ML082730784
| ML082730784 | |
| Person / Time | |
|---|---|
| Site: | Salem  |
| Issue date: | 07/02/2008 |
| From: | Public Service Enterprise Group |
| To: | Brian Haagensen Operations Branch I |
| Hansell S | |
| Shared Package | |
| ML080030005 | List: |
| References | |
| 2008-301 | |
| Download: ML082730784 (129) | |
Text
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~~~ istrati0 n Exam Level K4 MaterialRequiredfo rExamination Exam section R 000009KI 02 Q3 Steam Tables 1 OOWEI 1K202 Q25 1-EOP-LOCA-5 flowchart 1 005000K509 Q31 S1.RE-RA.ZZ-0016 Curve Book 2 1940016223 Q67 Tech Spec 3.7 1.2 3 194001G450 Q75 A-5-500-EEE-1686 Rev 8, Tables and Unit 2' Curves. 3 Friday, June 27,2008 Page I of 1
Question Source-RO Question Source Modification Method RO Number Facility Exam Bank Concept Used 3 Facility Exam Bank Direct From Source 8 Facility Exam Bank Editorially Modified IO Facility Exam Bank Significantly Modified 2 New 43 Other Facility Concept Used 2 Other Facility Direct From Source 2 Other Facility Editorially Modified 2 Previous 2 NRC Exams Direct From Source 2 Previous 2 NRC Exams Editorially Modified 1
RO Cognitive Level Cognitive Level Number of Questions Application 28 Comp rehension 12 Memory 35 Friday, June 27, 2008 7:31:32 AM
RO Answer Distribution Answer Nunrher of Qiiestinns a 19 b 21 C 17 d 18 Friday, June 27, 2008 7:31:39 AM
r Crop1'1 r&=s. k-ctI C 7 p&oy U.S. Nuclear Regulatory Commission c*nvd
""7 T)c S ite-Specific IqCclb Written Examination LA 1I Applicant Information l/j Name:
Date: 8/25/2008 I Region: I
' Facility: Salem 1 & 2
)ILicense Level: RO 1 ReactorType: W I1 Time:
1 Finish Time:
Instructions Use the answer sheets provided to document your answers. Staple this cover sheet on top of the answer sheets. The passing grade requires a final grade of at least 80.00 percent. Examination papers will be collected SIX hours after the examination starts.
Applicant Certification All work done on this examination is my own. I have neither given nor received aid.
Applicant's Signature II IF- Results
~-
Examination Value Points Applicant's Score Points
)II-I Applicant's Grade Percent I'
,Given the following conditions:
1- Unit 2 is operating at 95% power.
1- Control Band "D" rods are at 215 steps.
Rod Control is in AUTO.
1-Which of the following describes how the plant will be affected if PT-505, Turbine Steamline inlet Pressure, fails HIGH with NO operator action?
iThe reactor will trip on Over Power D/T. -~
~
/The reactor will t r b on Over TernDerature D/T.
FDll control bank rod motion will stop at Group Demand Counter indication of 227 steps (ARO)
From
_ Control
_ ~ Grade
_ ~interlock C-I I , and the Rx will NOT trip.
~~~~
_-__ ~-
bank rod motion will stop at Group Demand Counter indication of 227 steps (ARO)
Grade Interlock C-2, and the Rx will NOT trip.
/Continuous Rod Withdrawal I
r---
i 1 R u e s d a y . July 15, 2008 12:27:44 PM I _ 1_ of88 Page 1
Given the following condition:
1- Unit 2 is operating normally at 100% power when the unit is manually tripped.
Compared to their pre-trip values, which of the following indications will be present one minute following the uncomplicated Rx trip? _
__ _______ _______ ~ -_
[PZR leveF45% and lowering.
_ _ _ _ ~ ______
[Seal Injection flow to all RCPs has lowered.
~_ __
[Charging System flow has ~ lowered
_ _ _ to _ 65_ gpm.
~ -
@71, Letdown-- HX CC flow control valve demand~- has lowered.
R ! Application /Salem I& 2
- 1 , 8125/200(
n v a n d Abnormal Plant Evolutions ___ It WOTkK 7 2 1
~~
[El1- Abllltyo operate and I __
~~ or monitor t h e~ g_ t_ m_ R
_ _p _ _:
~ _ - - I-I EA1.091 CVCS IS incorrect because the initial shrink following the trip will reduce PZR level well below 45%(
s incorrect because since PZR level will still be above the programmed level derived from the I auctioneered Thot, charging flow will be lower due the charging system Master Flow Controller demand lowering, and driving charging flow lower. The CV71 valve will not have changed position, and backpressure to the seals will remain the same, so seal injection flow will remain essentially constant.
IC is correct. PZR level program will lower in response to the Auct hi Tave signal creating reference setpoint lowering from full power Tave. D is incorrect because normal letdown come from 23 loop cold lleg. The cold leg temperature will RISE following a Rc trip as SG pressure rises. The letdown
,temperature will RISE, and the Letdown HX CCW flow will need to RISE to maintain temperature. This will cause CC71 valve demand to RISE. When verified in the simulator, the rise in valve demand was lverv small. 1%. But it will definitelv NOT lower.
~
1 1 CVCSOOE015
________ LOR NCT Given plant conditions, relate the Chemical and Volume Control System with the following, Pressunzer Level Control System RCS Temperature Control Main TurbinelGenerator Reactor Coolant Pump seal injection flows Automatic Control Rod Control VCT Makeup Nuclear Instrumentation Emergency Core Cooling System Residual Heat Removal System Component Cooling Water System Pressurizer Pressure Control System Pressurizer including Pressure Relief Tank Waste Gas Waste Liquid Service Water 4 Kv Vital AC System 480 V Vital AC System 240 V Vital AC System 125 VDC System _ _
-_ - _ - _ _ ~- ~
p u e s d a y , July 15, 2008 12:27:44 PM I Page2of88
___ 1
7-
, Tuesdav, Julv 15, 2008 12:27:44 PM
IGiven the following conditions:
Unit 1 expericenced a SBLOCA.
A manual Rx trip and Safety Injection were initiated.
RCS pressure is 1085 psig.
The hottest CET is 554.0 degrees.
Operators are determining whether conditions are present to allow a transition to EOP-TRIP-3, SI Termination, at Step 9, SI Flow Reduction Criteria.
Due to a concern with the indication of the Subcooling Margin Monitor, the CRS asks the RO to determine RCS subcooling using Steam Tables.
I I
Which of the following identifies current RCS subcooling, and whether the transition to TRIP-3 is appropriate? (Assume all other conditions required to make the transition are SAT.)
subc cooling is degrees, and the transition to TRIP-3 be made. _ -
mk.25, Should NOT.
. - __ ._ __ __ - - -- -- - . . - - jo~oo.oo.gKlb.2.......~1 Emergency and Abnormal Plant Evolutions 1 1 1009 Ismall Break LOCA ~ ~
IEKI. IKnowledge of the operational implications of the following concepts as they apply to Small
__ ~~ ~_ ~ _ - . Break LOCA:
r- ____
lEK1.Oa_____ Use of steam tables
~--
_._______~___-
-3.5,4.2 1085 psig equals GOO psia. Saturation temperature for II 0 0 psia is 556.28 degrees per table 2 of t h e 7
!Steam Tables. This would indicate that subcooling is 2.25 degrees, and the criteria of >O degrees 1 subcooling is met, and the transition is warranted. If the candidate uses 1085, the subcooling would be very close to zero. . -. .- . .. . . . . ... . ..
I
_ _ ~ ~ - ~ _ . _ _ _ ~ I 1
L LOCAOIEOI1 A. Determine a discrete path through the EOP B. Determine an appropriate transition out of the EOP ________
(Tuesdav, Julv 15, 2008 12:27:44 PM- I Page4of88
/Salem Unit 2 has experienced a rupture of a RCS cold leg which has resulted in containment
/pressure peaking at 18 psig.
IWith all systems actuating as expected, which of the following choices identifies the containment iisolations which have occurred. and the reason whv thev have occurred?
I
~~
Phase A to ensure non-essential containment penetrations are isolated; Feedwater to isolate ALL feedwater to containment to Dreclude an excessive RCS cooldown event.
... ~~~ ~~ ~~~~~~~~~~~~~~~~~~~~~~~~~~~~~~~ ~~~~
[Phase B to isolate potential injection paths to containment; Containment Ventilation to ensure inon-essential containment ventilation Penetrations are isolated.
~ ~- ~
IMain Steamline to minimize potential primary-to-secondary-leakage; Feedwater to prevent luncontrolled fillina of anv SG.
phase A to ensure non-essential containment penetrations are isolated; Phase B to isolate
,additional potential release paths from containment.
_______ _____ ~~~ ~~
IEmergency and Abnormal Plant Evolutions F
~
-A -
[000011A107 -
I
' [Large Break LOCA
~~
- - - -- I I 4 IEAI. IAbility to operate and / or monitor the following as they apply to Large Break- LOCA:
-~
~~
~- ~ _ _ _ _ _ - ~
lEA1.04 __________
Containment isolation system ~~
_ _ _ _ _ ~
x 4 . 4 1 4 . 4 All of the isolations in the choices above do occur during a LOCA in which containment pressure goes above 15 psig. Only the correct answer d contains the correct reasons for its respective isolations. I Distractor a Feedwater Isolation only isolates Main Feedwater, it does not isolate ALL feedwater. AFW is still available for injection to SG's. Distractor b is incorrect because Phase B isolates leakage paths, not injection paths. Distractor c is incorrect because Main Steamline Isolation is designed to minimize and/or terminate the mass and energy releases associated with a high energy secondary line break.
~
~ -~~
~
~-
>*. L .
_L__L_ _ Y I __I - -..&..<zu-* .".."_ I A '__-_-
LOCAOlE007 Identify possible radioactivity release paths for a Loss of Cool the potential for a release I
r i u e s d a v , July 15, 2008 12:27:44 PM- 1 Page5of88 j
!With Unit 1 operating at 100% power, which of the following will cause RCP Standpipe level(s) to ke? ~
IFailure_of -a RCP ~ ~ #I _ Seal.
_ _ _ _ _ ~ ~ ~ _ _ _ _
12CV55;Charging FIOW Control Valve, fails _open. _ _ ~ ~ -
5 7 7 5 a l Pressure Control Valve, fails closed.
lant Pump Malfunctions
____ __ ~ ~ ~ _ _ _
b4A2.:.abbility to determine and interpret the following as they apply to Reactor Coolant Pump Malfunctions: -~
7 1~2.01 I
Cause of RCP failure ~ ~ ~~~~ ~ ~ _ _ _ _ _ _ _ _ ~ _ _ _
143.5*
(B) Level would lower; (D)2CV71 regulates flow to seals (C) More charging into RCS not through #I IA Correct - more flow throuah #2 Seal Y
I IReactor Coolant --Pump Abnormality ~ _ _ _ _
~Reactor
_ _ _ _ _ _Coolant
__-- Pump operationd ABRCPI EO01 - _
Describe the oDeration of the followina svstem as amlied
.. to S2.OP-AB.RCP-0001:
a) Basic RCP Construction b) Seal Injection and Seal Water Configuration
__c) RCP CW Configuration _ _ _ ~ _ _ _
F T u e s d a v , Julv 15. 2008 12:27:44 PM- 1 Page6of88
-~
1
/Giventhe following conditions:
Unit 1 is in MODE 6.
11 RHR loop is in service providing shutdown cooling.
12 RHR loop is in standby aligned for shutdown cooling.
RHR HX inlet temperature is 110 degrees.
The first fuel assembly has just been transferred to the Spent Fuel Pool.
11 RHR pump indications lead operators to believe it is cavitating, and the CRS orders the pump stopped.
The CRS makes the decision that local venting and normal restoration of the RHR system can NOT be completed prior to core boiling IAW S I .OP-AB.RHR-0001, Loss of RHR, and directs the RO to start 12 RHR pump.
I which of the following identifies how the 12 RHW pump will be operated, and why?
112RHR pump will be started with the _12RH18,
-_ ~-
_ _ _ _ _ _ ~ ~ ~-
lfull open and 1 RH20 full oDen to sweeD air from the RHR svstem into the vented RCS.
lull openand 1RH20, RHR HX Bypass, full shut to maximize cooling to RCS and prevent an lunwanted MODE change. - .__.__~___________
__ ____ ~ ~-
khut and 1RH20 shut.-T2RHl8 and 1RH20 will then be opened to establish full loop flow to isweep air~- from
__ the RHR system into the- vented ~- - RCS.
-____~____--
shut and 1RH20 open. 12RH18 and 1RH20 will then be adjusted to establish the maximum
!flow attainable while maintaining e25 deg/hr cooldown rate.
~
R , IFGjmIi2- J 8/25/2008 lfull flow restoration. It states to SHUT the RH18 and SHUT the RH20 prior to starting the pump. Then
'the operator is directed to operate the RH18 AND the RH20 to maintain flow to the RCS. Even when full Iflow for gas sweeping is required, the centrifugal pump is still started with its discharge valve shut. The RH29 recirc valve will open to maintain 500 gpm flow upon initial start. As per theebases document for AB.RHR-0001, page 6, full flow will be established, normal conditions verified, then flow will be reduced to 11800-3000 aDm.
ABRHRlE004 Describe, in general terms, the actio; taken in S2 OP-AB RHR-0001 and the bases for the actions in accordance with the Techni 1
~~
Tuesday, July
-- 15, 2008 12:27:44 PM 1 Page7of88
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.I 1
_ _. . . .... ....... .-.- .II_. ." ... " ". . ..."_..,--.. -....- L /
.t . I .
n u e s d a y , July 15, 2008 12:27:44 PM , I Paae 8 o f 8 8 1
IGiven the following conditions:
Unit 1 is operating at 100% power, steady state, with no surveillance procedures or testing in progress.
The peak outdoor temperature has exceeded 95 degrees for the past 7 days.
Service Water system problems combined with the high ambient temperature has caused Component Cooling Water system temperatures to rise.
The unit CRS is attempting to reduce CCW system flows.
which of the following components flows, if it was able to be reduced by HALF, would have the
,greatest impact on CCW system flow?
IAssume current flow could actuallv be reduced.
H~RHRHX.
~~ ~
ILetdown HX.
~_
[Spent Fuel Pit HX. ~ ~~ ~ -
~~ ~~
LRCP Thermal BarrierHX.
IM1. w t y to operate and __ / or monitor the following as they apply to Loss of Component Cooling Water:
- ~~~ ____ ____ - -.- -- ___ - ___ I k 0 7 , Flow rates to the components and systems that are serviced by the CCWS; interactions among the '-2.91 -3.0 comDonents i
_ 1 L i n c o r r e c t because with the conditions stated in the stem, there would be no CCW flow through the I RHR HX. B is incorrect because Letdown HX normal flow is 1,000 gpm. C is correct because Spent LFuelPool normal flow is 3,000. RCP thermal barrier normal flow is 40 gpm.
~ ~ ~~ ~~
___CCW System Operation - -_ ~ _ _ _ _ ~ -
~~
CCW System Lesson~- Plan - _ _ ~ _ _ _ _
~
I a) General arrangement of the Component Cooling Water system.
b) CC System loads c) d)
Making up to the CC Surge Tank Safety
- _ precautions for handling/worktng
_ _ ~
~ _
- ~
with chromated systems.
___ _ _ ~ ~ ___-_ _ -- -~ ~
1 Tuesdav. Julv 15. 2008 12:27:44 PM I Paqe9of88 1
/Given the following conditions:
1- Unit 2 is in MODE 4.
1- RCS pressure is 350 psig.
~- All wide range cold leg temperatures are 310°F.
1- Pressurizer Overpressure Protection System is ARMED.
IPredict the plant response to RCS wide range pressure transmitter 2PT-405 failing high with NO Joperatoraction?
1 - --
_ _ - _ _ _ _ - - _ _ _ _ _ _ _ _ _ ~ . _ _ ~ -
IOnlv PORV 2PR1 would oDen.
h t h PORVs 2PR1 and 2PR2 would oDen.
- - _ _ ~ - . ~ ~ _ _ _ _ _ ________
INeither PORV 2PR1 nor2PR2 would oDen.
_Emergency
_ _ and
~ Abnormal Plant Evolutions [ZT 1
~
.~
1000027K203 1 I
1027-- 1 IPressurizer Pressure Control Malfunction ~ -~ I
~ 81
_ _ _ _ _ - ~ - - ~ ~ . _ _ _ _ _ _ _ _ _ _ _ . ~ _ _ _ _ _ _ - _ _ ~
k 2 . [Knowledge of the interrelations between Pressurizer Pressure-Control-Malfunction - and the following: I j~2PT405feeds POPS channel I, and 2PT-403 feeds POPS Channel II. Fauilure of PT-405 would cause PR1 onlv to oDen.
1 NO. 1&2 Units PZR over press prot sys CH I control and indications I
1 ABPZRI EO01 Describe operation of the Pressurizer Pressure control system as applied to S2.OP-AB.PZR-0001(a).
_.__________ ~-
I Tuesday, July 15, 2008 12:27:44 PM -
IGiven the following conditions:
1- A 650 gpm tube rupture has occurred on 22 SG while operating at 100% power
~- 22 SG NR level rose to 93% before SI could be terminated.
1- 22 SG NR level is currently 89% and dropping slowly.
IWhich of the following describes how the CRS is allowed to utilize 22 SG during performance of
,SGTR-2, Post SGTR Cooldown?
lThe
-- CRS is ... ._-__
tallowed to steam 22 SG as long as its NR level remains ~ 9 2 % .
___ - - ~ ____________~__
\NOT allowed to steam 22 SG to prevent water ____ hammer in the steam line.
~ _ _ ~ __ ~- _______ -
allowed to steam 22 SG ONLY with TSC approval, AND an acceptable release rate calculation
_Emergency
___ ~ and Abnormal Plant Evolutions _____
~ ~-
team Generator Tube Leak b 2 . IAbility to___determine and interpret the following as they apply to Steam Generator Tube Leak: ,
- - _ _ _ ~ _ _ ~ _ _ -~
14A2.14 Actions to be taken if S/G goes solidand water enters steam lines ___ -____.-
1244.4
- -~ -.-
ithe ruptured SG to prevent water hammer in the affected SG steam line, it is NOT prohibited under all icircumstances. C is incorrect because 10CFR20 limits, while they always apply to the release from a ,
Post SGTR Cooldown - - ~~ -
~ . . .. . . .. .- ..... . .
" 2'
.. ..\ ..... 3.. ....... :. .
. -~.."
-Ai30
............ L. ... 1.
I SGrR02E006 Describe the basis for each step, caution, note, and Continuous Action Summary item in EOP-SGTR-2
[1 ~ ~ ~ 1
--Tuesday, July 15, 2008 12:27:44 PM I Page 11 of 88
- _________ ________ - ~ _ _
~- ~-
IWhen responding to a large SGTR, which of the following identifies an Operator Action time, and lfhe-reason for it? - - ~ _ _ _ _ _
shut the SG Blowdown Isolation Valves, GB4s, within 10 minutes to limit the spread of 1 contamination.
_ _ _ _ _ _ ~ _ _ _____
[Reestablish letdown with 45 minutes to prevent PZR overfill and water relief through lPORVS/Safeties.
-~
\Isolate feedwater into and steam flow out of a ruptured SG within 10 minutes to minimize the hoss of mass from the RCS.
/Terminate Safety Injection flow by isolating the BIT within 50 minutes to prevent SG overfill land potential introduction of water into the steamlines.
and Abnormal Plant Evolutions Generator Tube Rupture
__- -~ __ _ _ ~ _ _ _ _ _ _ _ _ _
IEK3. m d g_-__ e of the reasons for the following responses as they apply to Steam Generator _______ Tube Rupture: 1 L-
_ _ _ _ _ _ . _ _ _ _ ~ _ _ _ _ _ ~ _ ____ _____-__~ __
EK3.061 Actions contained in EOP for RCS water inventory balance, S/G tube rupture, and plant shutdown 1 1-4.2, 4.5 Drocedures I A is incorrect because although the action and reason is correct,there is no time associated with it. B is 1 incorrect because it is the action, time, and reason to terminate an Inadvertent SI. C is incorrect because 1 lthe action is correct but the reason is tp prevent SG overfill. D is correct because FSAR takes credit for 1
~ 5 0minute isolation. I 1 SGTROlE007 Describe the basis for each step, caution, note, and Continuous Action Summary item in 2-EOP-SGTR-1
)-%esday, July 15, 2008 12:27:44 PM I Page 1 2 o f 8 8
~~
1
-.___________ _ _ _ _ _ _ _ ~ _ ________ _ - ____
I
~~~
Which of the following describes the reason why steam dumps are blocked from opening on a condenser when vacuum lowers to 20" vacuum?
'Steam entering -- the condensers with low____-____ vacuum ~ causes..
L, -____
(substantialpitting of the condenser tubes. ____-_______
[uneven heatina and 'Dremature L v -
failure--__of the LP turbine exhaust hood. ______
__ - - _ _ - ~ _ _ __-___
vacuum to degrade further causing a reduction ~ - ~ _ of_ NPSH to-the
- - _ _ ~ ~ condensate
_ _ _ _ pumps. __ ~. .--
-~ -
[condensate depression to lower to zero, and flashing in the condensate system will occur.
R I and Abnormal Plant Evolutions o E o f Condenser Vacuum r 11)
IAK3.- [Knowledge of the reasons for the following _ _ responses
_ _ _ _ _ _as__ they apply to Loss of-~COIidenser Vacuum:
-~_-______ ____ ___-
la 3.1*
L - -
1 __-~-
lAK3.01 I 'Loss of steam dump capability upon loss of condenser vacuum
____ ~~
IA is incorrect because it will not cause substantial pitting. C is correct because as condensing rate goes Idown in the condenser, the condensate will rise in temperature and lead to a lowering of NPSH to pondensate pump suction. This is why AB.COND has operatoers monitor cond pump suction temp
'because it is expected to rise. D is incorrect because it will not lower to zero. B is incorrect because while exhaust hood boot temp may rise, it is not the reason for stomina steam flow I I \ c, -.". . x
<.', .*"
- g .Y
&milag
_ I *^_ILL -.. A uL.. B I ^- ___-- -.. A I _-lbLblUi_*..IXI^lA_..' I__
ABCONDE004 Describe, in qeneral temsrrhe actions taken in S2 OP-AB COND-0001 and the bases for the actions in accordance with the Technical Bases Document.
I STDUMPEOOG Turbine Bypass Spray/Steam Dump System _ _ ~ ~ __- ~
I - 7 1 Tuesday, July 15, 2008 12:27:44 PM 1 Page 130f 88
__-- ~-
IGiven the following conditions:
~- Unit 1 is operating at 100% power.
I-I-
~-
11 AFW pp is C/T.
The unit trips and auto SI is actuated due to an unisolable Main Steamline rupture on 11 SG.
12 AFW pump did not start when demanded, and can NOT be started.
13 AFW pump tripped 2 minutes after starting, and can NOT be immediately reset.
1- 11 SG has blown dry.
(Whichof the following describes the mitigation strategy when attempting to feed a SG with the icondensate system? - ~ - - _ _ _ _
_--____-_-~____. - _ _ - _ _ ~ _ _ _ _ ---_______
~ - _ _ _ _
rate as soon as the required lineup is established until CETs are
.--_____._-__. ~
pressure is at 0 psig- and feed can be rapidly established
_ _ _ - - ~ - _ _ _ - _ _ _ _ to_ it.
_ _ _ ~ _
Feed 11 SG at a rate between 1-5 E4 lbmlhr as soon as the required lineup is established,
,since its pressure is at 0 psiq and feed can be rapidly established to it.
