ML13032A193
| ML13032A193 | |
| Person / Time | |
|---|---|
| Site: | Brunswick  |
| Issue date: | 01/29/2013 |
| From: | NRC/RGN-II |
| To: | Duke Energy Corp |
| References | |
| 50-324/12-301, 50-325/12-301 | |
| Download: ML13032A193 (86) | |
Text
{{#Wiki_filter:SRO Written Exam Reference Index
- 1. OEOP-01-UG, Users Guide, Attachment 5, Figure 3, Heat Capacity Temperature Limit
- 2. OEOP-O1-UG, Users Guide, Attachment 5, Figure 5, Core Spray NPSH Limit
- 3. OEOP-Ol-UG, Users Guide, Attachment 5, Figure 6, RHR NPSH Limit
- 4. OEOP-O1-UG, Users Guide, Attachment 5, Figure 7, Pressure Suppression Pressure
- 5. OPT-07.2.4a, Core Spray System Operability Test - Loop A
- 6. OOP-06.4, Section 5.7, Recirculation and Sampling of Unit 1 Saltwater Release Tank
- 7. OPEP-02.1, Brunswick Nuclear Plant Initial Emergency Actions
- 8. OPEP-02.6.28, Attachment 1, PAR Flowchart
- 9. OPEP-02.6.28, Attachment 2, Page 1 of 2, Evacuation Zones/Time Estimates/i 0 Mile EPZ Map
- 10. TS 3.5.1 ECCS - Operating ii. TS 3.6.3.1, Primary Containment Oxygen Concentration
- 12. TS 3.8.1, AC Sources - Operating
- 76. S209001 1 Unit Two is at rated power with RHR Pump 2D 5 00- under clearance.
400a OPT-07.2.4A, Core Spray System Operability Test Loop A is in progress. 300- GE The Core Spray Full Flow Test Byp Vlv, E21-FO15A, 5000 is fully opened and the 1ST personnel have recorded the appropriate values on Attachment 2, Unit 2 Core 200-Spray Pump A Test Information Data Sheet. 100 (Data Sheet is provided) 2000 c Based on the performance of OPT-07.2.4A, which one of the following identifies the most limiting required action, if any, lAW Technical PUMP 2A PUMP 2A Specifications? DISCH DISCH PR ESS FLOW E2 1PIR6004 E2 1 FIFSO 1A A. No action required; the LCO is met. B. The unit must be in Mode 3 in 13 hours. C. Requires returning a low pressure injection system to OPERABLE in 72 hours. D. Requires returning a low pressure injection system to OPERABLE in 7 days. Answer: C K/A: 209001 Low Pressure Core Spray System G2.01 .07 Ability to evaluate plant performance and make operational judgments based on operating characteristics, reactor behavior, and instrument interpretation. (CFR: 41.5 /43.5 /45.12 /45.13) ROISRO Rating: 4.4/4.7 Objective: LOl-CLS-LP-018, Obj. 18 Given plant conditions, determine the required action(s) to be taken in accordance with Technical Specifications including the Bases, TRM, ODCM and COLR associated with the Core Spray system. (SRO only)
Reference:
OPT-07.2.4A, acceptance criteria and Attachment 2 markup. Cog Level: High
I!I!II!L &L L V!IIIPIr IiiI Explanation: Based on performance of the PT the loop is inoperable. TS 3.5.1 condition B applies (1 LPCI pump and 1 CS loop inoperable). Condition J does not apply because condition B does apply. Distractor Analysis: Choice A: Plausible because if condition A or B did not apply this would be correct. Choice B: Plausible because if condition J were chosen (inoperability for reasons other than condition A or B) Choice C: Correct Answer, see explanation Choice D: Plausible because this is an action for an inoperable RHR or CS loop only (condition A). SRO Basis: Facility operating limitations in the technical specifications and their bases. (10 CFR 55.43(b)(2)) This question requires knowledge of TS bases to analyze the required action/terminology. From TS 3.5.1 CONDITION REQUIRED ACTION COMPLETION TIME A. One low pressure ECCS A. 1 Restore low pressure 7 days injection/spray subsystem ECCS injection/spray inoperable, subsystem to OPERABLE status. OR One [ow pressure coolant injection (LPCI:i pump rn each subsystem inoperable. B. One LPCI pump inoperable. B.I Restore LPCI pump to 72 hours OPERABLE status. AND OR One core spray (CS) subsystem inoperable. 5.2 Restore CS subsystem to 72 hours OPERABLE status. (continued) ACTIONS (continued CONDITION REQUIRED ACTION COMPLETION TIME J. Two or more low pressure J. 1 Enter LCO 10.3. Immediately ECCS inection?spray subsystems inoperable for reasons other than Condition A orB. OR HPCI System and two or more required ADS valves inoperable.
