worker's department manager
- 3. The shift manager l- (3 9 0.33 pts. EACH)
B. When the component is Safety related or Fire Protection related. (0.5)
I REFERENCE WNP-2 EQUIPMENT CLEARANCE AND TAGGING PROCEDURE 1.3.8 pg 6 t
i I
I i
i .. h
H. - BDd1NISIB9112E_BBggEgyBES2_CgNplIigNS2_@Np_ Lid 11611gNS PAGE 37 ANSWERS -- WNP-2 -86/02/03-MILLER, L.
ANSWER 8.06 (1.00)
- 1. If the clearance order is a multiple clearance order, temporary lifting of the tags is not permitted.
.2. Only checkouts or tests that should take a short time (less than one hour) are allowed.
- 3. The checkout or test may not run through a shift change.
(ANY 2 OF 3 0.5 PTS. EACH)
REFERENCE WNP-2 EQUIPMENT CLEARANCE AND TAGGING PROCEDURE 1.3.8 pg 8 ANSWER 8.07 (1.00)
- 1. 'The cause of the reactor trip has not been determined and corrected. (0.5)
- 2. There are reportable occurrences, other than the reactor trip itself, associated with the reactor trip. (0.5)
REFERENCE WNP-2 REACTOR TRIP AND RECOVERY,pg 2
a___8DMINISIBOIIME_EBgCEggBES _CgNg111gNS 2 _@ND_
2 LIM 11@IlgNS PAGE 38 ANSWERS -- WNP-2 -86/02/03-MILLER, L.
ANSWER 8.08 (3.00)
A. 1. The procedure deviation shall not change the intent of an approved procedure.
- 2. The deviation should be marked-up in the appropriate text sections of the affected procedure.
- 3. If a deviation is too extensive to be easily understood or cannot be markedup on the applicable pages, a procedure revision is mandatory.
- 4. If more than one procedure is affected by a procedure deviation, each procedure shall have a seperate deviation.
- 5. Shall not alter any of the criteria items listed 1 through 5 on the Procedure Revision form.
6.- Approved by two members of Plant Management /
Supervisory Staff, at least one of whom holds a Senior Reactor Operator's license for the plant; both of whom are knowledgable in the areas affected.
- 7. Deviation is documented with the " Procedure Deviation form"
- 8. Reviewed by the POC
- 9. Approved by the Plant Manager within 14 calendar days of implementation.
(ANY 4 OF THE 9 0.5 PTS. EACH)
B. No, Providing it has been approved verbally by two members of Plant Management / Supervisory Staff (1.0)
REFERENCE WNP-2 USE OF PLANT PROCEDURES 1.2.3 pg 2 AND 3 WNP-2 TECHNICAL SPECIFICATIONS 6.8.3 4
1
Ez__8DdlNISIEBIlyE_EEQQEDUEE@2_QQNQlliQNgz_6NQ_LldlI611QNS PAGE 39 ANSWERS -- WNP-2 -86/02/03-MILLER, L.
ANSWER 8.09 (3.00)
A.. Operators are not to override the automatic actions of ECCS and other safety features, unless confirmed by at least two independent indications of a misoperation in the automatic mode or adequate core cooling is assured.
B. An operator may place a controller in the manual mode from the automatic mode whenever, in the judgement of the operator, continued automatic operation is undesirable.
f C. When a safety related motor operated valve has been manually seated or back-seated, the valve shall be declared inoperable until the motor operation can be demonstrated.
D. Instructions for aligning more than 2 valves or circuit breakers should be_ written on a Component Status Change Order and carried by the operator performing the change unless the operation is performed using the procedure or checklist.
( O.75 EACH)
REFERENCE WNP-2 STANDING ORDERS / NIGHT ORDERS, 1.3.1 ATT 1 pg 2-5 ANSWER 8.10 (3.00) 1 - Shift Manager - SRO (0.6) 1 - Control Room Supervisor - SRO (0.6) 2 - Reactor Operators - RO (0.6) 2 - Equipment Operators - none (0.6) 1 - Shift Technical Advisor - none (0.6)
REFERENCE WNP-2 TECHNICAL SPECIFICATIONS TABLE 6.2.2-1 pg 6-6
. ;9z__eDd1NigIB811ME_BBDCEgyBE@2_CQNpil19Ng2_6ND_ L idlI@IlgNS PAGE 40 i
ANSWERS -- WNP-2 -86/02/03-VILLER, L.
(
ANSWER 8.11 (2.00)
A. No, the instrument can not be relied on. (1.0)
+
- b. The instrument is unquailified due to enviromental conditions in the containment. (1.0)
]
REFERENCE
- WNP-2 PPM 1.3.1 pg. 2,7 L
! ANSWER 8.12 (1.50)
Yes,(0.5) the procedure change does involve an unreviewed safety concern.
Tha change decreases the margin of safety as defined in the basis for the Technical Specifications. (1.0)
REFERENCE
- 10CFR50.59 (A)(2) r WNP-2 TECHNICAL SPECIFICATIONS 2.2.1 pg. 2-4 1
i 1 4 L 4
4
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TEST CROSS REFERENCE PAGE 1 bOESTION VALUE REFERENCE 05.01 1.50 LZMOOOOOO1 05.02 2.00 LZMOOOOOO2 05.03 1.50 LZMOOOOOO3 05.04 2.00 LZMOOOOOO4 05.05 1.00 LZMOOOOOO5
.05.06 2.00 LZMOOOOOO6 05.07 1.50 LZMOOOOOO7 05.08 3.00 LZMOOOOOOB 05.09 2.50 LZMOOOOOO9 05.10 3.00 LZMOOOOO10 05.11 2.00 LZMOOOOO11 05.12 3.00 LZMOOOOO12 25.00 06.01 2.00 LZMOOOOO34 06.02 2.00 LZMOOOOO35 06.03 1.50 LZMOOOOO36 06.04 1.50 LZMOOOOO37 06.05 2.00 LZMOOOOO38 06.06 2.00 LZMOOOOO39 06.07 4.00 LZMOOOOO40 06.08 3.00 LZMOOOOO41 06.09 3.00 LZMOOOOO42 06.10 3.00 LZMOOOOO43 06.11 1.00 LZMOOOOO44 25.00 07.01 2.00 LZMOOOOO13 i
07.02 2.50 LZMOOOOO14 I
.07.03 2.00 LZMOOOOO15 07.04 2.50 LZMOOOOOl6 07.05 1.50 LZMOOOOO17 l 07.06 3.00 LZMOOOOO18 07.07 1.50 LZMOOOOO19 07.08 1.00 LZMOOOOO2O 07.09 3.00 LZMOOOOO21 07.10 3.00 LZMOOOOO22 07.11 3.00 LZMOOOOO23
- j. _--
25.00 08.01 2.50 LZMOOOOO24 08.02 1.50 LZMOOOOO25 08.03 3.00 LZMOOOOO26 08.04 1.00 LZMOOOOO27 08.05 2.50 LZMOOOOO28 08.06 1.00 LZMOOOOO29 C3.07 1.00 LZMOOOOO30 I
i E.
2 QUESTION VALUE REFERENCE
_=__ -- = --____ __________
- 08.08 3.00 LZMOOOOO31 08.09 3.00 LZMOOOOO32 08.10 3.00 LZMOOOOO33 08.11 2.00 LZMOOOOO45 08.12 1.50 LZMOOOOO46 25.00 100.00