ML20195D544
| ML20195D544 | |
| Person / Time | |
|---|---|
| Site: | Wolf Creek  |
| Issue date: | 06/15/1988 |
| From: | Mccrory S, Pellet J NRC OFFICE OF INSPECTION & ENFORCEMENT (IE REGION IV) |
| To: | |
| Shared Package | |
| ML20195D518 | List: |
| References | |
| 50-482-OL88-01, 50-482-OL88-1, NUDOCS 8806230142 | |
| Download: ML20195D544 (72) | |
Text
- ___ - _______ _ _
APPENDIX U.S. NUCLEAR REGULATORY COMMISSION REGION IV
.1 Operator Licensing Exam Report:
50-482/0L88 Operating License: NPF-42 l
Docket No: 482 Licensee: Wolf Creek Nuclear Operating Corporation (WCNOC)
P.O.' Box 411 Burlington, Kansas 66839 Facility Name: Wolf Creek Generating Station (WCGS)
Examination at: WCGS, Burlington, Kansas Chief Examiner:
e/
6 /> ((
p L7iMcCrory; Lead Examing, Operator D(te /
Licensing Section, Division of Reactor Safety Approved by:
d-M d[
[.f. Pellet, Chief,~0peratdrLicensing Date Sectilon, Division of Reactor Safety Summary NRC Administered Examinations Conducted During the Week of April 13, 1988 I
(Report 50-482c0L88-01)
NRC administered examinations to two (2) applicants. Both applicants passed I
all portions of the examination and have been issued the appropriate license.
8806230142 880615 PDR ADOCK 05000482 V
DCD DETAILS 1..
Pe, sons Examined SR0 R0 Total
-License Examinations:
Pass -
0 2
2 Fail -
0 0
0 2.
Examiners S. L. McCrory, Chief Examiner 3.
Examination Report Performance results for individual examinees are not included in this report as it will be placed in the NRC Public Document Room and these results are not subject to public disclosure, a.
Examination Review Comment / Resolution In general, editorial comments or changes made'during the examination, or subsequent grading reviews, are not addressed by this resolution section. This sectior, reflects resolution of substantive coments made by WCGS.. The enly coments addressed in this section
.are those which were not acce9ted for incorporation into the examination and/or answer kej, Those coments accepted are incorporated into the master examination key which is included in this report. Coments may be paraphrased for brevity. The full text of the concents is attached.
(2.08)
Normally, the #2 seal injection is not taught as a path for the #1 seal leakoff since the majority goes to the seal water HS or PRT.
(No recommendation made.)
Response
The point distribution in the answer was assigned to account for the relative importance of the flow paths.
The fact that this flow path is not covered in training lectures does not invalidate it from being tested.
(2.12)
CRVIS can also be manually actuated.
(No recomendation made.)
Response
Reject.
The question asked for conditions which lead to actuation. Manual actuation is considered an action, not a condition.
(2.13)
If applicant interprets item g to be the seal water
' return containment isolation valves, then "close" is the correct response.
(No recomendation made.)
Response
Reject.
The wording was determined to be sufficiently clear to indicate that a single valve other than the containment isolation valves.
(3.01)
C will generate both - Hi 1 will generate SI - Hi 2 will generate main steam line isolation.
Response
C was credited only if the dual condition nature indicated above was given.
(4.08)
"Removing RHR from service" may have been given as an item related to the mode change.
(No recommendation made.)
Response
This is not considered safety significant in the context of the question since the OPERABILITY status of the system is not affected, b.
Site Visit Sumary (1) At the end of the written examination administration, the facility licensee was provided a copy of the examination and answer key for the purpose of comenting en the examination content validity.
It had been previously explained to the facility licensee that regional policy is to have examination results finalized within 30 days. Thus, a timely response is desired to attain this goal.
j (2) No exit meeting was held with the facility licensee since the examination visit was only for retake written examinations and no results could be finalized until grading was completed, c.
Generic Coments Several coments provided in the examination review did not make recommendations as to what to do with the information provided in the coment.
Specific recommendations should accompany comments unless they are for future consideration prior to reuse of the question.
d.
Master Examination and Answer Key Master copies of the WCGS license examinations and answer keys are attached.
The facility licensee comrrents which have been accepted are incorporated into the answer key.
e.
Facility Examination Review Comments The facility licensee coments regarding the written examination are attached.
Those comn.ents which were not acceptable for incorporation into the examination answer key have been addressed in the resolution section of this report.
U.
S. NUCLEAR REGULATORY COMMIS$10N REACTOR OPERATOR LICENSE EXAMINATION FACILITY:
_WOLE_CREEE-REACTOR TYPE:
_EWR=WECd DATE ADMINISTERED _88ZQdZ11
-- =
EXAMINER:
_UCCBGBY._S. _
CANDIDATE:
INSIEUCIlOUS_IO_CeUDIDGIE1 Use separate paper for the answers.
Write answers on one side only.
Staple question sheet on top of the answer sheets.
Points for each question are indicated in parentheses after the question.
The passing grade requires at least 70% in each category and a final grade of at least 80%.
Examination papers will be picked up six (6) hours after the examination starts.
% OF CATEGORY
% OF CANDIDATE'S CATEGORY
__UALUE_ _IDIaL
___SCDRF
_VeLUE__
__CGIEDDRY
_25.00__ _25.00 1.
PRINCIPLES OF NUCLEAR POWER PLANT OPERATION. THERMODYNAMICS.
HEAT TRANSFER AND FLUID FLOW
_2$.QQ__ _25.QQ
________ 2.
PLANT DESIGN INCLUD*NG SAFETY AND EMERGENCY SYSTEMS
_25.00__ _25.00 3.
INSTRUMENTS AND CONTROLS
_25 QQ__ _25.QQ
_______,_4.
PROCEDURES - NORMAL. ABNORMAL.
EMERGENCY AND RADIOLOG1 CAL CONTROL 10Q.Q0__
Totals Final Grade All work done on this examination i s my own.
I have neither oiven nor received aid.
=
Candidate's Signature
1 __EBINDIELES_DE_UUCLEeB_EDWER_ELeUI_DEEEeIIDU.
PAGE 2
IBEBUDDXUedlDS _UEeI_IBeOSEER_edD_ELUID_ELDW
' QUESTION 1.01 (1.50)
Expl ai n HOW and WHY rod wor th varies:
a.
with radial position in the core.
b.
If another rod is inserted adjacent to it.
c.
if the moderator temperature changes (neglect boron effects).
ANSWER 1.01 (1.50) a.
A control rod's worth is proportional to the relative flux it sees.
Since flux density drops off as you move away from the core centerline so does rod worth.
(0.50) b.
Placing a second rod adj acen t to the first will depress the flux in that area and decrease the worth of the first rod.
(0.5) c.
The rod worth will increase as temperature increases.
This is because the neutron migration length increases as temperature increases, al l ow i n g the control rod to see more neutrons.
Since the moderator is less dense, a neutron will travel farther while slowing down and diffusing through the core increasing chances of encountering'a control rod.(0,5)
REFERENCE WCGS LP 11 306 33. EO B.9 192005K105
...(KA'S)
QUESTION 1.02 (1.00)
Exp?ain why t h e mass f l ow r a t e in the secondary system is substantially lower than that in the primary system.
ANSWER 1.02 (1.00)
The secondary side of the steam generator undergoes a phase change.
Large amounts of energy are stored in the latent heat of vaporization.
Therefore less mass of water is needed to store the same thermal energy.
REFERENCE WCGS LO 12 311 47 EO B.2 193003K104
...(KA'S)
(***** CATEGORY 01 CONTINUED ON NEXT PAGE
- )
1.__ERIUC1ELES_DE_NUCLEaE_EDWER_ELauI_DEEEaIlOU.
PAGE 3
..IBElit!ODYuatd1CS._B EaI_IEaNS EER_eUD_ELU1D_ ELOW QUESTION 1.03 (2.50) a.
-Def ine di f f erenti al boron worth'
- ( 0. 5 )'
b.
Expl ai n HOW and WHY the following changes affect differential boron worths (2.0) 1.
Boron concentration-INCREASES.
2.
Moderator temperature INCREASES.
3.
Fission product concentration INCREASES.
4.
Core age !NCREASES.
ANSWER 1.03 (2.50) a.
Change in reactivity due to a change in boron concentration.
(0.5)
(pcm/ ppm [Bl>
b.
1.
Differential boron worth-(DBW) decreases (0.2) because the boron atoms are competing with each other for neutrons-(0.3).
-2.
DBW decreases (0.2) as moderator density decreasing (moves boron atoms farther apart) (0.15) decreasing neutron capture probability in boron atems (0.15)' OR Fewer B atoms in core for=
same ppm concentration (0.3).
J 3.
DBW decreases-(0.2) because poisons are competing with bcron a t o'.as. (0.3) 4.
DBW increases (0.2) due to boron depleteen (0.3).
REFERENCE
192007K105
...(KA'S) a 4
(***** CATEGORY 01 CONTINUED ON NEXT PAGE *****)
(
=
312__ER1NCIELES_DEiUUCLEeR_EDWER_ELeNI_DEEReIlDU.