4p s nig1s,t IL aLbs) aajAaajtIjt-ATTACHMENT 2 Page 2 of 2 Unit 2 Core Spray Pump A Test Information Data Sheet
- 1. The ubroant leve (pumo running) is normal:
- 2. Calcuate pump dP as fol ows:
Pump discharge pressure suction pressure (run) = punc dP NOTE: Pump vibration measurement is required only during CPT. Vbration is measured at the test point marked on the pump for the correct bearing number and direction as indicated by The Test Position number as follows:
- the number indicates the bearing number from Attachment .5
- for oositon, N=North, S=SouTh. E=East, W=West
- for direction, A=Axial, H=Horizontal, V=Vertical NOTE: Reference values for pump suction and dscharge pressures are provided for determining the suitab Fty of altemate test gauges, if used.
NOTE: Pump stopped suction pressure should normally be between £ and 6 psig. This parameter s a function of torus pressure and suppresson pool water level. Values outside of this range may also be :ndicat;ve of air in the instrument line. UNIT 2 CORE SPRAY PUMP A TEST DATA MLERT RANGE REQUIRED ACTION
. ACTUAL REFERENCE ACCEPTANCE RANGE 0 MRA VALUE VALUE VALUE RANGE LOW HIGH LOW HIGH uotion Pres.
o:opped}_osig 45 psig S.D NM MA MA MA N?A Suction Press. Running) psig 2 . C pci 4.0 N/A N/A MA MA MA L Discharge Press. 2)0 N/A N/A MA MA MA Quartely urnp DP N/A 236.0 26C0 to 316.8 N4 MA < 26C;3 > 316.2 psid 253.0 CZT °unip OP os:d 236.0 287.8 to 26.e to MA < 260.0 > 26.f
<257.6 Flow Rate gpm 4,700 MA N/A N/A NM MA b
,i ra.ion-ve (ins
> 0.325 fl 000 0.230 0 to 0.325 N/A to MA > 0.700 peak) Position b-I 0,7)0
,.. . . > 3.325 0.486 022 0 to 0.325 N/A to N/A > 0.700 peaWiPo!itionlW A -
. u.Ju ibra:ion-el (ins 0.638 peak) Position 1 . H 0:55 3:0 0.325 N/A to N/A > 0.700 0 730 Performed By (Signature): Date: Time:
Reviewed, 1ST Group (Signature): Date: OPT-07.2.4a Rev. 71 Page 33 of 41
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LAI11!!IIJ? LJLL dIIIILi
- 77. S215002 1 Unit Two is performing a control rod sequence exchange.
The following annunciator is received while withdrawing control rod 26-15 from position 12 to 24: RBM UPSC/INOP RBM Channel A indicates: 109 on the 125 scale LPRMs in RBM Average: 7 APRM 1 indicates 68% Which one of the following choices completes the statements below? RBM Channel A is (1) Placing Rod Select Power control switch to OFF and then to ON lAW (2) will clear this condition. A. (1) inoperable (2) 20P-09, Neutron Monitoring System Operating Procedure B. (1) upscale (2) 20P-09, Neutron Monitoring System Operating Procedure C. (1) inoperable (2) RBM UPSC/INOP annunciator procedure D. (1) upscale (2) RBM UPSC/INOP annunciator procedure Answer: D K/A: 215002 Rod Block Monitor System A2.01 Ability to (a) predict the impacts of the following on the ROD BLOCK MONITOR SYSTEM ; and (b) based on those predictions, use procedures to correct, control, or mitigate the consequences of those abnormal conditions or operations: Withdrawal of control rod in high power region of core: BWR-3,4,5 (CFR: 41.5 /45.6) RO/SRO Rating: 3.3/3.5 w
L Lz aA9 - 2 Objective: LOI-CLS-LP-09.6, Obj. 14h Given PRNMS settings for abnormal conditions or operation, use the Annunciator Panel Procedures (APP) to determine the probable causes for the following alarms: A-6 4-8, RBM UPSC/INOP. LOI-CLS-LP-09.6, Obj. 24b Given PRNMS settings during normal or abnormal operation, predict the effect on the following: RBM System. LOl-CLS-LP-09.6, Obj. 25d Given PRNMS settings during normal or abnormal operation, use procedures to determine the required action to correct, control, or mitigate the consequences of the following: RBM alarm and trip signals.