Iselect ANY SG other than 11 to depressurize and feed since feeding 11 SG could cause a ltube leak or rupture.
'Select 12 or 14 SG to depressurize and feed since feeding 11 SG could cause a tube leak or I rupture. _ _ ~ _ _ ~__--._______-.~--______ _ _ _ _ _ _ _ ~ ~
9 Emergencyand
__ Abnormal Plant Evolutions~-
1054 1 /Loss of Main Feedwater
-~-- - ~- ~-__-_____
lAK1.- IKnowledge of the operational implications of the following concepts as they apply to Loss of Main I Eeedwa
_____ ter: I
__- ~ _ _ _ .__~____
IAKI
_ _ .O
~g_Effects of feedwater introduction on drv
~
S/G I_--
77a4.2 - - 1 2 of FRHS-1 states that if at least one intact or ruptured SG is available, then doi not feed a faulted
/SG. The ERG basis states that the thermal shock of feeding a dry SG could cause a tube leak or rupture Ithat would be unable to be isolated until the secondary boundary was restored. This is why 11 SG will NOT be fed, and A and B are incorrect. C is incorrect because the procedure states that if other SG are lavailable, then 11 and 13 SGs should be steamed last to maximize steam supply for TDAFW pump. The (stem states that the 13 TDAFW pp cannot be immediately reset, which infers that it may be able to be ireset, so conserving inventory in 13 SG is correct. That leaves 12 or 14 SG to be selected to idepressurize.
~- -
[-Tuesday,
-~ July
~ 15, _ 12:27:44 PM
_ 2008 _.
1 Page14of88-
IWhich of the following describes the basis for why Functional Restoration Procedures (FRPs) are h T implemented until directed in EOP-LOPA-1, Loss of All AC Power?
_ _ _ _ _ _ _ _ ~ . ~ _ _ ~ _
ALL FRPs are written on the premise that at least one 4KV vital bus is energized.
--____- ~ _____-____.___ - ___________~____
LOPA-I actions must be performed in sequence. FRPs interrupt the sequence and timing of steps.
-- ~- __-_____- ________
LOPA-I includes all key actions of RED path FRPs so performing FRPs would be redundant land
I prolong the time until RCS depressurization was performed. --
. _ _ _ _ _ _ _ _ _ _ ~ _ ~ _ _ _ _ _ _ _ _ _ _ . _ ~ _ _ _ _
hince only one FRP (FRHS) could affect plant operation due to operation of the TDAFW Ipump, ALL FRPs are not~implemented _ _ _ _ to preclude
~ _ confusion over their implementation.
I IEmergency
_____Procedures / Plan _ _ _ _ ~ . . _ _ ~ _ -________ _ _ _ _
d 1 Knowledge of the bases for prioritizing emergency procedure implementation during emergency 12.81, 3.8 operations.
____- ______ 1 I
IFRPs are written under the asumption that at least 1 4KV vital bus is energized to provide power for icontrolling equipment to provide mitigating functions for the loss of power. All 3 distracters contain i incorrect reasons for not implementing FRPs
_______.___________~______ . ~ - _ _ - _ . _ _ _ ____
B. Loss of all AC power
- 1. Describe the analysis assumptions.
- 2. Describe the protective features that mitigate the event (N/A for a loss of all AC power).
- 3. Describe the analyzed plant response.
- 4. State whether the analysis indicates fuel damage and, if so, describe the expected fuel failure mechanism--
___-__~
-I 7-Tuesday, July 15,2008 12:27:44 PM 1 P a a e 1 5 of 88 I
- Off site power has not been restored 15 minutes after loss.
~
'150' Elevation Wind Speed - 15 mph.
~ _ _ _ _ _ _ _ ~ ~ _ _ _ . _ _ _ _ _ _ . _ _ _ _ _ _ _ _ _ _ _
r--
IMain Steam Dump valve position - 8%.
Containment Averaqe Temperature - I 15 deqrees.
to Loss of Off-Site Power:
IAA2.82 Temperatures displayed on plant computer CRT
_ _ _ _ _ _ ~ ~ - monitor
~ _ _
Containment average temperature of 105 is a reasonable reading for August, and is displayed on the P- '
250 computer. Main steam dump valves will be blocked closed because of the loss of Group bus power 1 lto Circulating Water pumps. Meteorlogical tower data is not displayed on the Unit 2 P-250. RCS loop Tc's in the loops will be at saturation temperature for the SGs at 1015 psig because of natural circulation 1 iflow in the IOODS. P-250 is Dowered from Hope creek substation # I and will remain powered UD since 1
[stem states _ .Hope
___ Creek
___ remains
~ _ _ at power, and
_ __ _____ would have tripped- ~ if_ it lost
_ _ off-site power also. '
--___ I 2-EOP-TRIP-2 1~P25000E008
_ _ Identify
_ and describe the Control Room controls, indications, and alarms associated with the P-250 Computer, including:
- a. The Control Room locations of the P-250 Computer terminals (N/A NEO)
- b. The function of each P-250 Computer Control Room control and indication
- 1) USE all of the function buttons provided at the top of the process diagram windows
- 2) ACCESS the System Status Diagram
- 3) DIAGNOSE the status of the system using the functions provided by the System Status Diagram
- 4) EXPLAIN the significance of having a drop 254 on the highway
- 5) ACCESS the Base Alarm System
- 6) DEFINE alarm pnorities and EXPLAIN how to distinguish them on the display
- 7) DEFINE point Quality and DETERMINE the quality of any points displayed on the screen
- 8) MODIFY the alarm screen to change between displaying a current alarm list, an alarm history and a list of unacknowlec alarms
- 9) USE the alarm filtenng capabilities to focus alarm displays on user definable parameters IO) DEFINE the differences between analog and digital point records
- 11) DISPLAY point information on desired points by using at least 3 methods
- 12) EXPLAIN all of the information being displayed in the Point Information window when in the Reduced mode of display
- 13) PERFORM a Point Search to find groups of points meeting user define attributes
- 14) EXPLAIN the purpose of the Trend package
- 15) BUILD a Mini Trend
- 16) DISPLAY a trend group
- 17) DESCRIBE the 5 types of trend layouts
- 18) BUILD and MODIFY multiple point trends, including adding shading (color)
- 19) BUILD and MODIFY trend groups
- 20) VIEW point values at various points on a trend
- 21) DISPLAY a tabular trend
- 22) DESCRIBE the difference between Live and Historical trends
- 23) MANEUVER through the customized Salem Station process diagrams to display desired plant information
-Tuesday, 1 - July 15,2008 12:27:44 PM - 1 Page16of88
~~~~~ ~ ~- -- - ~~~~ - ~~ -. .__
- c. The effect each P-250 Computer control has upon P-250 Computer components and operation (N/A)
- d. The plant conditions or permissives required for the P-250 Computer Control Room controls to perform their intended function (N/A)
-e. The setpoints associated with the P-250 Computer control room alarms. (NIA)
~~
T T u e s d a y , July 15, 2008 12:27:44 PM I Page 1 7 o f 8 8
~~
/Giventhe following conditions:
~- Unit 2 is operating at 100% power.
1- OHA B-18,ZC 125VDC CNTRL BUS VOLT LO, annunciates in the control room.
1- Operators identify that the 2C 125VDC bus is deenergized.
IWhich of the following identifies why the ARP directs operators NOT to transfer Vital Busses and
\Distribution Cabinets to emergency DC power?
iwould
_____-__ lead to an additional plant excusion. - ~ _ . _ _ _ _ _ _ _ _ _ _ _ _ _ ~ ~~
k single failure could result in cross connecting 2 vital power su-quired-
[mainta i-_ ned separate. __ - ___ _ _ _ _ _ _ _ ~ - -
the remaining energized 125VDC busses from theswapped
- . - -- - - components.- .
cy and Abnormal Plant Evolutions __ 1 r 3 ioo0058~302 I, n
1058- , Loss of DC Power __I
!AmL[Knowledgeof the reasons for thefollowing responses as they apply to Loss of DCPower:
I (AK304 Actions contained in EOP for loss of dc - power ~ - 1- 4.2 k h e AC power supply for the battery chargers normally in service to the DC busses comee from the 3 vital' lbusses. If DC control power is transferred from deenergized 125VDC bus to its alternate, then the loperability of the AC bus is affected, because a single failure could cause 2 AC busses to be x-connected I and violate the seperation requirement for Vital AC power. A is incorrect because while it may be true Ithat re-energizing loads may cause compoonent operation, it is not the reson for the precaution. C is I lincorrect because the DC busses are normally supplied from their respective chargewrs and have very 1 llow loading associated with them under normal conditions. D is incorrect because the availability of DC 1
/power is not the same as energizing the trip coil, which would not happen.
~-
1 1
DCELECE014
- - Given a DC Electrical System failure, predict the effect of the DC Electncal System failure on the following: (License Operator anc STA only)
Emergency Diesel Generators Components using DC control power _________
___ __ - _~~__ ___~_- _ _ _ _ _
r --
i G e s d a y , July 15, 2008 12:27:44 PM 1 Page18of88 1
/Whichof the following describes the primary reason why an Accidental Radiological Liquid Release Ithat exceeds the ODCM limit by a factor of 2 is given a 60 minute time limit BEFORE an Unusual IEvent is required to be declared?
____ ~~~~
p t the 60 minute point, the integrated dose will exceed the federal limit. . _ _ _ ~
_ _ - ~
i f the release cannot be terminated in 60 minutes, it indicates that plant control is in a degraded status.
- - ~
I
~~
60 minutes allows sufficient time for re-sampling and verification of initial results following identification of the release.
~_
IThe 60 minute limit allows for pre-emptive coordination of Emergency Plan response to ensure
/the proper notifications are made if the UE has to be declared.
Emergency and Abnormal Plant Evolutions 7- 2 2 i000059K30T 1059-7 h G e n t a l Liquid Radwaste Release >
I
~~
L__
~- 16' lAK3. /Knowledge of the reasons for the following responses as they apply to Accidental Liquid Radwaste Release
___~_ - ~ ~ - ____ - - __ ~~~~~
~~
lAK3.021 Implementation of E-plan ~ 1-3.2*i 4 5
/F==t h Technical Basis Document for EAL 6.2.1, UE for Liquid Effluent Release, ..."The final-lintegrated dose is very low and is not the primary concern. Rather it is the degradation in plant control
!implied by the fact that the release was not isolated within 60 minutes." A is incorrect because of above, I but contains wordage contained in the basis. C is incorrect because while there might be time to re-
!sample following initial sample results, it is not the reason for the 60 minute time. While prompt inotification of responsible agencies is always of utmost concern, allowing a condition to exist longer solely~
Ito allow prepaation for notifications is not correct.
L - .-~- _ _ _ ~ ~~~ -~ I Salem ECG Technical Basis Document
~ ~~ ~ _~
i EL0 29.e Demonstrate a working knowledae of the bases for erneraencv Drocedure reauirements.
7-Tuesday, Julv 15, 2008 12:27:44 PM 7
1 Page19of88 ~
Q ,?'
L -
b v e n the following conditions:
- Unit 2 is in MODE 4.
1- 21 RHR loop is in service providing shutdown cooling.
~- RHR HX inlet temperature is 325 degrees, RCS pressure is 290 psig.
I- A total loss of all Control Air occurs.
lboron that can be added before the cooldown commences will depend on the available space in the PZR.
IDue to the slow rise in level, this may become limiting. Therefore, the charging pump suction is itransferred to the RWST early in the event. This ensures that any addition to the RCS is at RWST concentration. C is incorrect because the charging pump suction will auto swap (MOV's) to the RWST Bases Document.
i Tuesday, July 15, 2008 12:27:45 PM - i Page20of88 J
Which of the following identifies the major concern with the automatic fire suppression systems L-designed to extinguish a Class "B" fires in the _DFOST _ ~
rooms?
~ ~ _ --________~_______
-- from displacement
________ of oxygen.
__________~~_____
~ _ ~ _ _ _ _ _ _ ~ _ _ _ _
[Flooding and subsequent loss of vital eaubment.
7
- _ _ _ _ _ ~
la R Application Emergency~ - _ and
_ Abnormal Plant Evolutions ~~
,067 1 /Plant Fire on Site
!AKI._ /Knowledge of the operational implications of the following concepts as they apply to Plant Fire on Si?
m.01Fire classifications, by type i _
______________.__~_________
~
_ _ _ _ ~ ~ __ _ _ _ _ ~ -~
~~
22.913.9 A Class "C" fire is an electrical fire. The automatic fire suppression system to combat electrical fires is
/theC02 suppression systems. C02 displaces oxygen and can cause asphyxiation. B and C are 1
iincorrect because electrical fires aren't fought with water. D is incorrect because Halon 1301, bromotrifluoromethane, is primarily used in fixed, total flooding systems:
A significant advantage of Halon 1301 is that when used in areas normallv occupied. personnel mav be
[exposed to low concentrations. .. _ ~ _for_ brief
_ ~ .. _ periods
- without serious risk.
Fire Protection Svstem Lesson Plan I
FIRPROE004 Describe the function and operatina - characteristics for the followina Fire Protection Svstem comDonents:
Fire Barrier Components:
Fire Doors Fire Dampers Penetration Seals Fire Proofing Mannite Walls Energy Shields Protective Wraps and Coatings
- b. Fire Detection Devices:
Ionization detector Thermal detector Smoke and Fire detectors
- c. Fire Protection Subsystems:
Water Supply System Preaction Deluge System Wet-Pipe Sprinkler System Foam System Carbon Dioxide System Halon System
- d. Fire Header Pressure Switches ~ _ _ _ . ~- __
. ~~.~~~~~
~ . ... _ ...... .......... .
.... ~ .
~
.. .....~ ~ . . ~...~
~
1, Tuesday, July-~
15, 2008 12:27:45 PM ~ I Page21 of88 I
Which of the following transients is analyzed to result in the highest containment pressure AND greatest leakage out of containment?~ _ _
- - _ _ p ~ ~ ~
ILarae Break LOCA.
_ _ p ~ ___
/Inadvertant containment spray actuation.
_ _ ~ _ _ ~ _ _ - _ _ _ _ _ ~ _
IDesign basis Steam Line Break inside containment.
-~ - - ..
~-
-~
Emergency-and Abnormal Plant Evolutions 2 1 7
, ~ 1 ILOSS of
_ Containment Integrity
_ _____..__________ _ _ p ~ . _
~ ~ _
1 _ _ _ _ ~ ~ _
P K l IKnowledge
~ of the operational implications of the following concepts as they apply to Loss of Containment 1 lntegri
__ ty-: - ~ - -- 1 IAKI .Ol Effect of r-pressure on __
leakrate __ . - 2 -263.1 IFSAR section 3.8.1.4 (page 3.8-16)states. ."The containment structure has ais-p _ _ _ _ _ _ _ _ ~ _ _ _ _ _ _ _ _
lincrease in design loads due to the postulated MSLBs. The evaluation shows that for the design of the 1 lcontainment structures LOCA is the governing condition." This makes C and D incorrect. A PZR space 1
/LOCA will be much smaller and have a much smaller effect on containment pressure. I 1
- 'Z h k r 0
- ------ _ - *.' ~
Describe-the purpose;a
- 1 I- I II CONTMTEOOI design basis for the following Containment and Containment S u ~ ~ oSvstems rl subsvstems Containments Containment Airlocks Containment Isolation System Containment Fan Cooler System Containment Iodine Removal System Rod Dnve Ventilation System Reactor Nozzle Support Ventilation System Reactor Shield Ventilation System Containment Pressure 0 Vacuum Relief System Hydrogen
_ _ Recombiner
__. System . . -
-- - __ _ _ _ ~ ~ - - ~
/Given the following conditions:
Unit 1 has experienced a Rx trip and subsequent Safety Injection.
The Rx tripped when all off-site power was lost.
I- Concurrent events have caused conditions to deteriorate to the point that a transition
~ to FRCC-1, Response to Inadequate Core Cooling has been made.
which of the following describes how RCS pressure will be lowered to the point where ECCS
!Accumulatorswill inject into the RCS?
llntact SGs will be depressurized by.. .
bumping - - _steam
_ _ _ ~ - at maximum rate using MSI 0 Atmospheric Relief valves.
~ _ _
~~ ~ ~
__ ~ ~ ~ - - _ _ _ _ ~~~
the Main Steam Dumps in MS Pressure Control MANUAL and dumping steam at
~ - _ _ ~ ~ _ _ ~ - - ..-- __ ___
hsing the Main Steam Dumps in MS Pressure Control MANUAL and dumping steam at 25%
ivalve demand. - -
dumping steam using MSIO Atmospheric Reliefvalves while ensuring any load limit. The EDG load limits are: <2600KW: None, 2600-2750KW: 2000 hrs. 2750-2860: 2 Ihours, 2860-3100: 30 minutes. With the EDG operating at 2710, there are a total of 2000 hours0.0231 days <br />0.556 hours <br />0.00331 weeks <br />7.61e-4 months <br /> available
!before exceeding a design limit. EDG L
has already been operating for 30 minutes.
1 1
1--__-EDGOOOE002 LOR Describe the design bases of the Emergency Diesel Generators. (Licensed Operator
- _ ~ _ _ _ _ -
~~
& STAnly) 7-1 Tuesday, July 15, 2008 12:27:45 PM
IEOP-LOCA-6, LOCA Outside Containment directs actions to verify valve positions for only ONE Isystem. This system is the most likely location for a LOCA outside containment.
h h a t system is addressed by EOP-LOCA-6? - - ~ _ _ _ _ _ _
~~
L- _-__________ - ~~ _.
!Safety
____-_ Injection. ___~~_______~__ - _ _ - ~ ~~ __
IChemical and Volume Control.
1 R
and Abnormal --__Plant Evolutions OCA Outside Containment IEKl. IKnowledge ofthe operational implications of the following concepts as they apply to LOCA Outside I
Wnment:
- ~ - , ____________~ __- __
IEKI .2 Normal, abnormal and emergency operating procedures associated with (LOCA Outside I
~
Containment).
_ _ ~ ~- ~ _ _ -
j 3*5'4.2 1
LOCA-6 checks RHR suction, hot leg injection, and cold leg injection flowpaths.
__________ I 1 LOCA06E001 Describe the EOP mitigation strategy for a LOCA OUTSIDE CONTAINMENT
- I IEditoriallyModified (Tuesday, July 15, 2008 12:27:45 PM 1 Page25of88
~~ . 1
I ~
____.____ -~ -- - ~ _ - _ _ _ _ . _ _ _ _ __ ~ _ _ _ _ _ _ _ _.- _ . _
IFRHS-I, Response to Loss of Secondary Heat Sink, Step 3 asks, "Is RCS pressure greater than IANY INTACT OR RUPTURED SG pressure" IWhich of the following statements is correct if the operator answers __________
NO?
L_
~-
~~~
IMMEDIATELY go to Step 23, Bleed and Feed Initiation, since there is no decay heat removal occurring through the SGs.
______- ~ - _ _ _ _ _ _
-______ _ _ _ _ _ _ ~ ~--___-___ _ _ _ - ~ _ _ _
Attempts to establish a secondary heat sink would be ineffective at reducing RCS temperature lsince_SGpressure is higher than RCS pressure. _
jThe RCS has experienced a LOCA large enough such thata secondary heat sink is NOT
[required,
_______ because core decay heat is being removed by break flow.
__ _ _ . _ _ _ ~ _ _ _ -. __
~ _ _ _ _ - -~ ______
IIMMEDIATELY trip all RCPs to prevent further loss of reactor coolant through the LOCA, since a LOOP later in the event could cause a more severe loss of coolant or two-Dhase RCS flow.
e of the interrelations between Loss of Secon be worried abo size is present, and break flow will be removing decay heat, along with ECCs injection. Distracter A is 1 lincorrect because the criteria for going to Bleed and Feed is SG WR level. Distracter B is incorrect 1 lbecause a secondary heat sink could actuallv be established, and could reduce RCS temDerature bv I
[dumpingsteam from-SG. ~ Distracter
~ _ D is ~ incorrect
_ because
_ it is~the RCP
_ trip_ criteria~for a SBLOCA:
___ - ~- ______
Loss of Secondary Heat Sink _ _ _ ____
_ - ~ - _._______
_ _ _ _ ~ _ _ _ _________
1 - Tuesday, July 15, 2008 12:27:45-~
- PM I Paye26of88 I
With Unit 2 in MODE 3, which of the following actions in S2.0P-AB.RC-0004, Natural Circulation, if performed, should be announced on the plant page system prior to performing it if conditions permit, for personnel
~- -- safety? ~~ ~ ~ _ _ ~ _ _ . _ _ _ _ _ _ _ _
ptarting or stopping a Rod Drive Vent - - Fan. -
llnitiating cooldown using the Main Steam Dumps.
__ ___ ~~
Blocking the Lo PZR Pressure SI during the RCS depressurization.
~ ~ _ _ _ _ _ _
Comprehension , 8/25/200d E09G114 1
~ hatural Circulation Operations n
k.1 [Cknduct Of Operations _
~- 1 ledge of system status criteria which require the
_~
-~
~
notification
. . . . . . . . . of plant personnel.
_ _ _ - p p 5 7
~~
All of the choices are actions contained within the AB. A is correct because the AFW pump is located in a normally accessible area of the Aux Building. B is incorrect because the rod drive vent fans are located in lcontainment. C is incorrect because steam dumps are already in operation, and doesn't require I Inotification. D is incorrect because it is a control board manip[ulation and doesn't affect personnel safety.
L- 1 Natural Circulation
_ _ _ ~ ~ _ _~_ _
__- _ _ _ ~ ~ ~ ~ _ _ _ _ ~
I L
ABRC04E002 p p a) Determine the appropriate abnormal procedure.
b) Describe the plant response to actions taken in the abnormal procedure.
--c) Describe the final plant condition that is established by the abnormal procedure.
~~
I i
~p I Tuesday, July 15, 2008 12:27:45 PM 1 Page27of88
IGiven the following conditions for Unit 1:
1- A reactor trip and SI occurred at 0700 I- RHR system problems resulted in a loss of recirculation capability
~- Current time is 1300 hours0.015 days <br />0.361 hours <br />0.00215 weeks <br />4.9465e-4 months <br /> iPrior to entering EOP-LOCA-5, the following conditions were present:
I- RCS subcooling - 10°F l - All RCPs are secured
~- 11 and 12 Charging Pumps are running 1- BIT flow - 350 gpm full range 95%
1- RVLIS11 SIPumpflow-110gpm I- 12 SI Pump flow - 100 gpm
- 1 Containment pressure 4.9 psig Which of the following identifies the ECCS pumps that should be run following determination of minimum SI flow for decay heat removal? (Assume equal flow from each Charging Pump and that
-. ~- ~ . ~- -- _- _-
I LEI 1 ~ Koss of Emergency Coolant Re&culation
-- -- ___~__
!EK2_ iKnowledge -- of the interrelations between Loss
-..--_---~
of Emergency
-~
Coolant Recirculation and LEK2.2 -, Facility's heat removal systems, including primary coolant, emergency coolant, the decay heat '3.914.3 removal systems, and relations between the proper operation of these systems to the operation of 1 cility. ~ ~ I During step 13 of LOCA-5, charging pumps will be reduced to ONE centrifugal, and SI pumps will be I reduced to ONE. Starting at Step 19 of LOCA-5, with RCP's secured with 6 0 degrees subcooling, will 1 iuse Figure A to determine th ECCS flow required vs. time after trip. 6 hours6.944444e-5 days <br />0.00167 hours <br />9.920635e-6 weeks <br />2.283e-6 months <br /> equals 360 minutes, which is 1
-225 gpm. With the stem stating that charging pump flows remain the same, you need both the 175 gpm 1
!and EITHER SI DUmD flow flow. I Loss of Emergency Recirculation -
~ LOCA05E007 Determine a discrete path through the LOSS OF EMERGENCY RECIRCULATION.
