ML20210V214
| ML20210V214 | |
| 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 | |
| ML20210V208 | List: |
| References | |
| 50-397-OL-86-01, 50-397-OL-86-1, NUDOCS 8606100234 | |
| Download: ML20210V214 (33) | |
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i U. S. NUCLEAR REGUIATORY COMMISSION REGION V Examination Report No. 50-397/0L-86-01, Rev. 1 Facility:
Washington Nuclear Plant No. 2 Docket No. 50-397 Examinations administered at Washington Nuclear Plant No. 2. Richland.
Washington from February 4-6, 1985.
Chief Examiner:
4[ /[J/
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R( J. Pate, Chief Date Signed Reactor Safety Branch Ek b Approved:
g.O.Elih, Chief Ddte $igned wperations Section Summary:
Examinations on November 6-8. 1984 Written and operating (oral and simulator) examinations were administered to
8606100234 860529 PDR ADOCK 05000397 V
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REPORT DETAILS 1.
Examiners
- Lenord Wiens, NRC L. Miller, NRC R. Pace, NRC M. King, EG&G J. Sherman, EG&G
- Chief Examiner 2.
Examination Review Meeting
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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 noted 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 l
l 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 C099 TENTS AND RESOLUTION
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REACTOR OPERATOR EXAMINATION GIVEN ON 2/4/86 1.
Facility Coment on Question 1.02 Also give credit for using doubling count rate the new Keff is half the distance to one i.e.
100 200 cos 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.
[k 3.
Facility Coment 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 cha,ged on the answer key to INCREASES for part
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4.
Facility Coment 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 higber power fuel bundles.
5.
Facility Coment on Question 1.07(b) i Also accept " resonance absorber build-up causing more resonance capture."
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Question did not state " list the isotope."
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Examiner Resolution f
The facility comment is correct, but the question asks for the primary effect.
PU-240 build-up is the correct answer.
6.
Facility Coment 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.
Examine. 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 Coment 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.
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.
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. SAMPLE PROBLEM:
(Continued) l{
k Solution:
//
['L The total NPSH on the recirculation pump is calculated by 1
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first determining the inlet pressure P; in Equation 7-43.
1 i
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i=Pdome + PH O -Plosses 2
where:
P 2
dome = pressure as measured in the steam dome (Ibf/ft )
P 2
H O = pressure due to the water column (Ibf/f t )
2 Plosses = 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 a change in units to Ibf/f t2). The density of the water is taken as the saturation density of 20-Btu /lb-subcooled water (47.3 lb/f t3) which con be found from o table of subcooled water properties. The irreversible' losses are o function of the square of the fluid velocity'oructhe effects of elbows, pipe fittings, and suction valve in the" recirculation pump suction line. It is normally 20 psia at rated conditions and decreases cs the square of the flow rate.
Equation 7-43 beccmes:
2 7
e00 lbf/in x 144 in' /f t" -
00 f t x 4 7.1 lbm/t t 1 x 32.2 f t/we' 32.J 7
lat ec 1
2
- 20 lbf/in.< ! M in /f t
?; : iM,000 lbf/f f 7-94 I
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C SAMPLE PROBLEM:
(Continued)
The saturation pressure is found in the following 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 subcoosed 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.
2 P = 875 M x 14410 = 126,000 M 2
in f,2 f,2 The NPSH is then:
(7-43) 9 NPSH = (P; - P ) x c
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0 9 NPSH = (144,000 - 126,000) M x 32.2 lbm-f t 2
2 ft lbf-sec f
lb 32.2 i 2 x 62.4 see ff NPSH = 18.000 lbf/ff 3
62.4 lbf/f t NPSH : 288 f t of H O 2
The cont ibution of the water column above the pump ccn be calcuic:e separately by first assuming no subcooiing and neglectin ; the suction line head toss.
7-95
C V+ 0' Y-VS&& '. ( k
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floA &/= + sd
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Mu!d ffp r
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 restriction?
To obtain a qualitative picture of the effect of core inlet
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orifices, first consider the BWR core without inlet orifices.
