ML031500625
| ML031500625 | |
| Person / Time | |
|---|---|
| Site: | Salem  (DPR-070, DPR-075) |
| Issue date: | 05/13/2003 |
| From: | Reid J PSEG Power |
| To: | Conte R NRC/RGN-I/DRS/OSB |
| Conte R | |
| References | |
| 50-272/03-301, 50-311/03-301 50-272/03-301, 50-311/03-301 | |
| Download: ML031500625 (127) | |
Text
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US. Nuclear Regulatory Commission Si te-Specific Written Examination Name:
Date: 5/13/2003 License Level: SRO Start Time:
Applicant Information Region: I Facility: Salem 1 & 2 Reactor Type: W 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.
Results Examination Va I ue Ap p I ica n t's Score Applicant's Grade Applicant's Signature Points Points Percent
- Given the following conditions:
- Salem Unit 2 is at BOL, and has been at 60% power for 3 days.
- Control Bank D is at 174 steps withdrawn.
- Rod Control is in automatic.
- RCS Auctioneered High Tave is 562 deg F and steady.
Control Bank D rods start withdrawing at 8 steps per minute. With Control Bank D at I77 steps, Rod Control is placed in Manual, and all rod motion stops.
Which of the following describes why control rods will be inserted in manual to 174 steps withdrawn IAW S2.OP-AB.ROD-0003?
'To prevent reduced charging flow from lowering pressurizer level.
1
'To restore Tave to programmed band.
I Rl Torestore 1
1 RCS pressure to normal.
To prevent Tavernref console alarm from annunciating.
I b
i
/Comprehension
~000001 K301.......
I Ij L-.
/AK3. ]:Knowledge of the reasons for the following responses as they apply to Continuous Rod Withdrawal:
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~
j
.it's pre-withdraw power, with Tave higher. Auct. high Tave feeds other systems that rely on Tave to
'remain on program.
Distractor A is incorrect because charging flow will rise to raise Pzr level to higher 1
Setpoint established from higher Tave.
Distractor C is incorrect because the Automatic Pzr pressure 1
- control system will raise spray flow temporarily to restore the higher pressure to 2235 psig. Distractor D is incorrect in that alarms are meant to convey information about parameters, and while operating below i
.alarm setpoints is desireable, it is not the reason to restore primary plant parameters by driving rods to 1
me-event Dosition.
i i
I.-.
S2 :
OP-AB.ROD-0003 Continuous Rod Motion
-1 b,_ Describe the final plant condition'that is.
established by the abnormal procedure.
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Page 1 of 126
Given the following conditions:
- Salem Unit 2 is at 80% power, performing a power reduction @ 1% / min due to degrading condenser backpressure.
- Rod control is in automatic.
As control rods are moving inward at 8 steps / minute, OHA E-24, "ROD DEV OR SEQ" alarms.
Control rod 1 D1 indicates 220 steps and stationary. All other control rods expected to be moving are inserting correctly.
The CRS directs entry into S2.OP-AB.ROD-0001,I1 IMMOVABLE / MISALIGNED CONTROL
'RODS", Rod Control is placed in Manual, and all rod motion stops with the remainder of control
'bank D at 200 steps.
IWhich of the following describes the actions required by Technical Specifications?
1 1
/IMMEDIATELY enter TSAS 3.1.3.2.1 "POSITION INDICATING SYSTEMS - OPERATING". 1 r
!After 1 hour1.157407e-5 days <br />2.777778e-4 hours <br />1.653439e-6 weeks <br />3.805e-7 months <br />, if rod I D I is still misaligned by > 12 steps from its group counter, enter TSAS 13.1.3.2.1 "POSITION INDICATING SYSTEMS - OPERATING".
I
/in motion. With a rapid power reduction in progress, rods would be continuously moving during the
/applicability is "within 1 hour1.157407e-5 days <br />2.777778e-4 hours <br />1.653439e-6 weeks <br />3.805e-7 months <br /> after rod motion" to allow a soak time position indication. That one hour time 1
/frame would make distractors A and B incorrect in that entry into either Tech Spec is not REQUIRED I
- IMMEDIATELY. Tech Spec 3.1.3.2.1 for IRPl also has a 1 hour1.157407e-5 days <br />2.777778e-4 hours <br />1.653439e-6 weeks <br />3.805e-7 months <br /> soak time allowance.
j
'incorrect because with power < 85%, the allowable misalignment is +/- I 8 steps, not +/- 12 steps.
idownpower, so this might be the first indication of a stuck rod in this condition. Tech Spec 3.1.3.1 1
i Distractor C is ITechnical Specifications-Movable Control Assemblies I
Technical Specifications TECHSPEOI 5 Describe the general component and parameter categories which are ad throuah 3/4.12 i
ABRODIEOOI RODSOOEOI 3 Tuesday, May 13,2003 4:28.46 PM 1 Page 2 of 126
7 for Operation The Bases for the LCO(s) d)
The LCO Action StatementsW
-~...........
.. TS3.1.3.1 and TS 3.1.3.2.1
- I I New
-.. -.... 1 I:-:
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Page 3 of 126
Given the following conditions:
- Salem Unit 2 is operating at 90% power.
- Pressurizer (PZR) is 2235 psig.
,- PZR Power Operated Relief Valve (PORV) 2PRl is leaking.
- Pressurizer Relief Tank (PRT) pressure is 5 psig.
- PORV discharge temperature has stabilized at 230 deg. F.
Which one of the following will DIRECTLY cause the indicated PORV discharge temperature to
,rise?
~
1 1
IPORV i
leakrate rises
__ by 5 gpm.
!The PRT rupture disk develops a leak.
~.___
1
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'PRT Dressure is allowed to rise to 10 psia
/Pressurizer L__..
Spray is removed from service.
I
-c Comprehension 1
000008K202 3
ors and detectors.
> i 2.71 l..2!7l lincorrect, leaking disk lowers PRT pressure. A. is incorrect, leakrate has no effect unless it is large of the interrelations between Pressurizer Vapor Space Accident and the following:
IC. is correct, PORV discharge is to PRT. If PRT press. Rises, saturation T for discharge rises. B. is
- enough to raise PRT pressure. D. is incorrect, Pressurizer Spray has no effect on PORV leakrate or on IPRT pressure.
1 I
i IPRESSURIZER PRESSURE MALFUNCTIONS 1
ABPZRl EO01 1 Describe operation of the Pressurizer Pressure control system as applied to S2.OP-AB.PZR-0001 (a).
I 1
Tuesday. Mav 13.2003 4:28:46 PM I
Page 4 of 126--
Which of the following is the basis for establishing / maintaining S/G Narrow 9%-33% (non-adverse containment) for small or intermediate LOCAs?
Maintains a static head of water to reduce any existing S/G tube leakage.
Maintain the water level above the top of the U-tubes to prevent depressurizing S/G.
Ensures adequate feed flow or S/G inventory to ensure a secondary heat sink.
'A RCP may have to be started if FRCC-1 is entered later in the event.
- 000009K203 4
EK2.
... - 1IKnowledge i ____
of the interrelations between
____. Small Break LOCA and the following:
I
- adequate feed flow / SIG level is to ensure a secondary heat sink. C is correct because EOP-LOCA-1
'Basis Document states the purpose of establishing 9% level is.."To ensure adequate feed flow or S/G
,inventory to ensure a secondary heat sink for small or intermediate size LOCAs and secondary break
.accidents." 33% is the S/G narrow range upper control band limit.
[Loss of Reactor Coolant Basis Document
, LOCAOI E009 Describe the bases for each step, caution, note, and continuous action summary item in 2-EOP-LOCA-I Tuesday, May 13,2003 4:28:46 PM I
Page5of 126
Which of the following describes the condition that limits flow that is diverted from the ECCS injection path following ECCS actuation?
The CV55 valve, CENT CHG PMP FCV, closes to its minimum stop position to ensure NO greater than 40 gpm is diverted to the Reactor Coolant Pump (RCP) seals.
RCP Seal Injection flow is limited to less than 40 gpm with Charging Pump discharge pressure
>/= 2430 psig and the CV55 valve full open.
iAutomatic isolation of ALL non ECCS flow paths except the CV139 and CV140, CHG PUMP I
IRECIRC STOP VALVES.
I IAutomatic isolation of ALL non ECCS flow paths except the SJ68 and SJ69, SJ PUMP MIN
'FLOW VALVES.
J 00001 1 G222
,0111
/Large Break LOCA jdesign accomplishes this. C. is incorrect although there is flow balancing of injection flow for this reason jelsewhere in Tech Specs. D. is incorrect, the charging flow valve does not get an open signal, although it I
!does fail open on loss of air. If a Phase A isolation is received the charginlcl line isolates also.
/Salem Unit 2 Technical Specifications
.. -. ~ - --.
~~~
1 i
c Svstem. includina:
a i b) c)
d)
The LirnitiG Condition@) for Operation (N/A NEO)
The applicability of the LCO(s)
The LCO Action Staternent(s) (N/A NEO)
The Bases for the LCO(s)
L
.~
A Tuesdav. Mav 13.2003 4:28:46 PM I
Paae 6 of 126
Given the following conditions:
Salem Unit 2 is at 100% power.
Total Seal Injection flow is 34 gpm.
- I Seal leakoff flows are:
- 21 RCP 2.5gpm.
- 22RCP 2.3gpm.
- 23RCP 3.1 gpm.
- 24RCP 2.6gpm.
~ weeks later, with unit remaining at 100% power, these indications are:
Total Seal Injection Flow is 32 gpm.
- I Seal leakoff flows are:
- 21 RCP 2.5gpm.
- 22RCP 2.4gpm.
- 23 RCP 1.2 gpm.
- 24RCP 2.8gpm.
ch of the following describes the condition that has developed?
Reactor Coolant Drain Tank pr
.--~
---I
.24 RCP #2 seal has failed.
i23 RCP #I seal clearance has deteriorated.
c3 c3
-- 3 1 ' Reactor CElantm! PP~Va~f!Etio!F
[Emergency and Abnormal Plant Evolutions imI:Knowledge of the interrelations between Reactor Coolant Pump Malfunctions and the following:
'pressure rising would affect #2 seal leakoff, and not to the extent indicated. 24 RCP #2 seal failure would iust 23.
not result in a slight lowering of RCP amps. VCT pressure rising would affect ALL RCP seal leakoffs, not I I
L
.~.............
I OP-AB RCP-00 a)
Basic RCP Construction b)
Seal Injection and Seal Water Configuration c)
RCP CW Configuration RCPUMPEOOB coo
~
1 a)
The Control Room location of Reactor Coolant Pump control bezels and indications (N/A NEO) 1 b)
The function of each Reactor Coolant Pump Control Room control and indication (NIA NEO)
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1 Page7 of 126
c) d)
function e)
The effect each Reactor Coolant Pump control has upon Reactor Coolant Pump components and operation (NIA NEO)
The plant conditions or permissives required for Reactor Coolant Pump Control Room controls to perform their intended The setpoints associated with the Reactor Coolant Pump control room alarms L.-.
Tuesday, May 13,2003 4:28:47 PM
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1 Paae8of 126
Given the following conditions:
- Unit 2 is at 75% power.
- OHA's D20,21,22,23, RCP 21-24 RCP BRG CLG WTR FLO LO alarms come in at 0810.
- Operator observes valve 2CC187, RCP CC OUTLET VLV has closed, he attempts to re-open the valve but it will NOT open.
- Valves 2CC117,2CC118, RCP CC INLET VALVES, and 2CC136 RCP CC OUTLET VLV are verified open.
- NO other alarms are in.
- S2.OP-AB.RCP-0001 "Reactor Coolant Pump Abnormality" is entered at 081 6.
'Which of the following actions is required by the procedure and why ?
-1 Trip the reactor then trip the RCP's to prevent damage to the pump seals.
ABRCPI E003 CCWOOOE008
~
i
- Pumps can remain in service since the seal water outlet temperatures have NOT exceeded their alarm setpoints.
,Trip the reactor then trip the RCP's to prevent damage to the pump motor bearings.
1 Describe, in general terms, the actions taken in S2.OP-AB.RCP-0001 and the bases for the actions in accordance with the Technical Bases Document.
Identify and describe the Control Room controls, indications, and alarms associated with the Component Cooling Water System, including:
a) b)
c) operation (N/A NEO) d)
intended function e)
The Control Room location of Component Cooling Water System control bezels, indications, and alarms (N/A NEO)
The function of each Component Cooling Water System Control Room control and indication (N/A NEO)
The effect each Component Cooling Water System control has upon Component Cooling Water System components and I
The plant conditions or permissives required for Component Cooling Water System Control Room controls to perform their The setpoints associated with the Component Cooling Water System Control Room alarms j
I I
'Pumps can remain in service since the motor winding temperatures have NOT exceeded their alarm setDoints.
~.
'AK3.1iKnowledge of the reasons for the following responses as they apply to Reactor Coolant Pump Ma1functions:l AK3.02: PCW lineup and flow paths to RCP oil coolers
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~-
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ic is correct because only the Motor Bearing Oil coolers are isolated. A&b. is incorrect because thermal
!barrier is not isolated and seals and pump bearings can still be cooled. D. is incorrect because only the
- pump bearing oil is affected by loss of cooling.
- All other choices are variations of procedure actions.
C is also the required action stated in the procedure.
Tuesday, May 13,2003 4:28:47 PM I
Page9of126 L
Tuesday, May 13,2003 4:28:47 PM 1
PaaelOoflZ6 '
Given the following conditions:
- Unit 2 is in Mode 4.
- Reactor Coolant System (RCS) temperature is being maintained at 250 deg. F.
Which of the following is the Tech Spec requirement regarding the operability of Charging and
- Safety Injection (SI) pumps?
'At least one Centrifugal Charging Pump must be OPERABLE to provide Reactor Coolant
'path
....... from the Refueling Water Storage Tank (RWST).
.and one SI Pump must be OPERABLE to provide a boron injection path from the RWST if iRCS pressure is >/= 165 psig.
1
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,If 23 Charging Pump is operating to provide RCP seal injection then NO other Charging or SI iPump are allowed to be OPERABLE.
123-Charging Pump can be operating to provide RCP seal injection if 21 OR 22 Charging Pump
~
~ d
/Comprehension
~
@00022G133.j
[Emergency and Abnormal Plant Evolutions hnical sDecifications.
SRO OKLV 5523(b)(2)--
I
- d. is correct for the reason stated. Tech Spec limitations are for Centrifugal Charging pumps only. A and
/b are incorrect because only one Chg or one SI pump can be OPERABLE for the stated plant condition.
,C. is incorrect since there is no flowpath from RWST to RCS via centrifuqal pump.
___ - I
--- - -1 alem Unit 2 Tech Specs i
ECCSOOEOIO I State the Technical Specifications associated with the components, parameters, and operation of the Emergency Core Cooling System, including:
a) b)
c) d)
The Limiting Condition(s) for Operation (N/A NEO)
The applicability of the LCO(s)
The LCO Action Statement(s) (N/A NEO)
The Bases for the LCO(s)
Tuesday, May 13, 2003 4:28:47 PM
'Page 11 of 126
Given the following conditions:
- Salem Unit 2 is in Mode 3.
- RWST level is 40.6'.
- RWST temp is 48 deg F.
- RWST boron concentration is 2344 ppm.
- 21 BAST is C/T.
1-22 BAST level is 95%.
1-22 BAST boron concentration is 6650 ppm.
1-22 BAST temp is 90.5 deg F.
I
- Which of the following describes the status of the Borated Water Sources?
IINOPERABLE due to Boric Acid Storage System parameters.
1
/INOPERABLE due to RWST parameters.
- Fullv OPERABLE.
1
- Available. but NOT reauired to be OPERABLE in Mode 3.
1 iAA2.JiAbility to determine and interpret the following as they apply to Emergency Boration:
AA2.04 [Availability of BWST
/SRO ONLY 55.43(b)(2)
'For Mode 3:
- is -97%
12300 PDm.
Minimum BAST boron concentration is 6560 ppm Minimum level with 1 BAST C/T Minimum RWST level is 364,500 gallons (40.5')
Minimum RWST boron concentration is
/Borated Water Sources I
I I
1
[Salem Technical Specifications
- CVCSOOEOIO
_ ~ _ _ _____
cvcsooEoo8-State the Technical Specifications associated with the components, parameters, and operation of the Chemical and Volume Control System, including:
a) b)
c) d)
The LCO Action Statement&)
ldenti e the 01s h th rol System, including:
a) b)
c) components and operation (N/A NEO) d)
their intended function e)
The Limiting Condition@) for Operation (N/A NEO)
The Bases for the LCO(s)
The applicability of the LCO(s) (N/A NEO)
The Control Room location of Chemical and Volume Control System control bezels and indications (N/A NEO)
The function of each Chemical and Volume Control System Control Room control and indication (N/A NEO)
The effect each Chemical and Volume Control System control has upon Chemical and Volume Control System The plant conditions or permissives required for Chemical and Volume Control System Control Room controls to perform The setpoints associated with the Chemical and Volume Control System control room alarms.
~-.
-~
TS Figure 3.1.2 Tuesday, May 13,2003 4:28:47 PM 1
Page 12 of 126
I S2.OP-ST.CVC-0010 A T. 2 Tuesday, May 13, 2003 4:28:47 PM
&Page 13 of 126
Given the following conditions:
- Unit 2 is in Mode 5 with 21 Residual Heat Removal (RHR) pump in service for cooling.
- The RO reports that Pressurizer (PZR) level is slowly lowering unexpectedly.
- NO Overhead Annunciator alarms have been received.
- Refueling Water Storage Tank (RWST) level is stable.
I-21 Waste Hold Up Tank level is rising slowly.
1
/S2.OP-AB.RHR-0001, LOSS OF RHR, has been entered and steps are being performed to locate
/and isolate the leak.
4 I
,Which of the following operator actions will isolate this leak?
[Close 2SJ69, RWST TO RHR, AND 2RH21, RHR
..... TO RWST.
- ~_______
1
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/Close 2CV8. RHR Letdown.
~R 1
, emove 21 RHR Loop from service and put 22 RHR loop in service.
tion and isolability of leaks
!All are steps to isolate leak in the procedure. B. is correct since it is the only isolation of a source above I
i84' el in the Auxiliary building. The level rise in the WHUT could only come from a source above el. 84'
~
'All others. while isolation strateaies in the txocedure are not above 84'.
~~~
Tuesday, May 13,2003 4:28:47 PM
Given the following conditions:
j-Unit 2 is at 88% power.
i-The crew is attempting to isolate a Component Cooling (CC) Water leak using S2.OP-AB.CC-jOOOl, COMPONENT COOLING ABNORMALITY.
- - 22 Charging Pump is in service.
+ 21 and 23 CC Pumps are in service.
- - 22CC Pump and 22 CC Heat Exchanger are isolated.
\\- CC Surge Tank level indication LI-628C has been raised to 42% and it is now lowering.
1-CC Surge Tank makeup is isolated.
1-OHA C2 CNTMT SUMP PMP START has actuated and NO other alarms are in.
IWhich of the following identifies the location of the leak?
! i I.........
I 121 Component Cooling header.
~
~
1 122 Component Cooling header.
~
~...~..
!Component Cooling Surge Tank.
~. -
~
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~
r..-..--...--_.____.__.
~
- AA2.
]!Ability to determine and interpret the following as they apply to Loss of Component Cooling Water:
- a. is correct. With 21 CC Hdr and Non-safeguards header in service, surge tank level lowering and cntmt sump alarm in, the leak is on non-safeguards header in containment. B&c are incorrect for same reason.
$urge tank is not source of leak if cntmt sump alarm is only alarm in.
~..
ICOMPNENT COOLING ABNORMALITY Describe, in general terms, the actions taken in S2.OP-AB CCOOOI and the bases for the actions in accordance with the Technical Bases Document.
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Page 15 of 126
Which of the following Pressurizer Level controlling channel leak locations will cause ACTUAL level 1 to rise ABOVE the programmed setpoint ?
Level detector reference leg.
'Level detector variable leg.
Pressurizer Vapor Space.