PAGE 4
IBERtdODYtW41CS._EEeI_IEoblSEER_eUD_ELUID_ELOW
_ 2.00)
(
QUESTION-1.04-0.-
For an operator tak i ng da ta f or - a 1/M pl ot,- how wi l l : the proximity to criticality affect the time elapsed before a, stable count rate can be obtained after: withdrawing rods?
(0.5) 6.
.How will the - ini t i al count rate affect the count rate at criticality?
(0,5) c.
If.the speed of the control rods were to somehow increase. What would be the effect be on:
1.
Rod height at criticality?
(0,5) 2.
Count rate at criticality?
(0.5) b 6
. ANSWER 1.04 (2.00) c.
The closer to criticality, the longer time required to reach a stable cou'nt rate. (0.5) b.
A higher initial count rate will resuit in a higher count rate at criticality. (0.5) c.
1.
Critical roc height fs not affected. (0.5) 2.
Critical count rate will be l owe r.
(0.5)
~ REFERENCE WCGS LO 11 306 41, EO B.7 WCGS Reactor Theory Re v i ew',
P.
21 192003K101
...(KA'S)
QUESTION 1.05 (1.00)
E.
Explain the term "isothermal Temperature Coefficient" (!TC).
(0.5) b.
WHEN and WHY is the ITC of !nterest c.r concern to the operator?
(0.5) l
(
i l-
l-- BRINElELES DE UUCLEeB BDWER ELeNI DEEBeILDN.
PAGE 5
IBERUDDYuedlCS._BEeI IRANSEER_eND ELUID ELDW ANSWER 1.05 (1.00) a.
The ITC is the reactivity change associated with a unit change in the temperature of the moderator, clad and the fuel.
(0.5) b.
During start up because core and moderator are heated uniformly. (0.5)
REFERENCE WCGS LP 11 306 24 EO B.22 192004K101
...(KA'S)
GUESTION 1.06 (2.00) 8.
During natural circulation cooldown, you notice pressurizer level increase after the initiation of pressurizer spray. Explain what is occurring.
(1.0) b.
Assume a small LOCA r e su l t s f r om the rupture of a pressurizer level transmitter sensing line. Compare the severity of the accident if the ruptured line is the upper (reference) or l owe r (variable) sensing line and explain wh>-
there is a difference.
(1.0)
ANSWER 1.06 (2.00) a.
Voids are forming elsewhere in the primary system (0.5) forcing water into the pressurizer as spray flow condenses the vapor space (0,5).
5.
The lower line rupture is more severe (0.5) as the mo.ss loss of water is greater than the mass loss of steam. (0.5)
REFERENCE WCGS LO 12 311 50 EO B.4, LO 12 311 35 EO B.7 193003K102 193004K115
...(KA'S)
(***** CATEGORY 01 CONTINUED ON NEXT PAGE *****)
F
- 11__ER1NCIEL E S_ DE_bluCLEaR_EDWER_ ELaNI_ D EERGIloti PAGE A
IBERtdODYNMICS _8EaI_IRaNSEER_auD_ELU1D_ELOW l
QUESTION-1.07 (2.00)
C.
Why will' observing-local power density limits ensure ~ that the fuel temperatureD1imit is not exceeded?
(0.5) b.
~What limit ensures that' the fuel clad temperature Ilmit (2200 o*g.F)-
is.not exceeded when :f uel temperature is above 2200 deg.F?
(0.5)~
c.
Why does.the fuel rod surface temperature tend to. peak towards th'e top of the core'rather than at the location of peak heat flux?
( 1. 0 ) -
ANSWER 1.07
( 2.' 0 0 )
a.
Local power density is directly proportional to heat flux which determines fuel
_t emper a t ure.
As long as LPD is.kept sufficiently low fuel-temperatures will not approach limits.
(CONCEPT) (0.5) b.
DNB (DNBF.)-or Hot Channel Factors (Fq, Fdh) (accept either) (0,5) c.
' Fuel surface. temperature is a function of heat flux and moderator temperature. (0.5) Moderature temperature is higher at the top of the core. (0.5)
REFERENCE.
WCGS LO 12 311 28, EO B.11, 12'
.192005K112 193008K105
...(KA'S)
QUESTION 1.08 (2.00)
O.
How do each of the f ol l owi ng parame ters change (Increase, decrease or no change) if one main steam isolation valve closes wi th the plant at 50*. load.
Assume all controls are in automatic and no trip occurs.
/
1.
Af f ected loop steam generator icvel (INITIAL change only) 2.
Affected loop steam generator pressure 3.
Unaffected loop cold leg temperature (1.5) b.
Which of the reactor protection system signals could be expected to cause a reactor trip? (If more than one, list the one that would reach the trip point first.)
(0.5) t
g L_EElblC1ELES_OE_hluCLEeE_EOWER_ELeblI_OREEaIlOble PAGE 7
IBEEdQDYNAM1CS._BEaI_IEablSEER_eUD_ELUID_ELOW
+
- ANSWER 1.08 (2.00)
.a.
'1.-
Decrease 2.
Increase 3.
Decrease (0.5 each) 6.
Lo-Lo S/G~ Level (0.5)
REFERENCE WCGS LO 12 311 23. EO B.10, LO 12 311 46, EO B.7 192008K121
...(KA'S)
QUESTION 1.09
(.50)
Both-Pu-239.and Pu-240 concentrations increase over core life.
Which of The f ollowing statements concerning the effects of these increases is correct?
(0.5) a.
The buildup of Pu-240 increases the average delayed neutron fraction, b.
The buildup of Pu-239 decreases the core Reproduction factor.
c..
The buildup 24 Pu-239 causes the Moderator coefficient to become more
-necative.
- d..
.The buildup of Pu-240 causes the Doppler coefficient to become more negative.
' ANSWEl<
1.09
(.50) d.
(0.5)
REFERENCE WCGS LO 11 306 24, EO B.13, 15 19200K107
...(KA'S) l l
l
[
l
(***** CATEGORY 01 CONTINUED ON NEXT PAGE
- )
I
1 __ERiblC1ELES_DE_N U C LEAR _20W ER_ ELablI_DE EEGIlOtl PAGE
.8 IBERMODYNAMICS._BEeI_IEablSEER_ablD_ELU1D_ELOW QUESTION 1.10 (1.50)
How wi l l ' the -f ol l owi ng af f ec t the Moderator Temperature Coefficient?
LBRIEFLY EXPLAIN your answer.
a.
The charging-pump suction inadvertently switches to the Refueling-Water Storage Tank.
(0.5) 6.
The core ages from BOL to EOL.
(0.5) c.
The RCS i s cool ed down f rom 550 F to 450 F.
(0.5)
ANSWER 1.10 (1.50) a.
LESS NEGATIVE (0.2) More boron to leave core area per degree temperature change.-(0.3) (Or equivalent answer) 6.'
MORE NEGATIVE (0.2) Less boron, opposite result as above. (0.3) c.
.LESS NEGATIVE (0.2) Water density changes are less as temperature is reduced. (0.3) t REFERENCE WCGS LO 11 306 24, EO B.6 192004K106
...<KA's)
QUESTION 1.11 (2.00)
How would the Estimated Critical Position (ECP) calculated for a startup to be performed 4 hours4.62963e-5 days <br />0.00111 hours <br />6.613757e-6 weeks <br />1.522e-6 months <br /> after a trip from 100% power differ from the Actual Critical Position (ACP) for each of the following events.
Consider each independently.
Limit your answer to ECP is HIGHER, LOWER, or SAME as ACP.
a.
.The-fourth coolant pump is started two minutes before criticality.(0.4) b.
The startup is delayed until 8 hours9.259259e-5 days <br />0.00222 hours <br />1.322751e-5 weeks <br />3.044e-6 months <br /> after the trip.
(0.4) c.
The steam dump pressure setpoint is increased to a value just below the Steam Generator PORV setpoint.
(0.4) d.
Condenser Vacuum is reduced by 4 inches of mercury.
(0.4) o.
All Steam Generator levels are rapidly being raised by 5% while approaching criticality.
(0.4)
.L.__EElblC1ELES_DE_biUCLE0 LEDWER_ELeblI_DEEEeI1CL PAGE
.9 IHEEtdODYelCS._WEaI_IEatJSEER_eblD_ELU1D_ELOW ANSWER 1.11' (2.00) la.
SAME
-b.-
LOWER c.
LOWER d.
SAME e '.
HIGHER (0.4 ea)
REFERENCE.
192008K107-
...(KA'S)
QUESTION 1.12 (1.00)
Select the statement below that is correct concerning DNB.
a.
A reactor power increase puts the plant FURTHER from DNB.
b.
A pressurizer pressure decrease puts the plant FURTHER from DNB.
i c.
A reac tor cool an t ' fl ow increase puts the pl an t FURTHER f rom DNB.
d.
A cold leg temperature increase puts the piant FURTHER from DN8.
ANSWER 1.12 (1.00) c.