Reference:
None Cog Level: High Explanation: The identified control rod is in the center of the core (higher power) where there is a stronger likelihood that a rod withdrawal may exceed the RBM trip setpoints. RBM trip setpoints vary based on simulated thermal power as provided by the reference APRMs to the RBMs. The Intermediate Trip Set Point (TSP) is active when STP is between the Intermediate Power Set Point (IPSP) and the High Power Set Point (HPSP) (between 62.7 and 82.7 % power). In this range, an upscale alarm would be received at 108.3% indicated on the RBM. The minimum number of LPRMs required is four for the rod selected (1/2 of the available LPRMs for that rod). If less than the minimum LPRMs are available the same alarm would annunciate but the reason would be for an mop condition. This rod is not a peripheral control rod, therefore the minimum LPRMs required is four. APP-06 4-8 requires confirming with RE that adequate thermal margin exists, then deselecting and reselecting the control rod to null the RBM to the current value. Distractor Analysis: Choice A: Plausible because alarm window alarms for upscale or mop conditions. There would normally be 8 LPRM inputs to the RBM for this rod. 20P-09 provides a section for bypassing an LPRM but not for bypassing a RBM. The guidance for bypassing a RBM is in APP-06. Choice B: Plausible because bypassing of a RBM channel lAW the APP is only performed if the conditions is caused by an malfunction. Choice C: Plausible because alarm window alarms for upscale or mop conditions. Choice D: Correct answer, see explanation. SRO Basis: Assessment of facility conditions and selection of appropriate procedures during normal, abnormal, and emergency situations.[10 CFR 55.43(b)(5)].
aca LeJtaS k a. a APart a dJLSt 2APP-A-06 Unit 2 APP A-06 4-8 Page 1 of 3 RBN UPSC/INOP AUTO ACTIONS
- 1. Rod Withdrawal Block, CAUSE
- 1. RBM Channel(s) upscale.
- 2. High local flux levels in vicinity of selected control rod.
- h. If an RBN UPSC/INOP alarm is received, then perform the following for each occurrence (1) Confirm (with Reactor Engineer or Core Performance Log) that adequate margin to thermal limits exist.
(2: Oonfirm S seconds have elapsed since the RMCS Settle light extinguished. (3) Confirm with Rod Verifier S seconds have elapsed since the RMCS Settle light extinguished. (4) Confirm with Reactivity Manager S seconds have elapsed since the RMCS Settle light extinguished. (5) Obtain Reactivity Managers concurrence and place Rod Select Power control switch in OFF. (6) Place Rod Select Power control switch in ON.
- i. If a REN malfunction is confirmed, bypass the affected channel and refer to Tech Spec 3.3.2.1 for the required actions, SD-09.6 2.7 RBM SYSTEM The RBM consists of two redundant channels: A and B. The primary function of the RBM is to:
- Inhibit rod withdrawal if the local power levels, as indicated by neutron flux levels, in the vicinity of a control rod that has been selected for withdrawal or insertion exceeds preset limits.
- Provide alarm signals for operator action if the local power level in the vicinity of a control rod selected for movement exceeds preset limit.
The RBM bypass switch allows the operator to manually bypass either of the RBM channels but not both at the same time. The switch provides an input to each of the RBM channels. Both RBM instruments separately receive the STP value from each of the four APRM channels. Based on this value, the RBM instrument selects one of three different RBM Average Flux Upscale set points or automatically bypasses itself.
aArss .a . Each REM channel designates a hierarchy of normal and alternate APRIV1 channels to use as their reference APRM channel. The alternate channels are used in hierarchical order when the preferred channels are not available- The primary reference APRM for REM A is APRM 1 with the first alternate as APRM 3 and the second alternate as APRM 4. The primary reference APRM for REM B is APRM 2 with the first alternate as APRM 4 and the second alternate as APRM 3. The REM circuitry will automatically transfer to an alternate APRM on failure of the primary reference APRM (Critical Self Test Fault). No operator action is required for this transfer REM Channel A REM Channel B Primary Reference APRM I APRM 2 First Alternate APRM 3 APRM 4 Second Alternate APRM 4 APRM 3 The REM channel automatically bypasses itself (its upscale, downscale, and rod inhibit outputs) when the reference APRM STP value is below the REM Low Power Set Point or if a peripheral control rod is selected (reference Fig 09.6-27 for edge rod identification).
- The Low Trip Set point (LTSP) is active when STP is between the REM Low Power Set Point (LPSP) and the Intermediate Power Set Point (IPSP)
- The Intermediate Trip Set Point (ITSP) is active when STP is between the Intermediate Power Set Point (IPSP) and the High Power Set Point (HPSP) 2.7.2 RBM Processing The A level LPRM detectors are not used for REM input processing, while both REM channels use all C level detectors.