Pressurizer Surge Line.
/000028K101 1
n
/AKl._J:Knowledge of the operational implications of the following concepts as they apply to Pressurizer Level Control Malfunction:
[AKl.OI i (PZR reference leak abnormalities
~ ]12:e!l charging flow and actual level higher than setpoint. A. Reference leg leak will cause higher indicated
~
- level and a lowering of charging and actual level. C. Vapor space leak will result in actual level rising and j
- lowering of charging and level to setpoint. D. Surge line leak will cause actual level to lower and this will cause charging to rise to restore level to setDoint but not above setDoint.
J Pzr Level Control simplified 1
L PZRP&LEOW Variable Heaters Backup Heaters Pressurizer Pressure Control and Alarm Channels Master Pressure Controller Spray Valves Power-Operated Relief Valves (PORVs)
Power-Operated Relief Valve Block Valves Code Safety Valves Pressurizer Overpressure Protection System Pressurizer Level Control and Alarm Channels Master Level Controller State the purpose of the Pressurizer Pressure and Level Control system Tuesday, May 13,2003 4:28:48 PM I Page 16 of 126
Given the following conditions:
- Salem Unit 1 has experienced an Anticipated Transient Without Trip ( A M ), and has transitioned out of 1 -EOP-TRIP-1 to 1 -EOP-FRSM-1.
- SI has NOT been initiated.
- Rapid Boration flow CANNOT be established through 1 CV175, Rapid Borate Stop Valve.
!Which of the following choices identifies the next sequence of steps to be performed?
/Start both Boric Acid Transfer Pumps in High Speed, start a second Charging pump and fully lopen:
1 1CV55, CHARGING FLOW CONTROL VALVE; lCV77, CHARGING TO LOOP 13 AND
'1CV79, CHARGING TO LOOP 24.
1
- Start both Boric Acid Transfer Pumps in High Speed, stop the Primary Water Make-up pump,
,start a second Charging pump and direct full make-up flow to the Charging Pump suction by lclosing I CVI 81, MAKEUP FLOWPATH VALVE and fully opening I CV185, MAKEUP IFLOWPATH VALVE.
isend Operator to establish Rapid Boration by opening 1CV174, BA BLENDER BYPASS IVALVE then start both Boric Acid Transfer PumDs in High Speed.
i I
- Open ISJI and 1SJ2, RWST TO CHG PUMP; align charging flow through the BIT; shut
,I CV40 and 1 CV41 VCT DISCH STOP VALVE; and isolate normal charging flowpath.
~
iC~~2gA103.
~......................... 13'
[Emergency and Abnormal Plant Evolutions 1
E.-..,
1 I::::
Anticpated Transient Without Scram i
AAI. I Ability to operate and I or monitor the following as they apply to Anticipated Transient Without Scram:
in Control Room Evac procedures.
~~~~
[RESPONSE TO NUCLEAR GENERATION I
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I 1
Tuesday
~ May 13,2003 4:28:48 PM 1 Page 17 of 126
In accordance with the Salem FSAR Accident Analysis, which of the following Fuel Handling Accidents could result in a HIGHER thyroid dose received at the Exclusion Area Boundary (EAB),
and why?
Fuel Handling Accident in the CONTAINMENT Building because all airborne activity that reaches the containment atmosphere is assumed to exhaust to the environment within 2 hours2.314815e-5 days <br />5.555556e-4 hours <br />3.306878e-6 weeks <br />7.61e-7 months <br />
[without filtration.
IFuel Handling Accident in the FUEL HANDLING Building because all airborne activity that
!reaches the FHB atmosphere is assumed to exhaust to the environment within 2 hours2.314815e-5 days <br />5.555556e-4 hours <br />3.306878e-6 weeks <br />7.61e-7 months <br /> without Ifiltration.
IFuel Handling Accident in the CONTAINMENT Building because the dilution rate of the
'unmonitored release is less.
1
- Fuel Handling Accident in the FUEL HANDLING Building because of a lack of physical barriers
!to prevent egress of airborne activity to the outside environment.
i000036A203 AA2. JlAbility to determine and interpret the following as they apply to Fuel Handling Incidents:
AA2.03 ' [Magnitude of potential radioactive release
~
~~ ____ ~.
~
...___-__...-.I ISRO ONLY 55.43(b)(7),(4)
- In chapter 15 of the FSAR, section 15.4.6 is devoted to Fuel Handling Accidents. The facts and assumptions used to determine radiation dose at the EAB are the same, except for 3 additional
'assumptions which are only specific for a FHA in the CAN. One of these is, "All airborne activity reaching the containment atmosphere is assumed to exhaust to the environment within 2 hours2.314815e-5 days <br />5.555556e-4 hours <br />3.306878e-6 weeks <br />7.61e-7 months <br /> WITHOUT filtration. (My emphasis). Also in section 15.4.6.2 in the Method of Analysis section, # I 1 states, "It is conservatively assumed that 25% of the radioactive effluent escapes unfiltered from the fuel handling
- building following postulated failure of one exhaust fan." In section 15.4.6.3.1, Radiation doses, the
'following doses are noted to be at the EAB for the postulated Accident: FHB Fuel Handling Accident Thyroid dose 10.4 Rem, Containment Fuel Handling Accident Thyroid dose 28.7 Rem. Therefore, A is correct because of the above noted FSAR discussion.
the airborne activity is expected to be released to the atmosphere unfiltered.
'because dilution rate is not a factor for containment accident and all the activity is assumed to be
,released to atmosphere.
.. by filtering exhaust air prior to release to atmosphere.
Distracter B is incorrect because only 25% of Distracter C is incorrect Distracter D is incorrect because ventilation system will reduce dose received 1
I 4
Determine the expected alarms and indications.
Describe the analysis assumptions.
Describe the protective features that mitigate the event.
- 3)
- 4)
Describe the expected plant response.
1 I
Tuesday, May 13,2003 4:28:48 PM 1-Page 18 of 126 i
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Tuesday, May 13, 2003 4128~48 PM 1
Page 19 of 126
Given the following conditions:
- Unit 2 is at 100% power.
- 21 Steam Generator has a verified tube leak.
- The crew has entered S2.OP-ABSG-0001, STEAM GENERATOR TUBE LEAK.
Trending of the leak shows:
j-1000 - 25 gpd
-1100 -40 gpd
~ -1130-40gpd 1-1230 - 75 gpd
'-1300 - 75 gpd
-1330 - 80 gpd
-1400 - 85 gpd h030 - 30 gpd 1-1200 - 55 gpd IAW the Continuous Action Summary of S2.OP-AB.SG-0001, STEAM GENERATOR TUBE LEAK, which of the followina identifies the time a Unit shutdown is reauired. and whv?
'1230; because Tech Spec limits have been exceeded.
- I 330; to prevent leak propagation.
............................. I
~ _____..........
-~
.I
,1330; because Tech Spec limits have been exceeded.
[Emergency and Abnormal Plant Evolutions 1
12 -1
- 2
-~
-1 15' IO37 t
1 ~Steam.Generatp_rrube_leak.
- 2.1 1 Conduct Of Operations 2.1.32
! IAbility to explain and apply all system limits and precautions.
'13.4133.81
- c. A tube leak >/= 75 gpd has existed for >I hr. and the unit is shutdown in anticipation of rapid propagation of the leak. B. is incorrect because 75 gpd is a shutdown criteria but not a tech Spec limit.
A. is incorrect, a 30 gpd increase has occurred but not in 77% OR RCS subcooling 0. The ipurpose of depressurizing the RCS, (from Basis document) is "To decrease RCS pressure to stop
.primary-to-secondary leakage and establish an indicated pressurizer level. B. is incorrect since RCS
.pressure is already below safety setpoints. C. is incorrect since the pressure difference will not result in
.more heat transfer to intact generators or less heat transfer to the ruptured generator. D. is incorrect since there is no boron dilution in progress until backflow conditions exist.
i ISTEAM GENERATOR TUBE RUPTURE 1
~
SGTROI E003 I Describe the EOP mitiaation strateav for a steam aenerator tube ruoture.
I Modified stem conditions to reflect required actions to be taken at Salem.
I Tuesday, May 13,2003 4:28:48 PM 1
Paae 23 of 126
,For a given steam line rupture size and location, which of the following sets of initial conditions will result in the SMALLEST reactivity addition rate IMMEDIATELY following the rupture?
IMiddle of core life with Reactor Coolant System temperature at 350 deg F.
\\
LOSCOI E006 i
'Middle of core life with RCS temperature at 547 deg F.
For the following FSAR transients stated as items A, 6, and C, perform the tasks specified in the below requirements 1,2,3, and 4:
A. Accidental Depressurization
- 6.
Main Steam piping break C.
Main Feed piping break
- 1) Describe the analysis assumptions.
- 2) Describe the protective features that mitigate the event.
- 3) Describe the analyzed plant response.
- 4) State whether the analysis indicates fuel damage and, if so, describe the expected fuel failure mechanism
~.
~
,End of core life with RCS temDerature at 350 dea F.
1 IEnd of core life with RCS temperature at 547 deg F.
1
~000040K105 18
- AKl.
-- ]!Knowledge of the operational implications of the following concepts as they
. apply to Steam Line Rupture: 1 lAKl.05 1 beactivity effects of cooldown j imm1 I
iSRO ONLY 55.43(
la. Middle of core life will have a lower moderator temperature coefficient than end of life due to higher
/boron concentration, and for a given change (lowering) in temperature, a higher amount of positive
,reactivity will be added at an initial temperature of 547 than at 350 degrees. (Figure 18 of S1M.RE-iRA.zz-0012 j
/Salem Updated Final Safety Analysis Report J
LOSS OF SECONDARY COOLANT IFiaures 1
ABSTMI E005 For the following analyzed transients/accidents:
a)
Excessive Load Increase Incident
- 1) Determine the expected alarms and indications
- 2) Describe the analysis assumptions
- 3) Describe the protective features that mitigate the event
- 4) Describe the expected plant response.
i IEditorially Modified 1
Tuesday. May 13,2003 4:28:48 PM 1
Page 24 of 126
Given the following conditions:
STDUMPEOIO i
1-An electrical disturbance resulted in a loss of all Unit 2 Circulators.
1-Unit 2 reactor has tripped from 50% power.
/Which of the following will be the RCS temperature I O minutes after the trip and why?
555 den F due to the value of the lowest Main Steam Safety Valve setpoint.
1 I
Describe the conditions that will cause the Steam Dump System to become Armed and Blocked, and the actions that occur as
~
a result of armina and blocking
- 552 deg F due to the Steam Dump Load Rejection Controller.
648 L
_ deg F due to the MSIO L Main Steam Atmospheric Relief setpoint at 1015 psig.
1 I
1
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1547 deg F due to the Steam Dump Plant Trip controller.
C a
I
[Emergency and Abnormal Plant Evolutions 000051 K1 01 r']
Loss of Condenser Vacuum 191 -
AKI. 1 Knowledge of the operational implications of the following concepts as they apply to Loss of Condenser
/Vacuum:
AK1.01 J [Relationship of condenser vacuum to circulating water, flow rate, and temperature 1 c2,4;1[247
- c. The MSIO setpoint of 1015 psig will result in 548 deg F RCS temp since that is saturation for that pressure. Also the MSIO setpoint is lower than the MS Safety setpoint. A. 555 is Tsat for 1070 psig which is the safety valve setpoint. B&c. 552 is no-load Tave plus 5 deg which is the controlling setpoint for the load rejection controller post-trip, 543 is the PI2 setpoint which blocks Steam Dump operation
,post-trip. Neither of these are correct since condenser is NOT available for steam dump operation in this case.
I IREACTOR TRIP RESPONSE I
ISTEAM DUMP SYSTEM OPERATION
.. I a)
Turbine Bypass Spray/Steam Dump System :
i 1
STDUMPEOIZ r State the setDoints for automatic actuations associated with the Steam Dumo Svstem Tuesdav. Mav 13.2003 4:28:49 PM 1
Page 25 of 126
Given the following conditions:
AFWOOOE006
- Salem Unit 1 has tripped from 100% power.
- I I and 12 Steam Generator Feed Pumps have tripped.
b)
Turbine-Driven Auxiliary Feedwater Pump c) d)
e) 9 g) h)
i)
Describe the interlocks associated with the following Auxiliary Feedwater System components:
a)
Turbine-Driven Auxiliary Feedwater Pump Start-Stop Valve (MS132)
Turbine-Driven Auxiliary Feedwater Pump Trip Valve (MS52)
Turbine-Driven Auxiliary Feedwater Pump Speed Control Valve (GOV) (MS53)
AFW Pump Alternate Suction Header Supply Valves (AF52s)
Motor-Driven AFW Pump Recirculation Flow Control Valves (AF140)
Motor-Driven AFW Pump Discharge Flow Control Valves (AF21)
Turbine-Driven AFW Pump Discharge Flow Control Valves (AFI 1)
Auxiliary Feedwater Pump Automatic Start While performing Auxiliary Feed (AF) flow verification at step 3 of 1-EOP-TRIP-2, REACTOR TRIP
/RESPONSE, the operator observes the following conditions:
I
'-All Steam Generator (S/G) Narrow Range (NR) levels are off scale low.
-1 1 through 14AFIl's indicate closed.
-1 1 through 14AF21's indicate 10% open.
-AF flow to each S/G is 4E4 Ibm/hr.
-1 IAF Pmp discharge pressure is 1380 psig.
-12 AF Pmp discharge pressure is 1350 psig.
-1 3 AF Pmp tripped on overspeed.
I
,Which of the following choices identifies the action required by 1-EOP-TRIP-2 for these conditions?
- Depress the Pressure Override Defeat pushbutton for 1 I AF Pmp to open 11 and 12AF2l's
'and for 12 AF Pmr, to or,en 13 and 14AF21's to establish feed flow >22E4 Ibm/hr.
1
\\Depress the I MS52, TRIP VALVE OPEN, for 13 AF Pmp to restart 13 AF Pmp.
1
!Depress I I through 14AF21 OPEN pushbuttons to establish feed flow >22E4 Ibm/hr.
/Emergencya_nd-Abnormal Plant Evolutions Dwmmlp T1
~
'054............
Loss of Main Feedwater 1
20; 1AK3. //Knowledge of the reasons for the following responses as they apply to Loss of Main Feedwater:
- AK3.03
~ [Manual control of AFW flow control valves r-;s ;--~;,1
- d. is correct because the AF2l's are not currently open enough to establish feedflow requirements. There I is no reason for the valves to partially closed. B. is incorrect since the feedflow requirement for these conditions is not met. C. is incorrect the MS52 cannot be reset and opened form the Console. A. is incorrect. there is sufficient discharae Dressure to meet the nressure override.
1 I
J 1
I_._
REACTOR TRIP RESPONSE I--
I I
Tuesday, May 13,2003 4:28 49 PM j P a g e 2 6 0 f 1 2 6
AFWOOOE008 b) c)
Motor-Driven AFW Pump Recirculation Flow Control Valves Motor-Driven AFW Pump Discharge Flow Control Valves Identify and describe the Control Room controls, indications, and alarms associated with the Auxiliary Feedwater System, including:
a) b)
c) d)
intended function e)
The Control Room location of Auxiliary Feedwater System control bezels and indications The function of each Auxiliary Feedwater System Control Room control and indication The effect each Auxiliary Feedwater System control has upon Auxiliary Feedwater System components and operation The plant conditions or permissives required for Auxiliary Feedwater System Control Room controls to perform their The setpoints associated with the Auxiliarv Feedwater System control room alarms Tuesdav. Mav 13.2003 4:28:49 PM I Page27 of 126 1
!Which one of the following correctly completes the operator action concerning Safety Injection
,actuation in the event of an extended loss of all AC power?
The SI signal will be manually actuated...
and reset after power is restored to at least ONE 4KV vital bus, to allow automatic loading of bus.
and reset while the 4KV Vital busses are de-energized, to prevent automatic loading of bus.
/only if an automatic actuation signal is present, and is reset after power is restored to at least
'ONE 4KV Vital bus. to allow automatic loadina of bus.
,only if automatic actuation signal is present, and is reset while the 4KV Vital buses are de-
!energized, to prevent automatic loading of bus.
i 000055K302 j
~
.EK3. ]/Knowledge of the reasons for the following responses as they apply to Station Blackout:
rEK3.02
__ 1 [Actions contained in EOP for
- loss of offsite and
^_____
onsite
- p i-------___
~
- SI initiation and reset is directed in EOP-LOPA-1 step 21. SI Initiation and reset is performed to ensure
'automatic loading does NOT occur upon restoration of a vital bus. SI signal reset is necessary to allow qecovery in EOP-LOPA-2 for as wide a range of RCS conditions as possible, Le., those conditions jwherein RCS pressure or secondary pressures are below their SI actuation setpoints but the recovery
'procedure selection criteria (RCS subcooling and PZR level) are above minimum values which permit irecoverv per EOP-LOPA-2 I
J 1L
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~
~.__-
oss of All AC Power LOPAOOE008 Given a Step, Caution, Note, or Continuous Action Summary Item in EOP-LOPA-1, state its bases transientslaccidents.
A.
B.
Loss of all offsite power 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 Tuesday. Mav 13.2003 4:28:49 PM 1
Page 28 of 126
~
- RCS Pressure lowered to 1750 psig during the cooldown
.Which one of the following identifies the next procedure to be implemented?
REQ"l RED.
1
~ _.____
k-EOP-LOPA-3, LOSS OF ALL AC POWER RECOVERY/SI REQUIRED 1
~ - -. ~. ~
~
/2-EOP-TRIP-I, REACTOR TRIP OR SAFETY INJECTION.
~
-~
~~
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~
I
.2-EOP-TRIP-2. REACTOR TRIP RESPONSE.
- SI
. 1 (Emergency and Abnormal Plant Evolutions 1
IO56 3 :Loss of Off-site Power
~
2.4 1 Emergency Procedures / Plan 2.4.6 IKnowledge symptom based EOP mitigation strategies.
rn.4.01 A is correct because when a 4kv vital bus is recovered in LOPA-1, a continuous action step sends you to step 42. After restoring appropriate SW pps and CFCU's, a check for subcooling is made. In this case, subcooling is > 0, which will not require going to LOPA-3. the next check is for adequate PZR level. In this case it is adequate (>l l%), which will not require going to LOPA-3. The correct transition is to LOPA-2 to recover from the loss of All AC. It is incorrect to transfer back to TRIP-1 (procedure in
'effect) because all of the SEC's are deenergized, and equipmenffsystems will need to be manually
'alianed in LOPA-2.
/LOSS OF ALL AC POWER I
EOP-LOPA-1 and a set of plant conditions: A. Determine a discrete path through the EOP.
B.
Determine an appropriate transition out of the EOP Tuesday, May 13,2003 4:28:49 PM 1
Page 29 of 126
Tuesday, May 13,2003 42849 PM I Page 30 of 126
/Given the following conditions:
I
- OHA B-3,2A VTL INSTR BUS INVRT FAIL, annunciates.
I-An NE0 sent to investigate reports that 2A Vital Instrument Bus (VIB) Inverter Static Rectifier has an INV AC OUTPUT LOW/FAIL Alarm illuminated.
- All other indications are normal.
iln accordance with S2.0P-S0.115-0011,2A VITAL INSTRUMENT BUS UPS SYSTEM
/OPERATION, which of the following is the correct action to be taken, and why?
Use the MANUAL BYPASS SWITCH to transfer the inverter to its DC POWER SUPPLY.
- Use the MANUAL BYPASS SWITCH to transfer the inverter loads to the AC LINE
'REGULATOR.
- Ensure the STATIC SWITCH TRANSFERRED indicator is illuminated to verify 2A VIB has itransferred to the AC LINE REGULATOR.
'Ensure the INVERTER DC INPUT AVAILABLE indicator is illuminated to verify 2A VIB has itransferred to the DC POWER SUPPLY.
[Emergency and Abnormal Plant Evolutions I !
23' 057 1
1LosS.ofVita! -4Crnstrurnent-Eus
_.................................... 1 2.1 - ],Conduct Of Operations 2.1.32 - [Ability to explain
_ _ _ _ and apply all system _ limits
__ and
_-__-- precautions.