(1.0)
REFERENCE r
193008K105
...(KA'S)
I
(***** CATEGORY 01 CONTINUED ON NEXT PAGE *****)
i I
i
L f
-1.__ERlNCIELES_DE_NUCLEeR_EDWER_ELeNI_DEEReI1ON.
PAGE 10 IMERtdDDYNICS BEGI_IReUSEER_eND_ELU1D_ELOW QUESTION 1.13 (2.00)
'O.
Why~does a single RCP pump running during hot shutdown draw more motor amperage than when one of four running at power?
(0.5) b.
Why does a RCP running at cold conditions draw more motor amperage than at hot conditions?
(0,5) c.
Why is RCP motor amperage higher when starting the pump than when running?
(0.5) d.
Why_should-operating a pump with too much fl ow and no di scharge pressure be avoided?
(0.5)
ANSWER 1.13 (2.00)
O.
A single pump running has a higher flow than when all 4 are operating (due to reduced discharge pressure) so more work is done and more amperage drawn. ' ACCEPT CONCEPT FOR ALL-PARTS)- (0.5) b.
At cold conditions, fluid density is higher-so more mass.ls moved so more work is done and more amperage drawn. (0.5) 1:..
It must accelerate more mass, which rtauires more work, and amps.<0,5) d.
Operating at runout may cause rump damage / trip on overcurrent. (0.5)
REFERENCE WCGS.LO 12 311 35 EO B.22-32 191004K107
...(KA'S) i
(*****
CATEGORY 01 CONTINUED ON NEXT PAGE
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1.__ER1NC1ELES_QE_NUCLEeR_EDWER_ELauI_OEERallOU.-
PAGE 11 IHElit!ODYhlAMICS._HEaI_IRANSEER_eUQ_ELU1D_ELOW QUESTION 1.14
'(2.00)
The reactor;is critical at 10E-6%.and no load Tave.
The folloaing conditions exist:
-Rod bank D is.at 80 steps
-Core burnup is 4000 MWD /MTU-
-Boron.is at HFP, ARO, eq. Xe conditions for time in core life
-Those systems normally in auto operation at this point during startup are functioning as prescribed.
a.
What startup rate would be achieved if bank D is pulled to 120 steps?
(Assume step change in rod position.)
Show al l work.
(1.0) 6.
With no operator action, at what level will reactor power stablize after the rods are pulled to 120 steps?
Show all work (1.0)
Selected plant operating curves have been attacted to assist you, it may not be necessary to use all curves provided to answer the question.
' ANSWER 1.14 (2.00) a.
From figure 2.5: 230-250 pcm added by rod movement (0.5)
From figure 2.91 250 pcm results in about 4 dpm. (mathematical calculat ion acceptabl e) (0,5)
.b.
From figure 5.1: boron concentration is about 500-510 ppm (0.5)
From figure 3.3: reactor power stabilizes at about 14-16% (0.5)
REFERENCE WCGS CYCLE 2 Operating Curves, LO 11 306 24, EO B.28, LO 11 306 33 19200fK120
...(KA'S) i
(***** CATEGORY 01 CONTINUED ON NEXT PAGE
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s
L._ER1NCIELES_DE_UUCLEaR_EDWEB_ELablI_DEEEeIlOU.
PAGE 12 IBEEtdODYM1CS._BEaI_IEaNSEER_ablD_ELU1D_ELOW 4
3
. QUESTION-1.15 (2.00)
.' Fkwg would xenon concentration change for the power history shown be' low
.(Elther a' sketch or explanation is acceptable.)?' Assume'pewer changes are step changes as shown.
Assume xenon-free i n i't i al l y,
t P'
100%-1 o
l w
't e
50%-l.************
')
r i
- i *
(X) 0%-!
- l-f
/
/
/
/
/
0 40 80 120 160 200 hours0.00231 days <br />0.0556 hours <br />3.306878e-4 weeks <br />7.61e-5 months <br /> i
l 1
4 r
)
i i
r a
[
{
t
(
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q ni '
EL_ERiblC1ELES_QE t9UCLEAR_EDWER_ELeblI_DEEReIl0th
.PAGE 13 IBElft:iODYhleM1CS eWEeI_IEeblSEER_ablD_ELUID_ELOW h.
's
' ANSWER 1.15 (2.00)
. * = xenon cencentration (d)
/Xe-peak 6~8 hrs / trip Xe-pk-t
r-------------------**----------------------------
t X
i--------------****---*--------(f)----------------
e
-l
- - (c)
/-
I Ca)-
e 50%-1 ---------*-*--*---------------***-------*------------- '
q-I (e) 1
'(b) 1 Xe-free 3 I *
- /60~80 hrs / trip-
~ (%)
OX-! *
(g)
/
/
/
0 40 80 120 160 200 hours0.00231 days <br />0.0556 hours <br />3.306878e-4 weeks <br />7.61e-5 months <br /> Grading:
Locations marked ( ) should'be shown or described at 0.286 each.
a.
Xenon buil ds in almost to equilibrium value.
b.
Xe. drops due to burnout increase when flux increases with~
i odi ne decay ~ i ncreasi ng more sl owl y.
c.
X&. builds in almost to equilibrium value.
d.
Xe. builds to peak since burnout drops sharply wi th iodine decay dropp i ng more sl owl y,
e.
Xe. drops to equilibrium value, f.
Xe. increases to peak after trip.
Time and value are dependent on power history (~SQRT of % power).
g.
About 60~80 hours to negligible xenon.
REFERENCE WCGS LO 11 306 37 192006K111
..(KA'S) l
^
(***** END OF CATEGORY 01
- )
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2.__ELaul_DESIGu_1UCLUDluG_SeEEIY_6UD_EUERGEUCY_SYSIEUS PAGE 14 QUESTION 2.01 (1.00)
Explain briefly what differentiates between the cooling lake and the ultimate heat sink.
ANSWER 2.01 (1.00)
UHS is contained within the lake by the normal l y submerged UHS dams.
UHS is a subset of the cool i ng l ake.
(Accept concept of subset contained by dams 3 1.0)
REFERENCE WCGS LO 14 089 00, EO B.5 076000K126
...(KA'S)
QUESTION 2.02 (2.50)
MATCHING.
Match each of the areas below with the appropriate water application method used by the Fire Protection System.
a.
Diesel Generator Rooms 1.
Wet-pipe sprinkler system b.
Computer Rooms 2.
Water spray system c.
Preaction sprinkler system d.
Turbine Driven Aux. Feed.
Pump 4.
Hal on 1301 system e.
ESF Switchaear Rooms 9
Auxiliarv Boiler Room h.
Hydrogen Seal Oil Unit i.
Cable Spreading Rooms J.
Diesel Fire Pump Room ANSWER 2.02 (2.50) a.
3 f.
4 b.
4 9
I c.
1.
4 h.
2 d.
2 i.
3 e.
2 j.
1 (0.25 ea)
NOTE:
For future use modify item c to clearly indicate whether the TSC computer room is included.
t k
\\
2.__ELANT_DESIGu_1NCLUDIUG_E6EEIY_eUD_EUEEDENCY_SYSIEUS PAGE 15 REFERENCE WCGS LO 14 086 00 086000K400
...(KA'S)
QUESTION 2.03 (1.00)
Explain how failure of one incore thermocouple reference junction box heater can be detected by observing thermocouple (core outlet) temperature indication.
ANSWER 2.03 (1.00)
Thermocouples from the box with the failed heater will drift (WRT the other indications).
REFERENCE WCGS LO 16 107 12. EO B.2 000074A116 000074K208
...(KA'S)
B GUESTION 2.04 (2.50)
List five (5) failures or problems in the Reactor Protection System (RPS) which would cause a general warning indication.
ANSWER 2.04 (2.50)
(any 5 at 0.5 ea) 1.
Loss of either 48vde p owe r supply.
2.
Loss of either 15vdc power supply.
3.
Printed circuit cards not properly inserted 4.
Input error inhibit switch in the inhibit position.
5.
Slave relay tester mode switch in the test position.
6.
Multiplexing inhibited.
7.
Logic A switch not in off.
8.
Permissive test switch not in off.
9.
Memories not set.
- 10. Bypass bkr.
closed & connected.
REFERENCE WCGS LO 13 012 00 EO B.3 012000A200
...(KA'S)
(***** CATEGORY 02 CONTINUED ON NEXT PAGE +++++)
2 __ELaNI_ DESIGN _1NCLUQ1NG_SeEEIY_aND_ENERGENCY_SYSIEUS PAGE 16 QUCSTION 2.05
(.50)
Concerning the cold leg accumulators (CLA's), pick the following state-ment that is true:
a.
The accumulator i solation valves f ail as is.
b.
An orifice is provided to extend the blowdown time of tne CLA during a LOCA.
c.
The CLA reliefs discharge to the PRT.
d.
Accumulator check valves are not considered pressure boundary valves.
s ANSWER 2.05
(.50) a.