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The pressure drop across all the fuel assemblies is the some f
since, as we said, they share a common inlet and outlet plenum. Assume forther that all the bundles have the some
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flow resistance characteristics so that, at zero power and
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minimum recirculation system flow, all the fuel bundles have the some 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 f
an increase in flow in the warmer bundles. In addition, c.s 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 bunJte increases as does the two-phase flow friction multiplier 2
(See Figure 9-20). The result is 24 a large increase in flow resistance as quality increases. Since the channel pressure drop is controlled by the inlet and outlet plenums (i.e. for constant t P), equation 9-20 indicates that the flow through the fuel bundle will declease 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 eeeks the path of least resistance. This is, of course, undesireable.
Flow orifices are provided at the #vndle inlet to minimize the undesirecole ef fect of a quality intrease on bundle flow. The inlet orifice has the effect of proviaing a Inrger resistance to flo.v so that any additionc! flow resistance caused by two-phase f' v.v is accontnbly small in comparison.
There is a classic analogy to this effect which can be iliistrate! bv the simple electrical example shown in Figure 9-22.
9-51
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ES-202 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-ments, and pressure / temperature limits.
4 Questions on health physics and chemistry procedures should be determined on the basis of the facilities' type of health physics coverage.
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Extended definitions questions (e.g., 6-factor formula) should be 2
avoided.
6.
Questions on detailed system characteristics or instrumentation, su.n as annunciator logic or setpoints, 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 l(
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Other sources of questions are g
a.
standard questions and answers b.
Examination Question Bank c.
examinations on similar facilities c.
perscnal file of questions and answers 9.
2 rule of thumb is a.
approximately 55 to 70 responses for a 6-hour examination b.
a response that requires about 3 to 4 minutes 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 tne facility as much as practicable.
12.
" Discuss"-type questions should be avoided; questiens should be specific to elicit short precise answers.
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Practical realistic cuestians that relate to cperator.':nowledge and l
required coeratinc practice should be used.
14 Multipart cuestions snould be troken down into logical sequential l
parts.
The answer theet should snow points assigneo for subparts l
of answers.
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10.
Facility Coment 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-1pulation is required. This switch is covered in our training material but memorization of each switch position is not reouired 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.
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 Coment 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 or RCIC.
l 13.
Facility Coment on Question 2.08 l
Answer #1 "RCIC equipment area and/or pipe routing area high temp" should be accepted as 2 separate signals if so listed.
Examiner Resolution l
Agree with comment and will give credit for two separate signals, if so listed.
14.
Facility Coment on Question 2.09 I
I b.
The stop control for HPCS in the control room is a switch, not a l
push button.
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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 dif ferential pressures of 1160/1130/200 psig.
Initially, water from the condensate storage tank is used instead of injecting water from the scopression pool into the reactor, but no credit is taken in the safety analyses for the condensate storage tank water.
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With the HPCS system inoperable, acequate core cooling is assured by the 3.7 j0PERABILITYofthe.redundantanddiversifiedautomaticdepressurizationsystem
( v j and both the LPCS and LPCI systems.
In addition, the reactor core isolation 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 OPERABILITY of redundant and diversified low pressure core cooling systems.
The surveillance requirements provide adequate assurance that the HPCS system will be OPERABLE when required. Altnough 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 injection requires reactor shutdown. The pump discharge piping is maintained
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full to prevent water hammer damage and to provide cooling at the earliest fM moment.
D Ucon f ailure of the HPCS system to function properly after a small break loss-of-coolant accident, the automatic depressurization system (ADS) automa-tic 211y 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 conserva-tively required to be OPERABLE whenever reactor vessel pressure exceeds 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.
l ADS automatically controls seven selected safety / relief valves although the safety analysis only takes credit for six valves.
It is therefore i
appropriate to permit one valve to be out-of-service for up to 14 days without materially reducing system reliacility.
3/4.5.3 SU:cRESSION CuPSER Th.i suo,ression en2moer is recuired to be OPERASLE as part of the ECCS to ensure :na; a suf fi::ent suoply of water is available to tne HPCS, LPCS, and l
LPCI systems in t e went of a LOCA.
This limit on suopression cna: Der minimum water volume ensures :nat sufficient water is available to permit recirculation cooling flow to tne a re.
The CPE MEILITY of the suceression enamcer in OPERATIONAL CCND!T!;N 1. 2, er 3 is required by Specification 3.6.2.1.