- c. is correct because the inverted output has failed and the system should automatically transfer to the AC Regulator as indicated. A. and b. are incorrect because Precaution and Limitation 3.3 of S2.0P-attempt to use the manual Bypass (MAN>BYPASS) switch to transfer the inverter load. Use of the Manual Bypass switch will result in inverter failure. Inverter loads will not be transferred with the VIB energized. D. is incorrect because the DC input has no effect at this point with the inverter transferred to the line reaulator.
S0.115-0011,2A VITAL INSTRUMENT BUS INVERTER UPS SYSTEM OPERATION, "DO NOT 12A VITAL INSTRUMENT BUS UPS SYSTEM OPERATION I
115VACE013 I Describe the procedures which govern the operation of the 115 Vac Electrical Systems, including:
1 15VACE007 a) b)
The purpose of each procedure Significant prerequisites and precautions associated with each operating procedure which are required to be considered by i)
Licensed Operators (N/A NEO) ii) Non-Licensed Operators (N/A LO & STA)
The major operational evolutions accomplished by each procedure which are primarily the responsibility of:
i)
Licensed Operators (N/A NEO) ii) Non-Licensed Operators (N/A LO & STA)
The location of 11 5 Vac Electrical Systems local controls and indications The function of 115 Vac Electrical Systems local controls and indications The plant conditions or permissives required for 115 Vac Electrical Systems local controls to perform their intended c) a)
b) c)
fiinctinn 1
Page 31 of 126 Tuesday, May 13,2003 4:28:49 PM
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Tuesday, May 13,2003 4:28:49 fi
[
Page 32 of 126-
Which of the following describes how a loss of 125VDC affects a Reactor Trip Breaker (RTB)?
The breaker is NOT capable of opening on a signal to the shunt trip coil.
!d The loss of voltage causes a shunt trip actuation and the breaker to open.
The breaker is NOT capable of opening on a signal to the UV trip coil.
The loss of voltage de-energizes the UV coil and the breaker opens.
~
I il3.5jBJ
[m.!..]
. Ability to determine and interpret the following as they apply to Loss of DC Power:
h 2. 0 3 ! /DC loads lost: impact on to operate and monitor plant systems
~
/125VDC power is used to power the shut trip co
!Undervoltage Trip which is also on each reactor trip breaker. Power is not required to trip the UV portion
~
of breaker. but Dower is supplied to eneraize the shunt trip coil as a backup to the UV triD.
[125Dc GROUND DETECTION J
r I
- RXPROTEO24
-1 Summarize each event 1
Identify the root cause(s) of the event Describe the events likelihood of occurrence at Salem Nuclear Generating Station Describe established or alternative actions which might prevent the events (re)occurrence at Salem Nuclear Generating Station
- RXPROTEOIO 1 Discuss the function, operation, and interlocks associated with the Reactor Trip Breakers and the Reactor Trip Bypass Breakers [
For plant or industry events associated with the Reactor Protection System :
Tuesday, May 13,2003 4:28:49 PM L-Page 33 of 126
lone of the Radiation Monitoring System (RMS) 2R13, Containment Fan Coil Unit (CFCU) Service Water, channels goes into ALARM.
Which of the following actions are required by S2.OP-AB.RAD-000, ABNORMAL RADIATION, and
/why?
Initiate Containment Ventilation Isolation to prevent an unmonitored release to the environment Stop the affected CFCU's and close the SW2's to prevent an unmonitored release to the environment.
1 i--
IStop the affected CFCU's and close the SW2's to limit the spread of airborne contamination hn the Containment.
hitiate a Containment Purge to filter and monitor a release of the Containment environment to
'the Dlant vent.
,000059K201 8
25
,environment is contaminated. C. is incorrect for the same reason. D. is incorrect, there would be no
'release or wrae made under these conditions.
IABNORMAL RADIATION i
._ 1 Technical Bases Document.
I I
I I
I Tuesday, May 13,2003 4:28:49 PM I
Page 34 of 126
- In response to an Alarm condition on Radiation Monitoring System (RMS) channel 2R16, Plant Vent Gross detector, S2.OP-AB.RAD-0001, ABNORMAL RADIATION, directs the Control Area Ventilation (CAV) systems for both units to be put into ACCIDENT PRESSURIZED MODE if there are NO RIB channels in service in the Unit 2 CAV Inlet duct.
Radwaste Release:
IWhich of the followina is the correct reason for taking this step?
/Insure Control Room personnel exposure is within 1 OCFRIOO limits during the accident.
'Prevent Control Room Area pressure from becoming too high during the
.......... accident.
'Prevent ALL outside air from affecting Control Room habitability.
1 1
1
- Prevent CAV exhaust flow from diluting the sample flow past the 2R16 detector.
- a. is correct per the design criteria for the CAV. B. is incorrect, Control Room Pressure is maintained
- higher than surrounding pressure by design in Accident Pressurized mode. C. is incorrect, a small
'amount of outside air is introduced even during accident mode. D. is incorrect, sample flow dilution 1
i
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I ABMDI E002 j
CAVENTEOOS the actions taken in S2.OP-AB.RAD-0001 (q)and the bases for the actions in accordance with the i HEAPHYE018 r Describe common sources of plant radiation and contamination.
j Tuesday, May 13,2003 4:28:50 PM 1
Page 35 of 126
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iWhich of the following choices identifies ONLY events which can be identified using Area Radiation I
~
Monitors (ARM'S)?
I LOCA inside Containment; Steam Generator Tube Rupture; fuel handling accident in the Fuel Technical Bases Document.
Handling Building (FHB); high radiation at the Condensate Polishers.
,LOCA inside Containment; fuel handling accident in the Fuel Handling Building (FHB); high
$radiation at the Condensate Polishers: rupture of a Gas Decay Tank.
~~
~~~~
/LOCA outside Containment; Main Steam Line Break; Reactor Coolant Pump Thermal Barrier
/leak; fuel handling accident in the Fuel Handling Building (FHB).
!LOCA outside Containment; Steam Generator Tube Rupture; rupture of a Gas Decay Tank;
- Reactor Coolant Pump Thermal Barrier leak.
\\AKj,iiKnowledge of the operational implications of the following concepts as they apply to Area Radiation
,Monitoring System:
~,
~
lAKl.01 L--
i 1 [Detector limitations I
@;Ti] E94
- a kick out to EOP-LOCA-1. A fuel handling accident uses R5 and R9 for identification and also for
/evacuation criteria. The condensate Polishers are located outside, and have no process monitors to (identify high radiation at the polisher, A. is incorrect, Tube Rupture is identified with process monitors on i (blowdown lines. C. is incorrect, thermal barrier leaks are identified using process monitors on
~~
1 E:..
. the actions taken in S2 OP-AB RAD-OOOl(q)and the bases for the act1 Tuesdav. Mav 13.2003 4 2 8 5 0 PM 1
Page 36 of 126 '
Given the following conditions:
components
- Unit 2 is at 100% power.
1-The crew has entered S2.OP-AB.SW-0001, LOSS OF SERVICE WATER HEADER PRESSURE, due to a large leak just downstream of the 21SW22 and 21 SW23 piping connection.
- The control room crew has closed 21SW22 and 21SW23 as directed by the procedure.
,Which of the following describes the Service Water availability for the Emergency Diesel IGenerators (EDG) and Containment Fan Coil Units (CFCU) if a Reactor Trip/Safety Injection
/occurs?
~
I
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J
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~ _
lAll 3 EDG's are supplied from 21 and 22 SW Header; 3 CFCU's are supplied from 22 SW
~.
!Header.
1 Salem I
&2 I L-xBSWEEOO4 I Describe, in general terms, the actions taken in SW.OP-AB.SW-0002 and the bases for the actions in accordance with the I
-L Technical Bases document.
ABSWOIEOOI I Describe the oDeration of the SW system as applied to S2.OP-AB.SW-0001:
1 a) b)
c)
Normal power operation configuration.
d) e)
Cross-tie of headerdvalve locations.
Loads supplied from the Nuclear Header Loads supplied from the Turbine Header.
Major valves operated from Control Console.
r I
3
[Editorially Modified
- - - - - --I
~-..
Tuesdav. Mav 13,2003 4,2850 PM
There is a break in the Control Air System header supplying the Mechanical Penetration area.
/Which of the following actions will S2.OP-AB.CA-0001 have operators verify?
svstem.
~
/High flow has automatically swapped all redundant air supply panels in the Auxiliary Building, jwhich completely isolates the leaking portion of the system.
/Air Operated Containment Isolation valves 21 and 22CA330 use the downstream pressure to open. The decreasing pressure has isolated the leaking portion of the system.
High flow has caused the header isolation valves, 21 and 22CA50 to trip closed.
I lAAl. ]!Ability to operate and I or monitor the following as they apply to Loss of Instrument Air:
94A1.02 ' komponents served by instrume F_.
to minimize drain
'A. is correct, that is the purpose of the excess flow check valve. B. is incorrect the air supply panels swap on low system pressure and do not isolate the leak. C. is incorrect because the CA330's do not function that wav. D. is incorrect. there is no hiah flow close on the CA50's.
i I LOSS OF CONTROL AIR
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\\Editorially Modified I__.____....__....___-.
Tuesday, May 13,2003 4:28:50 PM I
Page 38 of 126
Given the following conditions:
- Alarm OHA A-7, FIRE PROT FIRE, is received in the Unit I Control Room.
!- 1 RP5 Fire Panel Zone 59 light, CONTAINMENT PANEL 335, and the TROUBLE ALARM for that row are illuminated.
- NO other fire alarms are present.
- Using the A-7 window section of S I.OP-AR.ZZ-0001, OVERHEAD ANNUNCIATORS WINDOW A,
/the Operator will:
'OPEN valve 1 FP147, Fire Protection Containment Isolation.
~
'initiate i
Containment Ventilation Isolation.
1 I
/announce i...................
on the station PA "Attention all personnel, fire at Containment Panel 335, stay clear".]
jannounce on the station PA "Attention all personnel, disregard Unit I fire alarm".
]:Ability to determine and interpret the following as they apply to Plant Fire on Site:
09 ! khat a failed fire alarm detector exists L....
d is correct. Although the Zone 59 alarm is in, the trouble alarm means there is a detection problem, no
/fire has occurred. A, b and c are all actions that could be taken if there was an actual fire but are
!incorrect for
~ the
~~~ same reason.
J OVERHEAD ANNUNCIATORS WINDOW A
~
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I A:
The Control Room location of Fire Protection System control bezels and indications (N/A NEO)
B. The function of each Fire Protection System Control Room control and indication (N/A NEO)
C.
The effect each Fire Protection System control has upon Fire Protection System components and operation (N/A NEO)
D. The plant conditions or permissives required for Fire Protection System Control Room controls to perform their E. The setpoints associated with the Fire Protection System control room alarms.
intended function Tuesday, May 13,2003 4:28:50 PM _ _
I Page 39 of 126
I !Given the following conditions:
i 1-Unit 2 Control Room is being evacuated in accordance with S2.OP-AB.CR-0001, CONTROL 1-The reactor has been tripped.
- - Three control rods indicate fully withdrawn.
lWhich of the following describes the required actions to initiate a Rapid Boration?
I 1 ROOM EVAC UATI o N.
i
- Align charging through 2CV175, Rapid Borate Stop Valve, in accordance with 2-EOP-TRIP-2,
'REACTOR TRIP RESPONSE.
~.-.
- Open CV175 from the Hot Shutdown panel and isolate control air to CV55, Charging Flow
,Control Valve, to allow maximum charging flow in accordance with S2.OP-AB.CR-001.
IEnsure a safety injection is initiated in accordance with 2-EOP-FRSM-I, RESPONSE TO
!NUCLEAR POWER GENERATION.
~
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c____________
IManually open 2CV175, and control charging flow through CV55, using the Manual hand 7
loperator IAW S2.OP-AB.CR-0001.
I
- d
'Comprehension Emergency and Abnormal Plant Evolutions d is correct because the CV175 must be opened manually and the CV55 is controlled using the manual
[controls at the valve. a and c are incorrect because the EOP's are not used during the implementation of
'this abnormal operating procedure. b. is incorrect because the CV175 cannot be operated from the Hot Shutdown Panel and the CV55 air is not isolated in this evacuation procedure.
CONTROL ROOM EVACUATION.
r--- ----- --
1 I
Tuesday, May 13,2003 4:28:50 PM I
Paae 40 of 126
Salem Unit 1 has experienced a Large Break LOCA.
The following plant conditions are present:
!- Containment pressure peaked at 24 psig, and is currently 18 psig.
!- Containment Sump level is 65%.
- All CETs are between 820-860 degrees F.
- Subcooling is -40 degrees F.
- RVLIS full range reading is 11%.
- When permitted by procedure rules of usage, which of the following procedures will be entered first
- from I-EOP-TRIP-I?
[I-EOP-FRCE-2, Response to High Containment
.................... Sump Level, Purple Path.
- I
-EOP-FRCE-I, Response to Excessive Containment Pressure, Red Path.
/I-EOP-FRCC-2 1
Response to Degraded Core Cooling Purple Path.
- I-EOP-FRCC-1, Response to Inadequate Core Cooling, Red Path.
3 1
1 I
2
- SI
- Comprehension :
[Salem I
& 2
/d
.EK3.
_________ ]/Knowledge of the reasons for the following responses as they apply to Inadequate Core Cooling:
nce contained in EOP for Inadequate Core Cooling
'SRO ONLY 55.43(b)(5)
IDistracter A is incorrect since the containment sump level of 65% is below the entry condition for 1-EOP-i i IFRCE-2 Purple Path of 75% (adverse number due to containment pressure being > 4 psig).
Distracter 1
IB is incorrect since the containment pressure of 18 psig is less than the >47 psig required for entry to 1-1
.EOP-FRCE-1 Red Path. Distracter C is wrong because a RED path condition for FRCC exists.
.Answer D is correct since the RED path conditions of FRCC are met; 5 or more CETs >700 and a RVLIS full ranae value 39%
~~
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RESPONSE TO DEGRADED CORE COOLING r--
1 FRCCOOE004 I FRCCOOT002 aths for a response to inadequate core cooling, and describe how the actions In 2-EOP-L.____.
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~...
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Tuesdav. Mav 13.2003 4:28:50 PM I Page 41 of 126
!Unit 2 is at 80% power when RMS channel 2R31, Letdown Line-Failed Fuel Process Rad Monitor istarts a rising trend.
~
iAK1. ]'Knowledge of the operational implications of the following concepts as they apply to High Reactor Coolant
- Which of the following describes how S2.OP-AB.RC-0002, HIGH ACTIVITY IN REACTOR
- COOLANT SYSTEM (RCS), will have the operators distinguish between a crud burst and failed fuel
- as the cause of the rising 2R31 indication?
!By requesting a Shift Chemistry Technician perform a radiological analysis of the RCS. A crud
!burst will show different concentrations of certain radionuclides than will failed fuel.
/By monitoring 2R31. Fuel damage will cause the indication to increase at a higher rate than a icrud burst.
requesting Radiation Protection to survey the letdown pipe area in the Auxiliary building.
1 L
,-B_y 7
'Radiation levels will be higher due to failed fuel than from a crud burst.
I
- By increasing letdown flow rate to 120 gpm. The 2R31 readings will lower if crud burst caused 1 ithe rising trend but will NOT lower.--
if failed fuel caused the rising trend.
1
- a
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F I
/Memory lsalern I 8 2 7
i000076K106 I
33 -
L High Activity in the Reactor Coolant System
_J rABRC02E003 Describe, in general terms, the actions taken in S2.OP-AB.RC-0002 and the bases for the actions in accordance with the
I
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I Technical Bases Document.
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Tuesday, May 13,2003 4:28:50 PM 1
Page 42 of 126
Given the following conditions:
- Unit 2 is performing the actions of 2-EOP-TRIP-3, SAFETY INJECTION TERMINATION.
- - SI is reset.
'- SEC's are reset.
- All safeguards equipment has been stopped with the exception of 21 Charging Pump.
Subsequently a Loss of Offsite power occurs and the following conditions are present:
I-Reactor Coolant System (RCS) pressure is 1660 psig and lowering.
I-Pressurizer level is 1 I
% and lowering.
(Which of the following actions is required IAW 2-EOP-TRIP-3, SAFETY INJECTION i
i
'TERMINATION?
'Allow Blackout Loading to complete, reset SEC's, and start ECCS pumps as necessary to
!maintain i
RCS inventory.
IAllow Blackout Loading to complete then align normal charging and letdown to maintain RCS
.inventory.
. I 1
'Allow
....... Blackout Loading to complete then manually reinitiate Safety...... Injection.
1
/Allow Blackout Loading with SI to complete then ensure that safeguards equipment is
!operating L
properly.
ia,
- B I Application
'F/iKnowledge of the operational implications of the following concepts as they apply to SI Termination:
'EKI.2 1
~ [Normal. abnormal and emergency operating procedures associated with (SI Termination).
~
3.4::-39]
a is correct because ECCS actions are necessary due to pressure and level lowering and the SEC's must I be reset to aain control of eauipment. B. is incorrect because normal charging and letdown are not I
appropriatefor pressure and level conditions. C. is incorrect because SI isnof re-initiated while in
- termination activities and cannot initiate until P-4 is reset. D. is incorrect because Blackout with SI I
I i
8 I
Tuesday, May 13,2003 4:28:51 PM 1
Page 43 of 126
Given the following conditions:
i I,
- The crew is performing 2-EOP-LOCA-2, POST LOCA COOLDOWN AND DEPRESSURIZA
- 21 Charging Pump has been stopped.
- Conditions are met for stopping one Safety Injection (SI) Pump.
Which of the followina is the required PumD to be stormed and whv?
'ION.
..... injection flow if any single vital bus failure occurs.
121 SI Pump to ensure one full train of ECCS equipment remains in service for injection.
I 122 i SI Pump to equalize Emergency Diesel Generator
~
loading.
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___- I 1
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1 c___________________....--_---
121 SI Pump to ensure ECCS iniection flow if any sinde vital bus failure occurs.
,Comprehension 100WE03A103 1
ExIzl
'EAl.]iAbility i
to operate and / or monitor the following as they apply to LOCA Cooldown and Depressurization: -
r-__ ---
~-
i.EA1.3 J [Desired operating results during abnormal and emergency situations.
I[ 3.7,: 4.11
~~
!a is correct per the Basis document for the EOP. B. is incorrect as diesel loading is not an issue until limits are reached or exceeded. C. is incorrect because train relationship is not relevant. D. is incorrect since that would leave both injection pumps in service on the same bus and may result in loss of injection until other pumps can be started.
I i
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i Tuesdav. Mavl3. 2003 4:28:51 PM r Page 44 of 126
'Which of the following Radiation Monitoring System channels in alarm identifies an indication that
'requires transition to 2-EOP-LOCA-6, LOCA OUTSIDE CONTAINMENT from 2-EOP-LOCA-I,
LOSS OF COOLANT ACCIDENT?
2RlB-I, Unit 2 Control Room Intake Duct.
2R15. Condenser Air Eiector.
2R34, Mechanical Penetration Area 100 Ft. Elev.
2R13A1 21 -25 CFCU Service Water.
1 kmergency and Abnormal Plant Evolutions iOOWE04A201 1
T I
LOCA Outside Containment Ezzl
.EM. ]/Ability to determine and interpret the following as they apply to LOCA.
Outside Containment:
_J conditions and selection of appropriate procedures during abnormal and emergency 3.4:; 4.3) i SRO ONLY 55.43(b)(5)
IC is correct. this is the only RMS monitor in table (c) of LOCA-1 which directs entry into LOCA-6. The I
irernaining channels do nit involve systems or areas that would likely be affected by a LOCA outside Icontainment.
1 (LOSS OF COOLANT ACCIDENT LOCA OUTSIDE CONTAINMENT 1
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LOCAOIEOI 1 1 Given 2-EOP-LOCA-1 and a set of plant conditions: A. Determine a discrete Dath throuqh the EOP
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i 1
B.