(0.5)
REFERENCE WCGS LO 13 006 00, EO B.3 006000K602
...(KA'S)
OUESTION 2.06
(.50)
When the Reactor Makeup system is in AUTOMATIC, how does it know what strength boric acid to provide?
a.
The automatic controller has an input from the boronmeter.
b.
The analyses from Chemistry are au t oma t i c al l y put into the controller.
c.
The controller uses the integrated charging header flow signal.
d.
The controller uses the setting on the manual / automatic potentiometer.
ANSWER 2.06
(.50) d.
(0.5)
REFERENCE WCGS LO 13 004 00 004000K106
...(KA'S)
(***** CATEGORY 02 CONTINUED ON NEXT PAGE
- )
2._ELauI_DESIGu_1NCLUDluG_SaEEIY AND_Et!EEGEUCY_SYSIEtdS PAGE 17 QUESTION 2.07 (1.50)
O.
List the signals that will au t oma t i c al l y start the TURBINE driven auxiliary feedwater pump.
(Include coincidences)
(1.0) 6.
Which S/G's can supply steam to the turbine driven auxiliary feedwater pump?
(0.5)
ANSWER 2.07 (1.50) a.
2/4 low-low levels in 2/4 S/G's 2/4 low voltage on NB01 or NB02 C0.5 ea.)
b.
B and C.
(0.5)
REFERENCE WCGS LO 14 061 00, EO B.18 061000K101 061000K402
.(KA'S)
QUESTION 2.08 (2.00)
O.
The f l owr a t e through the RCP M1 seal is not constant for all plant conditions.
Explain WHEN and WHY the flow rate will be at it's highest and l owest val ue.
Consider only normal conditions and not failure.
(1.0) b.
What is/are the flowpath(s) for RCP M1 seal leakoff during Safety Injection?
(0.5) c.
What par ame ters de termi ne the differential pressure across the #1 seal?
(0.5)
i L_ELeUI_DESIGU_1UCLIJDIUS_SeEEIl_eUD_EUERDENCY_Si'SIEUS PAGE 18 ANSWER 2.08 (2.00) c.
As the plant pressure changes so will the delta-P acnoss the Mi seal, thus changing the seal f l owr a t e. ( 0. 5)
Fl ow is high at high pressures and low at l ow pressures. (0.5) (Accept variation in CCP discharge pressure to seals.)
b.
Through M2 seal to the RCDT (0.1) and the Mi seal r e t u r-n line relief valve t the PRT (0.4).
c.
RCS pressure compared to the backpressure createo by the VCT. (0.5)
REFERENCE WCSG LO 13 003 00. EO B.2 003000K103 003000K407 003000K602
...(KA'S)
OUESTION 2.09 (2.00) n.
Li st TWO re asons f or mai n tai r. i ng a mi n imum pre ssur i zer spray line flow during normal "at p ow e r " operations.
(1.0) b.
What indication or annunciation is available to alert the operator that minimum spray f l ow is not being maintained?
(0.5) c.
What creates the motive force for pressurizer spray flow?
(0.5)
ANSWER 2.09 (2.00) a.
1.
To reduce thermal stress to the spray line and spray nozzle.
2.
To maintain Pzr. chemistry uniform with the RCS. (0.5 ea.)
6.
Spray or surge line low temperature alarms. (one required) (0.5) c.
Differential pressure across the reactor vessel (accept RCP dP or pump head) (0,5)
REFERENCE WCGS LO 13 010 00 EO B.11 010000A102 010000K603
...(KA'S)
(*****
CATEGORY 02 CONTINUED ON NEXT PAGE *****)
?-
ELauI_ D ESIGN_1N CL UDIN G _Sa EEIY_eN D _ Eld ERG EN CY_SY SIEtdS PAGE 19 QUESTION 2.10 (2.50)
I o.
What additional purpose is served by Diesel Engine Starting Air? (0,5)
I b.
During load test surveillance, what is the position of the UNIT -
PARALLEL switch and WHY?
(0.5) c.
How is starting air blocked if the diesel should receive a start signal with the unit already running?
(0.5)
I d.
Describe the sequence of events that occur if tne EDG is paralleled and loaded and a LOCA and loss of p ower occur s.
Discuss only those events related to the EDG or e l ec tr i cal system.
(1.0)
ANSWER 2.10 (2.50) a.
Used to shutdown the diesel (0.5) 6.
Parallel or Normal. (0.2)
Used to provide the speed control governor with a droop characteristic for stability of p ar al l e l operation. (0.3) c.
When the EDG is sensed to be above a certain speed, the air start solenoid valve receives a shut signal. (0.5) d.
A load shed signal will trip all feeder and load breakers. (0.25) l When proper speed and vol tage is sensed, (0.25) the DG output br eaker l
will close on to the bus. (0.25)
Vital loads are sequenced on to the bus. (0.25)
REFERENCE WCSG LO 14 064 00, EO B.3, 8
064000A300 064000K105
...(KA'S)
(***** CATEGORY 02 CONTINUED ON NEXT PAGE
- )
2m__ELeUI_ DESIGu_ INCLUDING _SeEEIY_ eUD_EldEEGENCY_SYSIEMS PAGE 20 QUESTION 2.11 (3.00) a.
What are 2 specific instances where the operator may be required to initiate flow through the RHR System Letdown Control Valve (HCV-128)?
(1.0) b.
How does system design assure that the Positive Displacement Charging Pump is started unloaded?
(0.5) c.
Explain tae effect on Seal Return flow, should the Seal Water Heat Exchanger become plugged or restricted on the seal water side?
(0.5) d.
What are 2 possi bl e ou tl e ts f rom each of the f ol l owi ng (1.0) 1.
Seal Water Heat Exchanger?
2.
Excess Letdown Heat Exchanger?
NOTE: Part b is no longer applicable and should be removed prior to future use of thi s quest i on.
ANSWER 2.11 (3.00) a.
1.
To purify coolant when on RHR cooling.
2.
Provide additional letdown during heatup or bubble formation.
3.
RCS solid plant pressure control.
(Any 2, 0.5 ea.)
b.
The pump is interlocked to prevent starting unless a bypass.alve is open which closes ~ 2 minutes after starting. (0.5) (Accept ti y answer which indicates that the pump must not be deadheaded.)
c.
Con t i nued f l ow is assured by a relief valve which directs flow to the VCT. (0,5) d.
SW HX:
Ex. LD HX:
1.
Chg. pump suct.
1.
UCT 2.
RCDT (0.25 ea.)
3.
PRT (Any 2, 0.25 ea.)
REFERENCE WCGS LO 13 004 00. EO B.2, 3
o*******
004000K405 004000K604 004010K101
...(KA'S)
(***** CATEGORY 02 CONTINUED ON NEXT PAGE ++***)
'2.__ELeUI_DESISU_IUCLUDING_SeEEIY_eUD_EMERGEUCY_SYSIEUS PAGE 21 i
GUESTION 2.12 (2.00) a.
What are 5 specific conditions that place the control room heating, vantilating, and air conditioning system into the emergency mode of operation?
Setpoints are NOT required.
(1.0) b.
De scr i be how the control room atmosphere is maintained habitable during accident conditions.
(1.0)
ANSWER 2.12 (2.00) a.
1.
Chlorine 5.
CIS "A"
2.
Control rm. gas, act.
6.
Cont. purge hi gas, act.
3.
FB vent. i sol. sig.
7.
Cont. atmos, hi gas. act.
4.
SI Si gnal.
8.
Any other CPIS (Any 5, 0.2 each)
NOTE:
Manual initiation is not acceptable since it is not a condition but an action.
b.
Outside air is isolated and the system goes on internal recirculation (0.5).
S i mu l t an e ou sl y, the CR is pressurized with filtered air (0.5) l (to prevent inleakage).
REFERENCE WCGS LO 14 088 03 EO B.9, 10 072000A301 072000K104
...(KA'S)
I l
l l
(***** CATEGORY 02 CONTINUED ON NEXT PAGE
- )
2.__ELauI_DESIGu_1UCLUDING_SaEEIY_aND_EMERGEUCY_SYSIEdS PAGE 22 e
GUESTION 2.13 (2.00)
For the f ol l owi ng componen ts, indicate whether they will receive an OPEN, CLOSE, or NO signal upon a manual safety injection initiation, n.
Control Room outside air isolation valves b.
Main Feedwater bypass valves c.
Cold Leg Accumulator isolation valves d.
Charging header isol ation valves o.
Main steam isolation valves f.
RWST to centrifugal charging pumps suction valves g.
RCP seal water return isolation valve h.
CCW isolation valve f r om RHR h e a t exchanger 1.
CCW isolation valve from letdown heat exchanger J.
Steam supply valves to turbine-driven AFW pump ANSWER 2.13 (2.00)
(0.2 pts each) a.
CLOSE b.
CLOSE c.
OPEN d.
CLOSE e.
N0 f.
OPEN g.
NO h.
NO i.
NO J.
NO NOTE:
For future use.. specify the valve numbers in the question to avoid potential confussion over nomenclature.