WASHINGTON NUCLEAR - UN:T2 B 3/4 5-2 O
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c' 3/4.5 EMERGENCY CORE COOLING SYSTEMS 3/4.5.1 ECCS - OPERATING LIMITING CONDITION FOR OPERATICN I
i 3.5.1 ECCS divisions 1, 2, and 3 shall be OPERABLE with:
\\V a.
ECCS division 1 consisting of:
1.
The OPERABLE low pressure core spray (LPCS) system with a flow path capable of taking suction from the suppression chamber and transferring the water through the spray sparger to the rea:: tor vessel.
2.
The OPERABLE low pressure' coolant injection (LPCI) subsystem "A" of the RHR system with a flow patn capable of taking suction from the suppression chamber and transferring the water to the reactor vessel.
3.
b.
ECCS division 2 consisting of:
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1.
The OPERABLE 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 OPERABLE A05 valves.
ECCS division 3 consisting of the OPERA 8LE high pressure core spray c.
(HPCS) system with a flow path capable of taking suction front the suppression chamber and transferring the water througn the spray sparger to the reactor vessel.
APPLICABILITY: CPERATIONAL CONDITICNS 1, 2"#, and 3".
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t "ine A05 is not recuired to be CPERA2LE wnen reactor steam deme pressure is less than or equal to 123 psig.
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- See 5::ecial Test Exce::t ion 3.10.6.
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WASHINGTCN NUCLEAR - UNIT 2 3/4 5-1 e.e e
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EMERGENCY CCRE CCOLING SYSTEMS t
s LIMITING CCNDITION FOR OPERATION (Continuedi 1
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ACTION:! f
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For ECCS division 1, proviced that ECCS divisions 2 and 3 are CPERABLE:
>".s 1.
With the LPCS system incoerable, restore the inoperable LPCS l '3 I
,, - :,' i system to CPERASLE 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" ineperable, restore at least. the inoperable LPCI subsystem "A" or the inoperable LPCS systam to OPERABLE :tatus within 72 hours8.333333e-4 days <br />0.02 hours <br />1.190476e-4 weeks <br />2.7396e-5 months <br />.
4 Other4f se, 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 COLD SHUTCC%N within the following 24 hours2.777778e-4 days <br />0.00667 hours <br />3.968254e-5 weeks <br />9.132e-6 months <br />.
b.
For ECCS divtsion 2, provided that ECCS divisions 1 and 3 are OPERABLE:
l.
With either LPCI suesystem "B" or "C" incoerable, restore the
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incoerable LPCI suosystem "B" or "C" to OPERABLE status within 7 days.
2.
With both LPCI subsystems "B" and "C" incperable, restore at least the inocerable LPCI subsystem "B" or "C" to OPERABLE status within 72 hours8.333333e-4 days <br />0.02 hours <br />1.190476e-4 weeks <br />2.7396e-5 months <br />.
3.
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 COLD SHUTCCWN within the fo11cwing 24 hours2.777778e-4 days <br />0.00667 hours <br />3.968254e-5 weeks <br />9.132e-6 months <br />".
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For ECCS divisien 3, provided that ECCS divisions 1 and 2 and the e
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' RCIC system are OPERABLE:
-. r.
1)
With ECCS division 3 ineperable, restore the inceerable division to OPERABLE status within 14 days.
2)
Otherwise, be in at least MOT SHUTCC%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 COLD ShuTCC%N within the fo11 ewing 24 hours2.777778e-4 days <br />0.00667 hours <br />3.968254e-5 weeks <br />9.132e-6 months <br />.
c.
For EC:3 civisiens 1 and 2, previced that ECOS civision 3 is CPERABLE:
1)
With.?CI sucsystem "A" and either LPCI subsystem "S" or "C" trc:eracle, restore at least the inc e-3 Die LPCI subsystam "A" 7
or.ne ine eracie L?C: stesystem "S" er "C" to CPERAELE sta us s
.1 nin 72 nears.
"%nenever two or mere RHR su systems are inc=erable, if unable to attain COLO i
ShuTCC%N as recuirec Ov this ACT*CN, maintain reactor ::alant tam:eratare
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as icw as pra:tical Sy use of alternate heat removal metaccs.