I
~
Tuesday, July 15, 2008 12:27:45 PM ' Page28of88 1
I ----
-- - - - - " ~ ~ ~ ~ * * ~ u - L ^ v I I I *
---q--7 r- ~ ~~ - ~
Changed distracters because question had 3 choices with ONE charging pump and one choice with ONE SI pump Removed the ONE SI pump only distracter because it is not like the other 3 choices, and there is no place where a Ileast 1 charana DumD would be run l-- Tuesday, July 15,2008 1227:45 PM - I- Page 29 of 88- 1
IGiven the following conditions:
Unit 1 has initiated a MANUAL Rx trip and SI on a large Steam leak.
MSLl has succeeded in closing 14MS167 ONLY.
11-13 SG pressures are all -71 0 psig and dropping.
14 SG pressure is 830 psig and rising.
Total AFW flow is 24E4, with -6E4 to each SG.
ALL SG NR levels are off-scale low.
1-EOP-TRIP-1, Reactor Trip or Safety Injection, is in effect.
Which choice describes ___- the mitigation strategy required for these conditions?
~- ___________-
DO NOT isolate AFW toany SG due to more than 1 SG being faulted.
_. ~ ~ _ _ _ _ _ _ _ _-._
llsolate AFW to 11-13 SGs, transition to FRHS, then return to procedure in effect.
[RaiseAFW flow to >22E4 Ibm/hr on- 14 SG, ~and _ ~ _isolate
______ AFW
_ _ _ _flow
____to_ _11-13 SGs.
llsolate AFW flow to 11-13 SGs and lower 14 AFW flow to no less than 1E4 Ibm/hr. Do NOT Itransition to FRHS.
lEK3. 'Knowledge of the reasons for the following responses as they apply to Uncontrolled Depressurization of all L
sociated with (Uncontrolled le SG. Up to 3 "single" SG's can be isolated. AFW flow is directed-NR level is 9%. This means 14 SG flow must be raised to maintain
'TOTAL AFW flow >22E4 while the 3 faulted SG's are being isolated. There is no step in LOSC-1 if student looks ahead for AFW flow requirements, but the next procedure in line, LOCA-1 states to maintain >22E4 until at least one SG NR level is >9%. While the stem does not state where in TRIP 1 the ~
ioperators are, the CAS for maintaining AFW flow is at step 20, before
- - _ - - -any transition
- point.
IReactor Trip or Safety Injection ~~ ~ _
- .I c
- _ - _ _ - I _ _ - __ _ _ - ---- , -- .- - - - - . _I TRP001E009 select which (if any) transition should be made from a qiven procedure. in accordance with SC OP-AP ZZ-O102(Q) 7 - -- ---
~
Tuesday, July 15, 2008 12:27:45 PM Paae30of88
lHow do loop flow and core flow differ when operating THREE RCPs as compared to operating IFOUR RCPs?
/With THREE RCPs running, the active loops TOTAL flow will be.. .
13/4 of the value for FOUR RCPs, and flow through the reactor core will be 3/4 of the value for IFOUR RCPs. _____________________-- ~ _ _ _ _ ~
13/4of the value for FOUR RCPs, and flow through the reactor core will be LESS THAN 3/4 of Ithe value for FOUR RCPs.
ILESS THAN 314 of the value for FOUR RCPs, and flow through the reactor core will be LESS ITHAN 314 of the value for FOUR RCPs.
IGREATER THAN 3/4 of the value for FOUR RCPs, and flow through the reactor core will be ILESS THAN 3/4 of the value for FOUR RCPs.
Comprehension
[Reactor CoolantPump System
_- - ~ _ _ _ ~ _ _ -____
!AI. 'Ability to predict and/or monitor changes in parameters associated with operating the Reactor Coolant
~- - System IPump ~controls - including: -
I
. - .. . -_ ..._ _ ~
~. 13.413.5 6 w i n the loop without a RCP will be in the reverse direction of normal flow. When validated on Salem Isimulator, the reverse flow = 32%. The loop flow for each of the loops with an operating RCP will rise. 1
\Whenvalidated on Salem simulator, the operating loops flow was 106% each. The reverse flow through (theidle loop will bypass flow through the core. So while flow in each of the loops will rise, the core bypass1
!flow wit result in actual core flow being
_ _ . _ _~ _less _ _ _ _3/4
_ _ _ _than _ _ _of
_ _the flow with 4 loops in operation.
Reactor Coolant Pump Operation LOR NCT Given plant conditions, relate the Reactor Coolant Pump with the following:
Chemical and Volume Control System Component Cooling Water System Service Water System Containment Isolation Signal Reactor Coolant Drain Tank Reactor Coolant System Reactor Coolant Pump lube oil Reactor Coolant Pump seal system Reactor Coolant Pump motor_cooling and ventilation _ _
- ~ _ - . _ _ _ _ _-
~-
_ _~_ _-~
__ _ - _ ~ - _ ~_ ~
~
____~_____ _ ~ _ _ _ __-____ --
~~
I I Tuesday, July 15, 2008 12:27:45 PM i Page31 o f 8 8 1
/Giventhe following conditions:
~- Unit 1 is in Mode 6.
~- 11 RCP Motor is uncoupled from the pump.
1- RCS Loop 11 is full.
I- Maintenance is working on the 11 RCP pump.
~
[Which of the following describes how leakage of reactor coolant up the RCP shaft is minimized?
ISeal injection flow which is maintained during this condition. -
IBackseatingthe pump shaft with the top of the thermal barrier assembly.
L ________~___
-~ ~ _ _ - - - ~ - ~ _ _ _ _ _ ~ _ _ _ ~ _ _ _ _ _ _ -
[Nozzle dam installation prevents RCS
_-__ - water
__ from entering the RCP shaft area. ~-
-- ~ ~- - ~~~ ~-
Beal Leakoff-- collects any RCS leakage up the shaft and directs it back to the VCT.
R ' Memory 8/25/2008
. -~ ~~
Plant Systems - 1
- L 1 -1 IReactor Coolant Pump System
_ _ _ - ~ ~ ~ ~~~
I
~K4. iKnowledge of Reactor Coolant Pump System design feature(s) and or interlock(s) which provide for the 1
[following:
___-- - ~
izing RCS leakage (mechanical seals) ~ ---
~
_ _ ~
~ -13.213.4 When the RCP pump is uncoupled from the RCP motor, the pump shaft is backseated. This is
'accomplished by lowering the RCP shaft -1 inch, which allows the top of the shaft to mate with the top of ~
the Thermal Barrier. This will reduce the leakage up the shaft from 5-10 gpm to -1 gpm, which is lsufficiently low enough to keep the water level below the #I seal runner. A is incorrect because seal ~
injection is NOT in service. B is correct as described above. C is incorrect because nozzle dams are I lnot used to isolate and drain piping. D is incorrect because the leakage up the shaft is collected and idisposed of by breaking the seal injection ~- line and establishing a drain collection system.
,RCP Seal Disassembly, ___ Inspection, Repair and Assembly ~ _ _ _ - ~ ~ _ _ _ ~ _ _ -_
~- __ ~~ _________
1 RCPUMPE004 LOR NCT Describe the function of the following components and how their normal and abnormal operation affects the Reactor Coolant PumD:
Impeller Turning Vane Diffuser Diffuser Adapter Thermal Barrier and Heat Exchanger Pump Radial Bearing Controlled Leakage Seal Assembly Lower Motor Radial Bearing Upper Motor Radial Bearing Flywheel Anti-Reverse Rotation Device Oil Lift Pump Motor Space Heaters
-__ __ ~ _ _ ~ ~ _ _ _ _
r%esday, L-July--15, 2008 12:27:45 PM - I Page 3 2 o f 8 8 i
/During Unit 2 steady state MODE 1 operation, the Chemistry Technician reports that the level of IFlouride in the RCS is elevated at the CVCS demin outlet. A second confirmed sample places the p unit in an Action Level 2 per CY-AP-120-100, Reactor Coolant System Chemistry.
IWhich of the following identifies the major concern with continued operation at this level, and what lis the rewired course of action?
~~ __. _ _ ~ _ _ _ _
'Corrosion of RCS and components. IMMEDIATELY initiate an orderly unit shutdown and
[cooldown to below 250 degrees as quickly as permitted by other plant constraints.
~ ~~
_~ ~ - ~_____.___ ~ _~_ ___ __
_._ _ ~ -~ _____________~__
ICorrosion of RCS and components. If concentration can NOT be reduced below the Action ILevel 2 limit within 24 hours2.777778e-4 days <br />0.00667 hours <br />3.968254e-5 weeks <br />9.132e-6 months <br />, initiate an orderly unit shutdown and cooldown to MODE 5 as
[quickly as permissible.
____-. __ __ _ _ ~ _ ~_ ~ _ . ~______ _.
____~______ ______________
hccelerated depletion of CVCS demin resin. IMMEDIATELY initiate an orderly unit shutdown land cooldown to below 250 degrees as quickly as permittedb y other plant constraints.
_____-______~- _~_______ ~ _ _ _ _ _ ~____ _____ __
IAccelerated depletion of CVCS demin resin. If concentration can NOT be reduced below the IAction Level 2 limit within 24 hours2.777778e-4 days <br />0.00667 hours <br />3.968254e-5 weeks <br />9.132e-6 months <br />, initiate an orderly unit shutdown and cooldown to MODE 5 las quickly --
as permissible. ____~______
R 7 - ~
1004 , khemical and Volume Control System . I - -
I IA2. /Ability to (a) predict the impacts of the following on the Chemical and Volume Control System and (b) lbased on those predictions, use procedures to correct, control, or mitigate the consequences of those labnormal operation: I
__--_______ __ .__~______ _________________
~
1A21<? High secondary and primary concentrations of chloride, fluoride, sodium and solids ______
~~ ~~~ _~_______________
_1 2.8) 3.5 balem FSAR (5.2.3.4 Chemistry of the reactor Coolant) states that the reason for maintaining RCS I lchemistry within defined limits is to.. ."ensure that corrosion of the RCS is minimized and reduces the (potentialfor RCS leakage or failure due to stress corrosion." Procedure CY-AP-120-100 further I delineates the reason as ...."The chemistry limits and action levels presented herein are appropriate for 1 protecting system materials, ensuring fuel performance, and controlling radiation field buildup." In general, there are 3 action levels associated with chemistry at Salem. Action Level 1 is a condition where' lchemistry is elevated outside the norm, and may have long term adverse consequence, and only requires 1 lmonitoring and attempts at correction. Action level 2 is declared at levels which, if allowed to continue 1 lindefinately, would lead to increased incidence rates of corrosion, and require a unit shutdown if not corrected within a certain period of time. Action Level 3 is a condition where chemistry is well beyond the 1 boundary at which accelerated corrosion will occur, and requires propmt action to shutdown and cooldow I the plant below 250 degrees. A is incorrect because Action Level 2 does not reuire a unit shutdown 1 lunless the condition cannot be reduced below the Action Level 2 limits in 24 hours2.777778e-4 days <br />0.00667 hours <br />3.968254e-5 weeks <br />9.132e-6 months <br />. B is correct. C and D are both incorrect because although rncreded levels of impurities in the RCS will deplete CVCS IReactor Coolant System Chemistry-Salem UFSAR 1 RCSOOOEOI 1 LOR NCT Discuss the procedural requirements associated with the Reactor Coolant System, including an explanation of major
---precaution and limitations in the Reactor Coolant System procedures
_.__~_______
1S%OE013 LOR NCT Given planconditions, relate the Reactor Coolant System with the following:
Refueling Water Storage Tank Rod Control System Pressurizer Relief Tank Chemical and Volume Control System
-__ __ _ _ _ _ _ _ _ ~
~
Tuesday, July 15, 2008 12:27:45 PM -
~
1 Page33of88 1
~-~- ~
Reactor vessel Level Indication System Emergency Core Cooling System Pressurizer Reactor Coolant Drain Tank Main and Auxiliary Feedwater Systems Nuclear Instrumentation Reactor coolant Pumps Spent Fuel Pool Purification Refueling Canal Refueling Water Purification Main TubinelGenerator
_ _ _ _ _ _ _ _ _ ~~ __
-- ____ - _~ __ _ _
I 1 -- - ~ _1 _
1 Page34of 88 ,
IGiven the following conditions:
j:
1 Unit 2 is operating normally at 100% power.
2cc71, LETDOWN HEAT EXCHANGER TEMPERATURE CONTROL VALVE, fails to the full closed position due to its temperature sensor failing low.
IWhich of the following - .predicts the plant response - ~~
to this event?
own flow rises o lowering backpressure. -
_ _ ~ - - - - - __ - ~ _ _ _ _ _ _ ___- - ~~ ..- -~
b C T temperature rises causing a reduction in charging pump NPSH. _
bressurizer level will rise and VCT level will lower-when CV7 closes.
[RCS boron concentration will slowlv rise with the CVCS demineralizers bvpassed.
ib Comprehension- Salem 1 & 2 8/25/2008 I
1004000A405 1004 khemical and Volume Control System I -30,
_______ _ _ ~ _ _ _ ~ _ . - - ~-
lAblllty to-manually
.--._ operate and/or
__ monitor in the control room: 1
~ ~
I valves I A is incorrect because backpressure will not lower, it wil rise due to heating up of the fluid in the VCT. B ;
is correct because with no cooling water flow through the Letdown HX, letdown fluid temperature will rise.
The temperature sensor that controls the CC71 (2TE130A) is the same one that actuates the 2SV496, 1 which is what controls 2CV21 to bypass the demins. VCT temperature will rise, and available NPSH to the ICVCS pumps will lower. C is incorrect because the interlock between the CV7 and CC71 is the lopposite. CV7 closing will shut the CC71. D is incorrect because the same temperature sensor is used ~
for CC71 and demin bypass valve CV21. With the sensor failing LOW, the CV21 will never divert the 1 lhotter letdown fluid Dast the demineralizer.
- Charging,
~- Letdown,
__ and Seal Injection ~________ ___~____
I CVCSOOE016 LOR Given a Chemical and Volume Control Svstem failure. Dredict the effect of the Chemical and Volume Control Svstem failure the following: (License Operator and STA onl;)
Automatic Rod Control Component Cooling Water System Reactor Coolant Pumps Pressurizer Level Control System Reactor Coolant System Pressurizer Pressure Control System Reactor
~ _ _Coolant_ _ Pump
- ~ Seal Injection _ _ ~
I-Page35of88
IGiven the following conditions and S I .RE-RA.ZZ-O016, Curve Book:
~- Unit 1 is preparing to initiate RHR in shutdown cooling mode during a late-cycle forced outage to MODE 5 two months before a scheduled refueling outage.
~- The RCS has been borated to the required CSD boron concentration.
- - The RHR system was last in service during a forced outage at BOL.
IIf RHR system boron concentration is NOT adjusted during sampling, which of the following describes how RCS boron concentration will be affected when the RHR system is placed in shutdown Cooling?
IRCS boron concentration...
__ _ _ - _ ~ -_ _ -- .
remain atmroximatelv the same.
!may rise OR lower depending on the RHR-to RCS piping boron concentration.
~~
&2 1- -
8/25/2008
- II 1005000K509 1
-~
1 0 0 5 - 7 !Residual Heat Removal System
_ _ ~
v 1K5. IKnowledge of the operational impli&tions of the following concepts as they apply to the Residual Heat I IRemoval Svstern.
IBOL CSD boron concentration will be AT LEAST 1425 ppm, which is what the RHR system Iconcentration will be, since it was not in service since being taken out of service following the previous shutdown. The CSD boron concentration for this shutdown 2 months before refueling is 900 ppm at a -
core burnup of -1 1,300 EFPH. ( S I .RE-RA.ZZ-0016, Page 121, Table A) Without adjusting RHR boron concentration, placing RHR in service with the previous HIGHER concentration will BORATE the RCS further. The candidate does NOT need to know exact numbers, so no procedure is provided. They DO
$needto know that required CSD boron concentration LOWERS over core life. The last distracter is lincorrect because that DiDinq will be at the same boron concentration as the RHR svstem itself.
~~
Initiating RHR -- __ _ . ~ _ _ _ _ _ _ _ _ _
Curve Book ~
I l RHROOOE015 LOR NCT Given plant conditions, relate the Residual Heat Removal System with the following:
Component Cooling Water System Pressurizer Spent Fuel Pool Cooling Chemical Volume Control System Reactor Coolant Pumps Emergency Core Cooling System Service Water System Refueling Water Storage Tank
~~
r i u e s d a y , July 15, 2008 12:27:45 PM I Page36of88 1
I Tuesday, July 15, 2008 12:27:45 PM
__ 1 Page37of88 1
1Which of the following running Safety Injection pump discharge flows is consistent with the RCS pressure shown during a LOCA?
With RCS pressure at psig, a single Safety Injection pump will be providing approximately gpm flow to the RCS. ~~
'1000; 650.
[006000A305-1 mergency Core Cooling System automatic operations of the
~-
re responsible for knowing the basic pump characteristics, including tthe pump curve. A is incorrect because the pump shutoff head is -1,520 psig, so there should be no flow 1 lat 1,765 psig. (1765 psig is AUTO SI initiation pressure) Distracter C is incorrect because runout flow is 1 runo our 1650 gpm, but should not be present at 1000 psig, as runout pressure is 650 psig. Distracter D is below pressure, but runout flow. B is correct because it is the only pressure flow combination that falls~
,onthe DumD curve TSC - Integrated Engineering Response (SI pump curve in document) 1 ECCSOOE008 Identify and describe the Control Room controls, indications, and alarms associated with the Emergency Core Cooling System, including: (Licensed Operator 8, STA only)
The Control Room location of Emergency Core Cooling System control bezels and indications.
The function of each Emergency Core Cooling System Control Room control and indication.
The effect each Emergency Core Cooling System control has upon Containment Spray System components and operation.
The plant conditions or permissives required for Emergency Core Cooling System Control Room controls to perform their intended function. -_ _ _ __
_ _ _ _ _ _ ~ ~ ~ _ _ _ _ _ - -~- _ ~ _ _ _ ~ - - - .
~ _ _ _ _ _ _ _ _ ____
_ ~_-_ _ _ _
I p~
c- . . - I 1 Tuesday, July 15, 2008 12:27:45 PM I Paue38of88 1
Given the following conditions:
1- Unit 2 is operating at 100% power when a LBLOCA occurs.
~- When 21 RHR pump starts, the pump mechanical seal fails.
which of the following describes the effect this will have on 21 RHR pump and its ability to perform iits ECCS function IAW Salem FSAR?
The FSAR analyzed seal failure will result in.. .
~-- p- . _ _ _ _ ~ ~-~ ~ _ __
1 L
a leak to the RHR pump room o f < 50 gpm. It will NOT affect the pumps ability to perform its
,RECIRCULATIONphase p
pp-
~
function.
p
__ _ _ _ _ _ _ ~
~~ ~-
p -
leak to the RHR pump room of > 50 gpm. This will result in overflowing the RHR pump room and it WILL adversely affect the pumps ability to perform its RECIRCULATION phase
[function.
~ p - ~ - _
p
- -~ - ~-
LTeak to the RHR pump room of 50 gpm. It WILL adversely affect the ability of the RHR
/pumpto perform its INJECTION phase function.
1
'a leak to the RHR pump room of > 50 gpm. This will result in overflowing the RHR pump room isump, but will NOT affect the ability of the RHR pump to perform its INJECTION phase
,function.
R I 006 Emergency Core Cooling System
- ______- - ~ ~ p ~ ~ _ _ _ .
K6. IKnowledge of the of the effect of a loss or malfunction on t I
iCooling System: - I
~-
1 _ _ __ _ _ _p -
1-3.0 3.5 r
5 HPI/LPI cooling water - ~~
As discussed belo< the seal failure is analyzed to be LESS than 50 gpm. That makes distracters b and d incorrect. C is incorrect because the seal leak is not analyzed for the injection phase because cool RWST water is flowing through the pump as it is being injected to the RCS. The RHR pumps are provided mechanical seal cooling from the CCW system. Salem FSAR, Section 6.3.2.11, page 6.3-42, discusses the seal failure during recirculation. It is postulated to be 6 0 gpm, and the RHR sump is sized to accommodate it for 30 minutes. Subsequent to that, the leaking seal is expected to be isolated by operator action, and is NOT expected to impact the recirculation phase of ECCS. The seal leak is not analyzed for the injection phase because cool RWST water is flowing through the pump as it is being
' ECCSOOE002 Describe the design bases of the Emergency Core_ Cooling System.
_ (Licensed
~ Operator
-~ & STA only) -._..
r--
~
I PaG39of88 1
~~ -~ - . - ~_ ~ _.- -~
U n i l is ready to begin drawing a bubble in the PZR with a vacuum in the RCS.
Prior to drawing a bubble in the pressurizer which ONE of the following must be accomplished 1
IAW S I .OP-SO.RC-0002, Vacuum Refill of the RCS?
- ~ _ _ _ _ _ _ _ _ _ ~
-- - _ _ _ _ ~
[Bypass the PZR heaterlo level heater cutout.
_ _ _ _ ~ ~ . ~ -- _ _~ _ _ _ ~ _ _ _ ~
~-
_ _ _ _ _ _~ - ~ ~ _ _ ~ ~ _ _ -~ ~-
[Establish 40-60% cold cal level in the PZR.
~~ __ _ ~ - _ _ _ _ _ _ ~ _ _ _ ~
kerform 30 second bumps of ALL RCPs.
IRaise PZR cold cal level to > 100%.
v R ' Coyrehension
- 1
~
1 +_-I 1 izer Relief TanWQuench Tank ~~~
lK5. /Knowledge of the operational implications of the following concepts as they apply to the Pressurizer Relief ITanWQuench Tank System: ~ ~ _ _~ _ _ _
- ___ __ ~ _ _ _ ___ _ _ _ _ ~ ~ _~ _
_ _ - ~ .
d of forming a steam bubble in the PZR ~
Distracter A is incorrect because the PZR heater I velcitout is NOT requi 3 hot cal PZR level instruments were removed from service prior to establishing the bubble, which removed any lo level signal from circuitry. Distracter C is incorrect because RCP bumps are NOT performed with the RCS at a vacuum. Distracter D is incorrect it is the level at which a bubble would be
[drawn with NO vacuum in RCS. -
I1 " " 1 Vacuum Refill of the RCS I
hmht.g-Obieaiwas
...____.L-.-.. ^_I_ L -v - I - A I ..- i__-_---- *-_v- .-L- _-I ---d I ---"d-&- II _ " _
PZRPRTE012 NCT Discuss the procedural requirements associated with the Pressurizer and Pressurizer Relief Tank, includincl - an explanation o major precaution and limitations in the Pressurizer and Pressurizer Relief Tank procedures
/Tuesdav.
Julv 15,2008 12:27:45 PM 1 Page40of88 ~ 1
q-3s-Given the following conditions:
1- Unit 2 control room has been evacuated during a Security Event.
1-L- Which of the following components can be operated from the Hot Shutdown
- ~ ~ _ _ _ _
~ Panel?
~ - -
__ _ _ _ ~ ~
~
r - 7 - [Station Air Compressors.