REFERENCE WCGS LO 13 013 01, EO B.7 013000K100
...(KA'S)
(***** CATEG0F' 02 CONTINUED ON NEXT PAGE
- )
?_
PI MJT DESM thiri tiniblG_mp rv man miwhin v nses ygge. 23 QUESTION 2.14 (2.00) a.
What is_one preparatcry operator action required prior to automatic moitchover'to RHR Cold Leg recirculation?
(0.6) b.
lhe RHR pumps must be stopped during a manual switchover to prevent loss of pump suction.
Why-is this necessary?
(0.0) c.
How can an ' AUT 0 " switchover be prevented?
(0.b) d.
How i s con t i nu ed fl ow assured dur ing manual switchover?
(0.6)
ANSWER 2.14 (2.00) c.
Accept any one of the f ol l owi ng s (0.5) 1.
2.
Check RWST level.
3.
Check Sump level.
4.
Verify AC power available as required.
b.
An interlock prevents opening the cont.2inmeni sump value with the same side RWSr valve open (0.5).
c.
Depress the RWST SIS reset button.
OR Rack out the power to the appropriate breakers.
(0.5) d.
By switching over only one train at a time. (0.5)
REFERENCE WCGS LO 13 005 00, EO B.1, 3 005000K400.
...(KA'S)
(***** END OF CATEGORY 02 *****)
4
~
3.__1NSIBudENIS..AUQ_CONIROLS PAGE 24 QUESTION 3.01
(.50)
Which one of the f ol l owi ng si gnal s cause BOTH a saf e ty injection and a main steamline i sol a t i on?
a.
Low Steamline Pressure b.
High Steam Pressure Rate c.
High Containment Pressure d.
Low Pressurizer Pressure NOTE:
Prior.to reuse, specify item c as "High 1" to preclude it from being an accep tabl e answer under certain circumstances.
ANSWER 3.01
(.50) a.
(0.5)
REFERENCE WCGS LO 13 013 01, EO B.7 013000K101
...(KA'S)
OUESTION 3.02 (1.50)
Indicate whether the f ol l owi ng statemen t s concern i ng a re si st ance temperature detector (RTD) are TRUE or FALSE.
P a.
When an RTD fails due to an open, it always indicates a downscale (low) reading on its meter, b.
As temperature that is sensed by an RTD changes, a proportional change in the output uoltage (current) across a bridge occurs.
c.
If an RTD is completely submerged, its ability to accuratel y monitor temperature is unaffected 'y flow rate.
a i
i l
l t
3.__1USIBUMFMYS 6UD_COUIBOLS PAGE 25 ANSWER 3.02 (1.50) a.
FALSE b.
TRUE c.
FALSE (0.5 ea)
REFERENCE WCGS LO 13 002 02, EO B.2 016000A200 016000K601
...(KA'S)
GUESTION 3.03 (1.50)
Indicate whether the follaaing statements concerning operation of the reactor trip (RT) and bypass (BY) breakers are TRUE or FALSE.
a.
A solid state protection system (SSPS) train A reactor trip signal will trip RTA and BYA' breakers.
b.
If it is attempted to close both bypass breakers at the same time, then both bypass breakers will trip but the reactor trip breakers will remain closed.
c.
If one train is placed in test while the other train's bypass breaker is closed, then both reactor trip breakers and both bypass breakers will trip.
ANSWER 3.03 (1.50) a.
FALSE b.
FALSE c.
TRUE (0.5 ea)
REFERENCE WCGS LO 13 012 00. EO B.6, 7 LO 13 001 00, EO B.5 001000K603 012000A307 012000kl03
...(KA'S)
(***** CATEGORY 03 CONTINUED ON NEXT PAGE
- )
3.__1USIEUMENIS_6UD_CONIBOLS PAGE 26 QUESTION 3.04 (2.50)
Match the f ol l an i ng syn.p t oms or c au se s in eciumn "B"
to the specific Rod Control System failure or error.in column "A".
.A" "D"
c.
Power Cabinet Urgent Failure 1.
Caused by simultaneous zero current to stationar.y and movable grippers, b.
Regulation failure 2.
Unselected rod (s) having current f l ow in movabl e or lift coils, c.
Phase failure 3.
Caused by failure of redundant power supply modules.
d.
Logic error 4.
Caused by pulser or slave cycler
- failure, e.
Multiplex error 5.
Caused by full current being applied for excessive time.
(There is only I correct numerical 6.
Can be caused by regulation or answer for each lettered error or phase failure as well as logic failure) or multiplex errors.
7.
Occurs whc.n vol tage to coils has excessive ripple.
(2.5)
ANSWER 3.04 (2.50) a.
6 b.
5 C.
7 d.
1 e.
2 (0.5 ea)
REFERENCE WCGS LO 13 001 00, EO B.6, 7,
10 001000K403 001010K604
...(KA'S)
(***** CATEGORY 03 CONTINUED ON NEXT PAGE *****)
l
l 3.__INSIRUMENIS_aND_DONIRDLS PAGE 27 QUESTION 3.05 (2.50) a.
Fully explain how a trip is avoided when testing the Source or intermediate ranges of nuclear instrumentation and the reactor is critical at any power level.
(1.0) b.
Explain the function of the "BYPASS" switches or, the Nuclear Instrument Miscellaneous Control and Indication Panel.
(1.5)
ANSWER 3.05 (2.50) a.
Each channel has a bypass switch to prevent a trip at low power levels due to 1/2 trip logic.
[0.5]
At higher levels, the trips are blocked by the P o or P-10 permissives. [0.5]
b.
1.
The "Rod Stop Bypass Switch" [0.53 removes the overpower rod stop function for the selected channel to al l ow rod mot i on.
[0.25) 2.
The "Power Mismatch Bypass Switches" [0.53 remove the selected channel from the rod control system auctioneer circuit.[0.253 REFERENCE WCGS LO 13 012 00. EO B.13 012000K406 012000K600
...(KA'5)
QUESTION 3.06 (1.00)
If the reactor trip due to turbine trip fails to actuate, what are two other Reactor Protection System trips that would cause a reactor trip?
ANSWER 3.06 (1.00) 1.
High RCS Pressure.
2.
Overtemperature delta T.
3.
High pressurizer level.
(Any 2 2 0.5 each)
REFERENCE WCGS LO 13 012 00 EO B.4 012000K404
...(KA'S) f
(***** CATEGORY 03 CONTINUED ON NEXT PAGE *****)
3.__lUSIRubENIS_6ND_CDUIROLS PAGE 28 GUESTION 3,07 (2.00) a.
De scr i be how power c an be supplied to a typical Safety Rel ated 120VAC Instrument bus.
(1.0)
+
b.
Describe the interlocks associated with the different bus power supplies.
(1.0)
ANSWER 3.07 (2.00) c.
Each bus can be supplied from a static inverter (0,5) or a regulated transformer power supply. (0.5) 6.
The Inverter output breakers are interlocked with the transformer breakers so that only one can be closed at a time. (0.3)
The transformer output breakers are interlocked so that a transformer can supply only one bus at a time. (0.5)
REFERENCE WCGS LO 15 063 00, EO B.4 062000K410
...<KA'S)
QUESTION 3.08 (1.00)
What specific protective functions are supplied input from the Loop Tavg signal?
ANSWER 3.08 (1.00) 07dT and OPdT are supplied from Loop Tavg to provides Rod Stop Turbine Runback Reactor Trip P-12 blocks steam dumps.
P-4 FWIS i
(0.2 ea)
REFERENCE WCGS LO 13 002 02, EO B.6, 7
002000K606
...(KA'S)
(+**** CATEGORY 03 CONTINUED ON NEXT PAGE
- )
3.__luSIRudEUIS_eUD_CDUIROLS PAGE 29 OUESTION 3.09 (2.50) c.
List all initiating signals for the P-14 Permissive and a Feedwater Isol at ion Signal (FWIS).
(1.0) b.
List ALL automatic actions that will result from reaching the P-14 setpoint and receiving a FWIS.
(1.5)
ANSWER 3.09 (2.50) a.
P-14:
High S/G 1evel FWIS:
o High S/G 1evel o
Si o
Rx. trip and Low Yavg o
Manual o
Lo-Lo S/G level (0.166 ea.)
b.
P-14:
o Trips Feed Reg. Valve o
Trips Main Feed Pumps o
Trips Main Turbine o
Closes Main Feed Isolation Valves FWIS:
o Closes Feedwater Isol ation Valves o
Closes Main and Bypass Feedwater Reg. Valves (0.25 ea.)
REFERENCE WCGS LO 13 013 01. EO B.6 013000K115 013000K413
...(KA'S)
QUESTION 3.10 (3.00)
Give the RPS reactor trip (s) designed to protect against each of the events l i s t e d be l ow.
The number in parentheses by each event i ndicates the required number of responses for full credit.
a.
Loss of Cool an t Fl ow. (3) b.
Rod Ejection. (1) c.
Reactor Overpower. (2) d.
Departure from Nucleate Boiling. (2) e.
Multiple Rod Drop. (1) f.
RCS Integrity. (2) 9 Loss of Heat Sink. (2) h.