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WASHINGTCN NUCLEAR - UNIT 2 3/4 5-2
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Examiner Resolution Comment noted, but this does not change the answer to the question.
15.
Facility Connent on Question 2.10 Answei #1 also accept "undervoltage" on associated bus.
a.
Examiner Resolution Facility comment is correct and answer annotated to accept undervoltage.
16.
Facility Connent on Question 2.11 RWCU pump no lor.jer 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 rc: pense shnuld also be accepted.
Examiner Resolution Com.nent 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 closure [0.25].
Due to facility comments, reference is changed to:
WNP-2 Systems, Volume I, Tab. 9, pp. 7-8.
17.
Facility Connent 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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If a high pressure develops in the piping downstream of the
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FCV, a high pressure setting on PS-14 will close the FCV at
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140 psig.
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- 6> 3... pp.
- 3. o Closure of either 'V-1 or V-4 'will cause FCV-33 to close. -
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'.'. 7 FCV-33 closes to prevent system depressurization which causes the hot water in tne system to flash to steam (the water flashing to stearr. and resulting water hammer when ficw is y *943.$,','
,g.r re-established could possibly damage the system piping or heat exchangers).
When V-1 and V-4 are both open, FCV-33 will automatically
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reopen to the position determined by its manual controller on v
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The following will cause a pump trip:
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Inlet isolation valves (V-1 or V-4) not full open r
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' Low system flow as sensed by FE 70 gpm
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D Filter Demineralizers
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1.
The filter-demineralizers each have a ficw control valve (air 9f/M w
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operated) which se'nses effluent flow and maintains a constant
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' flow rate through the vessel for varying pressure drops across v
t e.
the fil ter.
The vessel dP should range from a low of.1 psid 4.
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to a high of 20-25 psid (Figure 3).
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2.
The flow rec e vr control stations 74A and 748 start and stop their respecF te vessel hold pump.
The hold pump will start at less tna.1 etal to 100 gpm and stop at greater than 100 gpm.
(FRCS
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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 b.
Should also accept - possible RCIC overspeed due attempting max, 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) are separate, but are arranged in parallel such that a break in the D/P
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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 Consnent: This section (03) was very well written.
Questions were clear and concise, answers were brief but complete.
Every question was something an operator should know!
19.
Facility Consnent 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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Iurbine Ilirottle Valve-Closure 5 5% closed 5 7% closed 2.
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Facility Coment on Question 3.05
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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.
22.
Facility Coment on Question 3.10
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b.
We don't have a "thumbwheel 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 Coment 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 misleading 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 Coment on Question 4.04 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.
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Examiner Resolution Agree with facility comment and 5 of the 11 steps required for full credit.
25.
Facility Comnent on Question 4.05 Should give full credit even if... " including the public.." is not included, since injury to personnel is all inclusive - includes any persons public or employees.
Examiner Resolution Agree with comment.
The words " including the public" not required for full credit.
26.
Facility Conment 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.
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Examine Resolution Agree with facility comments.
a.
Full credit will be given for " ensures adequate core flow coast-down following a LOCA".
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.: PPf1 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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4 ABNCRMAL C ITION PROCEDURES sacron 4.601.Al ANNUNCIATCR RESPONSE. P601 ANNUNCIATOR Al T IT b5
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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 Pump (HPCS-P-3) rurning (P601).
NOTE: Tre following step is designed to prevent pmp run-out and possi-ble loss if LOCA shculd occur when HPCS-P-1 is running in the test mcde.
3.
If the Water Leg Pump fails w en hPCS is required to be operable, start and run HPCS-P-1 in a test mode while maintaining approximately 1,200 psig discharge pressure; maintain these concitions until tre Water Leg Pmp is made operational.
If during a power interruptien, the HPCS Water Leg Punp Discharge Pres-l 4
sure Lcw ala:m is received and acecuate core cooling is assured, hold l
HPCS-P-1 centrol switch in the STCP position or ctrerwise prevent pumo start until tre system can be filled and vented.
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SAFETY RELATED CONTROLLED COPY
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Veners Manual _
GEK-71334 O.
FSAR Sections 6.3, 7.3, 8.3 I
2.4.4.3 Prerecuisites The Reactor Building Heating, Ventilation and Air Ccnditiening A.
System in operaticn to support HPCS System Cperaticn.