Determine an appropriate transition out of the EOP
_ _ ~
~--
Tuesday, May 13, 2003 4:28:51 PM 1
Paae 45 of 126
Given the following conditions:
I
'- Salem Unit 1 has experienced a large Main Steamline Break (MSLB) inside containment from 100% power.
,- Safety Injection was manually initiated, with all components operating as expected.
- 12 and 13 AFP's tripped after starting.
- RCS pressure is 11 00 psig.
I-All RCS Pps have been tripped.
1-Containment pressure is 16 psig and rising.
- All Wide Range S/G levels are 35% and dropping.
- All S/G pressures are 425 psig and dropping.
I-RCS Tc's have dropped from 540 deg to 440 degrees in 40 minutes.
11 AFP is C/T.
interlocks, failure modes, and automatic and manual features.
- Which of the following choices identifies the correct procedure to be entered and action to be take jupon transition out of I-EOP-TRIP-I, REACTOR TRIP RESPONSE?
!I-EOP-FRHS-I, RESPONSE TO LOSS OF SECONDARY HEAT SINK; i [ONLY when
__ S/G WR levels have dropped less than 32%.
- I-EOP-FRHS-I, RESPONSE TO LOSS OF SECONDARY HEAT SINK; initiate feed and bleed
- immediately because S/G WR levels are less than 36%.
'CONDITIONS; shut all MSI 0's and steam dump valves to minimize
~
cooldown.
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F~EOPFRTS-I, RESPONSE TO IMMINENT PRESSURIZED THERMAL SHOCK
~
I il-EOP-FRTS-1 RESPONSE TO IMMINENT PRESSURIZED THERMAL SHOCK
'CONDITIONS; reset Safeguards Actuation
~
and restore normal charging and
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letdown.
I Tuesday. May 13, 2003 4:28:51 PM 1 Page 46 of 126
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Tuesday. May 13,2003 4:28:51 PM 1
Page 47 of 126
,Given the following conditions:
- - Unit 2 tripped from 100% power due to a loss of all power.
- Operators entered 2-EOP-LOPA-1, LOSS OF ALL AC POWER.
- 23 AFP failed to start.
~I After 30 minutes, operators have been successful in restoring power to 2B vital bus.
1-22 Auxiliary Feedwater Pump is in service supplying 21 and 22 Steam Generators.
j-Due to a positioner failure on Power Operated Relief Valve 2PR1, the valve opened and will NOT
'close.
- - Pressurizer level is at 89%.
1-Reactor Coolant System Subcooling is e0 deg. F.
j-There are 7 Core Exit Thermocouples (CET) between 710 and 720 deg. F.
i-The remaining CET's are between 670 and 690 deg. F.
I-RVLIS is indicating 65% and is stable.
i IWhich of the following is the correct procedure transition to enter upon reaching the end of 2-EOP-i i LO PA-I
?
'2-EOP-LOCA-I, LOSS OF REACTOR COOLANT.
'2-EOP-LOPA-2, LOSS OF ALL AC POWER RECOVERY/SI NOT REQUIRED.
I 1
~ -__
12-EOP-FRCC-2, RESPONSE TO DEGRADED CORE COOLING.
I
- 2-EOP-FRCC-1 L
, RESPONSE....
TO INADEQUATE CORE COOLING.
mi EOWY-__1 38 I
[Emergency and Abnormal Plant Evolutions
'E06
'EA2 /'Ability to determine and interpret the following as they apply to Degraded Core Cooling:
'SRO ONLY 55.43(b)(5) id is correct, plant parameters indicate steam void in reactor vessel, no fuel uncovered. A. is incorrect, jalthough a LOCA is in progress the FRCC-2 Purple Path has priority. 8. is incorrect, for the same i
Ireason. C. is incorrect because conditions have not dearaded to Red Path for FRCC-1.
1
/RESPONSE TO DEGRADED CORE COOLING I
I Tuesday, May 13,2003 4:28:51 PM L a s e 48 of 126
Given the following conditions:
- Salem Unit 2 is responding to a Saturated Core Cooling condition IAW 2-EOP-FRCC-3, due to a loss of subcooling following a Reactor Trip and Safety Injection.
- RCS pressure is 1200 psig and stable.
/Under these conditions, which of the following choices identifies that ECCS flow is injecting to the
/RCS?
~
121 SI pump flow meter reads 60 gpm.
i 1
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bHR pump discharge flow reads 500 gpm on 21 SJ49 flow meter.
1
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[Charging flow reads 125 gpm on SI systems charging flow meter.
i
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~~~
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~~
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~~~
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1
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Ell SI Accumulator pressures dropping slowly.
c
'Application I
[Emergency and Abnormal Plant Evolutions I
jOOWE07A103 T
I Saturated Core Cooling j_________39;
'RCS, so distracter A is incorrect. RHR pumps shutoff head of 21 0 psid would not allow injection until iRCS pressure was much lower than 1200 psig, so distracter B is wrong. Accumulator normal pressure jband is 600-650 psig, so they would not be able to inject until RCS pressure was below that of the I
- accumulators, so distracter D is wrong.
i esponse to Saturated Core Cooling Conditions iP 1
1 FRCCOOE005 I Determine the indications that are monitored to ensure proper system/component operation for each step in the following:
A.
EOP-CFST-1, Figure 2 I
Tuesday, May 13,2003 4:28:51 PM
Given the following conditions:
iEK2.2
!- Unit 2 was operating at 100% power.
i-1 A loss of offsite power caused a reactor trip.
/Ten minutes after the trip the following conditions exist:
1-21 S/G Pressure - I010 psig and stable.
1-22 S/G Pressure - 1005 psig and stable.
1-23 S/G Pressure - 101 5 psig and stable.
1-24 S/G Pressure - 101 0 psig and stable.
1-All Reactor Coolant Pumps are OFF.
i I-Reactor Coolant System pressure is 2230 psig and stable.
1-Thot is approximately 575 deg F in all loops and lowering slowly.
I-Core Exit thermocouples indicate approximately 580 deg F.
I-Tcold is approximately 555 deg F in all loops and stable.
- Based on the above conditions, what is the status of RCS cooling in accordance with 2-EOP-TRIP
- 2, REACTOR TRIP RESPONSE, and what action is required?
I 1
I I
I 1
1
'Natural Circulation exists. The steam dumps are to be used to maintain heat removal.
Natural Circulation does NOT exist. It can be established by opening the steam dumps.
- Natural Circulation exists. The MSlO's are to be used to maintain heat removal.
~
Facility's heat removal systems, including primary coolant, emergency coolant, the decay heat ipml removal systems, and relations between the proper operation of these systems to the operation of the facility.
1 i
'Natural Circulation does NOT exist. It can be established by opening the MSlO's.
1 REACTOR TRIP RESPONSE I
1
. - ~
..-.~
I Quesgon SOUW:
Facility Exam Bank Direct From Source 1
Tuesday, May 13,2003 4:28:51 PM 1
Page 50 of 126
Tuesday, May 13, 2003 4:28:51 PM
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Page 51 of 126
/Which of the following actions are directed by 2-EOP-LOCA-5, LOSS OF EMERGENCY
/COOLANT RECIRCULATION?
IEK2.1 L
i jl.
Provide guidance on aligning the Safety Injection Pump suction directly to the Containment Sump.
- 2. Terminate Cold Leg Recirculation and restore Charging and Letdown.
i3. Cooldown and depressurize the Reactor Coolant System to allow Residual Heat Removal to be
'put into service.
- 4. Provide methods to make-up to the Refueling Water Storage Tank.
'I, 2 and 3.
.2 and 4 ONLY.
I 1
.___/
Components, and functions of control and safety systems, including instrumentation, signals,
?:!%
L33l interlocks, failure modes, and automatic and manual features.
1
/3 and 4 ONLY.
--1 ILOSS OF EMERGENCY COOLANT RECIRCULATION Tuesday, May 13,2003 4:28:51 PM EPage 52 of 126
Salem Unit 2 has experienced a Main Steam Line Break inside containment. All attempts to initiate a Main Steam Line Isolation from the Control Room have failed. All MSIV's remain open.
,EK1.2 The following conditions are present:
I
!- Containment pressure is 22 psig and lowering.
'- RCS pressure is 1400 psig and dropping. 24 S/G WR levels are 48% and dropping.
- Total Aux Feedwater flow is 24E4 Ibm/hr.
Based on these conditions, what will be the earliest point to lower Aux Feedwater Flow to 22E4 4bm/hr. and whv will Aux Feed flow be lowered?
Normal, abnormal and emergency operating procedures associated with (High Containment
'I3.2jj3.7 i
Pressure).
- When Narrow Range level in at least I S/G rises above 9%, aux feed flow can be reduced to
!prevent overfilling S/G's.
~
- Aux feedwater flow will be directed to be reduced to NO less than 1.OE4 Ibm/hr in 2-EOP-iLOSC-2 to minimize cooldown and prevent Thermal Shock to S/G components.
- 2-EOP-FRCE-l directs Aux feedwater flow to be minimized, and maintained > 1.OE4 Ibm/hr to
'keep S/G components wet, to minimize any Thermal Shock if feedwater flow is increased.
1 a When Narrow Range level in at least I S/G rises above ~ Y O,
aux feed flow can be reduced to prevent excessive cooldown of the RCS.
c Comprehension 1
~
E14Kl!2-__.!
42 I n
1 n
[Emergency and Abnormal Plant Evolutions j
- ~
E14
_____.................... High Containment Pressure
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[Distracter A is incorrect because with all 4 S/G's faulted, NR level will never rise to 9 % NR, and
/overfilling SIG's in this situation is not a concern.
Distracter B is incorrect, since when the transition
- from step 26 of 2-EOP-TRIP-I to LOSC-2, the PURPLE path for FRCE is present, and rules of usage
/require immediately performing this Purple path procedure. FRCE-1, step 7, states that if all S/G's are jfaulted, then minimize and maintain aux feed flow >1.OE4 Ibm/hr. Basis document for FRCE-1 states it
!is to prevent subsequent thermal shock if feedwater flow is raised.
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Distracter D is incorrect for same ireason Distracter A is incorrect regarding NR level
[Response to Excessive
. Containment Pressure 1
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.l.................
i Tuesday, May 13,2003 4 28.52 PM 1
Paae53of 126
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Tuesday, May 13,2003 4:28:52 PM I
Paae 54 of 126
While performing Emergency Operating Procedures, a step is encountered which states "Control
!PZR level between 25% (33% adverse) and 77% (74% adverse) by adjusting charging and
'letdown." Containment Pressure has risen to 5.5 psig and then lowered to 3.3 psig. Containment radiation levels have risen to 3.OE5 Whr and lowered to 6.7E4 R/hr.
jWhich of the following PZR levels are required to be maintained?
~
iMinimum of 25%; maximum of 77%.
/Minimum of 33%; maximum of 74%.
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~ A S
<...... directed by the Shift Technical Advisor (STA).
b s directed by the Operations Support Center (OSC).
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~ b
'Comprehension
,EKl
.J;Knowledge of the operational implications of the following concepts as they apply to High Containment I Radiation:
- being exceeded. a. is incorrect for the same reason. d is incorrect since the OSC never gives direction
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!to for plant status. However, the TSC does in certain situations.
A.-Immediate Actions B,Continuous Action Summaries C.-Communications D.-Log Keeping E.-Application of Notes and Cautions F.-Transitions G.-Adverse Containment Tuesday, May 13, 2003 4:28:52 PM 1
Page 55 of 126
Given the following conditions:
- Salem Unit I is operating at 100% power.
- I E 460 volt bus is deenergized following a trip of its feed breaker.
- Tagging is in progress to allow troubleshooting of I E 460 volt bus.
- The operator mistakenly opens the 1 F 460 volt bus feed breaker, deenergizing the I F 460 volt bus.
- Which of the following describes a consequence, if any, of this action? Salem Unit 1 Reactor...
1
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'will trip due to the loss of BOTH Rod Drive Motor Generators.
I
'will trip due to the loss of a SINGLE Rod Drive Motor Generator.
will NOT trip because ONE Rod Drive Motor Generator is sufficient to maintain power to the Rod Control system.
I
- will NOT trip because BOTH Rod Drive Motor Generators are still in service.
'm
'Control Rod Drive System
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05/21/2003,
/001000K205
$p-_._JiKnowledge of bus power supplies to the following:
r iK2.05 i 1 [M/G ____
sets
] b:d E3251 i
!Rod Drive Motor Generators are powered from 1 E and 1G 460 volt busses. A single RDMG set is i l l RDMG. the loss of 1 F 460 volt bus will not trit> both Rod Drive MG sets.
1 i,
sufficient to power Rod Control. Since the unit is still at 100% power subsequent to the loss of the 1 E 460V bus one line diagram 1G 460V bus one line diaaram hodcontrol System Operation State the Power Supplies to the following Rod Control and Position Indication systems components:
a)
Rod Drive MG Sets b) c)
Power Cabinets d)
Logic Cabinet Reactor Trip and Trip Bypass breakers Tuesday, May 13,2003 4:28:52 PM I Page 56 of 126
- Given the following conditions:
- Salem Unit 2 experienced a Large Break Loss of Coolant Accident from 100% power.
- Containment pressure has reached 15 psig.
- Which one of the following actions occur as a DIRECT result of this containment pressure?
~
/Containment Ventilation Isolation.
r
!ALL i
reactor coolant pumps lose
................... bearing cooling water.
7
[Control room ventilation swaps - to ACCIDENT PRESSURIZED Mode.
1 1
/K3. ]'Knowledge of the effect that a loss
_ or malfunction of the Reactor Coolant System will have on the following4
!b is correct because a Phase B Isolation occurs at 15 psig in containment, causing the RCP bearing
/cooling supply and return valves to close. The three distractors are all automatic actions that occur upon 1
'Safety Injection initiation, one of wehose initiation signals is 4 psig in containment. b is tha only choice
/that DIRECTLY occurs at 15 psig in containment, even though the other 3 actions WILL have occurred 1
iPRlOR to 15 mia. and therefore are not DIRECTLY attributable to 15 win.
i 4:d[_4.d L-CONTMTEOOI CONTMTEOOS State the purpose and design bases for the following Containment And Containment Support Systems subsystems:
Containment Containment Airlocks Containment Isolation System Containment Fan Cooler System Containment Iodine Removal System Rod Drive Ventilation System Reactor Nozzle Support Ventilation System Reactor Shield Ventilation System Containment Pressure - Vacuum Relief System j)
Hydrogen Recombiner System State the setpoints for automatic actuations associated with the Containment And Containment Support Systems:
a)
Containment Isolation Actuation bl Containment Ventilation Isolation Actuation 1
Tuesday, May 13,2003 4:28:52 PM
[
Page 57 of 126
Given the following conditions:
- Salem Unit 2 is operating at 100% power when a reactor trip occurs coincident with a loss of off-site power.
- Main steam dump control is in Tave Control -AUTO.
- Core burnup is 10,000 MWD/MTU.
Which of the following identifies the method of circulation through the reactor core 20 minutes after the trip?
1 Safety iniection flow into the cold legs will circulate up through the core.
I
!Charging i
. flow to loop 23
. - cold leg combined with letdown flow.
I
!Reflux cooling flow in the RCS hot leg to the steam generator.
~
!Natural L
circulation
- - flow will replace RCP forced flow circulation
_____.- through
.-. the
- -. core.
I
!002000K513 iK5. ];Knowledge of the operational implications of the following concepts as they apply to the Reactor Coolant
/System:
!K5.13 c _
[Causes of circulation.
[-:33 13:i IDistracter A is incorrect because SI will not be initiated. Distracter B is incorrect because boiling in the I
/Reactor will not be taking place. C is incorrect because charging flow of 87 gpm (normal) and letdown of,
I 175 gpm (normal) will not cause circulation. D is correct because natural circulation will be established 1
lapproximately 5-1 0 minutes following trip.
' ABRC04E003 For the following analyzed transientslaccidents:
a)
Salem Natural Circulation Test b)
Complete Loss of Forced Reactor Coolant Flow
- 1) Determine the expected alarms and indications
- 2) Describe the analysis assumptions
- 3) Describe the protective features that mitigate the event
- 4) Describe the expected plant response Tuesdav, May 13.2003 4:28:52 PM I
Page58of 126 ;
The plant is operating at 100% power with all control systems in AUTO. The following parameters
'are noted:
- Letdown Hx outlet flow (FI-132) 75 gpm.
- Charging Header flow (FI-121) 87 gpm.
- Total seal injection flow (FI-I42 -FI -45) 33 gpm.
/What is the initial effect on total seal injection flow if controlling Pzr level channel LT-459 fails LOW?
ITotal seal iniection flow will...
1
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llower to 0 gpm.
llower L_
to approximately 20 gpm.
1
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iremain approximately 33 gpm.
!rise to greater than 33 gpm.
!A3. ].Ability to monitor automatic operations of the Reactor Coolant Pump System including:
T r 3 3 ; i 321 iA3.01
[Seal iniection flow i
. j /.
L...........
i
---~
.---I lThe failure of the level instrument low increases charging flow and charging discharge header pressure.
- Since seal injection flow is normally increased by throttling close on CV71 to increase backpressure, the I
result is the same and seal iniection flow will increase.
IlNPO EXAM BANK I
~- a)
Basic RC.p Construction b) c)
RCP CW Configuration Identify and describe the Control Room controls, indications, and alarms associated with the Reactor Coolant Pump, including:
a) b)
c) d)
function e)
Seal Injection and Seal Water Configuration
/
RCPUMPE008 The Control Room location of Reactor Coolant Pump control bezels and indications (N/A NEO)
The function of each Reactor Coolant Pump Control Room control and indication (N/A NEO)
The effect each Reactor Coolant Pump control has upon Reactor Coolant Pump components and operation (N/A NEO)
The plant conditions or permissives required for Reactor Coolant Pump Control Room controls to perform their intended The setpoints associated with the Reactor Coolant Pump control room alarms 1--
Tuesdav. Mav 13.2003 4:28:52 PM A
I Page 59 of 126
~~~
/Given the following two sets of conditions:
- Salem Unit 1 is in Mode 3, HOT STANDBY, @ NOP, NOT.
- A cooldown caused by a Steam Dump malfunction caused pressurizer level to drop to 12%.
- Pressurizer pressure fell to 21 85 psig before the Steam Dumps were isolated.
- Pressurizer level was quickly recovered to 22%, and pressurizer heaters were returned to AUTO.
- Salem Unit 1 is in Mode 3, HOT STANDBY, @ NOP, NOT.
- A depressurization caused by a PORV malfunction caused pressurizer level to rise to 23%.
- Pressurizer pressure fell to 21 85 psig before the PORV was isolated.
1-Pressurizer level was quickly recovered to 22% and pressurizer heaters remained in AUTO.
i i /Which malfunction will take a longer time for pressure recovery from 2185 psig and why?
'The PORV because when the PORV opens the steam space needs to be reheated to raise
!pressure.
/The Steam Dump because the pressurizer heaters are less effective since they had tripped iand cooled off on low PZR level.
~
!The i
PORV because the pressure reduction will be faster for the failed
__ open PORV.
Steam Dump because the subcooled water insurge during refill reduced the Pressurizer I
I space temperature.
- malfunction. Distracter B is incorrect because it's effect is inconsequential. Distracter C is incorrect ibecause pressure reduction rate does not affect recovery time. D is the correct answer because after
'the outsurge due to lowering pressure, the insurge will be of cooler water, and will require a greater time to reach saturation and cause a Dressure rise.
IUFSAR-Accidental Secondary Depressurization Describe the function and operating characteristics for the following Pressurizer Pressure and Level Control system components:
a)
Variable Heaters b)
Backup Heaters c) d)
Master Pressure Controller e)
Spray Valves 9
Power-Operated Relief Valves (PORVs)
Pressurizer Pressure Control and Alarm Channels I
Tuesday, May 13,2003 428.52 PM 1
Page 60 of 126
Power-Operated Relief Valve Block Valves Code Safety Valves i)
Pressurizer Overpressure Protection System k)
Master Level Controller j)
Pressurizer Level Control and Alarm Channels I
I Tuesday, May 13,2003 4:28:52 PM I Page61 of 126
Given the following conditions:
j-Salem Unit 2 is at 100% power.