This gives an accurate representation of actual power around the control rod. The E and D detectors are distributed evenly between the two REM channels. An example of LPRM input to a both REM channels with a four-string rod selected is two B level LPRM5, four C level LPRMs, and two V level LPRMs for each channel. WA IIIOP NA - Crthca self-test faL?li is detected
- RBM key bck swilcfl in (NOP
- LPRM inputs <fbi the iequired number
- More than one control rod selected
- Watchdog twner has limed out
- Loss of input power
n t aStiaaS Laasaa&A t a.aU 2.7.4 Set points
- a. RBM Alarm Set Points High Trip Set Point (HTSP) 104.5 Iritemiediate Trip Set Point (ITSP) 108.3
- Low Trip Set Point (LTSP) 114.1
- Set Points Relative to a full scale reading of 125 (e.g.. 114.1 means 114.1/125 of full scale)
Downscale Alarm 2.4%
- b. RBM Power Set Points High Power Set Point (HPSP) 82.7% RTP Intermediate Power Set Point (IPSP) 62.7% RTP Low Power Set Point (LPSP) 27.7% RTP iarrIIaIraPIrr7
!IIi2I1IIFi2L £i JL b JP L (tJfJIJhIAI1dlz
- 78. S215003 I Unit Two is performing a reactor startup, prior to the point of adding heat..
IRM E is bypassed due to failing downscale. IRM A Upscale/mop alarm is received due to degradation of the high voltage power supply to IRM A. Which one of the following choices completes the statements below? The power failure to IRM A will initiate a (1) lAW Technical Specifications LCO 3.3.1 .1, Reactor Protection System (RPS) Instrumentation, the required action is to (2) within 12 hours. A. (1) rod blockONLY (2) be in MODE 3 B. (1) rodblockONLY (2) place IRM A in a tripped condition C. (1) rod block and 1/2 scram (2) be in MODE 3 D. (1) rod block and 1/2 scram (2) place IRM A in a tripped condition Answer: D K/A: 215003 Intermediate Range Monitor System A2.01 Ability to (a) predict the impacts of the following on the INTERMEDIATE RANGE MONITOR (IRM) SYSTEM ; and (b) based on those predictions, use procedures to correct, control, or mitigate the consequences of those abnormal conditions or operations: Power supply degraded (CFR: 41.5 / 45.6) RO/SRO Rating: 2.8/3.2 Objective: LOl-CLS-LP-009-A, Obj. 13 Given plant conditions and Technical Specifications, including the Bases, TRM, ODCM and COLR, determine the required action(s) to be taken in accordance with Technical Specifications, TRM and COLR associated with the Source Range and Intermediate Range Monitoring Systems. (SRO/STA only)
Reference:
None Cog Level: High Explanation: TS require 3 IRMs per trip system in MODE 2. The 4 channels that make up trip system A are IRM A, C, E, and G. With IRM E failed downscale this makes IRM A a required channel. With one required IRM mop TS require placing it in a tripped condition, 01-18 specifies placing the switch in Standby to accomplish this task.
ITI!IIIIII1I Distractor Analysis: Choice A: Plausible because on a downscale (which could be thought of from a loss of power) only a rod block would be generated. If the action statement cannot be met then being in mode 3 in 12 hours is correct. Choice B: Plausible because on a downscale (which could be thought of from a loss of power) only a rod block would be generated. Choice C: Plausible because if the action statement cannot be met then being in mode 3 in 12 hours is correct Choice D: Correct Answer, see explanation SRO Basis: Facility operating limitations in the technical specifications and their bases. (10 CFR 55 .43 (b) (2 Jt?NT JWEER DS1-lM-A, , C U F , - T!JPNT ME. TS EFEP.ENCE 3.?.ll; FJi Te 1-l.la b TTP C-tEL AlA.! AZ-C, Q 51-5,F ThIF fEt A-Al 39 AZ
-5la!T2 TIFLCGC: Al aA2 a-1 o2- Re cr craii eiarr If lpçe codln t: F1cVI; IRM cwe edorEtLCr If Sbn1ty potcn
-kNNE. lrnJr MJ1?EER TIF L?-IT ACTICN F/NEL UNCTON CIJ1A IA Cll ei.r nCrIICrl Irp CIWI1EI A ECCf i- i
. C-M-C NA Cl-l - NeLrrcrflCr;tr
-i ctaA2craw ClR/- Cl-1 CIM F4A C14l Nr CrItCr- Irp FfC E ECfTf C! -IRM-F IA Cl- 1 C3- M-D IVA CS1-Sl - NeLtrr mccIIcr Irp E. cl-a,nel a
,N- FA C!-l- I CCl,NEM: ,VeIy Ire aiTecte-I IRM 1 iL bseecI I Ire RTC-E.