Reactor Startup Power Excursion. (2) m
3.__1NSI2UBENIS_eUQ_CDUIROLS PAGE 30 ANSWER 3.10 (3.00) a.
1.
loss of sensed flow 2.
RCP bus undervol tage 3.
RCP bus underfrequency.
b.
4% increase in LT 2 seconds (high pos, rate).
c.
1.
p ower range h i gh flux 2.
OP dT.
d.
1.
O'
- dT 2.
Ic,w pressurizer pressure 3.
loss of flow
( A 4Y 2) o.
4% decrease in LT 2 seconds (high neg. rate).
f.
1.
high pressurizer pressure 2.
high pressurizer level.
9 1.
low SG 1evel 2.
h.
1.
SR NIS high flux 2.
IR high flux 3.
PR high flux.
(ANY 2)
(0.2 ea)
REFERENCE WCGS LO 13 012 00, EO B.4 012000SG7
...(KA'S)
QUESTION 3.11 (2.00)
List four (4) conditions which must be satisfied to manually shut the diesel generator output breaker from the MCB.
ANSWER 3.11 (2.00)
( ANY 4 A 0.5 EA) 1.
Generator up to voltage (4.16KV).
2.
Generator up to speed (471 RPM).
3.
No lockout relays energized for that ESF bus.
4.
Sync check relay satisfied.
5.
Synchronizing switch ON.
6.
Breaker control switch CLOSED.
7.
Normal or Alternate ESF feeder breaker closed.
REFERENCE WCGS LO 14 064 00. EO B.6 064000A401
...(KA'S)
(***** CATEGORY 03 CONTINUED ON NEXT PAGE
- )
r 3.__lUSIRubENIS_auD_CQUIROLS PAGE 31 QUESTION 3.12 (1.50)
The reactor is in hot shu tdown wi th RCS pressure at 1800 psig.
The reactor trip breakers have been reset and the shu tdown bani:5 are fully withdrawn.
I & C maintenance personnel want to perform a calibration on the Turbine Impulse pressure channels.
Explain why this should or should not be done at this time.
(1.5)
ANSWER 3.12 (1.50)
Should not be performed at this time. [0.53 Pl ac i ng in test will deenergize the interlock circuit P-13 which feeds P-7.
[0.5)
This will unblock the low pressurizer trip and the shutdown control rods will trip.
(0.53 (1,5)
REFERENCE WC LP 050, P.
18 012000K406
...(KA'S)
QUESTION 3.13 (2.00)
Reactor power is 100%. Pressurizer pressure instruments are as f ol l ows :
PT 455 supplying PT 455A PT 456 supplying PCV 456A While pressurizer pressure is 2235 psig, PT 456 faits high.
Describe the plant response (focusing on pressure control components) with no operator action and all systems in auto / manual as prescribed by normal operating procedures.
Continue your answer to stable plant conditions.
ANSWER 3.13 (2.00)
PCV 456A opens immediately (0.5)
Variable heaters are at full by 2220 psig (0.25)
Backup heaters energize at 2210 psig (0.25)
PCV 456A and bl ock valves close at 2185 psig (on 2/3 of channels left>(0.5)
Pressure will stablize around 2185 psig since both PCV 456A and the block valves will reopen when the 2/3 of the remaining channels go >2185 psig and shut again when pressure drops to 2185 psig. (0,5)
REFERENCE WCGS LO 13 010 00 EO B.4 010000K601
...(KA'S)
(***** CATEGORY 03 CONTINUED ON NEXT PAGE
- )
t
3.__1NSIEUMEUIS_eUQ_CONIROLS PAGE 32 OUESTION 3.14 (1.50)
List three (3) saf e ty-rel ated process radi at ion moni tori ng systems which generate control si gnal s when the alarm setpoint is reached, include a brief description of system responses to these control signal s f or each PRM listed.
ANSWER 3.14 (1.50)
(any 3 at 0.2 for PRM and 0.3 for response each) 1.
Control Poom Air Supply (GK RE04, 05) generates CRVIS 2.
Containment Purge Exhaust (GT RE22, 33) generates CPIS and CRVIS 3.
Fuel building emergency exhaust (GG RE27, 28) redirects fuel building exhaust through the emergency exhaust train (FBIS) 4.
Containment Air (GT RE31
073000K101
...(KA'S) l i
(*****
END OF CATEGORY 03
- )
d._ _ E R O C E D U E E S _=_N D Etial._e Bu O Rtie L._ Eti E R G EN CY._6U D PAGE 33 rad 10 LOGICAL _CDUIBOL QUESTION 4.01 (1.00)
Explain why the ALARA program does not always reduce exposure to the lowest achievable levels.
ANSWER 4.01 (1.00)
ALARA reduces dose based on a comparison of state of technology and oconomics of improvements vs. benefit of reduced exposure (accept content not specific wording).
(1.0)
REFERENCE WCGS ALARA Program, ADM 03-050, p.
1 194001K104
...(KA'S)
OUESTION 4.02 (2.00)
MATCHING.
Match the symptoms with the type of leakage indicated per OFN 00-007, "RCS Leakage High."
Terms may be used more than once or not at all.
a.
Increased level in recycle 1.
Common to all types of RCS leakage.
holdup tank.
2.
Steam Generator tube leakage, b.
Increased condenser vacuum 3.
Leakage into the auxiliary building pump vent activity.
4.
Leakage into the containment.
c.
Increased liquid waste 5.
Leakage into other systems.
holdup tank level.
d.
Increased VCT makeup frequency.
e.
Increased containment temperature, f.
Increased pressure in PRT.
g.
Increased Auxiliary building sump level, b.
I ncre ased chargi ng f l ow.
NOTE:
Prior to reuse, define "RCS leakage" as reactor cool an t leaking from the RCS or any of its interconnecting and support systems.
4
A__ERDDED1JR M UDNI
--A NO NL_E11ERGENDX_GUD PAGE 34 BuDIDLDDIDuL_CDUISDL ANSWER 4.02 (2.00) a -5 e -4 b -2 f -5 c -3 or 4 9 -3 d -1 h -1 (0.25 ea)
REFERENCE WCGS OFN 00-007, RO, LO 17 324 01, E01 (pg 10) 002020K401
...(KA'S)
GUESTION 4.03 (1.50)
MATCHING.
Match the symptoms of a reactor trip or safety injection with the description from the list below.
Descriptions may be used more than once or not at all.
a.
PRZR pressure 1900 psig.
1.
Reactor Trip Required.
b.
ECCS Pumps running.
2.
Reactor Trip Occurred.
c.
Pressurizer pressuer 2385 psig.
3.
Reactor Trip & Safety Injection d.
All Rod bottom lights lit Required.
o.
Reactor trip breakers open.
4.
Reactor Trip & Safety injection f.
Pressurizer pressure 1849 psig.
Occurred.
5.
None of the above.
ANSWER 4.03 (1.50) a -5 i
b -4 c -1 or 2 d -2 e -2 f -1 (0.25 ea)
REFERENCE WCGS EMG E-0 PO, LO 17 323 01 012000A300 013000A403
...(KA'S)
(***** CATEGORY 04 CONTINUED ON NEXT PAGE *****)
a.__EBOCEDURES_=_UDEdeL._aBUORdeL._EdERGENCY_auD PAGE 33 RaDIOLOG1 del _CDUIROL QUESTION 4.04 (2.00)
Identify the evacuation exits from the fuel building.
NOTE:
For future use, modify the question to specify the number of routes required for full credit answer.
ANSWER 4.04 (2.00)
Door M61021 next to the roll-up door on the east side toward the south end of the fuel building el. 2000'.
Door #61011 located on the south side through the stairwell el. 2000'.
Door #15071, northwest door on el. 2047'.
Roll up door in truck bay el. 2000' Door through heat exchanger room (6104) through NW door to room 1119 of cux bldg on el. 2000' Through the switchgear room and the HVAC room to the aux bldg through NW door on el. 2026' (Accept any description which can be translated to the above locations.:
(any three at 0.5 ea)
REFERENCE WCGS LO 17 324 01. EO 5 (OFN 00-018) 000036SG6
...(KA'S)
GUESTI0tl 4.05 (3.00)
Answer the f ol l owi ng regardi ng the S/G tube rupture procedure EMG E-3.
A.
State the S.I.
termination criteria.
(2-0)
B.
Wha.t constitutes Reinitiation Criteria for 5.1.?
(1.0) i
A __RROCEDURES_=_NOCMeL+_uBUORt%L+_Et:lEaGENCY_68Q PAGE 36 red 10LOGICt&L_CONIROL ANSWER.
_4.05 (3.00)
A.
rmination criteria;
'S subcooling-based on RCS temp ind - greater.than required by 7.
1 or->30 deg.F ondary heat sinks Total - f eed fl ow to intact S/Gs-- greater than 320,000
'lbm/hr.
o.-
N.R.
level in at least one intact S/G - greater than 4%.
3.
RCS pressure - stable or increasing 4.
-Pressurizer level - greater than 4%
B.