1-Core Burnup is 11,000 MWDIMTU.
- Rod Control is in MANUAL.
- All other plant controls are in their normal configuration.
- AUTO makeup initiated to the VCT.
'- The boron addition rate is set 5 gpm higher than required for present RCS conditions.
i IAssuming NO operator action, what will be the effect on the following parameters 15 minutes after ithe auto makeup is complete?
- Reactor Power
- RCSTave
... I er, lower, higher I
I I-Main Generator Electrical Output I.--
- Higher, i
higher, higher.
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-. ~
]Higher, higher, lower. ____
iLower, lower, lower.
iA4.
Ability to manually operate and/or monitor in the control room:
1 c-3-67 f--3-d A4.15 poron concentration I __.__
-__L
/At a burnup of 11,000 MWWMTU, the operator should know the normal boron addition rate will be somewhere around 6 gpm. By having the addition rate set 5 gpm higher, the boron addition will add negative reactivity to the core as it is pumped into the RCS from the VCT. AUTO makeup to the VCT stops at 24% level, and normal charging flow is 87 gpm. The VCT contains approximately 20 gallons per percent, so 400 gallons should be pumped into the RCS in 5 minutes. This allows 10 minutes for the
/effect of the boron addition to be seen on the control boards. The negative reactivity will cause Reactor power to lower, and RCS Temperature will lower. With the lower temperature, S/G pressure will lower,
'causing a drop in Main Generator output. The Main Turbine Valve position limiter will prevent the Main Governor valves from opening further to attempt to raise first stage turbine pressure. Rod control is in MANUAL. which would Drevent outward rod motion if the operator were to assume that rods weren't ARO I
Boron Concentration Control I.............
I CVCSOOE008 Identify and descnbe the Control Room controls, indications, and alarms associated with the Chemical and Volume Control System, including:
a) b)
c) components and operation (NIA NEO) d)
The Control Room location of Chemical and Volume Control System control bezels and indications (N/A NEO)
The function of each Chemical and Volume Control System Control Room control and indication (NIA NEO)
The effect each Chemical and Volume Control System control has upon Chemical and Volume Control System The plant conditions or permissives required for Chemical and Volume Control System Control Room controls to perform their intended function i e)
The setpoints associated with the Chemical and Volume Control System
.- control
- room alarms.
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Tuesday, May 13,2003 4:28.53 P r
1 Page62of126
prerequisites and precautions associated with each operating procedure which are required to be considered by either Licensed or Non-Licensed Oaerators.
Tuesday, May 13,2003 412853 PM
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I Page 63 of 126
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Given the following conditions:
- Salem Unit 2 has experienced a Large Break LOCA. EOP-LOCA-3, "TRANSFER TO COLD LEG RECIRCULATION" is complete with NO abnormalities encountered.
!- Operators are currently at step 26, "Preparation for Hot Leg Recirc", of 2-EOP-LOCA-1, "LOSS OF REACTOR COOLANT".
- Off-site power is supplying all 4KV Vital busses.
1 If BOTH RHR pumps are operating, what would be the effect if 22 RHR Pp were to trip?
I22 Containment Spray Pump would lose NPSH.
/Containment Spray flow would drop to zero.
1
/b Application 005000K306
~
50/
'K3.
i
]!Knowledge of the effect that a loss or malfunction of the Residual Heat Removal System will have on the 1
!following:
~
r-----
r-----7
' 13: I-*i i 32*1
!Distracter A is incorrect because both CS pps would be secured by the time LOCA-3 was completed.
B
/is the correct answer. LOCA-3 explicitly states that if BOTH RHR pumps are operating, then 22CS36 is 1
i /opened to supply containment spray from 22 RHR pp discharge.
' Distracter C is incorrect because the 1
121 and 22RH19's are closed in LOCA-3 to prevent runout conditions on the operating pump if the other 1
/RHR pp were to trip will in cold leg recirc. Distracter D is incorrect because NPSH would still be supplied i I
Transfer to Cold Leg Recirculation I
scribe the EOP miti gy for a Transfer to Tuesdav. Mav 13.2003 4:28:53 PM 1
Page 64 of 126--
During performance of SI.OP-DL.ZZ-0003, Control Room Logs Modes 1-4, the operator notices 12 6 J Accumulator level is at the log limit of 51 %, with corresponding Accumulator pressure at 61 5 Ipsig. How will the operator restore 12 Accumulator level and pressure to proper values?
/Raise Accumulator pressure first, then raise level.
/Raise Accumulator level first, then raise pressure.
Raise Accumulator level ONLY.
[Lower Accumulator pressure ONLY.
Knowledge of the operational implications of the following concepts as they apply to the Emergency Core
,Cooling System:
Normal level for Accumulator level per S1.OP-DL.ZZ-0003, Control Room Logs Modes 1-4, is 5145%.
Normal pressure is 595.5 - 647.5 psig. S I.OP-SO.SJ-O002, Accumulator Operations, P&L 3.2 instructs operators to raise level first, then pressure if both need adjustment. In this case the rise in level will cause a rise in pressure such that no additional pressure rise will be required. Distracter A is incorrect because the sequence is backwards. Distracter B is incorrect because after the rise in Accumulator level, raising pressure further is non-conservative with regards to the maximum pressure to prevent N2 injection into RCS following LOCA. Distracter D is not addressed by procedure, hence it will not be performed in that manner. Only answer C is correct.
Describe the procedures which govern the operation of the Emergency Core Cooling System, including significant prerequisites and precautions associated with each operating procedure which are required to be considered by either Licensed or Non-Licensed Ooerators L
I J
I Tuesday, May 13,2003 4:28:53 PM
~
I Paae 65 of 126
Given the following conditions:
- Salem Unit 2 is in Mode 4.
- 22 and 23 CC pumps are in service.
- 21 CC pump is C/T for repair, and has been INOPERABLE for 24 hours2.777778e-4 days <br />0.00667 hours <br />3.968254e-5 weeks <br />9.132e-6 months <br />.
22 CC pump is declared INOPERABLE due to the discovery of an unqualified bearing being installed during the last maintenance window. It is estimated that it will be at least 3 days before a new bearing will arrive on site for 22 CC pump.
Which of the following actions is required to be performed?
Within the next hour commence actions to ensure Mode 5 entry in the next 24 hours2.777778e-4 days <br />0.00667 hours <br />3.968254e-5 weeks <br />9.132e-6 months <br />.
1
'IMMEDIATELY commence a RCS cooldown to ensure Mode 5 entry in the next 6 hours6.944444e-5 days <br />0.00167 hours <br />9.920635e-6 weeks <br />2.283e-6 months <br />.
ICommence a cooldown to Mode 5 within I hour to ensure Mode 5 entry in the next 30 hours3.472222e-4 days <br />0.00833 hours <br />4.960317e-5 weeks <br />1.1415e-5 months <br />. 1 CC pump is NOT returned to OPERABLE in the next 24 hours2.777778e-4 days <br />0.00667 hours <br />3.968254e-5 weeks <br />9.132e-6 months <br />, commence a cooldown to ensure Mode 5 entry in the next 24 hours2.777778e-4 days <br />0.00667 hours <br />3.968254e-5 weeks <br />9.132e-6 months <br />.
/2,2:22__! 1 Knowledge of limiting conditions for operations and safe ?.-:-_.:..LL limits 1
I SRO ONLY 55.43(b)(2)
Tech Spec 3.7.3 requires 2 OPERABLE component cooling loops in Modes 1-4. All 3 CC pumps must
/be OPERABLE for 2 loops of CC to be OPERABLE. Therefore, when 21 CC pp is declared
'INOPERABLE, TS 3.7.3 is entered for only 1 OPERABLE loop. When the 2nd CC pump is declared
/INOPERABLE, single failure criteria directs TS 3.0.3 to be entered since there is no ACTION for 2 INOPERABLE CC loops. Answer A is correct since it states that 3.0.3 is entered, and also states the
!actions of 3.0.3.
commenced to place the unit in a MODE where the TS 3.7.3 doesn't apply (MODE 5), and also because you have 24 hours2.777778e-4 days <br />0.00667 hours <br />3.968254e-5 weeks <br />9.132e-6 months <br /> to get to MODE 5. Distracter C is incorrect because MODE 5 entry must be made within 25 hours2.893519e-4 days <br />0.00694 hours <br />4.133598e-5 weeks <br />9.5125e-6 months <br /> of discovery of 22 CC pump being INOPERABLE, not 31 hours3.587963e-4 days <br />0.00861 hours <br />5.125661e-5 weeks <br />1.17955e-5 months <br />. Distracter D is incorrect
'because there is no Provision to wait 24 hours2.777778e-4 days <br />0.00667 hours <br />3.968254e-5 weeks <br />9.132e-6 months <br /> to Perform anv action.
Distracter B is incorrect because 3.0.3 allows one hour before actions must be Salem Technical Specifications I
Water System, including:
a) b)
c) d)
The Limiting Condition(s) for Operation (N/A NEO)
The applicability of the LCO(s)
The LCO Action Statement&) (N/A NEO)
The Bases for the LCOlsl I
i I
Tuesdav. Mav 13.2003 42853 PM I Page 67 of 126
I, A series of problems resulted in a reactor trip with RCS pressure peaking at 2785 psig a few
!minutes after the trip.
- Which of the following describes the required action (IAW Tech Specs), and the bases for the
'action for this situation?
/Reduce pressure to less than limit within 1 hour1.157407e-5 days <br />2.777778e-4 hours <br />1.653439e-6 weeks <br />3.805e-7 months <br /> to prevent overpressurization of Pressurizer
/Relief Tank (PRT).
Reduce pressure to less than limit within 1 hour1.157407e-5 days <br />2.777778e-4 hours <br />1.653439e-6 weeks <br />3.805e-7 months <br /> to prevent release of radionuclides in reactor
/coolant system from reaching containment environment.
1
-~
Reduce pressure to less than limit within 5 minutes to prevent overpressurization of
/Pressurizer Relief Tank (PRT).
/Reduce pressure to less than limit within 5 minutes to prevent release of radionuclides in
/reactor coolant system from reaching containment environment.
~
I Tech Spec Safety Limit 2.1 states that the RCS pressure limit in Modes 1-5 is 2735 psig. The stem of the
~
question states that the pressure exceeded the safety limit AFTER the reactor trip, so the action in MODE 3 is to reduce the pressure to less than 2735 psig in 5 minutes. The BASES for limit 2.1....."p rotects the integrity of the Reactor Coolant System from overpressurization and thereby prevents the release of radionuclides contained in the reactor coolant from reaching the containment atmosphere." A&C are incorrect because of the time limit. and B is incorrect because of the reason.
Salem Technical Specifications I
PZRP&LEOIO I State the Technical Specifications associated with the components, parameters. and operation of the Pressurizer Pressure and 1 TECHSPE007 TECHSPEOOG Level Control system, 'including:
a) b)
c) d)
Describe the consequences of exceeding a Safety Limit Explain the te Generating Station The Limiting Condition(s) for Operation The Bases for the LCO(s) (N/A NEO)
The applicability of the LCO(s) (N/A NEO)
The LCO Action Statement@) (N/A NEO)
~
Tuesday. Mav 13.2003 4:28:53 PM I
Page 68 of 126
Which of the following identifies the correct choices to fill in the blanks of the statement below?
'higher; pressurizer; raised.
1 those predictions, use procedures to correct, control, or mitigate the consequences of those abnormal
! operation:
lx-'I
_?
dicated level will be higher than actual level because the d e fluid at Cold Cal conditions is I
ihigher than that used at Hot Cal. Therefore, for the same M pzr, it will take up more SPACE in
~
/the pzr at a higher temperature. Actual level will need to be raised to bring pressurizer level up to I program.
i 1
Control Console CC2 1
~ ____...
including:
a) b)
c) components and operation d) their intended function The Control Room location of Pressurizer and Pressurizer Relief Tank control bezels and indications The function of each Pressurizer and Pressurizer Relief Tank Control Room control and indication The effect each Pressurizer and Pressurizer Relief Tank control has upon Pressurizer and Pressurizer Relief Tank The plant conditions or permissives required for Pressurizer and Pressurizer Relief Tank Control Room controls to perform I
Tuesday, May 13,2003 42853 PM I Page 69 of 126
Tuesdav. Mav 13.2003 42853 PM 1
Page 70 of 126
Given the following conditions:
- Unit 1 is at 100% power.
- Pressurizer Level Channel 1 is selected for control.
Pressurizer Level Channel 2 is selected for alarm.
1-Pressurizer Level Channel 2 fails LOW.
1 Which of the following completes the description of the immediate plant response assuming NO operator intervention?
- Charging flow...
,does NOT change, letdown isolates, and ALL Pressurizer Heaters shut off.
does NOT change, letdown isolation does NOT occur, and ONLY Backup Pressurizer Heaters shut off.
I
'will rise to maximum, letdown
~
isolates, and ONLY Backup Pressurizer Heaters
~ shut off.
'will rise to maximum, letdown isolation does NOT occur, and ALL Pressurizer Heaters shut off. I
~
~
~-
Application 0
b17066K603
~
55!
m
[Plant Systems I
~
iK6. ].Knowledge of the of the effect of a loss or malfunction on the following will have on the Pressurizer Level Control System:
lK6.031 /Relationship between PZR level and PZR heater control circuit
/12.9//3.31
.OHA E-36, PZR HTR OFF LEVEL LO will actuate when either the Alarm OR Control channel indicates 47%. In this case, the Alarm channel fails low, letdown isolates, ALL pressurizer heaters deenergize,
~
charging flow will remain the same since it is controlled by the CONTROL channel which is operating i
correctly.
Distractor B is incorrect because letdown WILL isolate.
Distractor C is incorrect because '
charging flow will remain the same, and ALL pzr heaters are deenergized.
Distractor D is incorrect I
'because charging flow will remain the same and letdown isolates.
loverhead Annunciators Window E I
1 c_..............
b) c)
d)
Low-temperature interlock with PORVs e)
Isolation of Letdown Orifices 9
Low-level Pzr Heater Cutout P-I 1 and the Pressurizer Low-Pressure Safety Injection Block P-7 and the High Level Reactor Trip Tuesday, May 13,2003 4:28:53 PM Page 71 of 126 1
~
Tuesdav. Mav 13.2003 4:28:54 PM 1
Page72 of 126 '
Which of the following states how the SSPS 48VDC and ISVDC power supplies are provided from the Vital Instrument Busses (VIB's) to ensure that a common power supply failure does NOT result in a reactor trip?
Train A 48VDC and I SVDC supplies are powered from...
'2A and 2B VIB's; Train B from 2C and 2D VIB's.
/2B and 2C VIB's: Train B from 2B and 2D VIB's.
1 12A and 2D VIB's: Train B from 2B and 2C VIB's.
/2A and 2B VIB's; Train B from 2A and 2C VIB's.
dge of bus power supplies to the following:
channels, components, and interconnections
'2B and 2D Vital Instrument Buses (VIB) have a common power source through 2B 230 VAC Vital Bus.
Id is correct because two of the four 115 V VIB supply each train of 15 and 48 VDC supplies and in order :
- to Drevent loss of a VIB. B and D VIB do not SUDD~V the same train.
I I A, B; C, and D i t a 1 Bus Essential Control Power System Essential Lighting Power System Unit 2 RMS Power System SPDS Power System Control Rod Control and Indication Power Systems Security Power System Telecommunications Power System RXPROTEOI 1 1 State the power SUDD~V to the SSPS Tuesday, May 13, 2003 4:28:54 PM 1
Page 73 of 126
Given the following conditions:
- Salem Unit 2 has experienced a Large-Break Loss of Coolant Accident (LOCA).
- RCS pressure is 40 psig.
- All ECCS pumps are running in Injection mode.
RWST Level is 14.9'.
IAW 2-EOP-LOCA-3, TRANSFER TO COLD LEG RECIRCULATION, which of the following conditions will prevent transferring to Cold Leg Recirc?
RWST level less than 15.2'.
I
~
IContainment sump level of 50%.
1
/Two (2) Service Water pumps in service.
-1
~
'One
- (1) Component Cooling Water pump in service.
~
Comprehension c___-__
~
~
!Al. ];Ability to predict and/or monitor changes in parameters associated with operating the Engineered Safety
/Features Actuation System controls including:
i c, iAl.08
. [Containment sump level
, 3.7ij34 lSRO ONLY55.43(b)(5) i Upon entrance to EOP-LOCA-3, at 15.2' RWST level, the first step asks is containment sump level is >
i
/62Y
- 0. If the answer is no, you immediately kick out EOP-LOCA-5, LOSS OF EMERGENCY COOLANT REC CIRCULATION. If sump level is inadequate (~62%)
the risk is present to cavitate or air bind RHR
'pps. Distractor A is incorrect in that 15.2' is the transition point to go to EOP-LOCA-3. Answer B is correct since there needs to be 62% level in the containment sump to transfer to cold leg recirc.
'Distractor C is incorrect since 2 SW pps would not preclude establishing cold leg recirc, it would,
- however, force to you to align as single loop recirc later in procedure. Distractor D is incorrect for the j
Tuesdav. Mav 13.2003 4:28:54 PM CPaSe 74 of 126
- An entry condition for S2.0P-AB.ROD-0004, ROD POSITION INDICATOR FAILURE, is "One or more Group Demand Counters (GDC) indicating +/-2 steps from Slave Cycler output".
'Which of the following describes how an operator will determine this condition exists?
c-~--
I---_-
~
KSr__];Knowledge of the operational implications of the following concepts as they apply to the Rod Position
- Indication System
OHA E-24 ROD DEV OR SEQ alarm actuated.
Plant Computer point Rod Position Deviation is displayed.
Compare the P/A Converter indication to GDC indication.
- Deviations can only be determined by voltmeter readings from the Rod Position Indication
'cabinets.
[--?
j014000K502 Rod Position Indication System J
58' 1
[Systems!
1 ABROD4E002 I Describe. in general terms. the actions taken in S2.OP-AB.ROD-O004(Q). in accordance with the Technical Bases Document.
I Tuesday, May 13, 2003 4:28:54 PM 1
I Paae 75 of 126
/Which of the following identifies the parameters that must be satisfied in order to transition from 2-
[EOP-FRSM-I, RESPONSE TO NUCLEAR POWER GENERATION?
'NO more than two control rods failed to insert.
P
/
FRSMOOE002
- The Cold Shutdown Shutdown Margin value is achieved.
/Either Reactor Trip Breaker and its associated Bypass Trip Breaker is open.
Describe the EOP mitigation strategy for a:
A.
B.
RESPONSE TO NUCLEAR POWER GENERATION.
RESPONSE TO A LOSS OF CORE SHUTDOWN
~~
Three Power Range Nuclear Instrumentation channels less than 5% and Intermediate Range Start UD Rate neaative.
' 12.4-L /Emergency i
Procedures / Plan ledge symptom based EO RO ONLY 55.43(b)(5) is correct in accordance with ERG. A. is incorrect although it is a criteria for Rapid Boration. B. is
[incorrect, it is not a transition point from FRSM. C. is incorrect, it does not mean the Reactor has isuccessf ullv t rimed.
IRESPONSE TO NUCLEAR POWER GENERATION I
Tuesday, May 13,2003 4:28:54 PM 1
Page 76 of 126
'Intermediate Range (IR) compensating voltage fails LOW on one of the IR detectors. The Reactor
/subsequently trips due to other causes, but the IR current on the failed detector does NOT go
- below 5.OE-5 amps.
Which of the following describes how the source range instruments will be energized as reactor power DECREASES below 7.OE-1 I amps?
1
/P-6 will be unblocked and the source range detectors will automatically unblock.
/The failed IR detector will be bypassed allowing the source range detectors to
~
energize.
~
I 1 I
'The source range manual reset pushbuttons will be used to manually reenergize the source
- range detectors.
'One source range detector will automatically reenergize and the other will be manually ireenergized i
using the reset pushbutton.