EFEfNCE DRAWINQ: l-FP-5C4S. -FP-!Cl! coi--a I Page4Dof1D8
SiL1aI dILl VlsI - S RPS Instrumentation 3.31.1 Tac 3.3.1 1-1 ct 3. RncIrP2cIsr S tnr h:ttrrn1t 1 app,5 MCCES CR csa_ SC prERSNr2: 2SR CRk..SLS ROM SPEC ED Z5 RSCIRS) StPvE ..LA1E F$SC D1 COO tINS SS1M AcjTlC? D/ SEILIPE9ENTS LUE
;2nesb1e Sans ti,nttrn Neulrcn Lx; 2 3 3 SR 11.1.1 2 21121 ci,ttrs cfM SR 3.3.1.14 SR 3.3. .15 SR 3.3.1.19 SR 3.32 IT SR 3.3.1.1 13 SR 3.3.1 .I IS 5 3 I- SR 3.2..I2 -21I2ScNNicrscfftE SR 3.3.1.14 SR 3.3.1.15 SR 3.32.1 13 SR 3.3.1.1 IS
- z. Irco 2 3 C SR 3.3.1.14 SR 3.3.1.15 SR 3.3.1.1 19 S 3 SR 3.3.1.14 SR 3.3.1.15 SR 3.32.1 iS RPS Instrumentation 3.3.1.1 3.3 INSTRUMENTATION 3.3.1:1 Reactor Protection System (RPS) Instrumentation LCO 3.3.1.1 The RPS instrumentation for each Function in Table 3.3.1.1-I shall be OPERABLE.
APPLICABILITY: According to Table 3,3.1.1-1. ACTIONS Separate Condition entry is allowed for each channel. CONDITION REQUIRED ACTION COMPLETION TIME A. One or more required A. 1 Place channel in trip. 12 hours channels inoperable. OR A.2 -----------NOTE--------- 12 hours Not applicable for Functions 2.a, 2.b. 2.c, 2.d, or 2.f. Place associated trip system in trip.
fld.flv it 4flJc w B. ------NOTE-------------- B.i Ptace channeL in one trip 6 hours Not applicable for Functions system in trip. 2.a, 2.b, 2.c. 2.d, or 21 OR One or more Functions with B.2 Place one trip system in 6 hours one or more required trip. channels inoperable in both trip systems. G. As required by Required G.1 Be in MODE 3. 12 hours Action D. 1 and referenced in Table 3.3.11-1. H. As required by Required H. 1 Initiate action to fully insert Immediately Action D.1 and referenced in all insertable control rods in Table 3.3.1.1-1. core cells containing one or more fuel assemblies. wiwra ztn
- 79. S218000 1 Which one of the following choices completes the statements below?
The impact of a loss of power to ADS from 1 25V DC panel 4B is that (1) (2) provides the guidance to reset 1 25V DC panel 4B, Ciráuit 11 (ADS Relay Logic A & B and SRV normal power supply breaker). A. (1) ADS will initiate if required from the A logic only. (2) OAOP-39.O, Loss of DC Power. B. (1) ADS will initiate if required from the A logic only. (2) 2APP-A-03, 2-2, Auto Depress Control Pwr Failure. C. (1) ADS will initiate if required from the B logic only. (2) OAOP-39.O, Loss of DC Power. D. (1) ADS will initiate if required from the B logic only. (2) 2APP-A-03, 2-2, Auto Depress Control Pwr Failure. Answer: D K/A: 218000 Automatic Depressurization System A2.05 Ability to (a) predict the impacts of the following on the AUTOMATIC DEPRESSURIZATION SYSTEM ; and (b) based on those predictions, use procedures to correct, control, or mitigate the consequences of those abnormal conditions or operations: Loss of A.C. or D.C. power to ADS valves (CFR: 41.5 /45.6) RO/SRO Rating: 3.4/3.6 Objective: LOI-CLS-LP-020, Obj. 1 5d Given plant conditions, predict how ADS/SRVs will be affected by the following: Loss of DC power
Reference:
None Cog Level: High Explanation: ADS Logic A is powered from 4B, ADS logic B is normally powered from 4B with a backup power supply from 4A. A loss of DC power from panel 4B will result in the loss of all power to ADS logic A. ADS logic B will also lose Div II power but will automatically swap to DC power from panel 4A thereby maintaining operability. The APP has the steps to reclose the circuit breaker. The AOP mainly deals with a loss of the whole panel Distractor Analysis: Choice A: Plausible because Div II DC provides power to the A logic and the AOP is for a loss of DC. Choice B: Plausible because a Div II panel provides power to the A logic Choice C: Plausible because the AOP is for a loss of DC power. Choice D: Correct Answer, see explanation SRO Basis: Assessment of facility conditions and selection of appropriate prccedures during normal, abnormal, and emergency situations (10 CFR 55.43(b)(5))
JC,WJt 4 4 t
£L J S a atac (lEsS JAt From OAOP-39:
3.2.5 Loss of Division II DC Panels 36, 1 lB. 9A(46, 126. 1OA)
- 1. IF loss of DC Distribution Panel OA(1OA) has occurred.