WCS service water system available to support HPCS diesel and WCS B.
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system operaticns.
The ccncensate storage tanks have the required amount of water to C.
support HPCS cperaticn (7'7" each tark or 13'3" in ene tark minir:un r-)
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per Tecrnical Scecificaticn 3.5.3).
t Must have at least minimun fuel supply (30,CCO gallens) cn site for D.
WCS diesel.
The suppressicn pcol level norr al (31 ft. 2 in. to 30 ft.10 in.).
E.
The HFCS pmo should not be started when the "HPCS WATER LEG PtNP f*.
The water leg pmo is designed to DISCH FRESS LCW" alarm is lit.
remain in service thrcughcut system cperation and during stancby
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Refer to Technical Specificatien 3.5.1.
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2.4.4.4 Limitaticns A.
Cbserve RWP requiratents per PPM 11.2.8.1.
The HPCS System shall not be removed frcm service anytim it is See Technical 8.
required to te coerable by Technical Scecifications.
Seecificaticn 3.5.1.
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k If during or folicwing a pcwer interrt.ption the HPCS WATER LEG FWP
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OISCH FRESS LCW alarm is received and adecuate core cooling 'is
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assured, hold the pmo centrol switch in the STCP or ctherwise
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prevent pro start tntil the system ccn be filled and vented.
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y The FPCS System shall te raintained full anytime the system is re-1 D.
If eater leg st m FPCS-P-3 f ails, start and Quired tc ce ccerable.
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ccerate 'FCS D-1 cn recirculaticn to the CST Until ccrrective action
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The ccccensate storage and sucoly system ay cnly be used to flush E.
It snall never_ te used to keep the system charged the FPCS Iystc1.
or 1. ired t.o to en mattencec system.
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Since the question did not state the cause of the water leg pump failure,
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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 Comment 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 -
g Also our Supp. Pool Chem, results show that it is well within j
the Chem. Specs. for the RPV.
k b.
Again - no reasol gnen, however, to prevent overheating of motor windings shoulc be given full credit.
Examiner Resolution 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 Consnent 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 147..
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 the list.
Point value is 20% of section, thus meeting requirements of ES-107.
Will accept other implicit items included as part of the items listed.
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WPN-2 Facility Review C
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.*.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 SDM.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 shutdcwn 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 Unabl'e 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)
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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 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 FK'LC 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 FWI.C 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 responte.
In Section 6.0 question #5 is the same as 5.8b, this double,ieopardy is not allowed by the examiners standard.
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C 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 Nocommentonanswer,however-questionstipulatesamode(flexauto) 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 C
similar response such as the " control rod" moves past the "00" position should deserve full credit as well.
6.05 Part 11 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 respond to the increase in level part d could also be acceptable (RPV
- level until turbine trip). Also this question is also asked in question 5.8c (FW response due to HPCS initial) Double 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 Ar wer Part d should also accept RHR pump 2b will not start
- due to suction valve interclock - no suction path, no injection 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 L
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C 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 do not anticipate other problems. Should accept any scram signal -
with justification.
Question Assumes only 3, but gives 4 as answer 6.11 General ccmment question #5: 6.02 (2 pts) 6.08 (3 pts) and 6.09 (3 pts) all referred to ti.e 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
[k 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 hour2.777778e-4 days <br />0.00667 hours <br />3.968254e-5 weeks <br />9.132e-6 months <br />) action statement.
C.
Same comment as "B" above.
7.04 General Comment Just because both loops of S/D cooling are INOP., per 5.3.5.
Tech Spec 3.4.9.1 or 3.4.9.2 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 pressere 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 "ccunt-off" for not listing RHR A or B as injecting into RPV.
d.
PPM 5.3.5 is used in accident conditions where cooldown is reouired 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 memorization of a limitation not an
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immediate action on a procedure which is not an abnormal or an
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emergency procedure.
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, 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, pcrt 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) 8.04 This answer requires memorization of an administrative precedure. 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 perform his test.
If you want the SR0 to know the restrictions placed on " Temporarily lifting danger tags," Lhen ask it that way! The PPM 1.3.8 states the S. M. authorizes the temporary lifting at taas, not systen 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."