-I Isatem I
& 2 El
'K6. IiKnowledge of the of the effect of a loss or malfunction on the following will have on the Nuclear
/Instrumentation System:
'range detectors will not occur. Distractor A is incorrect because source range detectors will not automatically reenergize. Distractor B is incorrect because the procedure does not direct bypassing IR
.detector. Distractor D is incorrect because both SR detectors need 2/2 IR logic to automatically reenergize. 2-EOP-TRIP-2 directs manually energizing source range (step 22.1) if undercompensation iof IR channel (s) is preventing SR automatic reenergization.
~
i
/Reactor Trip Response I
EXCOREE007 with the Excore Nuclear Instrumentation System components:
A.
C 1, Intermediate Range High Flux Rod Stop B.
C 2, Power Range High Flux Rod Stop C.
P 6, Source Range Block D. P 8, Enable 1/4 Loss of RCS Flow Trip E.
P 9, Reactor Trip / Turbine Trip At Power Trips ldentify and describe the Control Room controls, indications, and alarms associated with the Excore Nuclear Instrumentation System, including:
A.
The Control Room location of Excore Nuclear Instrumentation System control bezels and indications B.
The function of each Excore Nuclear Instrumentation System Control Room control and indication C.
The effect each Excore Nuclear Instrumentation System control has upon Excore Nuclear Instrumentation System D. The plant conditions or permissives required for Excore Nuclear Instrumentation System Control Room controls to components and operation perform their intended function the E Tuesday, May 13,2003 4:28 54 PM I Page 77 of 126
Tuesday, May 13, 2003 4:28:54 PM L
I Page 78 of 126
/Which of the following identifies the parameter that sets the High Steam Flow Safety Injection
- setpoint, and the logic for the SI actuation?
!Main Turbine first stage pressure instruments PT-505 and PT-506; 1/2 channels high steam
!flow on 1/4 main steam lines coincident with low steam line pressure or LO-LO Tave.
'Main Turbine first stage pressure instruments PT-505 and PT-506; 1/2 channels high steam
/flow on 214 main steam lines coincident with low steam line pressure or LO-LO Tave.
Main steam header pressure instrument PT-507; 112 channels high steam flow on 2/4 main steam lines coincident with low steam line Dressure or LO-LO Tave.
/Main steam header pressure instrument PT-507; 1/2 channels high steam flow on 1/4 main
/steam lines coincident with low steam line pressure or LO-LO Tave.
K1. -- ] Knowledge of the physical connections and/or cause-effect relationships between Non-Nuclear Instrumentation System and the following:
K1.07 ] [ECCS
@Tfi 13.71 1
The correct actuation signal is 1/2 high steam flow channels on 2/4 steam lines coincident with Lo-Lo Tave (543 deg.) on 2/4 channels or low steam pressure (600 psig) on 2/4 channels. The steam flow
~
setpoint is ramped from 40%-I 10% steam flow based on PT-505 (CH I) and PT-506 (CH II) pressure from 20%-100% first stage pressure. All the distracters have incorrect setpoint deriviatives or wrong 1
coincidence I
~~
- 7 ISalem Tech SDecs I
~UFSAR I
I Tuesday, May 13,2003 4:28:54 PM I Page 79 of 126
Given the following conditions:
CONTMTEOOI
- Salem Unit 2 is operating at 100% power.
- 21 Containment Spray pump has been declared INOPERABLE due to an oil leak.
- 24 and 25 CFCU's are INOPERABLE and isolated due to 22 SW Accumulator being
~
N
~
~
~
~
~
~
~
s equipment is OPERABLE.
/With the plant in this configuration, which of the following describes if the plant being operated
!within the Design Basis for containment cooling, and the BASES for your answer?
1
/No, two (2) Containment Spray pumps and five (5) CFCU's are required to be OPERABLE to
[meet the design basis for containment cooling.
State the purpose and design bases for the following Containment And Containment Support Systems subsystems:
al Containment
~
/No, one containment Spray pump and four (4) CFCU's are required to be OPERABLE to meet lthe design basis for containment cooling.
~
~
'Yes, a single OPERABLE Containment Spray pump meets the design basis for containment i lcooling
~
~~
- es, one (1) OPERABLE Containment Spray pump combined with three (3) OPERABLE CFCU's meets the design basis for containment cooling.
~
- y A
d Application 1
m 1
[Plant Systems
~
Containment Spray System
,K3,W.._JiKnowledge of the effect that a loss or malfunction of the Containment Spray System will have on the jfollowing:
~~
IK3.01 ICCS
/I 3.911 4.11 iSRO ONLY 55.43(b)(2)
!As per the UFSAR, "Any of the following combinations of containment spray and fan cooler equipment
!trains will provide sufficient heat removal capability to maintain the post-accident containment pressure below the design value, assuming that the core residual heat is released to the containment as steam:
- I.
All five containment fan coolers.
- containment fan coolers and one containment spray pump along with one train of the Emergency Core
'cooling System (ECCS)"
- 2. Both containment spray pumps
- 3. Three of the five IUpdated Final Safety Analysis Report
[Salem
...... Technical Specifications I
Containment Airlocks Containment Isolation System Containment Fan Cooler System Containment Iodine Removal System Rod Drive Ventilation System Reactor Nozzle Support Ventilation System Reactor Shield Ventilation System Containment Pressure - Vacuum Relief System j)
Hydrogen Recombiner System I- ___ - ___.
Tuesday, May 13,2003 4:28:54 PM 1
Page 80 of 126
Tuesdav. Mav 13.2003 4:28:54 PM I Pase 81 of 126
- Which of the following choices identifies a condition that will prevent starting of 21 Hydrogen
- Recombiner?
f# iOHA C-23, CNTMT H2 LVL HI annunciator in alarm.
/Hot Shutdown Panel 213, 21 H2 Recombiner Control Switch in OFF.
iPOWER OUT SWITCH red light illuminated.
1
'A4.03
~-
1 21 Recombiner Control Switch on 2RP5 in OFF.
Location and operation of hydrogen sampling and analysis of containment atmosphere, including alarms and indications
\\
53.31 Salem I 8 2 J
05/21/2004 id.. I Memory.........
[::I CONTMTEOl2
\\028000A403 1
63;
~_______I_._.___..___..__.._...__.__..._.___
Describe the procedures which govern the operation of the Containment And Containment Support Systems, including significant prerequisites and precautions associated with each operating procedure which are required to be considered by either Licensed or Non-Licensed Ooerators L - ___________ ____ - _____-----
Distractor A is an alarm function only. It alarms at a containment H2 level of 2%. Procedurally, when
/above 4% the recombiner will NOT be placed in service, but between 24% it will be.
Distractor B is incorrect because there are no controls or indications for H2 recombiners at Panel 213.
incorrect because the light will be illuminated when the switch is ON and power applied, which is an Distractor C is I
indication the recombiner has control power available.
ydrogen Recombiner Operation
- CONTMTE007 Identify and describe the local controls, alarms, and indications associated with the Containment And Containment Support CONTMTEOOB Systems, including:
a) b)
c) their intended function d)
Identify and describe the Control Room controls, alarms, and indications associated with the Containment And Containment Support Systems, including:
a) b)
c)
Systems components and operation d) perform their intended function e)
The location of Containment And Containment Support Systems local controls and indications The function of Containment And Containment Support Systems local controls and indications The plant conditions or permissives required for Containment And Containment Support Systems local controls to perform The setpoints associated with Containment and Support Systems local alarms The Control Room location of Containment And Containment Support Systems control bezels and indications The function of each Containment And Containment Support Systems Control Room control and indication The effect each Containment And Containment Support Systems control has upon Containment And Containment Support The plant conditions or permissives required for Containment And Containment Support Systems Control Room controls to The setpoints associated with the Containment and Support Systems Control Room alarms
~ ____........
~ --.
.. ~ - - ~
--j Tuesday, May 13, 2003 4:28:55 PM I Page 82 of 126
'Given the following conditions:
1-Salem Unit 2 is in Mode 5 entering a refueling outage.
I-New fuel is currently being placed into the Spent Fuel Pool.
1-Full Containment Purge is in service.
!- Fuel Handling Building (FHB) DIP LOWAlarm is received.
- Which of the following choices identifies the action that will allow fuel movement in the FHB to
/continue?
1 i
!Start another FHB exhaust fan to raise DIP in the FHB.
I Select Inlet Vane Open on the FHB Supply fan to raise backpressure in the FHB.
1
~
lPlace Modified Containment Purge in service to reduce backpressure in the FHB.
I
- Ensure 2ABV20 red light for DIVERSION DAMPER PURGE illuminated.
'A4.
_________ J Ability to manually operate and/or monitor in the control room:
A4.01 kontainment purge flow rate 55.43(b)7 S2.OP-WG-0006, CONTAINMENT PURGE TO PLANT VENT, P&L 3.1 1 identifies that full containment purge flow can result in FHB low D/P alarms due to backpressure. The
'stem of the question states that full containment purge is in service, vs. Modified containment purge with reduced flow. Distractor a is incorrect because in order to be moving fuel or loads over the spent fuel pool, FHB ventilation must be OPERABLE IAW TS 3.9.12, which requires ALL FHB ventilation fans to be in service. So there are no more fans to start. Distractor b is incorrect because the Inlet Guide Vanes are opened on the exhaust fans to attempt to raise FHB DIP, not the supply fans. (Supply fan does not have adjustable guide vane.) C is the correct answer because placing Modified Containment Purge in service will reduce purge flow low enough that it will not affect FHB backpressure. Distractor d is
'incorrect because during normal purge, the red light is illuminated, but when placing modified purge in service it will be shut (green).
I.
Containment Purge to Plant Vent I
J I
I a) b)
c)
Systems components and operation d) perform their intended function The Control Room location of Containment And Containment Support Systems control bezels and indications The function of each Containment And Containment Support Systems Control Room control and indication The effect each Containment And Containment Support Systems control has upon Containment And Containment Support The plant conditions or permissives required for Containment And Containment Support Systems Control Room controls to Tuesday, May 13,2003 42855 PM 1
Page 83 of 126
I Tuesdav. Mav 13.2003 4:28:55 PM 1 Page84of126
,Given the following condition:
1-22 Spent Fuel Pool (SFP) Pump is running providing SFP cooling.
iWhich of the following describes when 22 SFP Pump will lose suction if a leak develops in the i tpump suction line?
1
~
IWhen the spent fuel pool lowers to four feet below normal.
~
/Any i running pump trips when SFP LVL LO alarm (OHA C-35) actuates.
~
I 1
b h e n the anti-siphon hole in the return pipe uncovers.
1 Fhe fuel will uncover before 22 SFP loses suction on a suction line break.
- The spent fuel suction is located - 4 feet below normal SFP level to prevent gravity draining on the pool.
ib is incorrect because there are no automatic functions associated with the SFP level lo OHA. c is
!incorrect because the anti-syphon hole is on the return line, not the suction line. d is incorrect because ithe suction line is -20 feet above the fuel.
i I
l
]Spent Fuel Cooling Drawing E-::
I SFPOOOE003 I Explain the flowpaths on a one-line diagram of the Spent Fuel Pool Cooling System similar to AV 3416:
a)
Major Spent Fuel Pool Cooling System Components i) Spent Fuel Pool ii) Spent Fuel Cooling Pumps iii) Refueling Water Purification Pump iv) Spent Fuel Pool Skimmer Pump v)
Spent Fuel Pool Heat Exchanger vi)
Spent Fuel Pool Demineralizer vii) Spent Fuel Pool Filter viii) Spent Fuel Pool Skimmer Filter ix) Refueling Water Purification Filter x.)
Spent Fuel Pool Strainer xi)
Spent Fuel Pool Skimmer Strainer Major Spent Fuel Pool Cooling System Flowpaths i) Pool cooling flowpath ii) Pool filtration flowpath iii) Pool cleanup flowpath iv) RWST filtration flowpath v)
RWST cleanup flowpath vi) Refueling Canal filtration flowpath vii) Refueling Canal cleanup flowpath b)
Tuesday, May 13,2003 4:28:55 PM 1
Page 85 of 126
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TuesdavMav 13.2003 4:28:55 PM 1
Pase 86 of 126
Given the following conditions:
i-Salem Unit 2 is in Mode 6.
1-A fuel assembly has dropped onto the spent fuel racks as it was being transported from the upender to its destination in the spent fuel pit.
- The dropped fuel assembly has visible damage, and bubbles are rising from the assembly.
- Radiation levels in the Fuel Handling Building have risen to 1.2 R/hr.
1 1
I ABFUELOI EO0 i
3 Which of the followina identifies an action which must be performed?
I Given a set of initial plant conditions:
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.
121 and 22 Iodine Removal Units.
I 1
I Enter S2.0P-AB.FUEL-0001, FUEL HANDLING INCIDENT, press HEPA 22 PLUS CHAR
~ i pushbutton on Fuel Handling Building Ventilation.
- Enter S2.OP-AB.FUEL-0002, LOSS OF REFUELING CAVITY OR SPENT FUEL LEVEL, lsound the High Radiation Alert alarm.
~
~~
~
~
/Enter S2.OP-AB.FUEL-0001, FUEL HANDLING INCIDENT, initiate makeup to the Spent Fuel 1
'Pit from the RWST.
I I
.A2. ] Ability to (a) predict the impacts of the following on the Fuel Handling Equipment System and (b) based on
,those predictions, use procedures to correct, control, or mitigate the consequences of those abnormal
'operation:
~
w w11'4.41
[SRO ONLY5543(b)(4),(7)
/IAW AB.Fue1-001, depress the FHB ventialtion bezel pushbutton to realign the exhaust flow through the
/charcoal filter. Distractor A is incorrect because IRU will be started if accident is in containment, and
'wrong procedure is entered Distractor C is incorrect because the wrong procedure is entered, and there lis no requirement to sound the High Radiation Alert. Distractor D is incorrect because makeup is only lreauired in AB.FUEL-2. not AB.FUEL-1.
~
Tuesdav. Mav 13.2003 4:28:55 PM I
Page 87 of 126
Given the following conditions:
- Salem Unit 2 Reactor power - 25%, rising.
RCS Tave - 549 O F, lowering.
!- - Pressurizer pressure - 21 50 psig, lowering.
!- Pressurizer level - 22%, lowering.
,- Turbine load is stable.
- S/G pressures are - 950 psig (21S/G); 890 psig (22S/G); 950 psig (23S/G); 950 psig (24S/G), all lowering.
- MS1 Os, atmospheric relief valves, indicate closed.
- Steam Dumps indicate closed.
'- Sound heard in the control room indicates that a Main Steam Safety Valve may be open.
Which of the following describes the action to be taken IAW S2.OP-AB.STM-0001 for the above iconditions?
Containment temperature and pressure are normal.
[Immediately close 22MS167 and initiate a rapid unit shutdown.
/Immediately initiate Main Steamline Isolation (22 loop only); initiate a reactor trip; initiate IMANUAL SI.
/Trip the reactor; close 22MS167; initiate MANUAL SI (if necessary).
I
!Trip the reactor; initiate Main Steamline Isolation (all loops); initiate MANUAL SI.
,K3. I Knowledge of the effect that a loss or malfunction of the Steam Generator System will have on the following:
I steam leak, then a manual SI is required to prevent an automatic SI on low pressurizer pressure or Steam Line D/P. The distractor are wrong because the reactor must first be tripped prior to isolating the steamlines to identifv the leak, and the steamline isolation is on ALL loops, not iust the one you think it is.
Excessive Steam Flow I
ABSTM 1 E004 significant prerequisites and precautions associated with each operating procedure which are required to be considered by either Licensed or Non-Licensed Operators a) Determine the appropriate abnormal procedure.
b) Describe the plant response to actions taken in the abnormal procedure.
I c) Describe the final plant condition that is established by the abnormal procedure.
Tuesday, May 13, 2003 42855 PM 1
Page 88 of 126
Direct From Source 1
Tuesday, May 13, 2003 4:28:55 PM 1
Page 89 of 126
!Given the following conditions:
- Salem Unit 2 is operating at 12% power.
I-Main turbine is rolling up to normal speed.
,If Main steam dump AUTO setpoint is adjusted down to 940 psig, what effect will this have on Tave
'and Reactor power assuming NO other operator action?
Control Bank D rods are at 152 steps withdrawn.
Main steam dumps are set for 950 psig, in MS PRESS CONTROL-AUTO.
i_
1
~
iGVe will lower. RX power will lower.
kave L
will lower,
~
RX power will rise.
-I 1
lTave will rise, Rx power will lower.
1 iTave will rise, Rx power will rise.
Isteam Dump system Operation 1
I STDUMPEOI 1 Identify and describe the Control Room controls, indications, and alarms associated with the Steam Dump System, including:
a) b)
c) d)
function The Control Room location of Steam Dump System control bezels and indications The function of each Steam Dump System Control Room control and indication The effect each Steam Dump System control has upon Steam Dump System components and operation The plant conditions or permissives required for Steam Dump System Control Room controls to perform their intended
-7
'-1 Tuesday, May 13, 2003 4:28:55 PM I Page 90 of 126
iGiven the following conditions:
1 I
of the following describes how turbine load is affected and how the EHC controls will respond?
bemains constant. When the difference between REFERENCE and the input signal exceeds
/the setpoint, then operator ensures EHC transfers to IMP OUT.
[Increases until the difference between REFERENCE and the input signal exceeds the lsetpoint. An alarm then alerts the operator to select EHC IMP OUT.
1 IRemains constant. When the difference between REFERENCE and the input signal exceeds Ithe setaoint. an alarm then alerts the oaerator to select EHC MANUAL control.
ilMP OUT.
LONTROL CONSOLE 2cc3 ALARM RESPONSE 1
L.
J I
I I
L-EHCOOOE007 Identify and including:
a) b)
c) intended function.
d)
The location of Electric-Hydraulic Control (EHC) System local controls and indications.
The function of Electric-Hydraulic Control (EHC) System local controls and indications.
The plant conditions or permissives required for Electric-Hydraulic Control (EHC) System local controls to perform their The setpoints associated with the Electric-Hydraulic Control (EHC) System local alarms 1
I
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Tuesday, May 13,2003 4:28:55 PM 1
Paae 91 of 126
Tuesday, May 13,2003 4:28:56 PM I Page 92 of 126
Given the following conditions:
i ABCNOI E004
- Unit 2 is at 100% power with all major control systems in AUTO.
- 22 Heater Drain Pump is 00s for breaker replacement.
Given a set of initial plant conditions:
a) b)
_ c)
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.
Using the attached documents, which of the following specifies the REQUIRED action if 21 I
!Condensate Pump trips? Assume SGFP suction pressure remains above the trip setpoint.
1 Reduce reactor power to less than or equal to 75% at less than or equal to 5% / minute.
1 Reduce reactor Dower to less than or eaual to 80% at less than or eaual to 5% / minute.
Application 1
~
A2. ] Ability to (a) predict the impacts of the following on the Condensate System and (b) based on those predictions, use procedures to correct, control, or mitigate the consequences of those abnormal operation:
Pumps and I HD Pump; B. Reduction for 310 or 2/3; D. Reduction for 2/1 TierlGrou p 211 ilOCFR55 141.5
'Cognitive Level:
'Application
References:
LP NOS05ABCN01-01, Obi 4c S2.OP-AB.CN-0001 (Main Feedwater I Condensate System Abnormality I
J 1
Tuesday, May 13,2003 4:28:56 PM I Page 93 of 126
,Given the following conditions:
- Salem Unit 2 is operating at 85 % power.
- 21 Condensate pump CTT.
PT-506, Turbine Impulse Pressure Channel II is O/S.
- Main condenser backpressure is 2.1"Hg.
22 SGFP speed starts acting erratically, and quickly degrades to the point where the crew manually trips 22 SGFP.
!What other action is IMMEDIATELY required of the crew?
1
- Trip the Reactor, and GO TO 2-EOP-TRIP-1.
1 r
,Verify Automatic Turbine Runback has or is occurring.
!Ensure 1
Rod Bank Selector Switch is in AUTO.