THEN DISPATCH an operator to the BOP bus area of the Turbine Building to shift DC control power for DC Distribution Panel 9A(IOA) as follows:
- a. OPEN DC Distribution Panel 9A(1OA) normal power supply incoming feeder breaker.
- b. CLOSE DC Distribution Panel 9A(10A) alternate power supply incoming feeder breaker.
- 2. IF loss of DC Distribution Panel 38(48) has occurred, THEN REFERENCE Attachment 4 for plant effects.
- 3. IF loss of DC Distribution Panel 118(128) has occurred, THEN REFERENCE Attachment 5 for plant effects.
- 4. IF power can NOT be transferred for a panel OR emergency operations are required with Panels 36 or 118(46 or 128) deenergized, THEN REFER to 01-50 for specific load information.
ATTACHMENT 4 Page 1 of 1 Plant Effects from Loss of DC Distribution Panel 3B(4B) ADS: ADS Logic A is inoperable. ADS will initiate from ADS Logic B if Core Spray Pump A or both RI-IR Loop A pumps are running. From 2APP-A-03: ACTIONS
- 1. If cause of annunciator is loss of power to 125V DC Distribution Panel 4A or 4B, then refer to AOP-39.3, Loss of EC Power.
- 2. If ADS Relay Logic A & B and SRV normal power supply breaker, Circuit 11 on 125V DC Distribution Panel 48, is OFF or tripped, reset breaker.
- 3. If ADS Relay Logic B and SRV alternate power supply breaker, Circuit 11 on 12EV DC Distribution Panel 4A, is OFF or tripped, reset breaker.
4, If cause of annunciator is blown fuses, replace fuses. S. If a circuit nialfunction is suspected, ensure a WR/WO is prepared. 2APP-A-03 Rev. 55 Page 24 of 103 7flrESIr :rw E EE
s%nt SLaaL C&aa Ii da# (saa 4ZaJIUIL1. From 001-50: PANEL 4B LOCATION: NORMAL SUPPLY: ALTERNATE SUPPLY: Reference Drawing: LL-03024-7 Conhrol Building 49 South Switchboard 26 N/A CKTW LOAD EFFECT
- ADS Logo B Normal Power Loss no;;ar, a logo vail still fanotion wiei alternate power. (pane & oirouit ii;
- 2. Rsoeive annuno alor A-02 2-2.
ADSLcgoAPower 1. LossofADsLoqsA. ACSISRV Norrral Soei,oid Powe- . Loss of normal power, all valves v fully rcion with a:erna:e pow. rel A. ohouil 1 U From SD-20: NORMAL SUPPLY ALTERNATE SUPPLY
- EQUIPMENT Unit I (Unit 2) Unit I (Unit 2)
Auto Depressurization 125 VDC Distribution None Logic A Pnl. 3B CKT 11 MB CKT 11) from Bus 18 (28) Auto Depressurization 125 VDC Distribution 125 JDC Distribution Logic B PnI. 3B CKT 11 PnL 3ACKT 11 (48 CKT 11) (4ACKT 11) from Bus lB (28) from Bus 1A (2A) Safety/Relief 125 VDC Distribution 125 VDC Distribution Valves Solenoid Pnl. 3B CKT 11 Pnl. 3A CKT 11 / Power (48 CKT 11) (4ACKT 11) from Bus lB (2B) from Bus 1A (2A) L SD-20 Rev. 3 PAGE 21 of62 nrnr ry--
- 80. S241000 I Unit Two was operating at rated power when a loss of UPS Distribution Panel 2A occurs.
Subsequently a LOCA occurs. The following plant conditions exist: Reactor pressure 200 psig Torus pressure 10 psig Torus water level 4 inches Drywell pressure 12 psig Drywell temperature 31 0°F Which one of the following choices completes the statements below? Terminating and preventing injection (1) required. If all of the ADS valves fail to open during emergency depressurization the CRS will direct the opening of the remaining SRVs (2) A. (1) is (2) ONLY B. (1) is not (2) ONLY C.(1) is (2) and perform AEDP D. (1) is not (2) and perform AEDP Answer: C K/A: 241000 Reactor/Turbine Pressure Regulating System G2.04.09 Knowledge of low power/shutdown implications in accident (e.g., loss of coolant accident or loss of residual heat removal) mitigation strategies. (CFR: 41.10 /43.5 / 45.13) RO/SRO Rating: 3.8/4.2 Objective: LOl-CLS-LP-300-D, Obj. 3g Given plant conditions and the Reactor Vessel Control Procedure, determine if any of the following are appropriate or required: Alternate Emergency Depressurization.