4KV Vital
__ . Bus
- __ lnfeed Breakers, 12.4 iEmergency
__ Procedures / Plan -____
ding system geogra
~ _ _ - -
IAttachement 6 of A MOTE switches that are repositioned toallow lcontrol of those components from the HSD Panel. The only one of the above choices which can be 1
'operated
-- are the CCW pumps. I Control Room Evacuation 1 ABCROIEOOI Describe the operation of the followinq as applied tu S2.OP-AB.CR-O001(Q):
~
7 - --
Tuesdav. Julv 15. 2008 12:27:45 PM , Page4Tof88 i
IGiven the following conditions:
~- Unit 2 is in Mode 3.
RCS Pressure - 2235 psig.
1- RCS temperature - 547°F.
1- Channel Ill (PT-457) is the PZR PRESS controlling channel.
/Which of the following is the first response of RCS pressure control if PT-457 fails LOW and NO
/operatoraction _ _ _ is taken? -~ ________~______~___
1PZR Spray __ valves open. _--_ __-
~ --. __
A PZR Code Safetv lifts.
1010 ' IPressurizer Pressure Control System I
-- ~ ~ _ _ _ _ _ _ _ _ _
IA and C are incorrect because they are fed from failed channel. Actual pressure will sri-
'heaters ON from failed low pressure signal, no spray will occur because of failed LO signal, the 1unaffected PORV will open at its setpoint,_ which will_ keep pressure
~ from
~ rising to Safety setpoint.
-Tuesday, Jul~15,2008 12:27:45 PM 1 Page42of88 1
1-1 Operators are responding to an inadvertent SI from 100% power.
- Equipment malfunctions have severely slowed operator progression through the EOP network.
! Operators are at the step to transition out of TRIP-3.
1- PZR level is 93% and rising slowly.
(Whichof the following describes how FRCI-1, Response to High Pressurizer Level, should be lutilized for this situation?
kRCI-1.... -~ - _ ~ - ~ _ _ ___.________
- __-___ ~ ~~
~ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ . ~ -
NOT be entered since the criteria for its entry are not met.
lshould _--__ ~ - ~ _ _ _ _ _ _
_ ~ _ ~ _ _ _ _ ~ _ _ ~
khould NOT be entered since with the SI pumps still in service normal charging and letdown lcan NOT be established.
ICAN be entered at CRS discretion to establisha vent/drain path to prevent lifting PZR PORVs
!when PZR goes solid.
CAN be entered at CRS discretion to quickly secure ALL ECCS pumps except the one pharging _- pump required to establish ___-____normal charging and letdown.
Application ISalern 1 842 a12512ooa 10110006423 I 1011
, ~ ~ - -
' (Pressurizer Level Control System
- . ~ _ _ _ _ _
I I
~~
\ -37)
,2.4 /Emergency Procedures / Plan
~- 1~ --__________- - ________ -.
1- -
12.4.23 I Knowledge of the bases for prioritizing ure implementation during emergency 2.81 -3.8 1-
~~ --
- ._________ ____________~
New Rev. 2 Supplement 1 Actual K/A values are RO:3.4, SRO:4.4 FRCI-1 is a 1
'YELLOW path FRP, and as such is entered at the CRS direction. The condition IS met for entry with PZR!
Ilevel > 92%. The first step in FRCl asks if the SI pumps are running, and if so, returns to procedure in leffect. Since Yellow paths are not required to be entered when entry conditions are met, it is wrong to ~
lenter it and then leave it within a step. With the SI pumps in service, normal charging and letdown cannot!
Ibe established. I 1 Tuesday, 1 Page43of88 1 - July --15, 2008 12:27:45 PM
IWhich of the following describes how Logic Card testing is accomplished for the Solid State IProtection System-- when the unit is operating at 100% power?
Il&C technicians
___ perform___ scheduled surveillances
__ - which tests all Logic cards. _ _ _ _ ~ _ _ _ _ _
~
-- ~ ~~~ ~~
[Logic Card testing is done on a continuous i _ _ _ _ ~ ~ ~ _ _ _ _ -__-____ basis by the self test_ feature _ ~ _ _ of
_ _the
~ _ SSPS
_ _ _ system. .__
__ __-______. _ _ ~ ~ __
[Logic Card testing is done automatically on a daily basis by the self test feature of the SSPS i svstem.
-~ __ - -_
Control room operators perform scheduled surveillances which tests all Logic and Slave Relay Icards . ~~ -
(K4. Knowledge of Reactor Protection System design feature(s) and or interlock(s) which provide for the
!following:
~ -~
IK4.08 1 Logic matrix testing - -
--~___.____
SG s requires surveilla fferent frequencies for differen gic functions IAW i lSurveillance Table 4.3-1, performed during the Channel Functional Test. There is no specific selt test feature for logic, other than channel continuity or component failures, which would be identified by system 1
,alarms, but there is no automatic active testing circuitry. The Control Room operators perform SSPS Islave relay testing, but not logic testing. I&C technicians perform the functional tests when required.
- - I Solid State Protection
____ -_ System Train A Functional Test ~____ __
Tech Soecs 1
r----
RXPROTE013 a) b)
Analog testing SSPS Testing
~~
_. ___~_____
Describe the differentiate between the following types of RPS testing: (Licensed Operator and STA Only)
-~ ~~
~
~~
Tuesdav. Julv 15. 2008 12:27:45 PM I Page44of88_~ 1
iWhich of th&ollowing describes how Logic Card testing is accomplished for the Solid State Protection Syskm
____ x when the unit is operating at 100% power? -__
Il&C
_ _ _ _ _techniciah-perform
___~- ~- scheduled surveillances -~ which tests all Logic cards.
_ _ ~ p a continuous basis -by the self test feature of the SSPS system.
- _ _ _ ~ ~ ~
on a daily basis by the self test feature of the SSPS-lsvstem.
..-. ~ __ _ _ ~
pontrol room which tests all Logic and Slave Relay pards.
~alarms,but there is no automatic active tes islave relay testing, but not logic testing. I&
~
\-
~-Solid State Protection System Train A Functional Test Tech Specs - ___ _ _ _ - - ~ ~ _ _ _
1
\
RXPRQTEO13 List all Permissives and Blocks, including a) Name of the permissive or block b) SetDoint. and Coincidence (if anv) fNIA NEO)
I Page45of88 1 Tuesdav. Julv 15. 2008 12:27:45 PM
- ~- - ~ ~ ~ - _~ ~L__.Lp--_- -- ___
which of the following Rx trips is desinged to prevent exceeding local power density limits for the
~ R xfuel rods? --
__ ~ __-_______ _____
lover
-~ Temperature Delta Temperature. (OT'/DT) __ ___
~ _ _ _ _ _ _ _
following concepts as they apply to the Reactor Protectioy
~
system:
_ ~ _ _ _ _ _ _ ~ ~ _ _ -
iK5.02 ~ Power density
-~
~-
. . ~ _ _ _ _ _ _
~-
~-
j - 5 4 3.3 Tech Spec 2.0, Safety Limits and Limiting Safety System Setpoints shows the relationship between Tave, iDower and Dressure. The combination of these 3 characteristics are for DNB protection. The OP/DT h i e s Tavg,'and rate of change of Tave
-L ~ - to_develop_ _ a_trip setpoint that equates to-109% power.
1 1 RCTEMPE007 Describe the relationshiD between the Reactor Coolant Temperature Instrumentation System and the following protection and con system actuations:
a) OT DT Reactor Trip, Rod Block, and Turbine Runback b) OP DT Reactor Trip, Rod Block and Turbine Runback c) P-12 d) Feedwater interlock e) High Steamline Flow Safety Injection f)
__ High Steamline Flow Main Steamline Isolation
____-______- _ ~ _ _ _
_ _ _ _ _ _ ~ -
, RCTEMPEOIO Identify and describe the Control Room controls, alarms, and indications associated with the Reactor Coolant Temperature Instrumentation System, including:
m control bezels alarms and indications IConceDt Used T T u e s d a v . July 15, 2008 12:27:45 PM- I Page46of88 -
~ ~ . . _ _ _ _ . _
[Given the following condition:
i- Unit 1 is operating at 100% power.
- A momentary (1 second) Inadvertent SI signal is generated in the RPS system.
1i-- All systems function as expected for this condition.
AFW is reduced to 22E4 Ibm/hr right after the Immediate actions of TRIP-1 are complete.
/Which of the following temperatures is the CLOSEST to what actual 11 RCS loop Thot temperature lwill _ be_ 5 minutes
~ - ~ _ after this SI signal is sensed?
~ __________~_ ___~-___-____
1530 dearees.
--_- _ ~ -~_ _ ________
_ _ ~-~ -____ - ____-________~ -_
_ _ _ _ _ _ . _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ ~
The only ECCS pumps injecting to the RCS following an inadvertent SI will be the charging pumps. The 1 relatively small amount of cold RWST water being pumped into the RCS will have a very small effect on i Itemperature, since the RCPs will still be runnning and decay heat being generated. 603 degrees is the 1100% power Thot. Tavg of 547 degrees is the setpoint of the Plant Trip controller for steam dumps. I IThere will be a delta between Th and Tc, so Thot won't be right at 547. There will be very little power 1
,generation5 minutes after the trip. 530 would be a representative temperature for Th if a large injection I
/of RWST water were being injected with the RCS at a much lower pressure. 555 is representative of 1
~-IW/, Dower oeneretinn. I
'Rx Trio or Safetv lniection
. . ?
- - L - i RCTEMPEOOS Identify and describe thelocal controls and indications associated with the Reactor Coolant Temperature Instrumentation. indudin 1-- Tuesday, July 15, 2008 12:27:45 PM 1 Page47of88 ]
___ - -~ ~~ ~- - ~ - ~~
6 v e n the following conditions:
I1- Reactor Power is 75%
A failure of control rods to move in AUTO or MANUAL has occurred.
IWhich of the following identifies a consequence if control bank D rods are moved using the CBD I(Control Bank D) position of the Rod Selector Switch? -
~ .._________.___
step counters will not count.
monitor
___ __in -
the _
control
_ _ _ room:
_ ~ -__I ~
1 The Bank Overlap Computer (BOC), only receives input from control rod movement in AUTO or MANUAL1 control. When an individual bank is selected, the BOC is not seeing the steps in or out, and will not sequence rods. The Group step counters continue to count. The RIL computer gets its input from the 1 P/A converter and auct hi Tave(power) The P/A converter gets - it signal
- from the data loqqina card, which lis
__not affected by which mode of rod control selected 1 RODSOOE006 NCT Describe the function of the following components and how their normal and abnormal operation affects the Rod Control and Position Indication Systems:
Rod Cluster Control Assembly (RCCA)
Control Rod Drive Mechanism (CRDM)
Rod Drive MG Sets Reactor Trip and Trip Bypass breakers Reactor Control Unit Power Cabinets Logic Cabinet components:
Pulser Master Cycler Slave Cyclers Bank Overlap Unit
- j. Signal Conditioning Modules
- k. PulseOtoOAnalog (P to A) Converters I. Rod Bottom Bistables
- m. Rod Insertion Limit Comparator
- n. Step Counters 1
1L-- Tuesday, July_ _15,
_ 2008 12:27:45 PM 1- Page 48 of 88 -1
r u e s d a v . Julv 15. 2008 12:27:45 PM- I Paqe49of88 1
- Unit 2 is operating at 100.0% indicated NI power, and 3459 MWth indicated calorimetric power.
r-- __________ _.~
IRC LOOPSTave-Tref deviation alarm will be locked in on 2CC2.
[NI power will read lower than actual power due to colder feedwater entering the steam
\generators. ~- ~ ~ ~ ~
~- ~ ~ _ _ ~ - _ _ _ _ .~
IN1 power will read greater than calorimetric power due to the lower FW temp used in lcalorimetric calculation. ~ ~
(c IComprehension Isalemla2-7 L__
Plant Systems
-~
INuclear Instrumentation System L--
7 lcause Rx power to rise due to the negative temperature coefficient. The colder water will also act to allow lless thermal nuetrons to leak from the core and be seen by the Nl's. B is incorrect because with the (deviationat 0.0 prior to the event, the change in temperature will not be great enough to cause the alarm G Window Alarm Respone Procedure
_________~ -~
~
2CC2 Alarm Response Procedure ~ ~~~
-. L_--
EXCOSEEOOS Identify and describe the Control Room controls, indications, and alarms associated with the Excore Nuclear Instrumentation Systc including The Control Room location of Excore Nuclear Instrumentation System control bezels and indications The function of each Excore Nuclear Instrumentation System Control Room control and indication The effect each Excore Nuclear Instrumentation System control has upon Excore Nuclear Instrumentation System components an operation.
The plant conditions or permissives required for Excore Nuclear Instrumentation System Control Room controls to perform their intended function.
The setpoints associated with the Excore -
Nuclear
~ _ _Instrumentation
_ _ _ _ _ System control room alarms.
--- .______ ~-
1 EXCOREEOOI Describe the purpose-
__ of the Excore Nuclear
__ Instrumentation
_____ System
.________ - -_- ~
__ ___~_- I
-1 I
Concept Used do 1 Tuesday, July 15, 2008 12:27:45 PM I Page50of88
- 1
Given the following conditions:
- Operators have evacuated the Unit 2 control room due to a fire.
I which of the following indications powered from the ASDS inverter will be available to the operator jstationed at the Hot Shutdown Panel?
~- -
~~~ __-
ICCW Surge tank Level.
.~
IRCS Loop 21 Th and Tc.
121-24 SG Narrow Range Levels.
(LetdownOrifice 2CV4 position indication.
016 , b-orNuclear InstrumentationSystem inverter is the SG NR levels.
Hot Shutdown Station Panel 213 ABCROI EO01 Describe the operation of the following as applied to S2.OP-AB.CR-0001(a):
A. Hot Shutdown Panel
~~
I --
__ ~~
- ~
1
___1 IEditorially Modified 1 Tuesday, July 15, 2008 12:27:45 PM I Page51 of88 1
/Given the following conditions:
1 1- Unit 1 initiated a manual trip and safety injection coincident with a loss of off-site power.
1- I C EDG failed to start.
1- Prior to the event, 11-14 CFCUs were running in HIGH speed.
IWhich of the following contains ONLY CFCUs that will be running one minute after the event 1 hamens?
T-r 1 c _
yssms ~~ - L _1 [022000K201 I
1 /ContainmentCooling System ~-
I - 45; pplie The 11-15 CFCUs are powered from A,B,C,B,C 460 volt vital busses respectively. ALL CFCUs get a SLOW start signal upon a MODE Ill (SI plus Blackout) SEC initiation. With C bus deenrgized , only I I11,12, and 14 CFCUs will be running. With 5 CFCUs to choose from, there had to be 2 CFCUs that were 1 iin
-_ to the remaining 3 CFCUs, which were only selected in 2 of the choices. 1 3 of the choices, as opposed___._________
aF qJ ) B d I I
-_-\ LP1 CONTMTE004 State ;he er supply
. . to the followinq-Con
~
ent Support Systems components, includinq voltage level and 1ElNon 1E.
Containment Fan Cooling Units, including breaker alignment for Fast and Slow speed.
Containment Iodine Removal Fans (Licensed Operator & STA only)
Control Rod Drive Ventilation Fans (Licensed Operator & STA only)
Reactor Nozzle Support Ventilation Fans (Licensed Operator & STA only)
Reactor Shield Ventila$an Fans (Licensed Operator & STA only)
Hydrogen Recombiners (Licensed Operator
_ _ - - _ _ ~& _STA
_ _ only) _______
Bank -
Concept Used rGesday, L-- 15,2008 12:27:45 PM July-__ I Page52of88 1
IGiven the following conditions:
I
~- Unit 2 has experienced a Large Break LOCA.
- RWST level has reached the semi-automatic swapover setpoint.
1- 21 RH4 does not shut after 21SJ44 opens.
IWith the RWST at 8, which of the following identifies how this will failure will affect Containment Spray flow as compared to pre-swapover flow, and how will the failure be addressed in EOP-LOCA-13, Transfer to Cold Leg Recirculation?
containment-_-_spray flow~will.. _ _ _ ._ _ -~
~ _ _
~ ____ ________
hse due to the rise in NPSH. ODerators will manuallv close 21 RH4 at the valve.
~ _ _ _ _ _ _ ~
iower due to direct flow from the RWST to the containment sump. Operators will manually lclose 21 RH4 at the valve. -
~- - - __..- - ___ ~_ __ __________ ~- __ -. __ __ ~
h a y essentially the same. Operators will shut the 2SJ69 to isolate the flowpath from the IRWST to the RHR pump suction. - -
lrise or lower depending on containment pressure. Operators will shut the 2SJ69 to isolate the iflowpath
._ --from the RWST to the RHR pump suction. -
1 IC R Application jsalem I& 2
~ ~- 1 8/25/20081 1 1026000A202 1
~~ -_-
inment Spray System- - - . . . 46 will not be impacted by the valve malfunction. Step 5 of LOCA-3 directs that when the SJ44s are open Transfer to Cold Leg - - Recirculation
- - ~_ ~ _-_ _ _ _ ~
~- _ _ _ _ _ _
r y u e s d a v . Julv 15. 2008 12:27:45 PM i Page53of88 1
'Which of the following identifies the component(s) used for gaseous iodine removal from I
~- containment atmosphere?
I_
llodine Removal Units durina accident conditions and durina normal conditions.
~~~
~___._ -____-
Containment Spray during accident conditions, and Iodine Removal Units during normal conditions
~ -_____________.. ___-_____-
[Containment Sprayand Iodine Removal Units during accident conditions, and neither during Inormal
_____~
conditions. ~ ______ -~
I Containment Spray and Iodine Removal Units during accident conditions, and-Iodine Removal Units during normal_______ conditions. - --
lK1. Knowledge of the physical connections and/or cause-effectrelationships between Containment Iodine iK1.O1 IRemoval 1
1 css _-
System
- and the- following: _ ~~
- ~_ - _ . _ _
' 3.4*/
- 3.7*
Containment Spray system operation during accident conditions serves to remove iodine from t containment atmosphere. There is no direction in the EOP network to operate Iodine Removal Units.
I I
'IRUs would be placed in service in non-accident conditions at direction of Radiation Protection upon ldetection of iodine in containment. ~-
~-Containment Ventilation Operation 1
.~
mrning
' I
. ... .. L* -..__
n ^.__.
2:s _~~....___~".L.I LA...
CONTMTE003 Describe the function of the following - components and how their normal and abnormal operation affects the Containment and Containment Support Systems:
Containment Fan Cooler System Containment Iodine Removal System Rod Drive Ventilation System Reactor Nozzle Support Ventilation System Reactor Shield Ventilation System Containment Pressure D Vacuum Relief System Hydrogen Recombiner System I
~-______
____-_____ ~
- _ __ _ ~ -~
I Bank Editorially Modified IVision Q29202 Slightly changed stem from "mechanisms" to "component(s)"
_ _ ~ _ _ ~- -__ _______
7-I Tuesday, July15,2008 12:27:45 PM I - - Page54of 88 I
- - - ~ _ _ _ ~ _ _ ~ __- __ _ _ _ ~ - -__~__
[During Spent Fuel movement in the Spent Fuel Pool prior to a refueling outage, the Fuel Handling (Cranearea radiation monitor (2R32A) fails HIGH when the fuel handling tool and attached spent
[fuelassembly are being raised. The crane hoist has NOT yet been fully raised.
/Underthese conditions, which of the following correctly describes restrictions concerning jmovement of the fuel assembly attached to the crane?___ - -
[Movement of the fuel assembly must be terminated until an HPTechnician completes a lqeneral area survey.
r- ______- - ______- _ _ _ ~ ~ ~ _ _ _ _ _ _ _ ~
lThe fuel assembly can be lowered ONLY after pressing the BYP INT pushbutton on the crane I controls.
- ~ _ _ _ _ _ _ _ -_____ -__
[Crane controls are disabled until iumpers are installed to defeat the interlock.
-________ _ _ _ _ _ _ ~ ~ - -~
[Crane controls can only lower the fuel assembly.
R Memory 8/25/2008
-_- 2 I IFuel Handling Equipment System I (K6. Knowledge of the of the effect of a loss or malfunction on the following will have on the Fuel Handling
/EquipmentSystem:
I I
7 1
-.__-~~___ -
7 . -
lK6.02 1 Radiation monitoring systems - - J 2.61 3.3
_ _ _ _ _ -.._ _ _ ~ _ _ _
- ~ _
S2.0P-AB.RAD-0001, Att 3, NOTE: "High radiation indicated on 2R32, Fuel Handling Crane Monitor, will '
prevent crane hoist UP operation."
A I ---....AbC -__A I_ --___ -2 - -_i - - i d -
L -L-- i "
L-RMSOOOE005 . . -
NCT Outline the interlocks associated with the followina Radiation Monitorina- Svstem
, comDonents R1B, Control Room Inlet Duct Monitor R5, FHB I2 SFP Area Radiation Monitor R7, In-core Seal Table Area Radiation Monitor R9, FHB I2 New Fuel Storage Area Radiation Monitor RIOA, Personnel Hatch I2 Containment Elev 1 O O E Area Monitor RIOB, Personnel Hatch I2 Containment Elev 1 3 0 E Area Monitor R1IA, R12A, R12B, Containment Particulate, Noble Gas, and Iodine Monitor R13A, B, C D & E CFCU Service Water Monitors R17A and B, Component Cooling Liquid Monitor R18, Liquid Waste Disposal R19A, B, C, & D, Steam Generator Blowdown Liquid Monitors R32A, Fuel Handling Crane Area Radiation Monitor R36, Evaporator and Feed Preheaters Condensate Monitor R41D, Plant Vent Radiation Monitor 2R52, Liquid PASS Room Area Radiation Monitor
~ ~ _ _ _ .___-__________ _ _ _ _ _ _ ~
L--
I __ - - ~ i-- _ . ~ - _ _ _ _
IFacility Exam Bank 1 PaG56of88
given the following conditions:
- Salem Unit 1 is operating at 73% power.
- Main Generator output is 888 MW.
- A grid disturbance causes load to drop to 798 MW.
- The Rx does not trip.
!- The following indications are present:
- The "Block Cooldown" split bezel light on the Steam Dump control bezel is illuminated.
- The "Block Non-Cooldown" split bezel light on the Steam Dump control bezel is illuminated.
IWhich of the following identifies the status of the Main Steam Dumps?
IMain Steam dumps are.. . - -..
- ~ ~
brmed, and at least the 21-23TBIO's and 21-23TB20's will be open. - -
!armed, but the valves are blocked from opening.
INOT armed, ___ and SHOULD NOT be armed.
~ - ~ ~ _ _ _~-_ _- ~
INOT armed, but SHOULD be armed. ~-
L-I 8/25/2004 IC R I Application 1salem I& 2 j --
Systems 1 Ii ~ O A 4 0 7 ,
r---5o
~-
7 7
'Main and Reheat Steam System i
-~ ~ _ _- _
A4-w~ to manually operate and/or monitor in the control room:
- - -- ~
7 iA4.07 Steam
___ ____dump valves. ~~
L 1 2.8*) 2.9 A 10% (of full load) load rejection (-123 MWe) will arm the steam dumps instantaneously. In this case, ~
the load reduction is 90 MWe. While this is >IO% of current load, it is not >IO% of full load, and will not I
!arm the dumps. Therefore, the split bezel indication is correct, and the steam dumps are not armed, and ~
lare NOT required to be armed.
-~
_ -- _ . I i STDUMPEOOS -.
LOR Describe the conditions that will cause the Steam Dump Svstem to become "Armed" and "Blocked", and the actions that will occur as a resilt of arming and blocking. (Licensed Operator & STA only)
I r y u e s d a v , Julv 15, 2008 12:27:45 PM I i Page57of88 1
-_._____.~ __ _________~
IGiven the following conditions:
Unit Ihas just entered MODE I.