Meinitiate 91 if EITHER condition. listed below occurs:
1.
' RCS subcool ing based on RCS temp i r.d - l e ss than required by fig.1: or.<30 deg.F-2.
PressurizerLlevel - cannot be maintained greater than 4%.
0.5 pts for each of 6 major subparts.
REFERENCE WCGS EMG E-3, LO'17 323 04, E08 (pg 11) 000009A234
...(KA'S)
QUESTION 4.06 (2.00)
OFN 00-015 (Loss of RHR Cooling). instructs the operator to establish alternate decay heat removal if the RHR system should fail.under use.
Briefly describe the preferred alternate decay heat removal method fort a.
S/Gs available, S/G pressure 17 psia, condenser not available.
(1.0) b..
Vessel head removed.
(1.0) l.
1 l
l t
IL,
- A__BBDCED1JEES - blDBtdeL._eBUDBtdAL_EldERDENCY_eUD PAGE 37 BADlDLDDICAL.CDUIBDL i:
ANSWER 4.06 (2.00) a '.
' Use MD AFW pump - (f ol l owed by startup feed pump). ( O '. 5 ) and steam to S/G PORV's (0.5)
'b.-
Maximize charging / letdown flow (0.5) and pl a::e spent fuel pool cooling system in operation. (0,5)
NOTE -
Examinee may respond with. additional ac t i oris con t ai ned in the OFN.
These are to be treated as null responses, no credit awarded. 'The key indicates the essential elements which are required for full credit.
REFERENCE WCGS OFN 00-015 LO 17 324 vi, E03 (pg 26) 005000A200
...(KA'S)
QUESTION 4.07
(
50)
Which of the following external radiation exposuret would inflict the greatest biological damage?
a.
1 Rem of ALPHA.
b.
1 Rad of NEUTRON
. c '.
1 Roentgen of BETA d.
1 Rad of-GAMMA-
- ANSWER 4.07
(.50) b..
-(0,5)
-REFERENCE-WCGS LO-10 310 03, 10CFR20 194001K103
...(KA'S)
(***** CATEGORY 04 CONTINUED ON NEXT PAGE
- )
'di__2ROCi D'JRES_=_NORdeL+_eBNORheL+_ EMERGENCY _eND PAGE 38'
-red 10LOGICeL_CONIROL.
' QUESTION 4.08-(2.00)
N1 fo.
During-the performance'of GEN'00-002'(Cold Shutdown to Hot Standby),
what are two (.2) specific. actions taken by the control room operator to improve' secondary ~ chemistry or puri.ty.
(0.5) 6.
At what RCS temperature is-Operating Mode 3 entered,-and what significant_ step related to plant safety is taken at this time? (0.75) c.
Briefly' describe how-RCS is overpressure protected fron Mode 5-until startup is complete.
Disregard code safety valves.
(0.75)
NOTE:- Prior to reuse, reword part c to indicate that the desired answer is the' overpressure-device in service with the lowest setpoint for each condition.
ANSWER
-4.08 (2.00) a.
(ANY 2 )
1.
Place. condensate demineralizers in service. (0.25) 2.
S tar t S/G bl owdown. (0.25) 3.
Establish main condenser vacuum (which reduces oxygen) (0.25)
( Al l ow 1/2_ credit for. actions performed outside MCR) b.
-350 deg.F-(0.25)
Previously inoperable ECCS Pumps are made operable at this time. (0.5) c.
Initially protection is provided by RHR relief valve. (0.25)
When RHR is. isolated, Cold Over pressure protection actuation of PORV's provides protection (0.25) until temperature increases (above 310 deg's) and normal actuation of PORV's is available. (0.25)
REFERENCE WCGS GEN 00-002, LO 17 321 01, EO9 010000SG13 194001A114
...(KA'S)
(***** CATEGORY 04 CONTINUED ON NEX1 PAGE *****)
A __RBDCEDUREA -NDRet - 6BNDRet : =EldERDENCY_eND PAGE 39 RbD1DLDDlCAL_CDNTBDL
.GUESTION. 4.09 (2.50)
Arrange theifo11owing events in. order of occurrence during.a startup from hot standby to~ minimum load, a.
Block Power Range Low ~ Power trips
]
b.
-Pl ace ~ the Turbine Generator on the line.
~
c.
. Place Main Feedwater Pump Turbine Speed controls in auto.
'd.
Block ~ the Source Range ~ Fl ux Doubl ing Transf er trip.
e.
Start one Main:Feedwater Turbine Pump.
f.
Block the Sourcs Range High Flux' trip.
.g.
Place'the Steam Jump Mode Controller in Tavg mode.
h.
Pl ac e the Main Feedwater Control Valve Bypass Valves in manual and increase Steam Generator level to 55-65%.
' NOTE: Delete. item h prior to future use.
ANSWER 4.09 (2.50) d, f,
e, b,
a, g,
c There~ are six arrangement shifts possible for the worst case.
Each shift
'is wortn 0.417 pts when the minimum number of shifts are made to achieve
~
the correct order.
REFERENCE WCGS GEN 00-003, LO 17'321 02, TO 002000SG13
...(KA'S) 1
- GUESTION 4.10 (2.00) a.
Which critical safety function has the top priority?
'(0.5) b.-
What' is'the purpose of a Functional Restoration Guide?
(0.5) o c.
What-action is required for a red symbol in a status tree?
(1.0) l,
l
.. ~ - -
A __EROCEDURES_=_U0iitieL _eBNDEdeL _EMERGEblC'I_ablQ.
PAGE-40
= red 10LOGICeL_COblIROL ANSWER-4.10 (2.00) a.
'Subcriticality.
(0.5)
- b.. Direct operator action (0.25) to recover / restore the degraded CSF (.25)
(dependent on which CSF.is challenged & extent of degratation),
c.
-Operator should immediately stop EMG (f.5) and start FRG to restore the CSF under extreme challenge (0.5).
(1.0)
REFERENCE WCGS ERGS, LO 17 323 01 000007SG11
...(KA'S)
QUESTION 4.11 (2.50)-
Per the Radiation Protection Manual (WCHP-01) and 10CFR20:
k.-
Every effort should be made to maintain weekly exposure as far as.
. prac t i cabl e bel ow what level?
(0.5) b.
Heal th Physics should be i nformed of any unplanned weekly exposure greater than'what level?
(0.5) c.
What is the 10CFR20 whole body exposure limit, per quarter?
(0.5) d.
TRUE or. FALSE?
Exposures during medical treatment must be reported to Health Physics-for inclusion in dose cal cul at i ons.
( 0. 5)-
e.
TRUE or FALSE?
A "Calendar Quarter" i s defined as any time period consisting of'13 consecutive weeks within a calendar year.
(0.5) l l
_,t-o----
-m-*~
-r--,m----
-er---,
w eer--
wm%m--n-s
=wvw*
-w-e--'
e*--
d.__RROCEDURES_=_NORdeL._eBUDRdeL._EdERGEUCY_6MD PAGE 41
' rad 10 LOG 1CeL_CQUIROL ANSWER 4.11 (2.50) f a.
100 mrem.
I b.
300 mrem.
c.
1.25 rem /qtr. OR 3 rem /qtr w/ NRC-4 not to exceed 5(N-18).
d.
False.
o.
False.
(0.5 EA)
REFERENCE WCGS RPM, LO 10 310 07, 10CFR20 194001K103
...<KA'S) l 1
QUESTION o.12 (2.50) l l
List five conditions which indicate natural circulation flow exists as per EMG ES-03 (SI TERMINATION).
ANSWER 4.12 (2.50) 1.
RCS subcooling based on RCS temp ind - Greater than required per fig.1 or >30 deg.F 2.
S/G pressures - stable or decreasing 3.
RCS hot leg temperature - stable or decreasing 4.
Core exit T/Cs - stable or decrw. sing 5.
RCS cold leg temperatures - at saturation temperature for S/G pressure (0.5 pts each)
REFERENCE WCGS EMG ES-03 ATTACHMENT A, LO 17 323 01, EOS (pg 8) 000009A237
...(KA'S)
QUESTION 4.13 (1.50)
What three methods are prescribed to cool and depressurize a Steam Generator with a tube rupture?
- A.__EROCEDURES_=_SQRueL._eBSORMel+_ EMERGENCY _eUD PAGE 42 RADIOLOGICAL _CONIROL ANSWER 4.13 (1.50) n.
Blow it down to the bl owdown system.
6.
Blow it down to (backfill) the RCS.
c.
Cool down to steam dumps.
(0.5 ea)
REFERENCE WCGS EMG ES-31, 32, & 33, LO 17 323 04 000038K306
...(KA'S) i
w We ar s
,.y bmv M ulgk y
y R
y y
s w
4 N
4 inne: : 5 t.g/f um y
~
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X I
(
4
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40f f 3
,/,ol I, 5 2.01 40 3,01 n. T 1;o d
v.
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1102
/..
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.,7. o 2 a.f -;. a
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.i 1o
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u
/
?
4P5 1-71'
?