Reference:
None Cog Level: High Explanation: Loss of UPS requires use of LPC following the scram. LPC requires terminating and preventing injection sources prior to depressurizing. Inoperability of seven of the eleven SRVs results in less than five SRVs for depressurization and with reactor pressure greater than 100 psig above torus pressure, emergency depressurization using OEOP-01-AEDP, Alternate Emergency Depressurization, is required. With MSlVs open, use of turbine bypass valves to depressurize to the main condenser is available.
1!1III1ItIIIIPiILk L L JI L1L1IIII Distractor Analysis: Choice A: Plausible because if reactor pressure is less than 100 psig above torus pressure, use of AEDP is not required lAW LPC RCIP-46. Choice B: Plausible because use of LPC is required due to loss of UPS and unavailability of control rod position indication. If UPS were available and all rods inserted, RVCP would be utilized for depressurization. RVCP does not require terminate and prevent prior to depressurizing. If reactor pressure is less than 100 psig above torus pressure, use of AEDP is not required lAW LPC RCIP-46 and RVCP RC/P-24. Choice C: Correct answer, see explanation Choice D: Plausible because use of LPC is required due to loss of UPS and unavailability of control rod position indication. If UPS were available and all rods inserted, RVCP would be utilized for depressurization. RVCP does not require terminate and prevent prior to depressurizing. SRO Basis: Assessment of facility conditions and selection of appropriate procedures during normal, abnormal, and emergency situations.[10 CFR 55.43(b)(5)]. This question requires knowledge of the decision points in EOP that involve transition to EOP contigency procedures. 2EOP-01 -LPC r tF SEVEN ADS VALVES ARE NOT OP EN ThEN OPEN SRW . F. EOR GUNTILSEVEN &LVES AREOPEN 1 RC?P.45 I NO NI IS _ REACTOR 953 *
< GREATER ThAN 100 PMG ASOVE SUPPRESSION CI4MliER
.REsS RCIP40 IVES PERFORM ALTERNATE EWRGENCY DEPRESSURATION PROCEDURE EOP. Dl AEDP)
IRRESPECTWEOF OFFSITE RADIOAG1IVITY RELEASE RATE RC?P47 7i
Jwi&!I1gL ILkI OEOP-O1 -AEDP 2.8 RAPIDLY DEPRESSURIZE the reactor vessel. irrespective of the resulting cooldowri rate and offsite radioactivity release rate, using one or more of the following: NOTE: The systems listed below may be used in any order based on system availability and system capacity. NOTE: HPCI and RCIC can NOT be placed in the reactor pressure control mode if either system has a valid initiation signal present OR if any torus suction valve is full open. CAUTION Unit 2 Only: Steam flaw rates qreaterthan or equal to 3 x 106 lb/hr may cause Group 1 isolation. RO: - Main condenser (via the turbine bypass valves) 2EOP-01-RVCP I ERGBICY EPIS9IM1ON PROGEP D1A RAOAG1WflhILEASE RATE)
!Y
i iL ILZJ I
- 81. S2610001 During accident conditions on Unit Two the response of SBGT is indicated below:
SBGT A SGT B 2XE 2 I C C C Which one of the following choices completes the statements below? The SBGT 2A Fan (1) trip when the OH 210°F light illuminates. If a fire were to occur in the SBGT train, the deluge system is manually initiated lAW (2) A. (1) will (2) 2OP-1 0, Standby Gas Treatment System Operating Procedure B. (1) will (2) OOP-41, Fire Protection and Well Water System C. (1) will NOT (2) 2OP-10, Standby Gas Treatment System Operating Procedure D. (1) will NOT (2) OOP-41, Fire Protection and Well Water System Answer: A
K/A: 261000 Standby Gas Treatment System A2.03 Ability to (a) predict the impacts of the following on the STANDBY GAS TREATMENT SYSTEM; and (b) based on those predictions, use procedures to correct, control, or mitigate the consequences of those abnormal conditions or operations: High train temperature (CFR: 41.5 / 45.6) RO/SRO Rating: 2.9/3.2 Objective: LOI-CLS-LP-10, Obj. 6 - Describe the function of the temperature switches under abnormal operations.