,l-l- Rx power is 5.1%.
l- Power is being raised slowly in preparation for rolling the Main Turbine.
I- 11 SGFP is in service supplying FW to SGs.
ALL AFW pumps are aligned for normal standby operation.
I-
,- A spurious MSLl actuates.
Which of the following describes the effect this will have on the AFW pumps with NO operator i--action? -____ _________~_________________ ____ __ -_
~ _ _ ~ ~ ~ ~ ~ . _ _ _____________--___
b L L AFW pumps will remain in standby. The operating SGFP will remain in service since at Ithis power-____level it is being supplied ~with _ _ steam from the Heating Steam System.
_________ .-__ -_______-_____ ____ -~
'ALL AFW pumps will remain in standby. Sufficient steam will be supplied through the 11-
!14MSl8s. MS STOP BYP VALVES.
~ -~
1 -
The MDAFW pumps will start when 1ISGFP trips. The TDAFW pump wiil start whenSG llevels shrink followina the Rx triD.
__ ___ ~___________.
r The MDAFW pumps and the TDAFW pump will start when SG levels drop to the lo lo level
!setDoint.
-- - -~ ______ ____
7-IK3. [Knowledge of the effect that a loss or malfunction of the Main and Reheat Steam System will have on the ,
ifollowing:
r--- - _ _ ~ ~ . ~ _ _ ~ _ _ _ _ _ _ _ _ - _ _ _ _ _ ___ - ~
1 lK3.03 I AFW pumps ~ - ~
1 3.21 3.5 s incorrect because the operating SGFP(s) will be placed on Main steam supply prior to exceeding 5% 1
~
er (IOP-3, step 5.4.10), and will lose their steam supply when the MSLl signal closes the MSlVs AND 1 the MS18 bypass valves. D is correct because the MDAFW pumps and TDAFW will start on lo lo level in sGs as the SGFP coasts down after losing its steam supply.. B is incorrect because the MS18s shut on the MSLl also. C is incorrect because the sGFP will not trb.
Hot Standby to Minimum Load 1 AFWOOOE006 L. . . ~
NCT Outline the interlocks associated with the following Auxiliary Feedwater System components:
Auxiliary Feedwater Pump Automatic Start Matar-driven AFW Pump Recirculation Flow Control Valves Motor-driven AFW Pump Discharge Flow Control Valves 1
1 Tuesday, July 15, 2008 12:27:45 PM ~ Paae58af88 1
IWhich of the following indications would be present one minute following a normal manual Rx trip lfrom 100% ~x power?
/All BF19s -- and BF40s shut AND both SGFPs ~ __ tripped
__ due to the Feedwater Isolation signal.
_____~_ ~ ~ _ _ ~ _ _ _ _ _ _ ~
bll BF19s and BF40s shut due to the Feedwater Interlock signal.
~ ~ _ ~ ~
_ Steam
[Main _ _ Dump ~ ~ demand signal of -50%. ~ ~ _ _ ~ ~ . _ _ _ _ __~__
k x Dower is -1 X I 0-8 AmDs in the IR.
R
-~
~~
' IMain Feedwater System
~- .- _ _ ~ _ _ _ ___ ~ - __ --___
E - l m d g e of Main Feedwater System design feature(s) and or interlock(s) which provide for the following: 1
_ _ _ ~ _ _ ~ _ _ _-_______
feedwater reduction on plant _trip _ _ _ . __~ __ _ ___~
~ __________________ 7 3F A is incorrect becausea FW Isolation signal is generated at 67% SG NR level, and is not present. B is ,
correct because the FWI actuates with Rx trip breakers open and Tave c554,which occurs -1 7 seconds 1 ing a trip from 100Yo power. C is incorrect because steam dump demand will be much lower than 1 I50%, approximately 3-5%. D is incorrect because Rx power will be much higher thanlXl0-8A. The (promptdrop to 2-3 %, then -1/3dpm lowering. Rx power will be approximately 2 decades higher than 1
1x10-8 I RPS-Feedwater Control and Isolation Logic 1 CN&FDWE008 LOR Identify and describe the Control Room controls, indications, and alarms associated with the Condensate and Feedwater Svstem. includina:
The Control Rook location of Condensate and Feedwater System control bezels and indications. (Licensed Operator & STA only The function of each Condensate and Feedwater System Control Room control and indication. (Licensed Operator & STA only)
The effect each Condensate and Feedwater System control has upon Condensate and Feedwater System components and operation. (Licensed Operator & STA only)
The plant conditions or pennissives required for Condensate and Feedwater System Control Room controls to perform their intenc function. (Licensed Operator & STA only)
The setpoints associated
~-
with the Condensate __
and Feedwater System control___
room alarms. (Licensed
____ Operator & STA
-~
only) ~-
1 Tuesdav. Julv 15. 2008 12:27:45 PM r -
Page59of88
r~-- .
L-
~~ ~~~ ~
Which of the following describes the effect of 21 AFW pump failing to auto start on a normal Rx trip from 100% power? -_ _ _ ~ _ ~ _ _ _ _ _ ~ _ _ _ _ _ ~ _ _ _ _ _ _ _
lovercooling of the RCS during the initial 5 minutes folllowingthe trip. 23 AFW pump speed ishould NOT be reduced, and overfeeding _ _ ~_____~ of 23 and 24 SGs will occur. _.
_______ .~ __________ ~ _ _ _ ~ _ _
of the RCS during the initial 5 minutes folllowing the trip. 23 AFW pump speed be reduced, and overfeedina of 21 and 22 SGs will occur.
action to throttle the 21-24AF11, S/G LEVEL CONTROL VALVES, will be required to ~
the SGs, since 23 AFW pump will NOT be secured unless BOTH AFW EOP-TRIP-1 Rx Trip or Safety Injection.
/Operatoraction to throttle the 21-24AF11, S/G-LEVEL CONTROL VALVES, will be required to
'prevent overfeeding the SGs, since 23 AFW pump speed will NOT be lowered to minimum ispeed unless BOTH AFW pumps are running in EOP-TRIP-2, Rx Trip Response. ~-
R
~ IAuxiliary / Emergency Feedwater System 1K6. Knowledge of the of the effect of a loss or malfunction on the following will have on the Auxiliary /
__ 1 IEmergency Feedwater System: --- ~
_ _ _ _ ~ ~ -
iK6.02 Pumps
_ ~ ~-
~. ..
'262.7
-- -~ I 21 MDAFW pumpsupplies AFW flow to 23 and 24 SG. 23 TDAFW pp supplies all 4 SGs. Following a "normal" Rx trip, operators will transition to TRIP-2 after the Immediate Actions of TRIP-I are performed.
After stopping the SGFPs in TRIP-2, step 3, 23 AFW pp speed is lowered to minimum or 22E4 Ibmlhr.
Since there would be no flow to 23 and 24 SGs if speed was lowered to minimum, operators will throttle
/theAFI I s to balance flow to each of the SGs and maintain levels and pressures aproximate. A and B are incorrect because overfeeding will NOT occur since operators are directed to lower aFW flow. C is bcorrect because operators will have~transitioned _ _ _ _to_TRIP-2. ___________.~
1 AFWOOOE004 NCT Describe the function of the following components and how their normal and abnormal operation affects the Auxiliary Feedh System :
Motor-driven Auxiliary Feedwater Pumps Turbine-driven Auxiliary Feedwater Pump Turbine-driven Auxiliary Feedwater Pump Start-Stop Valve (MS132)
Turbine-driven Auxiliary Feedwater Pump Tnp Valve (MS52)
Turbine-driven Auxiliary Feedwater Pump Speed Control Valve (GOV) (MS53)
AFW Pump Alternate Suction Header Supply Valves (AF52s)
Motor-driven AFW Pump Recrrculation Flow Control Valves (AF140)
Motor-driven AFW Pump Discharge Flow Control Valves (AF21)
I
, T'uesday, July 15, 2008 12:27:45 PM___
I Page60of88
~ ___
h e n the following conditions:
I- 2A 4KV Vital bus experienced a loss of bus voltage.
1 - 2A EDG energized the 2A 4KV bus.
I- The SEC sequenced loads in accordance with MODE II*
I- The normal source to the bus is now available.
IWhich of the following describes the method for restoration of the normal power supply to the 2A 14KV Vital Bus in accordance with S2.OP-SO.DG-O001,2A DIESEL GENERATOR OPERATION?
IThe EDG is..
1 - _ ._
-_ ~ -
Itransferred to Droop Mode, placed in parallel with the normal feeder breaker closed and then lremoved from the bus.-
l_---___
-- ~ _____ _ _ _ ~
Itransferred to Droop Mode when the SEC is reset, unloaded and removed from the bus before lthe normal feeder breaker is closed.
1 -
lunloaded in Isochronous Mode and removed from the bus before the normal feeder breaker lclosed with the SEC deeneraized.
_ _ _ ~ - ____~__~
[unloaded in Isochronous Mode, the SEC is deenergized and EDG is placed in parallel with the normal feeder breaker closed and then removed from the bus.
R '
~
~-
A. C. Electrical Distribution
[-E m - d g e of the physic~connectionsa I
!Distributionand the following: _.____~_
7--
-- ~ _ _ _ _ _ ~
- 1 . 1 1 p A and B are incorrect because the EDG is NOT transfewrred to Droop mode after loading in lsochronus mode. Cis correc tbecause it is in accordance with the SO section 5.12. D is incorrect because there is
!no abilitv to Darallel across a 4KV vital bus feeder breaker.
1 1
1 EDGOOOE008 LOR NCT Identify and describe the Control Room controls, indications, and alarms associated with the Emergency Diesel Genera includina:
The Control Room location of Emergency Diesel Generator control bezels and indications. (Licensed Operator & STA only)
The function of each Emergency Diesel Generator Control Room control and indication. (Licensed Operator & STA only)
The effect each Emergency Diesel Generator control has upon Emergency Diesel Generator components and operation. (Licensc Operator & STA only)
The plant conditions or permissives required for Emergency Diesel Generator Control Room controls to perform their intended function.
I L--__--
Tuesday, July 15, 2008 12:27:45 PM 1 Page61 of88 j
1 Tuesday, July 15, 2008 12:27:45 PM 1 Page62of88 1Given the following conditions:
~- Unit 2 is operating at 100% power.
- 2C Vital 4KV Bus is aligned to 24SPT (breaker 24CSD closed).
1-- Power is lost to 2C Vital 125 VDC Bus.
Prior to restoring power to the 2C DC Bus, 24 SPT is deenergized.
[Which of the following describes the status of 2 6 4KV Vital Bus for these conditions?
~~ ~ ~~ ~ ~~
Application 8/25/2008'
. - ~~
Plant Systems 3
~
I
~
062 1A.C. Electrical Distribution lK1. knowledge of the physical connections and/or cause-effect relationships between A.C. Electrical I
,Distributionand the following: I
~_________ ~-
1K1.0T1, DCd.lsKution . - ~-~ i~G-4.o
~-
~ ~ ~
-- . .. ~ ~~
~ _ _ _ _ _ _ _
DC power is required to operate relays and contacts for the 4KV vital bus breakers. When DC power is lost, breakers will remain "as is". The EDG breaker can not close onto the bus even though it is
,deenergizedbecause one of the interlocks
~~ ~ ~ to _
shut the _
EDG output breaker is both infeedbreakers open.,
11 DCELECE013 NCT Given Dlant conditions, relate the DC Electrical Svstem with the followinq:-
AC Electrikl System Battery charger and battery Battery ventilation system
~-
I L--
I I 1 Tuesday, July 15, 2008 12:27:45 PM I -Page63 of 88
- ~~ ~- _______. - --
of the following choices identifies the threshold for UNSAT ground detection on a 125VDC bus, and the method in which operators perform ground isolation IAW S2.OP-S0.125-0004,
~125VDCGround -- Detection? ~ ~ _ ~~
_ ~
1500M Ohms. Transfer to the backup battery charger to determine if the I/S chargeris the lcause of--the ground. -
L __ _____
- ____ ~ ~ ~ _ _ ~-_ _ ~-
500M Ohms. Individually deenergize then reenergize each load on the bus to determine the iground location.
.. - .. ~ ~. ~ ~ . .... ~~
I/S charger is the Ohms. lndividuallydeenergize then reenergize each load on the bus to determine the 1D.C. Electrical Distribution ~
~
- p-p______-_p-p m-lAbilityo-(a)predict the impacts of the following on the D.C. Electrical Distribution and
/predictions, use procedures to correct, control, or mitigate the consequences of those abnormal operation:
~
~~ '
COT7 Grounds - _ _ _ ~ _ _ _
~ _ _ ~ _ _ _ _ ~
---12.5i3.2*
I-
~~
The ground detection procedure has operators isolate individual loads. The ARP for low battery voltage 1 has operators transfer to the standby battery charger if bus voltage is low, and battery current is present, ~
- XI .
- . I I L _ I y _ II_ %Ar.. -
rlCFLECEOOR
~
- -- - -~ ldentifv and describe the Control Room controls. indications. and alarms associated with the DC Electncal System, including The Chntrol Room location of DC Electrical System control bezels and indications. (Licensed Operator 8 STA only)
The function of each DC Electrical System Control Room control and indication. (Licensed Operator & STA only)
The effect each DC Electrical System control has upon DC Electrical System components and operation. (Licensed Operator & S' only)
The plant conditions or permissives required for DC Electrical System Control Room controls to perform their intended function.
~ (Li tor & STA only)
I I I 1
L- Tuesday, July .___
15, 2008 12:27:45 PM I Page64of88
~ ~~
IGiven the following conditions:
Operations Management has received information from a Vendor about certain components associated with the 28 VDC Electrical Distribution system.
The Vendor is performing testing at their manufacturing facility to prove compliance with applicable standards. This test data is expected to be received in 24 hours2.777778e-4 days <br />0.00667 hours <br />3.968254e-5 weeks <br />9.132e-6 months <br /> by Salem.
As a precaution, the Operations Manager wants the 28VDC system monitored much more closely than normally required for the next 24 hours2.777778e-4 days <br />0.00667 hours <br />3.968254e-5 weeks <br />9.132e-6 months <br />.
Which of the following identifies the method in which this additional monitoring should be made known to the shift operators IAW OP-AA-102-104, Pertinent Information Program?
~ ~
/issue astanding Order. ~ ~~ _ ~ _ -
~
lprepare___ a Temporary Modification (TMOD). _ _ _ _ ~ ~ _ _ - ~ _ _
__ ___ -~ ________ --
[prepare-__ a Current Operationally Related Events (CORE) report. -~
-___ .- - _ _ . ~ _ ______
r--
i2.1 _____---
IConduct Of Operations I I
~- ~ -~ ~ ~ -~ -
I 12.1.I5 1 IAbilitv to manaae
_- - short-term information~such _ _ as _
niahtand w _ _ _standha orders.
w --______ I231 30 New Rev. 2 Supplement IWA is actually.. ."KNOWLEDGE OF ADMINISTRATIVE REQUIREMENTS 1 FOR TEMPORARY MANAGEMENT DIRECTIVES, SUCH AS STANDING ORDERS. NIGHT ORDERS. I OPERATIONS MEMOS, ETC." Daily Orders: Are
!short-term,written philosophy and instructions from Operations IManagementto Shift Crews providing information which includes, but is not limited Ito:
1- Special plant operations,
!- Operating administrative requirements, l- Priorities, I- Manpower requirements and availability, 1- Equipment deficiencies, 1- Housekeeping, 1 - Special data taking, 1- Orders to cover backshifts and weekends, 12.1.1. Daily Orders are typically valid for one day, weekend or holiday period. B is incorrect because (StandingOrders are for longer term conditions. C is incorrect because there is no modification taking Mace. D is incorrect because it is information out to shifts. not direction to do anvthina.
__ _ _ _ _ ~ ~
- ~
limitations in the DC Electrical System
~_ procedures r- ~~
-Tuesday, i -
July 15, 2008 12:27:45PM 1- Page65of88 -!
1 Tuesday, July
-- 15, 2008 12:27:45 PM I Page66of88 1
Given the following conditions:
A total loss of all AC power occurred at Salern.
'1 Operators were successful in restoring power to a single vital bus with its respective EDG.
- The EDG output breaker just tripped on bus differential.
Which of the following describes how the EDG will operate following its output breaker trip?
IWith NO oDerator action. the EDG.. .
L- ~- --___________
~~- -
lwill run indefinatelv unless it is manuallv shutdown.
~ ~ _ _ - _ _ _ _ ~ - _.________
output breaker tripped on bus differential.
~____ __ -______---- . _ _ _ ~ _ _ _ _
hipped when the EDG local MCC lost power since the Fuel Oil pumps lose power.
__ . - ~ _ _ ________. __
b i l l continue to run until its Fuel Oil Day Tank empties due to the loss of power to its Fuel Oil ITransfer Pumps.
I power supplies to the following: - -
rrect because the Fuel Oil Transfer pumps are powered from the EDG local MCC, and have no power until AC power is restored to the bus. B is incorrect because the Bus Differential specifically does INOT trip the EDG, even though it is still an active trip after the SEC initiated EDG start. C is incorrect land D is correct because EDGs have shaft driven fuel oil pump specifically so they do not need external ~
2A Diesel Generator Operation 4 ~
-_i L
EDGOOOEO04'
- - - comDonents NCT Describe the iuT&on of the followina . and how their normal and abnormal oDeration affects the Emeraencv Dies Generators:
Lubricating Oil System Jacket Cooling Water System Fuel Oil System Starting Air System Turbo-charger Turbo Boost Air System Exciter 0 Regulator GovernodSpeed Control _ _ _- ~ - ____ - _ _ ~ ~ _ _ _ _ _ _ _ _ _ _
~~ ~
I I Page 670f 88 ~
- - __ - - - _ _ _ _ _ _ _- _~. -.-
p v e n the following conditions:
I1- Preparations are in progress to perform a release of 21 Waste Gas Decay Tank IAW S2.OP-SO.WG-0008, Discharge of 21 Gas Decay Tank to Plant Vent.
The 2R41A rad monitor is declared INOPERABLE due to an intermittent power supply failure.
Which of the following describes the effect this will have on releasing 21 GDT?
- ~ _ _ - _ _ _ _ _ _ _ _ _ . ~ _ _ _ _ _ _ _
/CAN be started ONLY if the 2R41D remains OPERABLE.
~. ____________-_________
CAN be started after double samples/ana&sis are performed, and double release rate L_______--_-
1 i L t i o n i a__r e performed.-
~ -- - . -
-___~_
_____ ~
~-
lcan NOT bestarted since there is no low range noble gas monitor available to automatically lterminate the release on high-radiation. --
- -~ - - --
lcan NOT be started since the 2NDl7572-2WG41 Waste Gas Decay Tank Block Switch is uired to be BLOCKED when 2R41A or D is INOPERABLE.
IWaste Gas Disposal System lA2. IAbility to (a) predict the impacts of the following on the Waste Gas Disposal System and (b) based on 1 those predictions, use procedures to correct, control, or mitigate the consequences of those abnormal operation:
___ ___~~____
I Ihrn
~- ~ - ~ _ _
by051 Power failure to the ARM i d PRM Systems - ---
- ~ - -
_____-_~
[TheGDT can bereleased without the R41A or D if the double sampling and analysis, and double release 1
'rate calculations are performed IAW Section 3.0 of Attachment 2. A is incorrect because the 2R41D does' hot have to be OPERABLE. C is incorrect because there does not have to BE a low range noble gas monitor as long as B above is performed. D is incorrect because while it is true that the monitor will be 1 1
blocked. it does not Dreclude startina the release.
)Tuesday, July 15, 2008 12:27:45 PM Page68of88 -
ns are in progress to perform a release of 21 Waste Gas Decay Tank IAW Gas Decay Tank to Plant Vent.
OPERABLE due to an intermittent power escribes the effect this will have on releasing 21 GDT?
~~
~-
remains OPERABLE.
- - ~ - _ _ _ _ _ ~ -
s/analysis are performed, and double release rate kaiculations
~ - are performe --__________
--- - - ~- ~~~ - -
o low range noble gas monitor available to automatically terminate the release on hiqh can NOT be started since the 2ND17\2-2WG41 Waste Gas Decay Tank Blockswitch is lrequired to be BLOCKED when 2R41A blr D is INOPERABLE.
The GDT can be released without the R41 rate calculations are performed IAW Section not have to be OPERABLE. C is incorrect b
- Y ; T" ,
-_-_ _
- _ _._ _- ~
WASGASEOl I LOKNCT ral requirements associated with the-Radioactive Waste Gas S g an explanation of
~-
F X J e s d a y , July 15, 2008 12:27:45 PM I Page69of88 I
_ _ _ _ _ _ ~ ~ _ _ _ ~ _ _ _ . . _ _ _ _ _ ~ . _ _ ~ _____ -
while removing a source, RP personnel drop it on the floor 10 ft. from an area monitor. If this area
_-I'monitor is reading
~ _ _ 2 Whr, what is the approximate .___ dose rate 1 ft. from the dropped source?
120 Whr.
L------__ _ ~ __ ~ _ _ _ _ _ _ ~
R .~
072 1 [Area Radiation Monitoring System 1 6:Fowledge
___~-
of the operational implications of the following concepts as they apply to the ARM system:-
. ~ _ _ _ _ _ _ _
___1--
1 Radiation theory, including sources, types,~units, _ _ and_ _effects
_ -________ ~- I 2.71 3.0 formula for a point source is Dose Rate 1 = Dose Rate 2 times the product of distance 2 squared divided by distance 1 squared. DR1= 2Wh times 100/1 DRI = 200 Whr. ____
1 RADCONEOOG Perform the following, using the principles of time, distance, and shielding:
a) Describe the methods used to shield personnel from alpha, beta, gamma, and neutron radiation.
b) State how time, distance, and shielding are used for dose reduction.
c) Define stay time and perform calculations to determine stay time or dose.
,~ d) Calculate the dose rate at a distance from a gamma point or line source.
e) Use the tenth thickness equation to solve shielding problems
~ ~ . . . _ _ _ _ _ _ _ _ ~ _ . ~ ~ ~ ~ ~ .
_____..___ _ ___..____~
- - ~ -
/Direct From Source
[ Tuesday, July__ 15, 2008 12:27:45 PM i Paae70of88 ~
---~__________ _________
piven the following conditions:
' Due to Main Condenser problems, a rapid power reduction on Unit 2 has been performed I_ from 100% at 15% per minute, IAW S2.0P-AB.LOAD-0001.
1- The power reduction is stopped at 22% when the problem is corrected.
- The Reactor Operator reports that RCS temperature has just dropped below 541O F 1
I and continues to slowly lower.
'Which L-- of the following identifies
_- ~-
the appropriate
_ _ _ _ ~action
_ _ _ _for restoring Tavg IAW AB.LOAD? _ .. -
/Place Rod Control in MANUAL and slowlv withdraw control rods:-
~ _ _ - ____.______ ~_ _____
ITrip the turbine and go to S2.OP-AB.TRB-0001.
~ __ _ ~ _ _ ~____
krip the Rx and ao to EOP-TRIP-I.
____ _ _ _ _ _ ~ ____
hitiate a dilution.
1 [Process Radiation Monitoring System
-~
1
[
1 62;
[Conduct Of Operations
__-- __~____ -~ _~
)2$ Ability to use plant computer to obtain and evaluate parametric informationon system or c o F q7 3.0: -
3.0 status.