/Initiate Main Turbine runback at less than or equal to 15%/ minute to less than 352 psig turbine
,first stage pressure.
i I
I iSRO ONLY 55.43(b)(5)
I
/Immediate action in S2.OP-AB.CN-0001 2.2, if either PT505 Or 506 is out of service, with turbine power
/greater than 70% and a loss of a single SGFP has occurred, then trip RX and go to trip-I. Distractors
!are all actions that would be performed if the SGFP tripped and the PT-506 was in service.
i J
1
[Main Feedwaterlcondensate System Abnormality i
I
' ABCNO1 E002 I State the immediate actions of S2.OP-AB.CN-0001.
a) b)
Determine the appropriate abnormal procedure.
Describe the plant response to actions taken in the abnormal procedure.
r- -_-.. -_.
- I Tuesdav. Mav 13.2003 4:28:56 PM 1
Paae 94 of 126
Given the following conditions:
.- Salem Unit 1 is operating at 85.0 % power.
!- During a manual bus swap prior to clearing and tagging a Station Power Transformer, the 1A
'4KV vital bus is inadvertently deenergized, and the SEC loads 1A bus in Mode 2*.
i-All other electrical bus transfers expected to occur from the loss of I A 4KV vital bus are
[successfu I.
/With NO operator action, which of the following identifies the plant condition 5 minutes after the
!initial loss of 1A 4KV vital bus?
'Reactor power is > 85.0%.
,The Reactor.............................. is tripped.
I ARNOOOE005
~
!The i
___ Main Turbine will have run back to 60%.
I State the power supply to the following Auxiliary Feedwater System components:
a)
Motor-Driven AFW Pumps
-I
~
peactor power has remained at 85.0%.
~_ ~~~~
061000K202 1
i I
72' I
!cause a positive reactivity addition due to temperature of the feedwater entering the SIGs lowering.
- Reactor power will rise to some value greater than 85.0%.
Distractor B is incorrect because with all 1
,other bus transfers occurring correctly as stated in stem, no reactor trip would occur. Distractor C is incorrect because no runback will occur as long as all bus transfers occur. Distractor D is incorrect 1
because reactor power will rise due to the cold feedwater addition from 11 AFP starting and injecting into I (Loss of 1A 4W Vital Bus I 1 Unit 4160V Vital bus one line diagram
... 1 L
1
--j Tuesday, May 13,2003 4:28:56 PM 1 Paqe 95 of 126
The Auxiliary Feed System is designed so that a minimum of AFW pump(s) can sufficiently remove decay heat and cooldown the RCS at 100% power.
deg/hr following a Reactor trip from
/I:
- 50.
12; 50.
1
/I; 100.
12; 100.
y heat sources and magnitude icooldown capability when necessary. Each pump has the capacity to remove heat from the steam
'generators at a sufficient rate to prevent over-pressurization of the RCS and to maintain steam generator
.levels to prevent thermal cycling. Once the normal steam generator level is re-established the AFW
- system can cool down the RCS at a rate of 50 deg/hr."
FSAR figures 15.1-6 and 15.4-47 identify fission
- product decay as the largest source of decay heat. Prompt neutrons will drop immediately upon a
'Reactor Tri~. Onlv 1 AFP is reauired.
/Salem UFSAR I
L i
I MOPERE024 1 Define decay heat.
Tuesdav. Mav 13.2003 4:28:56 PM 1
Page 96 of 126,
/I25 VDC breaker 2BDClAX12,2G 4KV Bus Control Power Supply (Reg) tripped due to a breaker malfunction.
DCELECEO07.
Which of the following describes the impact this malfunction will have?
Only motor amperage indication will be lost for 24 RCP.
124 RCP will trip immediately.
24 RCP will remain running but will NOT automatically trip if required.
Identify and describe the local controls, indications, and alarms associated with the DC Electrical Systems, including:
A.
The location of DC Electrical Systems local controls and indications B. The function of DC Electrical Systems local controls and indications C.
The plant conditions or permissives required for DC Electrical Systems local controls to perform their intended function D. The setpoints associated with the DC Electrical Systems local alarms I !
Emergency control power from the 2A 125 VDC bus will automaticallv be provided.
'063000K401 1
1
__ K4.
] Knowledge of D.C. Electrical Distribution design feature@) and or interlock(s) which provide for the
'following:
- automatic trips of any breakers will be possible, due to no power available to the trip coil. Transfer of the 1
,bus control power supply will be a manual evolution under procedure. Only manual trips of breakers will
- be Dossible.
~
J 14&24 Reactor coolant pumps and lift pumps schematic 1
i _______--
Tuesday, May 13, 2003 4:28:56 PM I
Page 97 of 126,
Given the following conditions:
- Unit 1 is operating at 100% power.
- I A EDG starting air receiver 11A is C/T.
- I 1A EDG starting air compressor is CTT.
,- A Loss of Off-Site Power occurs.
1-I A EDG starts and trips for an unknown reason.
!- I I B starting air compressor will NOT run.
!There is sufficient air remaining for how many more engine starts?
~
- 2
-. ~.
~
i4 1
/064000A304 75'
&3,___] Ability to monitor automatic operations of the Emergency Diesel Generators including:
_ _ _ ~
ISalem UFSAR I
~
I I
EDGOOOE002 1 Describe the design bases of the Emergency Diesel Generators r
Tuesday, May 13, 2003 4:28:56 PM 1
I Page 98 of 126
,Given the following conditions:
-1 I
- - Salem Unit 2 is operating at 100% power.
- 2C EDG is declared INOPERABLE at 0700 on May 1st.
'Which of the following describes the actions required to be taken IAW Tech Specs, assuming 2C iEDG remains INOPERABLE, and NO other equipment is INOPERABLE?
IRestore 2C EDG to OPERABLE status by 0700 on May 4th, or be in HOT STANDBY by 1300 ion May 4th.
Perform S2.OP-ST.500-0001, ELECTRICAL POWER SYSTEM AC SOURCE ALIGNMENT --I within 1 hour1.157407e-5 days <br />2.777778e-4 hours <br />1.653439e-6 weeks <br />3.805e-7 months <br /> and once per 24 hours2.777778e-4 days <br />0.00667 hours <br />3.968254e-5 weeks <br />9.132e-6 months <br /> after that.
I
~
- Within 4 hours4.62963e-5 days <br />0.00111 hours <br />6.613757e-6 weeks <br />1.522e-6 months <br />, declare 22 CS pp, 22 CVCS pp, and 22 SI pp INOPERABLE, then enter TSAS 13.5.2 for 1 INOPERABLE ECCS subsvstem.
1
~
- Be in HOT STANDBY by 1700 May Ist, and in COLD SHUTDOWN bv 2300 on Mav 1st.
'064 I \\Emergency Diesel Generators
~
-. 2.2
]/Equipment Control
'2.2.22 i IKnowledge of limiting conditions for oDerations and safetv limits.
/SRO ONLY 55.43(b)(2) 1 Distractor B is incorrect because ST-500-1 must be performed every 8 hours9.259259e-5 days <br />0.00222 hours <br />1.322751e-5 weeks <br />3.044e-6 months <br />, not 24. Answer A is correct because TS 3.8.1.I
.b action b4 states to restore the EDG to OPERABLE within 72 hours8.333333e-4 days <br />0.02 hours <br />1.190476e-4 weeks <br />2.7396e-5 months <br />, or HSB within the next 6 and CSD within the next 30. Distractor C is incorrect because as long as all other equipment is OPERABLE, there is no need to declare any other equipment INOPERABLE because of the
- EDG being INOPERABLE. Distractor D is incorrect because it is the action for 2 INOPERABLE EDG's.
I alem Tech Specs 1s.
~
~
.J I
a) b)
c) d)
The Limiting Condition@) for Operation (N/A NEO)
The applicability of the LCO(s)
The LCO Action Statement@) (N/A NEO)
The Bases for the LCO(s)
I Tuesday, May 13,2003 4:28:56 PM I
Page 99 of 126
... ~-
'Given the following conditions:
i 1-1 WMHUT is in recirc, a sample has been drawn and is in the process of being analyzed.
!- The RWO mistakenly places I WMHUT in service.
j One hour later, the RWO realizes hidher error, and returns 1 WMHUT to recirc.
/What effect, if any, will this have on the release preparations for 1 WMHUT?
!A 1
new sample must be drawn, with
...... NO
- minimum required recirculation.
1_
~
~
/The release preparations may continue as long as volume added to tank does NOT exceed 1
11% of total tank volume ONLY.
I
~
IThe release preparations may continue as long as volume added to tank does NOT exceed 1
11% of total tank volume AND double verification of sample analysis is performed.
I khe tank will require further recirculation and resampling prior to release.
I
- further recirculation time and resampling" i
/Release of Radioactive Liquid Waste
~ _ _
~
1
~
1 1..
~..........................
I I
I Tuesday, May 13,2003 4:28:56 PM I Page 100 of 126
/Given the following conditions:
~- A Unit 2 Pressurizer Safety Valve has been leaking.
' Procedural actions being taken to prevent overpressurizing the PRT are generating liquid waste.
!Assuming the containment isolation valves are open, which of the following describes the flow path
/of the water after the operator opens 2PR14, PRT Drain valve?
1
!The PRT gravity drains to the in-service CVC HUT.
~
c______.-_________-_______________________.___.
The PRT gravity drains to the RCDT. The RCDT pumps automatically cycle on RCDT level to the in-service CVC HUT.
I
'The
~
RCDT Pump in AUTO cycles to control PRT level --
whenever 2PR14 is open.
lThe RCDT pumps start on interlock with 2PR14, directing flow to the in-service CVC HUT. 1
/K1.
Knowledge of the physical connections and/or cause-effect relationships between Liquid Radwaste System and the following:
incorrect, no gravity drain. B. is incorrect the RCDT pumps do not automatically cycle on level. C. is incorrect the pump does not control PRT level.
12 Unit Waste Disposal Liauid P&ID c-::.......
I WASLIQE009 WASLIQEOOB State the setpoints for automatic actuations associated with the Radioactive Liquid Waste System Identify and describe the Control Room controls, indications, and alarms associated with the Radioactive Liquid Waste System, including:
a) b)
c) d)
intended function e)
The Control Room location of Radioactive Liquid Waste System control bezels and indications The function of each Radioactive Liquid Waste System Control Room control and indication The effect each Radioactive Liquid Waste System control has upon system components and operation The plant conditions or permissives required for Radioactive Liquid Waste System Control Room controls to perform their The setpoints associated with the Radioactive Liquid Waste System Control Room alarms
~-.
TuesdaT May 13.2003 4:28:57 PM I Page 101 of 126
While performing a review of 21 GDT release paperwork, S2.OP-SO.WG-0008 the CRS notices that the calculated maximum release rate is 28 scfm.
'Which of the following statements is correct regarding the GDT release?
I The GDT release CANNOT be initiated because release flow rates c 32 scfm will result in off-site dose exceeding 500 mrem/yr total body.
The GDT release CANNOT be initiated because a minimum calculated flow rate of 32 scfm is required.
1
,The GDT release CAN be initiated if a double, independent verification of the release rate has
/been performed.
1
~
!The GDT release CAN be initiated with NO abnormal restrictions.
!AI. Ibbility to predict and/or monitor changes in parameters associated with operating the Waste Gas Disposal bystem controls including:
Discharge of 22 Gas Decay Tank to the plant vent
~~~~-
r----
and precautions associated with each operating procedure which are required to be considered by either Licensed or Non-i Tuesdav. Mav 13.2003 4:28:57 PM j Page 102 of 126
!Which of the following identifies the Radioactive Gaseous Waste Gas Decay Tank (GDT) release 1
ipath?
'To the Fuel Handling Building exhaust ventilation header, then to the Plant Vent release point
'at the top of containment.
- To the Fuel Handling Building exhaust ventilation header, then to the Plant Vent release point at the top of the Auxiliary Building.
To the Auxiliary Building exhaust ventilation header to the Plant Vent release point at the top of containment.
I To the Auxiliary Building exhaust ventilation header to the Plant Vent release point at the top of the Auxiliary Building.
~
Waste Disposal-Gas Auxiliary Building -Ventilation r---
Draw a one-line diagram of the Radioactive Waste Gas System similar to TP-2 which indicates the following:
a)
Major Radioactive Waste Gas System Components i) Vent Header (Waste Gas Compressor suction header), including Pressure Regulating Valve 2WG8 and Vent Header inputs ii) Waste Gas Compressors, including Separator Tanks iii) Waste Gas Compressor Separator Backpressure Control Valves(21-22WG25) iv) Waste Gas Compressor Internal Pressure Control Valves (21-22WG21) v) Waste Gas Compressor Bypass Valves (21 -22WG22) vi) Waste Gas Compressor Separator Level Control Valves (21-22WG15 and 21-22WG13) vii) Waste Gas Decay Tanks viii) Waste Gas Decay Tank Inlet Valves (21-24WG29) ix) Waste Gas Dewy Tank Sample Isolation Valves (21-24WG42) x)
CVCS Holdup Tank Pressure Regulating Valve 2WG36 xi) Radioactive Gaseous Waste Release Valve 2WG41 xii) Radioactive Gaseous Waste Release Pressure Control Valve 2WG38 Major Radioactive Waste Gas System Flowpaths i) Waste Gases into the Radioactive Waste Gas System, through the Compressors, and into the Decay Tanks ii) Waste Gas release flowpath from the Waste Gas Decay Tanks to the Plant Vent iii) Waste Gas Cover Gas flowpath from the Waste Gas Dewy Tanks to the CVCS Holdup Tanks iv) Sample flow from the Radioactive Waste Gas System to the Gas Analyzer b)
~--.
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Tuesday, May 13,2003 4:28:57 PM I Page103of126 1
Which of the following lists the automatic functions performed by the New Fuel Storage FHB 1 (2IR9 Area Monitors?
!I (2)R9 diverts the ventilation for the FHB through charcoal filters, is interlocked to start FHB
!Exhaust Fans, activates radiation warning lights outside the building and emergency warning
/evacuation alarm inside the buildina.
2R9 and 1R9 have Unit differences. An interlock with the Unit 1 FHB Exhaust Fans exists on the 1 R9 monitors. Unit 2 2R9 monitors have NO automatic function.
1 (2)R9 activates radiation warning lights and emergency evacuation alarms inside the FHB AND Containment when limits are reached.
~
12R9 is interlocked with the FHB Exhaust Fans to prevent starting the fans if a high radiation
- condition exists. Unit I I R9 monitors have NO automatic functions.
L-7
/Memory Salem 1 &2
!072000K302
___. 81,
[ABNORMAL RADIATION a)
Supply Fan Unit b)
Exhaust Fan Controls I
e channel in warning or alarm.
1
[Editorially Modified Tuesday, May 13,2003 4:28:57 PM I Pase 104 of 126
!Given the following conditions:
Fire Protection System
- A defective fire protection wet head fusible leak in the Fire Pump House has caused both Diesel driven fire pumps and their associated Jockey pump to become INOPERABLE.
- Fire protection water header pressure has dropped to - 0 psig.
- The fire protection cross-tie valve to Hope Creek has NOT been opened.
I 82' 2
~
- Which of the following choices identifies a consequence of these events on other fire suppression jsystem capabilities?
/Fire detection capabilities outside the Fire Pump House will become degraded, but
/OPERABLE.
available for DFOST rooms even if deluge system is lost. Distractor C is incorrect because no dry or wet I head sprinklers will work without water. Answer D is correct because without fire suppression water to j
~
hLL fire suppression capability to the DFOST Room is unavailable.
i i
1
'ONLY dry head fire suppression sprinkler systems will remain available and OPERABLE.
I Salem Unit 3 Foam suppression capability is unavailable.
jd Comprehension 086000K604
~
Fire Protection System Malfunction I
i FIRPROE003 1 Given a P&lD of a Fire Protection System, Identify the following:
A.
Major Fire Protection System Components
- 1. Fresh Water and Fire Protection Water Storage Tanks
- 2. Diesel Fire Pumps
- 3. Jockey Pump
- 4. CNTMT Is01 Vlv 2FP147
- 5. C02 Storage Tanks
- 1. Water Supply System
- 2. Preaction Deluge System
- 3. Wet-Pipe Sprinkler System
- 4. Foam System
- 5. Carbon Dioxtde System B.
Major Fire Protection System Flowpaths Tuesday, May 13,2003 4:28:57 PM 1 Page 105 of 126 '
Tuesday, May 13,2003 42857 PM I Page 106 of 126 -
Which of the following Containment Personnel Airlock conditions must be met to ensure containment integrity is maintained in Modes I
-4?
!Inner Door is stuck open; Outer door OPERABLE.
/Inner Door has broken seal, closed and locked; Outer Door OPERABLE.
'Inner Door OPERABLE, closed, and locked; Outer Door has a broken seal.
ITSAS i
3.6.1.1, CONTAINMENT INTEGRITY, states that each containment airlock must be OPERABLE to lensure containment integrity. Distractor A is incorrect because the airlock is INOPERABLE with a stuck
!open door unless the other door is locked and sealed. Distractor B is incorrect because with a broken I
'Inner Door OPERABLE; Outer Door has a broken seal.
~
I-.-
Salem Tech Specs i
I J
CONTMTEOI 0 State the Technical Specifications associated with the components, parameters, and operation of the Containment And Containment Support Systems, including:
a) b)
c) d)
The Limiting Condition(s) for Operation The applicability of the LCO(s) (N/A NEO)
The LCO Action Statement@) (N/A NEO)
The Bases for the LCO(s) (N/A NEO)
CONTPOEOOl 1 List the five elements of a control system.
Tuesday, May 13,2003 4:28:57 PM 1 Page 107 of 126
Given the fo I lowi n g conditions:
- Unit 2 is in MODE 3.
- Both Reactor Trip Breakers (RTB) are closed for I&C testing.
- All Rod Drive MG (RDMG) Set breakers are open.
- 21 and 23 Reactor Coolant Pumps (RCP) are operating.
,- Electrical Maintenance is scheduled to test 21 RDMG.
/Which of the following describes the required conditions prior to starting 21 RDMG?
i E2 AND 24 RCP must be started before closing 21 RDMG set breakers.
1
~~
~
1 k21 AND 23 RCP's must remain in oeeration.while 21 RDMG set breakers are closed.
1 121 OR 23 RCP must remain in operation while 21 RDMG set breakers are closed.
I
land d are incorrect for the same reason.
/SALEM UNIT 2 TECHNICAL SPECIFICATIONS I
Indication Systems, including:
a)
The Limiting Condition(s) for Operation b)
The Bases for the LCO(s) d)
The LCO Action Statements(@
Tuesday, May 13,2003 4:28:57 PM I Page 108 of 126
A determination has been made that a compensatory reading must be taken on a normally logged
- parameter due to a safety related piece of equipment being out of service in Mode I.
I
- Which of the following is the requirement for the normal log reading while the compensatory
'reading requirement is in effect?
/Normal logging requirements are suspended.
'Normal loqginq requirements remain in effect.
Normal logging requirements are suspended provided the compensatory readings are verified I Iby the STA.
I
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hormal logging requirements are suspended provided the compensatory readings are verified PY a second RO l
.L
'194001G118 1
E z I E l i2.1 Jponduct of Operations
- a is correct in accordance with S2.OP-DL.ZZ-0003, Control Room Reading Modes 1 4. b is incorrect for 1
'the same reason. c and d are incorrect because there is not a requirement for verification.
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bontrol Room Reading Modes 1-4 Describe the type of information required to be recorded in the following narrative logs:
a)
SNSS Narrative Log b)
Control Room Narrative Log c)
NE0 Narrative Log
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Tuesday, May 13,2003 4:28:58 PM I Paae 109 of 126
During steady state 100% power operation, you are performing a "Plant Computer Online Calorimetric Validation" in accordance with Attachment 5 of S2.OP-I0.Z-0004, POWER 0 P E RAT1 0 N.
,The computer points for Steam Generator (S/G) Blowdown are all entered at 40K Ibm/hr.
iActual blowdown for 22 S/G is 42K Ibm/hr, all other S/G's are at 40K Ibm/hr.