Recommend deletion from exam.
(see attached PPM 1.3.1).
Ouestion 5.03 NRC Pesolution Point distribution in the answer key was corrected Question 5.04 NRC Resolution The question asked " Describe how the reactor power is increased with increasing recire flowrate? Core void content and core net reactivity were to be included in the discussion and not to have the discussion limited to core void content and core net reactivity. The Recirc Flow Control System Manual discussion includes boiling boundary; therefore, it should be included in the answer.
Question 5.06 NRC Resolution The question specified that the shutdown margin was measured ten hours following a scram from full power which would be during peak xenon. Any assumptions should be written with the answer and will be considered in the grading of this question.
PPM 7.4.1.1.2 was not included with reference material for exam or as reference material supporting coninents of not conducting shutdown margin test until after 50 hours5.787037e-4 days <br />0.0139 hours <br />8.267196e-5 weeks <br />1.9025e-5 months <br /> after shutdown or for temperature correction.
Question 5.07 NRC Resolution To answer the question requires the application of the principles in Section 7 of the reference. The last sentence in the answer is to complete the reasoning developed in the first sentence.
Question 5.08 NRC Resolutions The question asks for a description of the plant response for turbine loadi reactor water level and feedwater flow. Stating that there are " reactor power and pressure perturbation, reactor vessel level perturbation" is not adequate. The question states ASSUME N0 SCRAM occurs. The answer key was modified to accept turbine load decrease due to decrease in reactor pressure caused by cooler water spraying in upper plenum. Any stated assumptions will be considered with the answer.
Question 5.11 NRC Resolution The question states that either indicate on attached figure or discuss allowing the cardidate to choose either nethod. Answer requires discussion which only requires understanding of principle, not memorization.
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Question 5.12 NRC Resolution Facility answer is the correlary of answer key answer and will be accepted.
Question 6.02 NRC Resolution Coment noted. The mode of operation is discussed in-depth in plant procedures and system manual.
Question 6.03 NRC Resolution Facility answer is the correlary of the answer key answer and will be accepted.
Question 6.04 NRC Resolution The expression " control rod" moves past the "oo" position is an acceptable answer comparable to the answer key answer.
I(k Question 6.05 NRC Resolution Questions 6.05 and 5.08c are not double jeor,ardy questions because each question has different in'.tial conditions and stated assumptions and the answer for question 6.05 is not dependent on the answer to question 5.08c.
The system response is analyzed in WNP-2 FSAR 15.5.5.1.
Inadvertent HPCS startup and is examined in accordance with ES 402.A.2, which states
" Questions are asked about design' intent, construction, operation etc."
l Question 6.06 NRC Resolution Comment incorporated in answer key.
Question 6.07 NRC Resolution The answer key has been modified to also accept in Part d, that RHR pump 2 B will not start due to the suction valve interlocks, the electrical breaker will close then reopen and there will be no injection because no suctior path or pump running.
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,. Question 6.08 NRC Resolution Comment incorporated into answer key Question 6.09 NRC Resolution The mode of operation is discussed in the system manual and addressed in the plant procedures in depth.
Question 6.10 NRC Resolution WNP-2 Technical Specifications and WNP-2 systems manual for reactor protection system set points discuss the purpose for the three scram set point as anticipatory. MSIV closure, Turbine governor valve fast closure (turbine control valve fast closure - technical specifications) and turbine throttle valve closure (turbine stop valve - closure Technical specifications) each anticipate the pressure and flux transients which will occur. The answer key has been modified to remove the fourth answer.
Question 6.11 NRC Resolution Answer key changed to reflect correct units.
General comment Question 6.02's main emphasis is on the response of the operation of the DEH system.
The mode of operation of the recire flow control system is included for inte/ grated plant cperation.
Question 6.08 has two points of emphasis, one on the recire flow control's flux estimator. The other on the hydraulic pcwer unit.
l Question 6.09 main emphasis is system operation and plant response.
l With the diversity. of question topics as indicated and importance of each t
addressed area The point valve and question are considered appropriate.
y Question 7.02 NRC Resolution The facility answer for Part 'A' is the correlary of the answer key answer and will be accepted. The portion of the Technical Specification stating time requirements (i.e., within 24 hcurs or be in hot shutdown in next 12 hours1.388889e-4 days <br />0.00333 hours <br />1.984127e-5 weeks <br />4.566e-6 months <br /> and in Cold Shutdown within the following 24 hours2.777778e-4 days <br />0.00667 hours <br />3.968254e-5 weeks <br />9.132e-6 months <br />) is indicated as not required for full credit.