~ _ _ . _ ~ - ~__ _ _ ________________~__
.__~______ ~_
New Rev. 2 Supplement 1 WA is 2.1.39 Knowledge of Conservative decision making practices AB.LOAD Rev. 16 CAS, Att 2, gives 4 7
(thingsto do to raise temp. Reducing turbine load is that only choice which is in the list. Tripping the Rx is inot atmroriate since the Droblem which caused the load reduction has been cleared as stated in the stem. 1 1 Tuesday, July 15, 2008 12:27:45 PM I Paae-71 of88 1
1Given the following conditions:
- BOTH SW header pressures are 118 psig.
- 11,13, and 15 SW pumps are in service.
!Which of the following describes how the Unit 1 Service Water System will respond to a leak downstream of the SW26 in the underground piping outside the SW structure?
-- _ _ _ _ _ _ _ _ _ _ _ _ _ ~ ~ - ~ _ _ _
[The SW pump selected to auto will start when EITHER SW header pressure drops to 99.5 lPsigm ~- -~ ~- _ _ _ _ _ _ _ _ _ _ ~ _ ~ -
/ T h e W 7 0 8 , SW B a y l S W P R E S S - E N K K REG VALVE will throttle in the SHUT ldirection to raise SW header pressure.
- - - - _ _ _ - _ _ _ _ _ _ . ~ ~ _ ____
_ __ ~_ _ _ _ _ _ ~ .
~~ ~ ..~
.. ~ - . . ..
.~.~~~~~~~~
.......~ ~~~~ ~ ~ ~ . ~ ~ ~
~ ~ ......... ~
~~~~~~ . ....... ....~ ~.~~ .~
~
...... ~
..~ ~
k h e 1SW31I,SW Bay 3 SW PRESS CONT HDR REG VALVE will throttle in the SHUT ldirection to raise SW header pressure.
I toverDressure valves, and are not normallv oDen. 1 I ",
- d I d i n bmmg P- e',, '6 22-a _II_- ""M ate the setpoints. coincidence, blocks and permissives for automatic actuations associated with the Servi Bavs
)%esday, July 15, 2008 12:27:45 PM 1 Page72of88 1
r
~- - ~ __ _ _ ~_________.______.. _ _ _ _ ~ _ _ ~
Which of the following identifies how the Emergency Control Air Compressors (ECAC) respond to L-lowering Control Air header pressure on _~ both
_ _ the
_ _ "A" _ _ "B" headers?
_ _ and _ _ _ _ _ ~ _ -~
The Unit 1 ECAC will start at 85 psig on CA header "B". If ControlAir header "A" pressure continues to dearade and reaches 80 psis, the Unit 2 ECAC will then start.
_ _ _ _ _ ~
. _ _ ~ _ _ ~ _ _ _ ____
iThe Unit 2 ECAC will start at 85 psig on CA header "B". If Control Air header " A pressure lcontinues lo dearade and reaches 80 psia. the Unit 1 ECAC will then start.
1
~__.____________________._______
IBOTH ECACs will start at 85 mia on their respective CA header.
IBOTH ECACs will start at 80 &a on their respective CA header.
-1 R M
.~
nstrument Air System __- __-
I A ~ --
. [Ability to monitor automatic operations of ___-_______
the Instrument Airsystem including:-___ -
E 0 1 I Airpressure ~_
_ _ - _ _ ~
-13.13.2 EFAC operation is independent of the other ECAC. Unit 1 ECAC starts on a signal from,-and supplies, ICA header 8 , reaching 85 psig. Unit 2 ECAC starts on a signal from, and supplies, CA header A, reaching I85 miu.
1 I
CONAIRE006 NCT Outline the interlocks associated with the following Control Air System components:
Emergency Control k r Compressors (ECACs)
Station Blackout Compressor
_ . _ _ _ _ _ ~ ~ ~ _ _ _ _ _ _ ~ _ _
__________ I 1
L Tuesday, July 15, 2008 12:27:46 PM I Page73of88
- 1
IGiven the following conditions:
I
~- Unit 2 is operating at 100% power.
- OHA A-7, FIRE PROT FIRE, actuates.
- 2RP5 indicates the following:
- Zone 59 - Air and Water Deluge, Containment El. 100 Panel 335 is lit.
1 - Zone 74 - Smoke and Fire Detector, Containment El. 100 Panel 335 is lit.
Wi- c h of the-- following is the required action? ~~ ~~~~ ~~ ~
~~
Dispatch an NE0 to unlock and open 2FP239, Fire Protection Manual Isolation Valve, in the mechanical penetration area.
-__ ~ _ _ _ _ - ____.
kerify Fire Protection water flow to containment on 2RP5.
b p e n 2FP147, Fire Protection Containment Isolation Valve on 2RP5.
- - ~ _ _ - _______ ___
Verify 2FP147, has automatically opened.
stems 2
~
~~ l lFire Protection System I
lA3. IAbility to monitor automatic operations of the Fire Protection System including:
~ ~ ~
1 Actuation of fire detectors -_11'219133
~~
7- ~~~
lA3.03 ~~~
~ ~ ~ ~~ - -
[Containment Fire Protection water is normally isolated. When the 2 alarms are receievd for Zones 59
,and 74, this is positive indication of the fire in containment, and the ARP directs the opening of the lFP147. The manual isolation valve is locked open normallv. There is no fire Drotection flow indication on 12RP5.
-~~
FP147 is not an automatic valve. ~~~ I OHA Window A Response
- "I *-(s-wv* L *( ^.
I -
FIRPROE008 ldentifv and describe the Control Room controls. indications, and alarms associated with the Fire Protection Svstem. includina The Control Room location of Fire Protection.Systern control bezels and indications. (Licensed Operator & S f A only)
The function of each Fire Protection System Control Room control and indication. (Licensed Operator & STA only)
The effect each Fire Protection System control has iipon Fire Protection System components and operation. (Licensed Operator STA only)
The plant conditions or permissives required for Fire Protection System Control Room controls to perform their intended function.
(Licensed Operator & STA only)
~~~
The setpoints associated with the Fire Protection System control ~ room ~ alarms.
~ - (Licensed
- _ _Operator _ _& _ STA-only)
~ ~ - _ _ _ ~~ ~ _ ._ _ _. _ _ ~ _ _ ~ ~~
I .--
i __
-~ __
~~
r--
~ ~~~
_ ~ . _ _ _ _ _ _ _ _ ~ _ _ _ ~ _ _ _
- 7 I y I J e s d a v , July 15, 2008 12:27:46 PM l Page74of88 1
2 is in Mode 6 and core alterations are in progress.
Which ONE of the following conditions, taken by itself, will result in a loss of the ability to establish icontainment closure within the 1 hour1.157407e-5 days <br />2.777778e-4 hours <br />1.653439e-6 weeks <br />3.805e-7 months <br /> time limit if reauired?
__--___~__
bource Ranae NI count audible indication is lost in containment.
~ _ _ _ _
lThe containment 100' Airlock inside Door becomes cocked and can NOT be closed.
-- ~~~~ ____- ___-__
2 1 SG secondarv manwavs removed with 21 MS167 oDen and its oPerator disconnected.
Only 6 bolts can be found for securing the Containment Outage Equipment Hatch (OEH) inside door.
~- ~ _ _ _ -
1103: , Lontainment System I and through the MSLl valve outside containment, and no way to close the MSIV would result in icontainment closure beina unable to be established in 1 hour1.157407e-5 days <br />2.777778e-4 hours <br />1.653439e-6 weeks <br />3.805e-7 months <br />.
I L-__----
Tuesday, July 15, 2008 12:27:46 PM 1 Pase75of88 1
~ ~- _____ ____ ~~ - ~
~-
h h i c h of the following choices contains ONLY the information which is required to be read to lagencies being notified of an Alert at Salem through the Initial Contact Message Form (ICMF)?
I
- 11. Communicator Name 111. Current RX power level 1111. Radiological Release status IV. Wind speed and direction V. Description of event IVI.
A-Status of -_unaffected unit ~~- __________
~-
[I, Ill, IV.-- ---___ __-____
__ -~ - _ - __--____~-_____
Ill. IV. VI.
iiil, VI.
2.1_- Iconduct of Operations
.________ -~ ~
1 ake accurate, clear and concise-verbal reports.
__~________
~~ --=-jTJ3.6 ment 2 is usedfor an ALERT. It requires in Section I Communicator name, Salem Unit number.
ection II: time of declaration, date of declaration, EAL#, Description of event. Section Ill: Radiological Release status. Section IV: Wind meed and direction.Section V. No PAR recommendation.
7-j Tuesday, July 15, 2008 12:27:46 PM I Paye76of88 1
_ _ ~
. _ _ _ . _ _ _ _ _ ~ . - _______ -- ~
- ______~
which of the following identifies the purpose of the Containment Spray System?
Maintain containment pressure less than the design pressure of 47 psig following a Loss of ICoolant Accident (LOCA).
L - -
_ _ _ _ _ _ . ~ - ~ _ _ _ _ _ ~ _
7 - - _ _ _ ~ _ _ ~ ~ ~ _ _ _ _ ~ _ _ _ _ _ _ _ _ _ _ ~
'Maintain containment pressure less than the test pressure of 54 psig following a Main Steam Line Break (MSLB) inside containment.
Inject a mixture of borated water and Sodium Chloride (NAClGnto the containment atmosphere following a LOCA to-_____ minimize exposure-to the public following a LOCA.
_ _ _ _ _ _ _ _ _ . ~
(Inject a mixture of borated water and Sodium Hydroxide (NAOH) into the containment
,atmosphere following a MSLB with failed fuel, to minimize exPosure to the Public.
I
/ensure that containment Dressure does not exceed the desian Dressure of 47 Dsia....". I Tuesday, July 15, 2008 12:27:46 PM i Page77of88 ]
1 Given the foIlowing conditions:
~- Unit 2 is operating at 100% power.
1- 21 and 22 AFP's were declared inoperable at 0405 on April 1st due to being sprayed with leaking SW and failing a subsequent motor megger.
A unit shutdown is commenced at 0615 to comply with the associated Tech Spec I Action Statement.
Which of the following identifies a situation where the shutdown would be stopped prior to taking the unit off line?
paintenance reports that they are certain at least one of the AFW pump motors can be dried
- and restored to OPERABLE status by noon of that day.
r~ ~ _ _ _ ~ _ _ _ _ _ _ _ _ _ _ .~
'The Electric System Operator (ESO) calls the Control Room and orders the unit shutdown be haced on hold to maximize aeneration.
[More than one Control Bank D rod is identified as beina stuck with the Rx >50% Dower.
not stop the shutdown. If the candidate uses the 0615 commencment of the shutdown as the start of the 6 (hour clock, they may think the shutdown could be stopped. B is incorrect because the ESO does not have ithe authoritv to order anvthina related to Rx oDeration.
Misaligned/lmmovable __ Control Assemblies _ _ _ . ~ - _ _~.~ ~
~- - ~
1 Y __-- ---
b *-.h i wmjectivrri;
-*-_1 _$*A " Ad&"
I -
n dealing with Auxiliary Feedwater System operability, examine the situat Specification action. (License Operator and STA only)
NCT State the Technical Specification associated with the component, parameters and operation of the Auxiliary Feedwater Syste including the Limiting Condition for Operation(s) (LCO) and the applicability of the LCO(s) (Non-licensed Operator)
I -
1 Tuesday, July 15, 2008 12:27:46 PM 1 Paae78of88 1
- Unit 2 is in MODE 6.
- Containment Closure is established IAW S2.OP-ST.CAN-0007, Containment Closure.
/How are remotely operated valves used to ensure containment closure identified IAW ST.CAN-
~0007?
is affixed to the Control Console bezel.
_ _ _ _ . _ ~ _ _ ___________-_
Generic Knowledge and Abilities 3 [I 94001G226- ~ 1 RIC 11 - - 1 I 70; (2.2
__-_ (EquipmentControl __~________ ~
I v
12.2.26 Knowledge of refueling administrative requirements.
1-37 New Rev. 2 Supplement 1 WA is 2.1.40.Different number, but same statement. WA value is RO-2.8, ISRO-3.9 Procedure P&L 3.5 states that an INFO tag will be used to identify a remotely operated valve I used. to ensure containment closure. The distracters are all types of control console "tags" that are
- applied in different circumstances. i CONTMTE007 L-____
ldentifv and describe the Control Room controls. indications. and alarms associated with the Containment and Containment SUDD(
I ,
Systems, including:
The Control Room location of Containment and Containment Support Systems control bezels and indications. (Licensed Operatoi STA only)
The function of each Containment and Containment Support Systems Control Room control and indication. (Licensed Operator &
STA only)
The effect each Containment and Containment Support Systems control has upon Containment and Containment Support System components and operation. (Licensed Operator & STA only)
The plant conditions or permissives required for Containment and Containment Support Systems Control Room controls to perforr their intended function. (Licensed Operator & STA only)
The setpoints associated with the Containment and Containment Support Systems control room alarms. (Licensed Operator & Sl
~ ~ -_ _ _ _ _-
only)
- _ __ _ _ __ _~ _ ~ __ __ ~_ ~- _ __ _ ___ _ ~ _ _ ~ _ . _ _ _ ~ _
L_- _ _ _ _ _ _ _ _ _ _ _ ~ ~ _ _ - ~ _ _~_~
. _ _ _ _ _ _ _ _ _ _ _ _ ~ ~ _ _ - _ _
_~
1 r y u e s d a v . Julv 15. 2008 12:27:46 PM I Page 79 of 88 -1
~- Unit 2 is operating at 100% power, with no active Tech Spec LCOs in effect.
1- 2C EDG is being C/T for a scheduled 30 hour3.472222e-4 days <br />0.00833 hours <br />4.960317e-5 weeks <br />1.1415e-5 months <br /> maintenancewindow.
In addition to the TSAS for the INOPERABLE EDG, 3.8.1 .I .b, which of the following identifies a TSAS, if any, required to be entered?
~~ ~
ems.
13.1.2.4 for 22 Charging pump. ~ ~ ~ ~
for Boron Injection Flow Paths.
3.1.2.2 ____ _~ - ~ _ _ ~-_____-
[No additional TSAS entries are required. ~ _ _ ~ ~
Id R ' Comprehension ISalem I& 2
-~ j 8/25/2008j i2.2 Equipment
____-______ Control ___ _ _ ~ ~_________
/2.2.34 IKnowledge
__- of the process for determining the internal andexternal effects on core reactivity.
~~
New Rev. 2 Supplement IWA is 2.2.36 Ability to analyze the effect of maintenance activities,
'degraded power sources, on the status of limiting conditions for operations. When addressing an IINOPERABLE EDG in TSAS 3.8.1.l.b, Action b.2 states "Within 4 hours4.62963e-5 days <br />0.00111 hours <br />6.613757e-6 weeks <br />1.522e-6 months <br />, declare required systems or I components supported by the inoperable diesel inoperable WHEN A REQUIRED REDUNDANT SYSTEM OR COMPONENT IS INOPERABLE. . . . I ' (My emphasis) This precludes the necessity for "cascading" I ltech specs to the components who have their emergency power supplied from the inoperable EDG, but lhave all redundant systems and components remaining operable. In this case, no other indication is provided in the stem that would lead to the conclusion that any required systems were not operable, and no other TSAS entry~is_required. _ _ _ _ _ ~ _ _ _ _ _ _ I I TECHSPE012 Describe the general requirements associated with Specifications 3.0.1 through 3.0.6 relating to implementation of the Technical SDecifications 1I Tuesday, July 15, 2008- 12:27:46 PM _ _.
I Page80of88
] -
L-Which of the following statements is TRUE regarding personnel exposure limits during an ALERT Emergency at Salem IAW NC.EP-EP.ZZ-0304, OSC Radiation Protection Response?
~ _ _ _ ~
iThe EDO's authorization for an Emergency Exposure can NOT be delegated, and MUST be made in Person.
-~ ~~~
b n operators yearly dose limit is automatically raised to 4500 mremregardless of their NRC
[Form 4 status. ~ ~ _ _ _ - ~ ~ ~
-~ ~~~
'An operators yearly dose limit is automatically raised to 4500 mrem ONLY if their completed NRC Form 4 is on record. - ~
~
[Emergency exposure can be authorized for someone who has previously received an
[Emergency --
Exposure, as long as the
- ~ -
_ previous
_ ~ _ ~ exposure _ _ _ _ resulted _ in _ less
~ than
~ _ 10_Rem.
12.3
~~-
[Radiation
~
' ' Control
____~~___~__ ~ ~ _ _ _ _
____ ~~
- J 12.3.4 j Knowledge of radiation exposure limits and contamination control, including permissible levels in 125)3-lexcess
_____ of those authorized. _ _ ~ _ _
A is incorrect because while it is true that the ED0 cannotdelegate his authority, he CAN authorize via telephone. B is incorrect because a current NRC Form 4 is required to be on file. C is correct 1
because of B above. D is incorrect because the rule for NOT considering an Emergency Exposure as an Emergency Exposure is if the resultant exposure for the year is less than or equal to 4.5R after lfinishing the exposure.
-~
'OSC Radiation Protection RADCONE003 L ,%A&"
Discuss the emerqency exposure limits in accordance with aDplicable station procedures. Dertainina to.
A. Automak dose extensions during an emergency.
I B. Position responsible for authonzing emergency exposure.
C. Guidelines for personnel receiving emergency exposure.
D. Emergency Exposure Limit for life saving E. Emergency Exposure Limit for Accident~Mitigation.
~ - ~_
- ~ _- ___ _
_ _ __ _ ~ - _ _ ~ _ _ _ _ _
I_--
-~___-___ _________-- ~
I - 7 T T u e s d a y , July 15, 2008 12:27:46 PM 1 Page 81 of88 1
-- - -~ _ -. -~ ~ _ _ _ _ _ _ _ _ _ _ _
p n the following conditions:
- Unit 1 is operating at 80% power.
- Radiation Protection technicians are transferring a used SFP filter cartridge from the filter to its packaging point.
Which of the following identifies a method used to reduce the amount of dose received by personnel while transporting the filter?
[Personnel not involved in the transport of the filter are restricted from bekg in Unit 1 OR Unit l2Aux Buildings during transfer of the filter.
____~-___ -~ ~~ _______
Full protective clothing (PCs) is worn by personnel transporting the cart to minimize likelihood of a discrete particle becoming ~- lodged in personal-clothing.
~
- ~ ~ ~ _ _ ~ ~ _ _ _ ~ _ _ _ _ _ ~ _ _
huxiliary Building ventilation supply fans are secured during removal and transport of filter to lminimize the sDread of anv loose surface Contamination.
Generic Knowledge and Abilities
~- -~
12.3 IRadiation Control _ _ _ ~ _ _ _ ~ -~-~
1 3.1
~
12.3.4 Knowledge of radiation exposure limits and contamination control, including permissible l e v r l 2 . 5 /
s of those authorized. -- .I I
- ~
New Rev. 2 Supplement 1 KIA is 2. ination hazards that may arise during normal, abnormal, or emergency conditions. A is correct. Spent filters are moved inside a water filled transfer cart. B is incorrect. Personnel will be warned by page announcement when highly radioactive filter is being moved, but access to the building will not be secured. C is incorrect because wearing PCs is not required. D is incorrect. Engineering controls like minimzing ventilation supply forced 1
onto a surface contamination is a good idea - in some cases, but ABV will not be realigned.
uul a) Describe the methods used to shield personnel from alpha, beta, gamma, and neutron radiation b) State how time, distance, and shielding are used for dose reduction.
c) Define stay time and perform calculations to determine stay time or dose d) Calculate the dose rate at a distance from a gamma point or line source e) Use the tenth thickness equation to solve shielding problems
-%esdav. JUIV 15. 2008 12:27:46 PM [ Paqe82of88 1
-- _______~~- __
r -- - ~ p
,The Unit 2 CRS has directed a Unit shutdown based on RCS activity exceeding TS 3.4.9 limits, Which of the following Tech Spec required actions performed after the Rx is shutdown is designed
,to limit the release of radioactivity in______-_______
the event of a subsequent SGTR? __-
L _ _
____- _ _ _ ~ ~ _ _ _ _ _ _ _ _ _ _ _ _
lMSlVs are closed.
IRCS is cooled down below 500F.
[Main steam dump valve setpoints are raised.
~
[Maximum condensate polishers are placed in service.
R l Knowledge and Abilities
-- _ _ ~ _ _ _ _ _ _ ~
A is incorrect because closing MSlVs would contribute to radiation release through SG a-t reliefs and/or Safeties if cooldown and depressurization was not performed in a timely manner. B is
[with the actions of TS.3.4.9
~
_ I I 6 Laam r__LLL - -u__--. -". _ 1 -. -
1 SGTROI E005 ldentifv Dossible radioactivitvr; hs for a steam senerator tube rupture. and describe how the actions in 2-EOP-SGTR-1 rninirn'ke the release potential I
L-Tuesday, July 15, 2008 12:27:46 PM
~~- _-
given the following conditions:
- Unit 1 is responding to a LOCA, and due to a Core Cooling PURPLE path is taking action IAW FRCC-2, Response to Degraded Core Cooling.
- While performing actions IAW FRCC-2, the STA identifies a Containment Environment PURPLE path is present due to containment pressure, and been verified by control console indication.
- No other RED or PURPLE paths are present.
IWhich of the following describes how the crew should proceed IAW OP-AA-101-111-1003, Use of
~Procedures? -~ - __ ~ _ _ _ _
rfMMEDlATELY transition to FRCE-1 since it is a hiaher order PURPLE Dath FRP.
- - ~~~
kransition to FRCE-1 at the completion of FRCC-2 even if the FRCE PURPLE condition has lcleared.
~ ~ _ _ ~
IF the PURPLE path for FRCC is currently clear due to actions performed in FRCC-2, THEN transition to FRCE-1.
-~
-- ~
[Transitionto FRCE-1 at the completion of FRCC-2 ONLY if the FRCE PURPLE condition is ktill present -- AND no RED or higher PURPLE path exists.
R 8/25/2004 Knowledge and Abilities 1194001G414-- I
~- ~ - -- ---- - 75)
(2.4 k r g e n c y Procedures / Plan - .
. __ I 2.4.14 Knowledge of general guidelines for EOP flowchart use. ~~
I OP-AA-101-111-1003, Use of Procedures states .....I' Once a RED or PURPLE FRP is I operating crew remains in the FRP until either the FRP directs a transition to another procedure (which 1 1
may be theprocedure in effect), or a higher priority challenge is detected, in which case thecurrent FRP is suspended and the operating crew transitions to the FRP directed by the higher-priority challenge." A is incorrect because FRCE is not a higher order PURPLE path. B is iincorrect because the transition is not made if the condition is not currently PURPLE. C is incorrect because once entered, a FRP is performed to completion OR when a higher order FRP entry condition is identified. !
A.-immediate Actions B.-Continuous Action Summaries C.-Communications ~
D.-Log Keeping E.-Application of Notes and Cautions
~ F.-Transitions G.-Adverse Containment
.~~~.. ....... . .....
~ ~ ...... . ~ ~ ~ ~ .
I y u e s d a v . Julv 15,2008 12:27:46 PM 1 Page84of88 1
I Tuesday, July 15, 2008 12:27:46 PM I Paae 85 of 88- 1
!Given the following conditions:
- Unit 1 Control Room has been evacuated due to a fire in the Relay Room IAW S I .OP-AB.CR-0002, Control Room Evacuation Due to Fire in the Control Room, Relay Room, 460/230 Switchgear Room, or 4KV Switchgear Room.
- The Reactor Operator assigned to perform Attachment 4 places the EDG Fire Emergency Bypass Switches 69/1, 69/2, and 69/3 in BYPASS while the EDG is running unloaded.
,Which of the following choices identifies how this will affect 2C EDG operation?
L The EDG...