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'7D 9M s11 90.'3
1.01 NO CCMGE 1.02 NO CCMBE 1.03 A. ALSO ACCEPT PCM/ PPM [B)
B. 2.
AGO ACCEPT ECR REASON LESS BORON IN 'IHE CORE FOR 'IHE SAME BORON CONCENIRATION. (SEE ATTACHED LESSON PLAN).
4.
'IHE FIRST PART OF 'IHIS ANSWER KEY IS INCORRECT. DIFFERENPIAL BORON NCR'IH (DBR) INCREASES OVER CORE LIFE DUE 'IO BCRON DEPIEfION. DBW DOES NOT INITIALLY DEREASE. SEE ATIACHED GRAPH FRCH NO3S CCRE DATA BOOK, CYCLE 3.
ALSO SEE EXCERPT FRCH REFERENCED LESSCN PLAN.
1.04 10 CmMDE 1.05 NO CCEMENT 1.06 10 CCEMENT NCGS HAS 10 LIMIT 1.07 B.
AG O ACCEPT IDP CHANNEL FAC70RS. (F, FdH).
O ON DNBR. BY OBSERVI1G IDT CHANNEL FAC'IOR LIMITS, NE DEURE
'IHAT DNB DOES NCTP OCCUR. SEE EXCERPT FRCH NOGS TECH SPEC BASES.
1.08 10 CCEMENP 1.09 NO CCEMENT 1.10 10 CCEMDE 1.11 10 CCFRENT 1.12 10 CTMENT 1.13 NO CCEMENT 1.14 10 COMENT 1.15 NO CCEMDE l
1 QUESTION:
2.02 a, b, and d - j, No response c.
TSC - cmputer rocxn in TSC is protected'by halon (#4) therefore i+ cans
- 1 or 4 should be acceptable responses Ref: Fire Protection Iasson Plan, ID 14 086 00 Rev. 000, page 20 or 36.
4 l
i
QUESTION:
2.08 a.
'Ihe amount of flow thru the #1 seal is determined in part by the
'c amount of seal injection flow provided frcm the discharge of the operating CCP(s) therefore this explanation should be accepted as to why seal flow varies.
b.
Normally, the #2 seal injection is r.ot taught as a path for the #1 seal leakoff since the vast majority of #1 seal leakoff is directed to the seal water HS or PRT.
Ref: 10 13 003 00 Rev. 000, lesson Text, page 27 c.
No response
QUESTION:
2.09 a.
No response b.
No response c.
Should accepts discharge head of RCP (A & B) (this could be construed to be understood as part of RCP dP).
l l
l l
l I
1.
. -. =._.
\\
QUESTION: '2.11 a.
No response-b.
2 min interlock retoved c.;
No response d.
Accept "VCr" in lieu of "SWHX" since Excess letdown HX Outlet Valve BG IN 8143 switch is engraved "RCDT/VCT."
Reft Westinghouse bCB RL001 layout.
i I
(
I i
L
QUESTION:
2.12 a.
CRVIS can also be m nually actuated.
Ref: J104-0390-03 b.
No response l
l t
I l
l l
l
t i
CUESTICN: 2.13 a-f.
No_ response
- g.
if interpreted to be seal water return Containment Isolation valves (BG HV 8112 and 8100) then receive close signal frm SIS induced CISA signal.
Ref: M02BG01 h-j.
No response i
QUESTICN: 3.01 "A" is nost correct, however C will generate both - Hi 13.5 peig will generate SI - Hi 217 peig will main steam isolate.
I e
l l
l l
t
b t
GESTION: 3.08 With P EWIS l
l i
l l
l l
l
, - - -, -, - ~, - - -
1 i
QUESTION:
3.09 a.
Under FWIS need to add S/G In-Io level, i
(
l l
i I
r l
l
QUESTION: 3.10 d.
Should accept "Ioss of ficw" trips.
l t
i I
1
(
l t
k
QUESTION:
3.11 1.
Typo should te 4.16 KV need to add "1bumi or alternate bkr closed."
QUESTIm: 4.02
'Ihe response to this question depands on the interpretation of iten #1 (Camon to all Types of RCS Isakage)
Does this mean strictly RCS or does this inply any inventory loss fran any location?
In interpret this to be RCS and hence a knowledge of plant systans would indicate alternate answers. Since the candidate does no',have OFN 00-017 marorized he nust rely on plant knowledge.
'11EREKRE:
Itans d and h in addition to being etched with #1 could also etch up to the #3 and #4.
VCT mkeup or increased charging could be a result of BG systen leakage in the Aux or Containment Building.
Iten c in addition to being mtched with #3 could also etch up to #4 and #5.
RCDT leakage and floor drains as well as various valvesten and misc.
systan leakage could be linal up to the waste hold up tank. Refer to M02's.
~~
C wa I h@
tvo <
U
$(C>
- l-
i QUESTICN 4.03:
ANSNER:
a.
Should be 5 not 1 because the reactor trip for low PRZR pressure is
<1875 psig, see DG E-0, Pg.1 (attached) b.
No response DG E-0, Pg. 2 c.
Should be 2 if reactor trips on high PRZR pressure of 2385 psig, but could also be 5 because DG E-0, Pg.1 (attached) states >2385 psig.
d.
No response DG E-0, Pg.1 e.
No response DG E-0, Pg.1 f.
Should be 1 because 1849 psig is less than the PRZR low pressure trip of 1875 peig. See DG E-0, Pg.1 (attached). SI does not occur until PRZR pressure is <1830 peig per DG E-0, Pg. 2 (attached).
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i QUESTION 4.04:
Alternate escape routes need to include the (1) roll up door 2000' level truck bay, (2) 2000' level thru heat exchanger reczn (6104) thru the northwest door to roczn 1119 of the Aux Building, (3) on 202G' level thru the switchgear roczn and the IWAC Roczn to the Aux Building thru the northwest door.
(See attached drawings) l
QUESTION 4.05:
AlENI!R A.
DG E-3, Pg.17, Step 20a (attached) states:
1.
RCS Subcooling:
Based on RCS tEUperature (Curves USe CoIB exit Es and Rtd's, not Core exit E s only).
Figure 1 should be per zero subcooling curves.
Another possible answer ist based on subcooling nargin nonitor -
greater than 300F subcooling.
2.
Secondary Heat Sink Total feed flow should be 320,000 lbn/hr instead of 300,000 lhn/hr.
Narrow range level should be 4% (28% for adverse containment) instead of 5%.
B.
DC E-3 foldout page 39 (attached) 1.
RCS Subcooling See response to A.1 above.
2.
PZR level Should be greater than 4% (28% for adverse containment) instead of 5%.
QUESTION 4.06:
r It s a may have additional responses i.e.:
Page 8 of Om 00-015 (attached)
Step 10.e. Increase charging and letdown Step 10.f.. Increase A m flow It s b may have additional responses Refer to page 11 & 12 of Om 00-015 (attached)
Step 13.d. Containment Cooling max Step 13.e. Proper level in cavity Step 13.1. Attspt to restore RHR t
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QUESTION 4.08:
MENER:
a.
No response verified per steps 4.14, 4.16, 4.17, and 4.19.7 of GEN 00-002, Rev. 12 -(attached).
b.
See steps 4.36 of GEN 00-002, Rev.12 (attached).
If this question is interpreted as things related to the mode change an additional answer of - Raoving RHR fran service - could have been added by the student. See Procaution 2.2.1 of GEN 00-002 (attached),
c.
Initial protection includes: RHR relief, COPPS, PZR bubble and letdown. When RHR isolated protection is provided by COPPS, PZR bubble and L/D tmtil taperature is above 3680F (Step 4.37). Above 3680F protection is provided by nonnal operation of the PZR PORVs, PZR 1rbble and letdown. See steps 3.1, 3.2, 3.4, 3.6, 3.15, 4:22, 4.23.1, 4.23.2, 4.34, 4.35, 4.37, 4.37.1, and 4.44 of GEN 00-002, (attached).
QUESTION 4.09:
'Ihe following answer is correct per GEN 00-003, Rev.16, (attached).
Note that it s h should be excluded because GEN 00-003 does not have such a step. 'Ihe steps which allude to SG level control are 3.14 and 4.25.4 which place the bypasses in auto.
GEN Page:
9, 11, 13, 15, 16, 16, 18 GEN Step: 4.5 4.20 4.25 4.32 4.35 4.37 4.42 Test Its: d f
e b
a g
c
QUESTION: 4.10:.
NEliMER:
a.
_No response b.
No response c.
Our people will probably use DG rather than ORG.
'Ihey my also qualify their response (which should etch the test key answer) by starting the FRs are nonitored only in DG C-0, and not implanented until the procedure directs than to do so. See attached pages of DGs C-0, CS-01, and CS-02.
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QUESTICN 4.12:
Att9ER:
1.
Should read based on RCS tgmture '(not core exit 'ICs) greater than required per zero subcooling curves, or Based on subcooling margin nonitor - greater than 300F subcooling.
See Attachnent D to DG ES-03 (attached).
2.
No response 3.
No response 4.
No response 5.
No response F
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