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( Question 7.04 NRC Resolution The question makes a statement that alternate shutdown cooling has been established to remove decay beat and continue the cool down. The answer key has been modified to also accept for Part 'C' that 'A' and 'B' RHR pumps may be in suppression pool coolirg. The answer key has being rrodified to accept for Part 'D' a value for reactor pressure greater than containment pressure that will indicate an understanding of why there are limits established.
Acceptable vales between 76 to 120 psigs.
Question 7.06 10 CFR 55.21 and 55.22 state the written exams will include standard and emergency operating procedures for the facility and plant. Although this procedure is included with the General Operating Procedures it is recognized that this procedures will only be used following a condition requiring a manual scram or an automatic scram.
ES 202 B 4 (also applicable to SR0's) - states ---- be able to explain reasons', cautica's, and limitations of normal operating procedures.
Question 7.08 NRC Resolution PPM 3.1.4 minimum startup checklist Part 3.1.4.2 Discussion states:
A.
The Minimum Startup Checklist is to be used for plant startups following outages or shutdowns that have not been extensively disruptive to the normal alignment of systems.
It is to be used only where the active items during the outage were easily tracked and controlled by the Shift Staff.
For outages which have been disruptive to the normal alignment of systems, or where the amount of work items could not be easily tracked and controlled by the Shift Staff, PPM 3.1.1, Master Startup Checklist shall be used.
B.
This minimum startup checklist need not be completed in full if:
1.
The Reactor Trip and Recovery Report requires no corrective action prior to restart and, 2.
Verbal concurrence is obtained from the Operations Panager or his designee to omit specified portions of the checklist.
( Question 7.09 This question and answer were developed from the reference material which was provided by the facility.
It is the responsibility of the facility to provide current references to the examiners to ensure that questions and answers are correct. The answer key has been modified to also accept increaselcooling to RWCU system heat exchanges and operate a CRD pump and reject via RWCU.
Question 7.10 Facility Comment NRC Resolution The question states that a number of control rods are immovable, no reason is clearly identified or asked for. The statement is there to set up the question on the criteria for standby liquid control initiation.
Entry into PPM 5.1.3 Reactor Power Control 1srequired.klfany condition which requires a reactor scram and reactor power remains above 5% on the APRMs or cannot be determined.
In addition to the entry conditions,if suppression pool temperature reaches 110 F boron injection is required. Boron injection is dependent on both power level >5% (or cannot be determined) and suppression pool temperature >110 F.
Question 8.03 Facility Comment NRC Resolution The question asked for the radiological limits that require the use of a RWP.
This restricts the answer to the 3 conditions specified on the answer key.
Question 8.04 Facility Comment NRC Resolution This question tests a candidate on general knowledge of a normal administrative procedure allowed ES 402.A.4.
Question 8.05 NRC Resolution Fedundant verification is required when tagging either safety related components (systems) or fire protection components (systems) as stated in PPM 1.3.8 Equipment Clearance and taggina procedure. Both Safety related and fire protection are required for full credit.
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I(f Question 8.06 NRC Resolution The shift manager authorizes the equipment checkout by authorizing the temporary lifting of the clearance tags. Two of the three conditions are test conditions, not tagging conditions.
Question 8.08 NRC Resolution Answer 7 of this question will be considered as three individual parts.
Examinee must answer any 4 of 9 for full credit.
Question 8.09 Facility comment NRC Resolution This question and answer were developed from the reference material which was provided by the facility.
It is the responsibility of the facility to provide current references to the examiner to ensure that questions and answers are correct. Answer key has been modified to accept the new referenced answer.
Question 8.11 ES 402 A.4.
Specifies that this category contains questions on administrative, procedure, and regulatory item that affect safe operation of this facility, A standing order restricting use of reactor water level instrumentation and ECCs flow indication during a LOCA clearly falls within this area.
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