1
_ _ _ ~~
_ _ ~ _ _
~~
_ _ ~ ~ .-_____
~ - ~ ~
voltage regulator willswap into lscochonus Mode regardless of its previous mode, since the EDG
___ __ can NOT be operated in parallel with the
~~ ______..____ ~-
grid when the control ~ _ _room _ _ _has _ _been
_ _ evacuated.
-~ ~
lwill trip due to the break-before-make switch characteristic, and must be manually restarted if ineeded for power.
room ventilation i l l automatically shutdown due to insertion of a "fire detected" signal into its operating circuitry.
~~ ___ _ _ _ _ _ ~ -
bill continue to run, and all non-SEC trips will be reinstated.
&2 1 1
GENERIC ,
~-
~
~
12.4 IEmergency Procedures / ___ Plan bx%]
L- Knowledge
~~
of fire protection procedures.
____~____________ ~
__._ -~
~ - _ _ _ _ _ _ _ _
Theswitches isolate the EDG from the SEC. A is incorrect because it does not affect voltage regulator
~ J:434 operation. B is incorrect because it does not trip the EDG. C is incorrect because it does not affect room ventilation.
Control Room Evacuation Due to Fire in the Control Room,
____Relay Room, 460/230Switchgear-Room, or 4KV Switchgear Room.
~~
1 LOR NCT Identifv and describe the lo ed with the Emeraencv
" , Diesel Generator incluc The location of Emergency Diesel Generator local controls and indications. (Licensed Operator & Non-licensed Operator only)
The function of Emergency Diesel Generator local controls and indications. (Licensed Operator 8 Non-licensed Operator only)
The plant and conditions or permissives required for Emergency Diesel Generator local controls to perform their intended function (Licensed Operator & Non-licensed Operator only)
The setpoints associated with the Emergency_
Diesel Generator
~
local alarms. (Licensed
_ Operator & STA_only) ~
L-Used Vision Q70480 as a bases for this question, and made 3 new distracters.
_____~ _____~________~
~- __-_ 1 Page86of88 1 F T u e s d a v , Julv 15, 2008 12:27:46 PM ~ ~~
I Given the following conditions:
- Unit 2 is operating at 80% power.
Unit 1 and Hope Creek are operating at 100% power.
ii 1-Unit 2 Main Generator is operating normally, with output of 1000 Mwe, 325 MVAR OUT.
Unit 2 Main Generator H2 gas pressure is 75 psig.
The Electric Sytem Operator (ESO) calls the Unit 2 control room and reports a grid disturbance has tripped several non nuclear units off line.
- All transmission lines remain in service.
- The ESO requests Salem Unit 2 to adjust its MVAR loading to the maximum allowed per IEngineering Evaluation A-5-500-EEE-1686.
Which of the following identifies how the change will be accomplished, and identifies the MVAR loading CLOSEST to what the maximum MVAR loading will be? -~
~-
~ ~
[Utilize the Voltage Regulator AUTO ADJUSTER -~~~ to obtain 400 MVAR OUT.
~~ _ ~
[Utilizethe Voltage Regulator AUTO ADJUSTER to obtain 650 MVAR OUT. ~~
!Transferthe Voltage Regulator to MANUAL IAWS2.OP-SO.TRB-0001, Turbine Generator IStartuo OPerations. Utilize the MANUAL ADJUSTER to obtain 400 MVAR OUT.
__ __.__ -~ __
r lTransfer the Voltage Regulator to MANUAL IAW S2.OP-SO.TRB-0001, Turbine Generator IStartuD ODerations. Utilize the MANUAL ADJUSTER to obtain 650 MVAR OUT.
gas pressure is 75 psig. Using 1000 MW and the 75 psig line the intersection for maximum MVAR loading is 650 MVAR OUT. To reach this, the operator djusts the Auto Volts Adjust PB to RAISE which raises excitation to the Main Generator and causes MORE MVAR to be sent to the grid. The distracters are either to use the manual volts adjust, or the limit for 100% electrical loading, not the 80%/1OOOMWe I
!as stated in the stem. 1 Engineering Evaluation A-5-500-EEE-1686
_ ~
~ _ ~ . . _ _ _ ~~
~~Turbine Generator
~ ____Startup Operations ~~~~
I)- a #.Gh*dw p - $8 1
-_____. 1 - ,
GEN002E012
_~ IdentiW and describe the Control Room controls, inoications, and alarms associated with the Unit 2 Main Generator and Hydrogen Cooling System, including:
a) The Control Room location of Unit 2 Main Generator and Hydrogen Cooling System control bezels and indications b) The function of each Unit 2 Main Generator and Hydrogen Cooling System Control Room control and indication c) The effect each Unit 2 Main Generator and Hydrogen Cooling System control has upon Unit 2 Main Generator and Hydrogel Cooling System components and operation l d) The plant conditions or permissives required for Unit 2 Main Generator and Hydrogen Cooling System Control Room control perform their intended function e) The setpoints associated with the Unit 2 Hydrogen Cooling System control room alarms Tuesday, July
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15, 2008 12:27:46 PM
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Q75 A-5-500-EEE-1686
.. Rev.
. _Tables and Unit 2 Curves.
8, ._
L J u e s d a y , July 15, 2008 12:27:46 PM , ! Paqe88of88 1
U S . Nuclear Regulatory Commission Site-Specific Written Examination Name: ~ Region: I
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Date: 8/25/2008 Facility: Salem 1 & 2
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License Level: SRO Reactor Type: W
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Instructions
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Use the answer sheets provided to document your answers. Staple this cover sheet on top of the answer sheets. The passing grade requires a final grade of at least 80.00 percent. Examination papers will be collected SIX hours after the examination
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Applicant Certification
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All work done on this examination is my own. I have neither given nor received aid.
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Given the following conditions:
l- Unit 2 is operating at 100°/o power.
- A total loss of off-site power occurs, and the Rx automatically trips.
l- ALL Control Bank "A" control rods fail to trip and remain fully withdrawn.
l- ONLY 2C vital bus energizes from its EDG, 2A and 2B vital busses remain deenergized.
- As 21MSI0 starts to open as expected for RCS temp control, it fails full open.
1- The following indications are presents during the second pass through the Immediate Actions of EOP-TRIP-1, Rx Trip or Safety Injection:
- ALL PR Nl's indicate 0% power.
- IR NI indication is 3x10-6 A and stable
- IR SUR is 0.0 dpm on BOTH channels.
Which of the following identifies the current status of the Rx , and how operators __ should respond?
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The __
reactor trip is NOT confirmed, go to~
LOPA-I
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, Loss p of ~All AC
_ Power.
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IThe reactor trip_ is
- _~ _ _confirmed,
_ - ~ continue in TRIP-I , after Immediate
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-__ are completed.
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The reactor trip_is_NOT
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confirmed,
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go to FRSM-1, Response to Nuclear Power Generation.
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The reactor trip is confirmed, transition to TRIP-2, Rx Trip Response after verifying SI not
~ 2 1 - - bonduct Of Operations
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~~ - - p emistry within allowable Ii
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I
.37 Knowledge of procedures, guideline associated with reactivity management. SRQ 4.6 55.43(5)55.43(6)The criteria for confirming the Rx trip are 3/4 power range NI <5%, IR NI level dropping, IR SUR NEGATIVE. With a SUR of ZERO, Rx Itrip is NOT confirmed. Transition is to FRSM. A is incorrect because a single Vital Bus is energized and the transition to LOPA-I is made if NO vital busses are energized. B is incorrect because the Rx trip is L ~
Friday, August
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08, 2008 9:00:54 AM
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Page 1 of 35
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IGiven the following conditions:
I- Unit 1 is operating at 100% power.
- Charging system problems result in NO Unit 1 charging pumps running OR being available
~ to be run.
1- The decision is made to align and run 23 charging pump (Unit 2) from Unit 2 RWST to
~ Unit 1 CVCS system.
- The proper lineup is completed, and 23 charging pump is now supplying the Unit 1 1 charging system.
- Unit 1 PZR level is rising very slowly.
which of the following identifiesthe nextaction to be performed,
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- __ ~_~ and why?
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'Commence a Unit 2 shutdown IAW S2.0P .ZZ-0004, Power Operation, due to
-~2 RWST Unit ~ __ - level. ~ ~ ~ -- ~ ~ _~.__ _~
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ICommence a Unit 1 shutdown IAW S I .OP-IO.ZZ-O004, Power Operation, due to the boration of the RCS from Unit 2 RWST.
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IReestablish normal letdown on Unit 1 IAW S I .OP-S0.CVC-0001, Charging, Letdown, and lSeal
-~ Injection ~ -~ ~~~
to control PZR ~ -~
level.
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Initiate makeup to Unit 2 RWST IAW S2.OP-S0.CVC-0006, Boron Concentration Control, to iensure minimum Tech Spec required
- ~~ _- _ _ level ~in Unit 2 is maintained. - _ ~
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ep 3.52) A is incorrect ed operation of 23 charging pump will be evaluated or other contingency actions implemented (AB.CVC-1 Step 3.49 NOTE)
IC is incorrect because normal letdown can NOT be restored since the letdown isolation valves are interlocked with the charging pump breakers such that at least ONE charging pump breaker has to be phut to open letdown isolation valves. Excess letdown may be placed in service to control PZR level. D is incorrect because while it will probably be done, it is not the next step. It will be done after the ABCVCIE003 ~_
Given a set of initial plant conditions:
Determine the appropriate abnormal procedure.
Describe the plant response to actions taken in the abnormal procedure.
Describe the final plant condition that is established by the abnormal procedure I Page2of35 1
Page3of35 I
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IUsing the following list of components and assuming them to be% service, selectthe choice which contains ONLY those locations of a CCW leak that could cause Radiation Monitors 2R17A and i2R17B indications to rise. There is no RCS primary to secondary leakage.
I. Excess Letdown HX Ill. 22 CCW HX
'Ill. Letdown HX lIV. Seal Water HX IV. Spent Fuel Pool HX Rad Monitor HX
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~ determine
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- - and interpret_ _ the following as they apply to
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-~Loss of Component
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Cooling
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- Water:
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2.9 3.6
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use of possible CCW ~ ~-~
loss ~~
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55.43(4) The excess Letdown and Letdown HX have primary coolant flowing through them at higher pressure than the CCW system. The CCW HX is cooled by SW, so inleakage would not be radioactive.
IThe RCP Seal Water HX seal water is at higher pressure than CCW system. The SFP HX is at a lower lpressure so leakage would be out of CCW system. The R19A SG Blowdown Rad monitor, would not lcause counts to rise, since normal ops assumed in stem does not infer SG tube leakage. The distracters ABCCO 1E006 For the following analyzed transientdaccidents:
a) Determine the expected alarms and indications b) Describe the analysis assumptions.
c) Describe the protective features that mitigate the event.
d) Describe the expected plant re
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I -Gday, August 08, 2008 9:00:54 AM
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I Page4of35 I
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Given the fo IIowing conditions:
Unit 2 was operating at 100% power, steady state.
The following indications have changed as indicated over the past 2 minutes.
- PZR Master Pressure Controller output is rising slowly.
- Charging flow has lowered from 86.4 gpm to 84 gpm and continues lowering slowly.
- Letdown flow has lowered from 80 gpm to 79 gpm and continues lowering slowly.
- PZR level has risen from 57.3% to 57.8% and continues rising slowly.
Which of the following identifies
~ ~ ~~~ blem occurring, and how will the malfunction be addressed?
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ntrolling PZR pressure channel has failed at 2238 psig. Enter S2.OP-AB.PZR-0001,
,the_ _controlling
_ _ _ _ _PZR_ ~pressure IPressurizer_Pressure
_ ~ _ _ _ _
~ _ ~ ~channel
_ _ _ Malfunction.
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-~. has failed at 2232 psig. Enter S2.0P-AB.PZR-0001,
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IPressurizer Pressure Malfunction.
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The controlling PZR level channel has failed at 58.5%. Enter S2.OP-AB.CVC-0001, Loss of Charging.
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The controlling PZR level channel has failed at 55.5%. Enter S2.0P-AB.CVC-0001, Loss of
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channel failed slightly above actual pressure at beginning of transient, the MPC output will rise, since it thinks pressure is too high, and will act to deenergize heaters and open spray valves. As actual pressure drops due to the rising spray flow, the RCS will expand into the PZR, causing PZR level to rise. This will icause charging flow to lower as PZR level rises above program. Letdown flow will lower as RCS I pressure lowers.
I Refwenwe** I Pressurizer Pressure Malfunction
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ABPZR1 E003 a) Determlne the appropnate abnormal procedure.
b) Describe the plant response to acitons taken in the abnormal procedure.
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- AB 1 Friday. August 08, 2008 9:00:54 AM 1 Page5of35
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Given the following conditions:
I- Unit 1 is performing a Rx startup.
i- Rx power is 2E4 cps.
- SUR is .2 dpm.
- During the refueling outage, BOTH IR NI detectors were replaced.
1- IR NI indication for both N35 and N36 is flashing at 1x10-II A .
Which of the following describes the status of the IR instrumentation, and the required action(s) that will be performed?
I
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,are under compensated. Stabilize power, block SR Hi Flux trip, and correct compensating Ivoltage problem for BOTH IR Nl's prior to exceeding 5% Rx __
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power.
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ilnstrumentation:
n at 20, 000 counts, the decade of overlap should already be lpresent. With the Hi Flux Trip at 100,000 counts, there can't be proper overlap. With no other information iin the stem to provide inference of any other problems with the Nl's EXCEPT that both IR detectors were lreplaced, the IR Nl's should be declared INOPERABLE. There is only an action in 3.3.1.Ifor ONE INOPERABLE IR NI, with BOTH INOPERABLE TS 3.0.3 is entered. C is correct because even if the candidate though they were reading low due to overcompensation, the P-6 block would not be manually lperformed without the power above P-6 interlock to allow P-6 to be blocked. Under compensation would
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EXCOREEOOS ldentifv and describe the Control Room controls, indications, and alarms associatedwith theExcore Nuclear Instrumentation System, including:
The Control Room location of Excore Nuclear Instrumentation System control bezels and indications.
The function of each Excore Nuclear Instrumentation System Control Room control and indication.
The effect each Excore Nuclear Instrumentation System control has upon Excore Nuclear Instrumentation System components and operation.
The plant conditions or permissives required for Excure Nuclear Instrumentation System Control Room controls to perform their intended function.
ts associated with the Excore Nuclear Instrumentation System controoroomarms.
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Friday, August 08, 2008 9:00:54 AM
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I Page6of35 1
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'Given the following conditions:
- Unit 1 is operating at 15% power.
- Main Steam Dumps are in MS Pressure Control-AUTO.
I- PT-507 Steamline Pressure Transmitter, fails high, and an automatic Safety Injection initiates on Hi Steam Flow and Lo Steam Pressure after all the steam dumps open fully.
'- The 24MS167 valve status light on 2RP4 is flashing.
IWith NO operator action, which of the following would identify that the 24MS167 has failed to shut? ~ -~ ~ ~~
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-~~- R level
~ __ would shrink out low. ~ _~ ~ - ~ -- ~~ - -_ ~ ~_ --
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Conduct Of Operations -
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o maintain primary and secondary nt chemistry within allowable limits.
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tual KA is 0040 Steamline Ruptu 1.45 Ability to identify and interpret div e response of another indication. SRO Value 4.3 55.43(5) The auto SI on high stm flow coincident with lo steamline pressure will initiate a MSLI. There is some period of time in seconds in lwhich ONLY the failed open 24MS167 will continue to pass flow through to the steam dumps before they iautomatically shut on lo lo tave at 543 degrees. This will cause NR level, WR level, and steam pressure for 24 SG to all be lower than the other 3 SGs. The NR level is a good distracter if the operator thinks it '
will shrink low out of the indicating band, which it won't because of the low initial power level. The High
'Steam flow bistable does not "lock in" its indication on 2RP4, and will clear when the high flow condition clears.
I Refenkce T&
Overhead Window F L m i qO b j ~ v e s
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NCT Draw a one-line draaram of the Main Steam System which indicates the following Major Component Main steam System Componenis Steam Generators Steam Generator Flow Restrictors Safety Valves Atmospheric Steam Dump Main Steam Warm-up Valves Main Steam Drain Valves Main Steam Isolation Valves Main Steam Mixing Bottle Main Steam Stop and Governor Valves Turbine Bypass Valves b Major Main Steam System Flowpaths Main Steam Flow Auxiliary Feedwater Pump Steam Flow Turbine Bypass Steam Flow Main Steam line Drains
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Fiidav. Ausust 08, 2008 9:00:54 AM Page9of35 1
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Given the following condition:
- Unit 2 lost off-site power while in MODE 5 1- When attempting to reset the SEC IAW S2.0P-AB.LOOP-0001, Loss of Off-Site Power, 2A SEC would not reset.
~- An operator deenergized the 2A SEC cabinet by opening its power supply breaker from
' the Vital Instrument Bus.
I- Off site power was subsequently restored, and AB.LOOP-0001 has been completed and exited.
I Which of the following describes how this breakers status will be tracked until it is restored to lTagged/Not~-
- ~ - ~~ Tagged Discrepancy sheet.
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When S2.OP-AB.LOOP-0001 was completed, the Work Clearance Module (WCM) was iupdated IAW OP-AA-108-101-1002.
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IAW OP-AA-108-101, Control Of Equipment and System Status, Attachment 1, Abnormal Component Position Sheet.
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[OP-AA-108-101 directed updating the breaker position in the WCM during performance of
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Fnday, August 08, 2008 9:00:54AM
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Page 10of35 ,
- whichof the following describes a situation which is considered an unmonitored radioactive liquid release? ~ -- ~ ~ ~ - - ~ -- ~~- ~
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'21 CVCS Monitor Tank is in recirc IAW S2.OPSO.WL-0001, Release of Radioactive Liqui waste from 21 CVCS Monitor Tank, and the 21WR25, 21 CVCS Monitor Tank drain valve is opened.
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IA Radioactive Liquid release is performed IAW S2.OP-S0.WL-0002, Release of Radioactive ILiquid Waste from 22 CVCS Monitor Tank with the knowledge that 2R18, Liquid Radwaste Effluent Line Monitor,
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is INOPERABLE.
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IWith the 2FR1064, Radwaste Overboard Discharge Flow Monitor INOPERABLE, sample flow iis lost to 2R18 during an authorized radioactive liquid release from 2 WMHUT IAW S2.0P-S0.WL-0003, Release of Radioactive Liquid Waste from 2 Waste
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~ -~ ~~ ~- Monitor Holdup Tank.
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1000059G421 ldischarged. The Spent Fuel Pool Liner Leak Detection system consists of channels under the Spent Fuel lpool, which lead to 17 tell tale drains. The small (20 dpm) amount of leakage will be contained in the 1 channels, and the tell tales collect in a common sump, which is then pumped to in service Waste ~
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appropnate Technical Specification action. (License Operator and STA only)
NCT State the Technical Specification associated with the component, parameters and operation of the Radioactive Liquid Waste System including the Limiting Condition for Operation(s) (LCO) and the applicability of the LCO(s) (Non-licensed I Page 11 of 35
Given the following conditions:-
'r-Unit 2 is operating at 100°/b power.
A containment pressure relief in in progress IAW S2.OP-SO.CBV-0002, Containment Pressure-Vacuum Relief System Operation.
- 2VC5 and 2VC6, Containment Isolation Valves, are open.
- An RCS leak occurs, leading to a Rx trip and Safety Injection.
I IWhich of the following identifies how containment integrity will be restored after the SI is initiated, and what action is required if the 2VC5 and 2VC6 do not shut, and can not be shut, when idemanded?
IReference provided. - - -~ ~ ~
6 will automatically shu
_ _ - _ _ _ _ _ _ - ~ lare e Area
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the 2VC5 and 2VC6 will automatically shut directly from the SI signal. Declare an Alert.
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,The2VC5 and 2VC6 will automatically shut directly from the CVI signal. Declare a Site Area IEmergency.
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The 2VC5 and 2VC6 will automatically shut directly from the CVI signal. Declare an Alert.
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occurred based on an RCS leak, so the leak has to be bigger than what a centrifugal charging pump can lhandle to maintain PZR level > 17%. This would be at least 3 and could possibly be 4 points on ECG 3.2 IRCS barrier. Additionally, the open containment as stated in the stem would be 2 points on ECG 3.3 the Reactor Protection System, including: (Licensed Operator and STA Only) a) The Control Room location of Reactor Protection System control bezels and indications b) The function of each Reactor Protection System Control Room control and indication c) The effect each Reactor Protection System control has upon Reactor Protection System components and operation d) The plant conditions or permissives required for Reactor Protection System Control Room controls to perform their intended function SRO 9 ECG
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Friday, August 08, 2008 9:00:54AM
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I Page 13 of 35 1
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IGiven the following conditions:
- Unit 1 is operating at 100% power when it receives an inadvertent Safety Injection signal from the RPS system.
1- The control room crew has progressed through the EOP network.
I- Operators are preparing to exit EOP-TRIP-3, Safety Injection Termination.
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dge of limiting conditions for oper
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nd safety limits.
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is incorrect because the transition would be to TRIP4 if a cool down was requir RCPs were operating, or if no cool down was required it would be to IOP-8 Maintaining Hot Standby. B 1 is incorrect because the EDGs are secured in TRIP3 if off-site power is powering the vital busses. D is correct because TSAS 3.7.10 requires 2 chilled water pumps, and there is no action for 2 INOPERABLE lpumps, which is TS 3.0.3. With the unit already in Hot Standby following the trip, the unit must be placed iin Hot Shutdown within the next 6 hours6.944444e-5 days <br />0.00167 hours <br />9.920635e-6 weeks <br />2.283e-6 months <br />. C is incorrect because it is onlya 72 hour8.333333e-4 days <br />0.02 hours <br />1.190476e-4 weeks <br />2.7396e-5 months <br /> TS and would not r- ~~~
Elday, August 08, 2008 9:00:54 AM
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IGiven the following conditions:
I--- Unit 2 was operating at 100% power when 21 SGFP tripped.
The Main Turbine ran back as expected.
Control Rods did not respond in AUTO, and the RO is inserting them in manual to restore Tavg Ito program.
1- PZR pressure is 2238 psig.
1- OHA E-20, PZR HTR ON LVL HI is in alarm.
1 - BOTH PZR Backup Heater groups are OFF.
which of the following explains how the heaters could be OFF in this situation, and the action that lshould be performed?
The Backup Heaters are OFF because -~
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the PZR Master Pressure Controller ( output is demandin ZR heaters off in lresponse to the PZR pressure > 2235 psig. Ensure Charging System Master Flow Controller is responding to reduce charging flow IAW S2.OP-S0.CVC-0001, Charging, Letdown, and Seal Injection.
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Ithe PZR MPC is a PID type controller, and the signal to turn on the heaters has not been developed. Cycle MPC from AUTO to MANUAL to remove the reset windup, and return to AUTO IAW S2.OP-SO.PZR-0005,
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~ ~~ ~ ~ -~- ~ Pressurizer
~~ ~ -~ Pressure Control
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System
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operation.
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the operating charging pump has tripped, causing a Letdown Isolation, which blocks the heaters from energizing. Start an available charging pump IAW S2.OP-AB.CVC-0001, Loss of
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lthose predictions, use procedures to correct, control, or mitigate the consequences of those abnormal cause the Hi LVL on function for the PZR heaters is i MPC. The MPC WILL be above the setpoint for heater on demand. B is incorrect because the demand will be to turn OFF the heaters. Use of the term reset windup refers to the way of removing saturation conditions in the controller and was used in the past on Salem's MSIO controllers. C is incorrect because there is no interlock between charging pumps and heaters, the interlock is between charging pumps and lletdown isolation. D is correct because MANUAL mode of operation overrides any other automatic signal Ito operate heaters.