'All other parameters are SAT.
Which of the following is the correct response to these conditions?
I 1
Enter a value of 42K Ibm/hr for 22 S/G Blowdown and insure the online calorimetric remains below 100%.
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,Declare the online calorimetric SAT since the entered value for 22 S/G is less than the actual iblowdown.
-l the online calorimetric UNSAT AND perform a confirmatory calorimetric in accordance
!Declare the online calorimetric UNSAT AND lower 22 S/G Blowdown Flow to </= 40K Ibm/hr.
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12.1 IjCOnduct of Operations to use plant computer to obtain and evaluate parametric information on system or component.L.3La [...3:0]
I
- d. is correct per the procedure attachment. A is incorrect, per the caution statement values of >40K Ibm/hr cannot be entered into the plant computer, that is out of the range that the online calorimetric uses. C. is incorrect, a confirmatory calorimetric cannot be performed with S/G blowdown >40K Ibmlhr.
,B. is incorrect, although an entered value less than actual is more conservative, the attachment does not permit calorimetric with S/G flow >40K.
I POWER OPERATION 1
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IOP004E003 Summarize the actions of S2.OP-IO Z-O004(Q), Power Operation, which are required or may be taken when plant parameters are at or reach the values stated below:
a) b)
c) d)
e) 9 Reactor power >40% during power ascension Reactor power S O % during power ascension Reactor power >75% during power ascension Reactor power <75% during power reduction Reactor power 60% during power reduction Reactor power ~ 3 5 %
during power reduction
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Tuesday, May 13,2003 4:28:58 PM 1 Page 110 of 126
Unit 2 is at 100% power when a low pressure feedwater heater string is automatically bypassed.
In response to these conditions, which of the following is correct concerning any Operator actions taken?
L CRS approval is required prior to reducing load.
!CRS approval is NOT required prior to reducing load.
/The CRS must be present at the controls for a power change greater than 5%.
/The CRS must verify all boration and dilution calculations prior to any power changes.
194001G120 !
12.1 JlConduct of Operations -
j2,1..20_) hbility to execute proced
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F------------__----_-_--------------_-__._~----..
/b is correct, the RO may take immedi without prior approval of j
ithe CRS. A, c and d are incorrect for the same reason although a and d are correct in a normal condition. '
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PROCEDEOOl I State the purpose of SC.OP-AP.ZZ-O102(Q), Use Of Procedures T
i ISignificantly Modified
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i Tuesday, May 13,2003 42858 PM 1 Page111 of126 1
During a plant heatup from Cold Shutdown in accordance with S2.OP-I0.Z-0002, COLD SHUTDOWN TO HOT STANDBY, the 21 thru 24SJ54 Accumulator outlet valves are opened. Once the valves are fully opened how will the control power be aligned and why?
Left in the ON position and C/T to prevent a single failure from rendering the accumulator INOPERABLE.
/Placed in the OFF position and C/T to prevent the valve from closing on a spurious actuation Jsig n ai.
- Placed in the OFF position with the 2RP4 Lockout Switch in Valve Operable to allow
,safeguards i
actuation -
of the valve.
- Left in the ON position with the 2RP4 Lockout Switch in Lockout to prevent the valve from
- closing on a spurious actuation signal.
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COLD SHUTDOWN TO HOT STANDBY I
plant parameters are at or reach the values stated below:
a)
RCS Temperature c2002F b) c)
d)
RCS Temperature >3503=
RCS Pressure 61 000 psig SG Pressure >680 psig Permissive P-I 1 Permissive P-12 RCS Temperature 547$$F and Pressure 2235psig RCS Temperature >200@F and ~ 3 1 2 F RCS Temperature >3128? and c3503F I
Tuesday. Mav 13.2003 4:28:58 PM Page 112 of 126
Which of the following On-The-Spot-Change (OTSC) requests could the OS/CRS sign for approval in accordance with NC.NA-AP.ZZ-0001, NUCLEAR PROCEDURE SYSTEM?
/Change of a test gauge to a different type of gauge which has the identical range and
!accuracy.
/Change a QA Inspection Hold Point to a QA Inspection Notification Point.
'Change a drain path for a section of piping which could cause an increase in general radiation levels.
,Change to the acceptance criteria for 23 Auxiliary Feedwater Pump overspeed test.
- 2.2
]/Equipment Control J
in the safety analysis
]E-2>3[:31 SRO ONLY 55.43(b)(3)
- a. is correct since it does not cause a change in intent of the procedure. The remaining choices do 1
- change intent and are specifically listed in NAP-I as examples of intent changes.
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NUCLEAR PROCEDURE SYSTEM I.
J Describe the situations which permit deviation from written procedures, and the type of documentation of the deviations required by SC.OP-AP.ZZ-OlOZ(Q), Use Of Procedures 1
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I Editorially Modified nk 1
Tuesdav. Mav 13.2003 4128159 PM 1 Page 113 of 126
Given the following conditions:
- A normally scheduled surveillance test run of I B EDG is in progress IAW S I.OP-ST.DG-0002.
- SC.OP-PT.DG-0001, Diesel Generator Manual Barring has been completed SAT.
Durina the test. at what point hashill 1B EDG be declared OPERABLE?
jWhen the 1 B EDG barring procedure was complete with second verification.
lWhen the I B EDG Lockout Switch (LOSW) is placed in the In-Service position.
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'When 1 B EDG acceleration, voltage, and frequency meet Acceptance Criteria of S I.OP-
/ST.DG-0002, and output breaker is shut.
_-__-_..._______________c_____--.---_______..------_~~~~~~-~~_._.._________
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When S I.OP-ST.DG-0002 is complete, and ALL Acceptance Criteria are met.
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-. -.- of pre-and post-maintenance operability 1 E,, 2 3 r3---]
i_L5 55.43(b)(2)The EDG Barring procedure specifies to enter the TSAS for INOPERABLE EDG kout switch (LOSW) is placed in lockout. Upon completion of the barring procedure (as
- stated in stem), the LOSW will be returned to the Inservice position. IAW SC.OP-PT.DG-0001, the
- OPERABILITY RETEST requirements which will satisfy TSAS termination criteria are: 1. sc.op-pt.dg-
- 0001 ATT. 1 INDEPENDENT VERIFICATION is complete. This is true because the stem states that this
'procedure has been completed SAT; 2. Diesel Generator voltage, acceleration, and frequency meet the I
'Acceptance Criteria stated in the Surveillance procedure used for retest; 3. The EDG is synchronized
,to the bus.
'required for OPERABILITY. D is incorrect because of the above explanation.
A and B are incorrect because each is not complete, it is only 1 of the 3 conditions
/Diesel Generator Manual Barring 7
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OPDETRE002
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EDGOOOEOIO a) b)
Degraded Condition c)
Nonconforming Condition d) e)
9 Single Failure g)
Consequential Failure Structures, Systems, and Components (SSCs)
Conditions Adverse to Quality (CAQ)
Justification for Continued Operation (JCO)
J cal Specifications associated with a) b)
c)
The Limiting Condition(s) for Operation (N/A NEO)
The applicability of the LCO(s)
The LCO Action Statement(s) (N/A NEO)
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Tuesday, May 13,2003 4:28:59 PM I Page 114 of 126
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Tuesday, May 13,2003 4:28:59 PM I Page115of126 1
Which of the following describes the primary reason for maintaining Refueling Cavity water level greater than 23 feet (127'1 1/2" elevation) over the top of the reactor pressure vessel flange with the reactor head removed?
cooling the core, in the event that RHR flow is lost Sufficient water volume is available to provide time for the operator to recognize the indications of a dilution accident before Keff can exceed 95% deltak/k
$he assumed 10% iodine gap activity released from the rupture of an irradiated fuel assembly. The basis
- of water above the pressure vessel flange is that a large heat sink is available for core cooling. Thus, in ithe event of a failure of the operating RHR loop, adequate time is provided to initiate emergency
- procedures to cool the core. Water does have the function of reducing the radiation levels, but this is NOT ithe bases for the 23' requirement. The required boron concentration has a basis for maintaining
!su bcriticalitv.
ISalem Technical Specifications kefuelina OPerations 1
L including:
a) b)
c)
The Limiting Condition@) for Operation The Bases for the LCO(s)
The applicability of the LCO(s) ment(s) (N/A NEO)
Tuesdav. Mav 13.2003 4:28:59 PM
[ Page 116 of 126
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'Which of the following parameter limits is established to ensure that radiation releases will remain within the limits of IOCFRZO?
k&l [Liquid Waste discharge activity.
n
/Primary system activity.
- Secondary system activity.
1 TECHSPE015
,Primary to secondary leakage.
Describe the general component and parameter categories which are addressed by Technical Specification Sections 314.1 314.,2 I
[I 94001 G301 j
92!
ITechnical Specifications 1
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Tuesday, May 13,2003 4:28:59 PM I Page 117 of 126
During a Unit 2 shutdown a decision has been made to enter the containment to inspect an I I N OP ERAB LE component.
!The Unit is being shutdown at 10% per hour.
!Whose specific approval is required to enter containment under these conditions?
/Nuclear Technical Supervisor (ALARA).
Radiation Protection Manager.
Vice President-Nuclear.
Operations Manager - Salem.
'to access containment is provided by the OS/CRS/Designee. However, to access certain areas of
/containment, or when power is being changed >5%/hr, the Radiation Protection Manager's (RPM)
/approval is required. Since the OS/CRS/Designee is not one of the available choices, the only correct
/answer is the RPM. Since the stem asked whose specific approval is required, the Operations Manager lis not correct. even thouah he/she sians the aoDroval of all ODS orocedures.
ICONTAINMENT ENTRIES AT POWER I
/Salem Containment Entries in Modes 1-4 RADCONE004 In accordance with NC.NA-AP.ZZ-O024(Q), Radiation Protection Program:
A.
Define the following terms:
- 1. Protected Area
- 2. Radiologically Controlled Area (RCA)
- 3. Restricted Area
- 4. Contaminated Area
- 5. High Contamination Area
- 6. Airborne Radioactivity Area
- 7. Radiation Area
- 9. Locked High Radiation Area I O. Exclusion Area
- 11. Very High Radiation Area Describe the following posting requirements, as well as any applicable restrictions:
- 1. Contaminated Area
- 2. High Contamination Area
- 3. Airborne Radioactivity Area
- 4. Radiation Area
- 7. Exclusion Area
- 8. Very High Radiation Area B.
C.
D.
E.
(SRWP)with a General Radiation Work Permit (GRWP)
F.
Alternations Describe when a whole body count is required Describe why protective clothing is used Describe the purpose of a Radiation Work Permit and compare and contrast a Specific Radiation Work Permit Describe the restrictions associated with Containment entry, including restrictions during Mode 1, Mode 2, and Core
- Tuesday, ay 13, 2003 4:28:59 PM I Page118of126 :
Tuesday, May 13,2003 4:28:59 PM 1 Page 119 of 126
The Nuclear Equipment Operator reports that the remote dose rate monitor is reading 22Whr.
/Continue sluicing, radiation levels are within the capacity of the High Integrity Container. 1
/Stop the sluicing operations and get Operations Shift Manager approval before continuing. ~_ ]
,Continue sluicing while the Duratek Technician raises the Dewatering Pump flow rate.
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'2.3 IIRadiation
.......... Control J
- 2.3.3._
j Knowledge of SRO responsibilities for auxiliary systems that are outside the control room (e.g.,
I [-::SI riig
[waste disposal and handling systems).
~SRO only
- 55.43(b)4
- a. is correct per the caution in the sluicing procedure Dose Rate is limited to 20Whr without Radiation Protection approval.. B. is incorrect for the same reason. C. is incorrect, the procedure requires Radiation Protection approval. D. is incorrect, raising dewatering pump flow rate is done if dose rate is ~20Whr.
1
~LUICING RESINS TO A HIGH INTEGRITY CONTAINER 1
and precautions associated with each operating procedure which are required to be considered by either L&ensed or Non-Licensed Operators i
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Tuesday, May 13,2003 4:28:59 PM 1 Page120of126 ;
,When performing a 13 Waste Gas Decay Tank Release with the I R41A INOPERABLE, which of ithe followinn is NOT rewired IAW SI.OP-S0.WG-OOIO?
'Obtain grab samples of release at least every 8 hours9.259259e-5 days <br />0.00222 hours <br />1.322751e-5 weeks <br />3.044e-6 months <br />, and analyze with 24 hours2.777778e-4 days <br />0.00667 hours <br />3.968254e-5 weeks <br />9.132e-6 months <br /> for gaseous I jprincipal gamma emitters.
/Estimate the Plant Vent Flow Rate at least every 4 hours4.62963e-5 days <br />0.00111 hours <br />6.613757e-6 weeks <br />1.522e-6 months <br />.
Two indeDendent samdes of I 3 Waste Gas Decav Tank have been analvzed.
iRelease rate calculations have been independently verified by at least two technically qualified lmembers of Facility Staff.
i2.3
]/Radiation Control performing a planned gaseous radioactive
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- Distractor B is incorrect because the estimated plant vent flow rate is only required when the plant vent
'flow rate monitor is INOPERABLE. All the other conditions are valid IAW P&L 5.2.3 i
I ISignificantly Modified 1
Tuesday, May 13,2003 4:29:00 PM I ~ a a e 121 of 126
Given the following conditions: EOP-FRCC-2, RESPONSE TO DEGRADED CORE COOLING, was entered due a PURPLE path condition for Core Cooling Safety Function.
- While performing this procedure the STA informs you that concurrent RED path conditions exist for both the Heat Sink Critical Safety Function and the Containment Environment Critical Safety Function.
- NO other abnormal conditions are noted.
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/TO A LOSS OF SECONDARY HEAT SINK.
/Complete the actions of 1-EOP-FRCC-2 and then transition 1-EOP-FRCE-1, RESPONSE TO
/EXCESSIVE CONTAINMENT PRESSURE.
Stop performing 1 -EOP-FRCC-2 and immediately transition to I
-EOP-FRHS-I.
istop performing
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-EOP-FRCC-2 and immediately transition to I
-EOP-FRCE-1.
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,ernpted by a higher priority critical safety function condition. The RED paths are higher priority than the
'PURPLE path, and a RED path on heat sink has a higher priority than a RED path on containment
/environment accordina to the CFST Hierarchv Tuesday, May 13,2003 4:29:00 PM I Page 122 of 126 '
Given the following conditions:
- - The crew is performing EOP-TRIP-1, REACTOR TRIP OR SAFETY INJECTION.
- The PO has been directed to perform EOP-APPX-1, COMPONENT COOLING RESTORATION.
,- While performing APPX-1, the CRS determines that a transition to EOP-LOCA-1, LOSS OF IREACTOR COOLANT, is necessary.
- Which of the following actions are required?
I IDiscontinue action in APPX-1 until directed by EOP-LOCA-I.
/Transition to LOCA-1 and continue action as necessary in APPX-I.
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bontinue action in APPX-1 but do NOT block or reset SECs until directed by EOP-LOCA-4.
1 homplete the action required by APPX-1 prior to transition to LOCA-I.
1 beneric Knowledge and Abilities 1940016412 1
1 IIIEl 1,~
- 3,4_._3291 c__--_______________.---....-_--_-~-______...
/2-.4_jFmergency Procedures / Plan ledge of general
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operating crew responsibilities
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during emergency operations.
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]"When an inter-procedure transition step is reached, the reader should perform the following:
- I. Review the procedure for steps in progress, these must still be completed. "
Action in APPX 1 is considered a step in progress, and will continue unless the action was not in line with
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'procedural action directed by an FRP. Do not hold up or postpone any action required by an EOP unless sPecificallv directed.
lUse of Procedures I
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I Tuesday, May 13,2003 4:29:00 PM 1
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Page 123 of 126
Unit 2 is at 100 % power and Unit 1 is in CSD when improper operation of a mobile crane results in a puncture in the # I FireIFresh Water Storage Tank (FWST). The System Manager estimates there is a total of 260,000 gallons remaining in the FWST's when level stabilizes below the p u nct u re.
Which one of the following describes the correct course of action for the operating shift IAW S2.0P-AB.FP-0001, FIRE PROTECTION SYSTEM MALFUNCTION?
NO compensatory action is necessary since 260,000 gallons meets the minimum requirement ifor fire protection purposes.
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/Initiate actions to place Unit 2 in HSB within the next 6 hours6.944444e-5 days <br />0.00167 hours <br />9.920635e-6 weeks <br />2.283e-6 months <br />.
'Place the fire pumps in manual and direct Site Protection post a watch for starting the pumps if i
/a fire is confirmed.
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/Direct Site Protection to lineup backup fire protection from HoDe Creek.
194001 G425 98 i2.4 JlEmergency Procedures / Plan i,I--;_-:
Knowledge of fire protection procedures.
2.9 [-3..d I
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SRO ONLY 55.43(b)(5)
Fire protection tanks are normally cross-tied, so puncture on one will i
Idrain both. With the Salem fire Protection tanks not having a minimum of 300,000 gallons, the AB directs I
- that Site protection shall establish backup supply from Hope Creek. Distractors a and c are incorrect 1
- because 260,000 gallons does not meet the minimum fire protection requirements. Distractor b is
!incorrect because only if the backup water supply cannot be established within 24 hours2.777778e-4 days <br />0.00667 hours <br />3.968254e-5 weeks <br />9.132e-6 months <br />, initiate a unit shutdown. and be in HSB within 6 hours6.944444e-5 days <br />0.00167 hours <br />9.920635e-6 weeks <br />2.283e-6 months <br />.
IFIRE PROTECTION SYSTEM MALFUNCTION
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J Tuesday, May 13, 2003 4:29:00 PM I Pane 124 of 126
- An explosion and fire at the RAP tank area has resulted in a possible large spill of radioactive water
'in the area. The Fire Department has determined that off-site assistance from the local fire department is needed.
,Which of the following choices identifies who must authorize requesting off-site fire department
/assistance?
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1 Security Duty Supervisor.
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- Emergency Duty Officer (EDO).
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- Radiological Assessment Coordinator (RAC).
I iNuclear Fire Protection Supervisor.
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'2.4 ]'Emergency Procedures / Plan 1
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'2.4.27 i
knowledge of fire in the plant procedure.
IES-401 page 6 of 46 d. states that SRO level questions can be 55.41(b) if they evaluate knowledge and I
jabilities at a level that is unique to the SRO job position Caution 3.41 in SC.FP-EO.ZZ-0001 states, "In i
ithe event of a radiological emergency, the Nuclear Fire Protection Supervisor should obtain permission
'from the EDO/OS Prior to callina for off-site assistance."
IFire and Medical Emergency Response Manual Salem Station - Control Room Fire Response 1
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Tuesday, May 13,2003 4:29:00 PM i Page 125of 126
,Given the following conditions:
- Salem Unit 2 is operating at 100% power.
- A Salem NE0 has discovered 21SJ35 and 22SJ35,21 and 22 SI PUMP DISCH VLVS, shut with
,their handwheels removed, and apparent valve stem damage.
I
'As the OS, which of the following actions must you perform IAW SH.OP-AP.ZZ-0007,
/SUSPECTED TAMPERING?
/Notify Shift Security Supervisor, NRC Resident, and NJ State Police.
Perform valve position verification on ALL systems required for safe shutdown, starting with I
- high head injection system.
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ilmmediatelv quarantine area with ANY available personnel.
1
- Notifv Shift Securitv Sur>ervisor, Hope Creek OS, and NRC resident.
12.4
]:Emergency Procedures / Plan 12.4.28 _i [Knowledge of procedures relating to emergency response to sabotage.
I pz -514 lSRO ONLY 55.43(b)(5)
ISH.OP-AP.ZZ-0007, SUSPECTED TAMPERING identifies who needs to be contacted after identifying
- event.
Distractor A is incorrect because NJ State police not required by procedure to be notified Distractor B is incorrect because Valve position verification should start with affected system(SJ)
!Distractor C is incorrect because Quarantine is Derformed bv Securitv due to safetv concerns.
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[Suspected Tampering
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J Tuesdav. Mav 13.2003 4:29:00 PM 1 Page 126 of 126