1 PZRPXlLE006
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Outline the interlocks associated with the following Pressurizer Pressure and Level Control System components:
P-7 and the Pressurizer Low-Pressure Reactor Tnp P-I 1 and the Pressurizer Low-Pressure Safety Injection Block P-7 and the High Level Reactor Trip
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Friday,
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August 08, 2008 9:00:54AM
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l Page 16 of 35
Given the following conditions on Unit 2:
I- A LBLOCA has occurred.
~- Operators are performing 2-EOP-LOCA-5, LOSS OF EMERGENCY RECIRCULATION.
~- Containment pressure is 15.1 psig and is rising slowly.
Which of the following describes how the Containment Spray system will be operated, and why?
!The Containment Spray ---
System is operated as directed in.. .
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CE-1 "RESPONS SSIVE CON restoration of the critical safety function takes precedence.
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,LOCA-5 because it establishes minimum required containment spray flow and conserves IRWST inventory.
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12.EOP-FRCE-1_ because
- _ _____ Containment Spray operation is independent of recirculation.
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~LOCA-5since FRPs are NOT implemented during-~
_ - ~ ____ the performance of LOCA-5.
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Procedures / Plan -~ --
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'2.4.22 Knowle are to be operated IAW LOCA-5. The basis document states that this is because in FRCE, maximum available heat removal system operability is warranted to reduce containment pressure, whereas in LOCA-5 a less restrictive criteria permits reduced spray pump operation depending on RWST level, containment pressure, and # of CFCU's operating. The less restrictive criteria in LOCA-5 is used I because recirculation flow to the RCS is not available, and it is very important to conserve RWST water, if lpossible, by stopping containment spray pumps. So while the operator WILL enter FRCE-1 due to IPURPLE path of containment pressure > 15 psig, the containment spray pumps will be operated IAW
~LOCA-5.
on Source: Facility Exam Bank Qu&fbn Modwlcation Y e t W Direct From Source IViston Q 48927
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Friday, August 08, 2008 9:00:54 AM Page 1 7 o f 3 5 I
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,Unit 1 is i n o d e 5 with a containment purge inprogress.
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IFailure of the Plant Vent Flow Monitor with all Auxiliary Building Exhaust Fans operating.
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Failure of RMS Channel R I I A , Containment Particulate, with Channel R41D, Piant Vent Noble gas Release Rate Composite, operable
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Failure of RMS Channel R12A, Containment Noble Gas, with Channel R41D, Plant Vent Noble IGas Release Rate Composite, operable to control radiation releases.
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KA is 3.14 knowled f radiation or contamination hazards that may arise d abnormal, or emergency conditions or activities. SRO 3.8 55.43(4) If the Aux Building is not maintained at a negative pressure, the potential exists for an unmonitored release from the Aux Building. P&L 2.10 istates that if the plant vent flow monitor is unavailable, then ALL ABV fans must be in service. As long as lthe R41D remains OPERABLE, the 11A or 12A are not required.
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CONTh4TE012 _._____
nisrliss .
the procedural requirements associated with the Containment and Containment Support Systems, including an n and IimiLatEns inthe Contarmentand Ccnorinment Support Systems pIo_cedures-
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- Friday, August 08, 2008 -~
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9:00:54 AM~- Page 1 8 o f 3 5
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'Given the following conditions Salem Unit 2 is operating at 73% power.
An electrical panel in the CW building has been exposed to the rainy outdoors environment when the roof panel above it is removed.
Operators perform a power reduction to 40% power in response to Circulator malfunctions associated with the now wet panel.
Operators trip the Main Turbine, and continue lowering power in response to lowering condenser vacuum.
23 SG NR level detectors (2 of 3) see a Bad Quality input, and the 23 loop ADFACS swaps to manual. Actual NR level is 30% and rising when the swap occurs.
'Which of the following describes ~- how the 23BF19 and 23BF40 will be operated?
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g in S2.0P-AB. Turbine Trip Below P-9, B
~23BF40will re&ire operator
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control to prevent a high level in -~
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~~ ~ 23 SG. ~~ ~~ ~~ ~~ -~ ~ ~~~ ~
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iUsing S2.OP-AR.ZZ-0007, Overhead Annunciators Window G for OHA G-7, ADFCS Swap to iManual, BOTH 23BF19 and 23BF40 will require operator control to prevent a low level in 23 SG.~~~~ - ~ _ _ _ _ ~~ ~ -~ ~ ~ ~ _ ~ _ ~~ _ - ~ ~~~~ ~ ~ -~ - __ ~
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After entering S2.OP-AB.CN-0001, Main Feedwater/Condensate System Abnormality, ONLY the 23BF40 will require operator control to prevent a low level in 23 SG since 23BF19 closed on~-Feedwater Interlock.
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Using S2.OP-AR.ZZ-0007, Overhead Annunciators Window G for OHA G-I 5, ADFCS Trouble, ONLY the 23BF19 will require operator control to prevent a high level in 23 SG since 23BF40 flow is small
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compared to 23BF19 flow.
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IO59 iMain Feedwater System reduction continued. The stem states that the power reduction continues after the turbine was tripped.
AB.TRB is required to be entered for the turbine trip, and operators stay in AB.TRB to perform the power reduction since vacuum is still lowering (Steps 3.1 1-3.12) With actual SG level at 30%, below iprogrammed level, the demand on the BF19s and 40s will be higher than that required to maintain normal llevel. This will cause NR level to rise without operator action to correct it. All of the distracters contain iincorrect actions, but their procedures are applicable to the conditions.
Technical Bases Document.
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ABTFRBE005 Determine the appropriat
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~~ b) Describe the plant response to actions tak I
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Fndav. Auaust 08, 2008 9:00:54 AM
Paye20of35 1 Given the followingconditions:
1- Unit 2 is operating at 100% power.
- 2B EDG is running in parallel on the 2B 4KV Vital bus.
'The 500 KV ring bus loses all off-site power.
'Which of the following identifies how this will affect the operation of 2B EDG?
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n co 28 vital bus, and Blackout Loads will sequence on the bus if inat already running.
Lp __ Enter __ EOP-TRIP-I Rx Trip or Safety Injection.
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I ill remain connected to 2B vital bus. NO Blackout loads will start since 2 6 SEC still lsenses voltage on -
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2B 4KV - ~
Vital bus. Enter S2.OP-AB.LOOP-0001,
~ ~ - ~ - Loss of Off-Site Power. -~
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r p ~ _ - ~~ ~_ ~ ____ ~p ~ -__ ~ - ~ _~ ~ ~_ _~ p _ - ~ ~_
'26 EDG output breaker will trip, 26 bus will strip, the EDG output breaker will close, and Blackout - Loads will sequence on the bus. Enter EOP-TRIP-1 Rx Trip or Safety Injection.
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'2B EDG output breaker will trip, 2B bus will strip, the EDG output breaker will close, and Blackout
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on the bus.Enter S2.0P-AB.LOOP-0001,
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Loss of Off-Site Power.
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8/25/2008
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K2. A b z y to (a) predict the impacts of the following on the Emergency Diesel Generators an those predictions, use procedures to correct, control or mitigate the consequences of those abnormal 2 (Blackout). This will act to deenergize the respective busses if energized by opening EDG output breaker, strip all loads, energize the bus from the EDG, and sequence loads on due to start current iconcerns for the EDG vs. starting all loads at once.
2A Generator Endurance Run Generator, including:
The Control Room location of Emergency Diesel Generator control bezels and indications (Licensed Operator & STA only)
The function of each Emergency Diesel Generator Control Room control and indication. (Licensed Operator & STA only)
The effect each Emergency Diesel Generator control has upon Emergency Diesel Generator components and operation.
(Licensed Operator & STA only)
The plant conditions or permissives required for Emergency Diesel Generator Control Room controls to perform their intended function.
ro 1 Page21 of35
Given the following:
- 1 WMHUT is in recirc, a sample has been drawn and is in the process of being analyzed.
1- The RWO mistakenly places 1 WMHUT in service.
- One hour later, the RWO recognizes his error when the high level alarm for 1WMHUT lannunciates at the Unit 1 104 Panel.
- The RWO immediately returns 1 WMHUT to recirc.
What effect, if any, will this have on the release preparations for 1 WMHUT IAW S I .OP-S0.WL-i0003?
The CRS will direct the.. .
b 1-1 S 7 1 Application 8/25/2008
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The stem gives the high level alarm as annunciating, this will give a tank level from the tank curve, and then will yield a recirc time IAW Att 2 of WL-0003, of 306.67 minutes prior to sampling. This is 5.11
~hours.Any extrapolation error would be in the lower tank level based on the way the rectangle is placed on the graph for the high level alarm. The 4.33 hour3.819444e-4 days <br />0.00917 hours <br />5.456349e-5 weeks <br />1.25565e-5 months <br /> distracter is if the candidate uses the wrong tank
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(Y-1 Requlrn for Examlnatlon Q16 S I .OP-S0.WL-0003, Att's 1 and 2. S I .OP-TM.ZZ-0002 Tank Curves Page 22 of 35
Page 23 of 35 whichof the following describes why FRCE-3, Response to High Containment Radiation, i s a YELLOW path FRP performed at the discretion of the CRS, instead of a RED or PURPLE path lcompletion.
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The actions performed in FRCE-3 are redundant to actions which are expected to have loccurred automatically.
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IConditions requiring entry into FRCE-3 would not result in the release of radiation outside the lcontainment buiIding. ~-
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If conditions have degraded to where the entry conditions are met for FRCE-3, then higher priority RED or PURPLE path FRPs will be present, and must be performed
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I"actions" taken in FRCE-3, to ensure the containment pressurehacuum relief path is isolated through VC5 and 6, and starting CFCUs in low speed to ensure the filters are realigned. D is incorrect because the entry conditions for FRCE-3 is containment radiation >2R/hr. This does not always mean that other '
FRPS will be in effect. A is incorrect because of B above, and the CRS should direct entry into a IYELLOW path only when its performance will not impede other actions critical to plant recovery.
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I Fridav. Auaust 08, 2008=0:54 AM i Page24of35 ~
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IGiven the following conditions:
- Unit 2 is operating at 100% power.
1- 21 Control Air Dryer is in service and operating in the AMLOC mode.
I- When automatically switching from the left to right desiccant chamber, the CAI 884, Purge 1 Exhaust Valve for the left dryer opens as expected.
'Which of the following describes the consequences of this valve failing to close when the desiccant
[chamber drying cycle is complete, and
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- if any, will-~
_ __ be required
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The continued loss of air through the on the desiccant chamber, even as it Iremainsout of service, will lower control air header pressure to the point of starting the Unit 2 IECAC. Enter S2.0P-AB.CA-0001,
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lThe affected desiccant chamber will still be automatically be placed in service when the timing sequence demands it, and a rapid lowering of control air header will occur. Enter S2.0P-IAB.CA-OOOI, Loss of Control Air.
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lThe affected desiccant chamber will be interlocked from going in service, and the 21 Control Air Dryer will be removed from service IAW S2.OP-SO.CA-0004, 21 Control Air Dryer loperation.
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The continued loss of air through the purge valve will lower control air slightly. Use S2.OP-
~SO.CA-0001,Control Air System
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-- - -_ _- 21-Control Air Dryer. _- ___-_ __-_____ _
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A2._- Ability to (a) predict the impacts of the following on the Instrument Air System and (b) ect, control, or mitigate the consequences of those
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I ldryers are designed to operate normally with purge flow in service. B is incorrect because the chamber will be interlocked from being placed in service with the purge valve open, and additionally because of A above. C is correct because of B above, and the local panel alarm will alert operators to the fact that the lsequence has not completed. The SO provides direction for removing a dryer from service as the I redundant dryer(22) is normally in service also. D is incorrect because the procedure referenced will not ldirect CA drver oDeration. l System:
Control Air Dryers Control Air Receivers Emergency Control Air Compressors (ECACs)
Emergency Control Air Dryers Excess Flow Check Valves (EFCVs)
Station Blackout Compressor CA Containment Isolation Valves PORV Control Air Accumulators Redundant Air Su els
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Page26of35 1 Chemistry reports that the lithium concentration in the RCS is approaching its upper limit. What is lthe preferred action that should be taken to restore the lithium concentration to its proper level?
letdown flow should be.. .
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L minimized to improve the DF_of _
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the_ CVCS
_ _ mixed---
beds. - ~ ~ -_
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diverted--to the HUT
~ -~ to establish a bleed
~ -~ and feed for the RCS.
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'maximizedto accelerate the cleanup using the CVCS mixed beds.
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secondary plant chemistry
~ ~ - within allowable ~.limits. -
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Actual KA says Knowledge of primary and secondary plant chemistry limits 55.43(4)M not provide lithium control. Placing the cation bed in service per chemistry direction will remove lithium 1 lfrom the RCS. B is incorrect because the cation bed is placed in service at normal flow (75) gpm, and the mixed bed does not lower lithium. C is incorrect because it would take a long time to lower lithium this CVCS Demineralizer- Normal Operation Leardng Objectives
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CVCS%%004 LOR NCT Describe the function of the followinq
- components
. and how their normal andabn&nalopeGtion affects the Chemical and Volume Control System:
LetdownlCharging Letdown lsolaiton Valves, CV2, CV277 Regenerative Heat Exchanger Letdown Orifices Letdown Orifice Isolation Valves, CV3, CV4, CV5 Letdown Releif Valve, CV6 Letdown Line Containment Isolation Valve, CV7 RHR Flow Control Valve, CV8 Letdown Heat Exchanger Low Pressure Letdown Control Valve, CV18 Temperature Control Valve, CV21 Demineralizers (Mixed Bed, Cation, and Deborating Inlet Valve to Deborating Demin, CV27 Reactor Coolant Filter Diversion Valve, CV35 CVCS Holdup Tanks Volume Control Tank VCT Isolation Valves, CV40, CV41 Chernical Mixing Tank Charging Pumps (Centrifugal and PD)
Miniflow Recirc. Valves, CV139, CV140 Seal pressure Control Valve, CV71 Chg. Line Containment (sol. Valves, CV68, CV69 Charging to Loop 3 Valve, CV77, Loop 4 Valve, CV79 PZR Auxiliary Spray Valve, CV75 CCP Flow Control Valve, CV55
- b. RCP Seal Water Seal Water Injection Filters Seal Bypass Flow Valve, CV114 Seal Water Return Isolation Valve, CV104 Seal Water Return Relief Valve, CV115 Seal Return Cont. Isol. Valves, CV116, CV284 Seal Return Filter Seal Water Heat Exchanger
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Fridav. Ausust 08.2008 9:00:54 AM Page 27 of 35
- c. Excessletdown Excess Letdown Isolation Valves, CV278, CV131 Excess Letdown Heat Exchanger Excess letdown Flow Cotrol Valve, CV132 Excess Letdown Diversion Valve, CV134
- d. Makeup Primary Water Storage Tank Primary Water Makeup Pumps Boric Acid Batch Tank Boric Acid Tanks Boric Acid Transfer Pumps Boric Acid Filter Boric Acid Blender Primary Water Flow Control Valve, CV179 Boric Acid Flow Control Valve, CV172 Charging Pump Suction Valve, CV185 VCT Makeup Isolation Valve, CV181 I
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k s t i o n Modification Method: Editorially Modified 1
~~ ~P Fnday, August 08,20089:00:54 AM Page28of 35 I P P ~~ ~P P~ ~~
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Given the following conditions:
1- Unit 1 was operating at 100% power when 11 Condensate Pump tripped.
- One minute later, 11 SGFP pump tripped.
perations
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o maintain primary and s
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1 2.3 Actual KA is 2.1.39 Know1 ve decision making practices SRO 4.3 Section 5.20 NAP-5 is Conservative Decision Making. A is not correct because inserting rods at a faster speed than what is inserting in auto following a load reduction is conservative. B is not correct because the operator lis expected to take action prior to reaching an automatic trip setpoint. C is correct because tripping the lturbine in this situation is forcing an automatic Rx trip on the turbine trip, and is not an expected action when power is >P-9 (49%) power. D is incorrect because reducing load is required for a SGFP trip, and additional load reduction beyond that inserted by the SGFP trip is conservative if required. I I "ReferenceTitie I
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Friday, August 08,2008 90054 A M- ~ Page 29 of 35 L - ~ - ~ -~ ~ __ ~~
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During a refueling outage, who is responsible for the FINAL authorization prior to entry into a P D Safety Status of Or
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Manager.
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- Shift Outage Manager.
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Shutdown Safety Manager.
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'Director of Work Management.
ldoes not provide final approval.
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Learning Objecuvits CONDOPEOOT-State the purpose of the Conduct of Operations guidelines.
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given the following conditions:
1- A male radiation worker at Salem Station returned 3 weeks ago from outage support
~ at Limerick Station.
f- His Total Effective Dose Equivalent (TEDE) received at Limerick was 75 mrem.
,- He received an exposure of 50 mrem to his right hand during an inspection while wearing special ldosimetry.
- The worker's current TEDE from Salem for this year is 75 mrem.
1- The worker had an chest x-ray one week ago estimated at 25 mrem exposure.
Learnin$ Objectivrts
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RADoNE062-
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List the followina external radiation exposure Iim&, in accordance with StationFrocedures, ?bCFR50, and Reg Guide 8 13 A. 10CFk20 dose limits for external, internal, and total whole body, skin, extremities, and eyes, as well as extension limits and requirements B. Administrative dose control levels for Category 1 and 2 Workers, as well as extension limits and requirements C. Reg. Guide 8.13 limits and administrative dose control levels for Declared Pregnant Women D. 10CFR20 and Administrative limits for members of the general public and minors E. Category 1 Radiation Worker
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F. Category 2 Radiation
~- Worker-
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Fnday, August 08, 2008 - AM 9:00:54 ~ Page31 of35
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Given the following conditions:
- Unit 2 is operating at 100% power.
I- 23 & 25 CFCUs have been C/T for emergent corrective maintenance for 24 hours2.777778e-4 days <br />0.00667 hours <br />3.968254e-5 weeks <br />9.132e-6 months <br />.
~- A crew of 5 people entered containment at 1415 to investigate a rise in the RCS leakrate, with a Heat Stress stay time of 1 hour1.157407e-5 days <br />2.777778e-4 hours <br />1.653439e-6 weeks <br />3.805e-7 months <br />.
- At 1416 22 CFCU breaker trips.
10 minutes after the crew entered containment, the NCO reports that 2R12A CONTAINMENT GAS
!EFFLUENT is reading double what it was when the crew entered containment, and containment average air temperature is rising slowly.
'Which of the following describes the effect, if any, on the personnel in containment IAW SC.SA-JST.ZZ-0001SALEM CONTAINMENT
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ENTRIES -
IN MODES - __
1 THROUGH 4?
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p____
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p~ ~
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_pp p
The control room will contact the crew in containment by flashing the containment lights, and direct them to exit the containment.
__ ~~~ __ ~~ -~ _ _ __ ~ p p - ~- ~~~ ~ - - _ ~ p~ -p ~ ~ _ _ ~
~_ ~ ~- p~ ~~ -~ ~ ~p
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Since the 2R12A is expected to rise with an RCS leak, the crew may remain in containment
~_ ~- - ~ -~ ~- ~ p p - ~ ~ ~ - ~ ~
ANY increase in radiation levels in containment while it is occupied REQUIRES dispatching a lRadiation Protection technician
__ ~ p p~p ~ -~ ~
into containment
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to evacuate p~ ~- ~ _
containment.
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Personnel in containment may continue their inspection while monitoring for any continuing rise in radiation level. If radiation levels on 2R12A increase by a factor of 4 from original level, use the page system to direct
~ p p p_ ~ p
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in containment
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to exit.
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ctual 1122 Rev. 2 supp 1 KIA is 2.3.13, Knowledge of radiological safety procedures pertaining to ensed operator duties, such as response to radiation monitor alarms, containment entry requirements, fuel handling responsibilities, access to high radiation areas, aligning filters,etc. SRO value 3.8 55.43(4) SC.RP-TI.ZZ-1102 states that the SRPT, upon discovering a 50% rise in RMS data, shall (prohibitany subsequent entries to containment and DIRECT the control room to contact any work parties
,and have them exit containment. IAW SCSA-ST.ZZ-0001, 3.2.1.. .."The containment lighting, when flashed, is the preferred method the Control Room uses for requesting communications with the work party." ~p ~ _~~ ~ - _ _ ~ ~p ~ ~ __ ~ ~ p p p~ - _ ~ ~- ____-~ -
Reference mt4e I Previous 2 NRC Exams
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~ Given the- foIlo4 ng conditions:
I1- Salem 1 and 2 are operating at 100% power.
1- Hope Creek is operating at 100% power.
1- Fire Brigade manning consists of 6 qualified personnel, which includes one Fire Brigade Leader.
- The site ambulance is involved in an accident during a training exercise, and two 1, Fire Protection Operators require off-site transportation to Salem Memorial Hospital.
Which of the following describes the status of the Fire Brigade with four members per the Salem
,FSAR,and action(s), if any, which are required to be performed IAW the appropriate Fire IDepartment Procedures? ___
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(T main ately ur members are req FSAR. No --__-
-__ compensatory
~- measures are required.
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The Fire Brigade remains adequately staffed. Notify the Superintendent- Fire Protection Operations if callout is initiated for
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any Fire Protection Operators..
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kd 'The Fire Brigade staffing is inadequate. Initiate call-out of qualified personnel to ensure
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manning is restored to six members within 2 hours2.314815e-5 days <br />5.555556e-4 hours <br />3.306878e-6 weeks <br />7.61e-7 months <br />, otherwise submit a 24 hour2.777778e-4 days <br />0.00667 hours <br />3.968254e-5 weeks <br />9.132e-6 months <br /> report to the NRC.
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!!! The Fire Brigade staffing is inadequate. Initiate call-out of qualified personnel to ensure
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manning is restored to five members within 2 hours2.314815e-5 days <br />5.555556e-4 hours <br />3.306878e-6 weeks <br />7.61e-7 months <br />, otherwise initiate a Notification and review far licensina commitment violation.
GENERIC
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2.4- /Emergency Procedures / Plan ~- ~- -
nt fire bri nd portabl equipment usage.
iteo sa callout if less than that. With 2 of 6 operators requiring transportation off site due to accident staffing is INADEQUATE. NC.FP-AP.ZZ-0001, 5.2.5, states that if brigade manning falls below FIVE for more than 2 hours2.314815e-5 days <br />5.555556e-4 hours <br />3.306878e-6 weeks <br />7.61e-7 months <br />, initiate a notification and review for licensing commitment violation.
LearningObjectiv~s FIRPROE012 Discuss the procedural requirements associated with the Fire Protection System, including an explanation of major precaution Pr m Friday, August 08, 2008 9:00:54 AM~- Page34of35 1 ~~
Which of the following Security Events reported to the Shift Managerby theproper authority
'REQUIRES full staffing of the Emergency Response Organization IAW the Salem ECG?
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'The VP-Nuclear has been kidnapped, and the kidnappers are demanding PSE&G pay a ransom. ___- _ _~ ~~~
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The discovery of a pipe bomb in a car being searched at the Site Access Road Security
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The StatePolice officer assigned to Artificial Island has been shot at from a boat in the
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9.1.I.The correct answer is classified as a ALERT under EAL 9.1.2 since it is.. ."an armed attack, explosive attack, airliner impact, or other hostile action is occurring or has occurred within the Owner Controlled Area." The candidate must also know that the full ERO is not required to be activated until the ALERT level.
Hostage / Duress Situations mining Qbiectrvss
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1 Friday, August 08, 2008 9:00:54 AM
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