Reference:
None Modified question that was last used on the 10-2 NRC exam. (S261000_1) Changed the temperature reading with panel indications which changed the distractors. Cog Level: High Explanation: f the charcoal area reaches 210°F without the suction temperature greater than 180°F the SBGT Fan will trip. With the Emerg Oper light out this indicates that the suction temperature is less than 180°F so when the 210°F light illuminates the fan will trip. The steps to actuate the deluge system which is normally valved out are contained in the SBGT procedure while the steps to reset the deluge are contained in the fire procedure. Distractor Analysis: Choice A: Correct Answer, see explanation Choice B: Plausible because the fan will trip when the temperature reaches 210°F and the steps for resetting the deluge are contained in the fire procedure. Choice C: Plausible because if the inlet temperature reaches 180°F then the trip is bypassed and the steps to initiate the deluge are contained in this procedure. Choice D: Plausible because if the inlet temperature reaches 180°F then the trip is bypassed and the steps for resetting the deluge are contained in the fire procedure. SRO Basis: Assessment of facility conditions and selection of appropriate procedures during normal, abnormal, and emergency situations (10 CFR 55.43(b)(5))
La da a w a4t - From 20P-10: 8.0 INFREQUENT OPERATIONS 16 8.1 SBGT System Operation to Reduce Humidity 16 8.2 Venting Containment Via the SBGT 18 8.3 Local Deluge System Manual Operation 24 8.4 Filling SBGT Train Drain Trough 27 From OOP-41: SECTION PAGE 8.26 Resetting Unit 2 Main Transformer Phase C Deluge Valve 2-FP-DV3 61 8.27 Resetting Unit 1 Caswell Beach 7.5 MVA Transformer Deluge Valve 2-FP-DV36 63 8.28 Resetting Unit 2 Caswell Beach 7.5 MVA Transformer Deluge Valve 2-FP-DV37 66 8.29 Resetting Unit 1 SBGT A-i Deluge Valve 1-FP-DVA-lAi 69 8.30 Resetting Unit 1 SBGT A-2 Deluge Valve 1-FP-DVA-iA2 71 8.31 Resetting Unit 1 SBGT B-i Deluge Valve 1-FP-DVA-lBi 73 8.32 Resetting Unit I SBGT B-2 Deluge Valve i-FP-DVA-i B2 75 8.33 Resetting Unit 2 SBGT A-i Deluge Valve 2-FP-DVA-2A1 77 8.34 Resetting Unit 2 SBGT A-2 Deluge Valve 2-FP-DVA-2A2 79 8.35 Resetting Unit 2 SBGT B-i Deluge Valve 2-FP-DVA-2B1 81 8.36 Resetting Unit 2 SBGT 8-2 Deluge Valve 2-FP-DVA-282 83 8.37 Resetting DG Air Filter #4 Foam Deluge Valve 2-FP-DV27 85 flh&t/,.
- JIE IIILLL ULI EIi
- 82. S263000 I Unit Two is at rated power when the following conditions are discovered:
December 12 at 12:00, the A RHR pump is declared inoperable. December 14 at 12:00, the C RHR pump is declared inoperable. December 17 at 06:00, the A RHR pump is restored to OPERABLE status. December17 at 08:00, HPCI FIC Power Loss annunciator is received. Including any extensions permitted by Technical Specifications, which one of the following describes the LATEST time to place the unit into MODE 3? (Reference provided) A. December 17 at 21:00 B. December 19 at 24:00 C. December 20 at 20:00 D. December 20 at 24:00 Answer: C K/A: 263000 D.C. Electrical Distribution G2.04.45 Ability to prioritize and interpret the significance of each annunciator or alarm. (CFR: 41.10 /43.5 /45.3 /45.12) ROISRO Rating: 4.1/4.3 Objective: LOl-CLS-LP-01 9, Obj. 26f Given plant conditions and one of the following events, use plant procedures to determine the actions required to control and/or mitigate the consequences of the event: E41-FIC-R600 failure in automatic and/or manual LOI-CLS-LP-019, Obj. 25b Given plant conditions associated with the HPCI system determine the required action(s): to be taken in accordance with Technical Specifications or the TRM (SRO/STA only)
Reference:
Tech Spec 3.5.1 Cog Level: High Explanation: The power loss (24 VDC) to the FIC would cause the HPCI to be declared Inop. When HPCI is declared inoperable, entry into Condition D and Condition E is required. After 72 hours expires (Condition E) which is 12/20 at 08:00, entry into Condition is required. Condition I requires that the unit be in MODE 3 within 12 hours. This is 12/20 at 20:00.
I1aLIIIfI1zJ 2F LII Distractor Analysis: Choice A: Plausible because it assumes entry into Condition J and LCO 3.0.3 when HPCI is declared inoperable. LCO 3.0.3 requires being in MODE 3 within 13 hours of entry. This is 12/17 at 2100. Choice B: Plausible because it assumes entry into Condition C following the 7-day completion time for the inoperable pumps. This is incorrect because a 24-hour completion time extension is permitted for Condition A because the first inoperable component is fixed first. This time without the extension is 12/19 at 12:00. Then entry into Condition C which allows 12 hours to be in MODE
- 3. This time is 12/19 at 2400.
Choice C: Correct Answer, see explanation Choice D: Plausible because it assumes entry into Condition C following the 7-day completion time for the first inoperable pump plus a 24-hour extension for the second pump. This time with the extension is 12/20 at 12:00. Then entry into Condition C which allows 12 hours to be in MODE
- 3. This time is 12/20 at 2400 which is greater than that for Condition E and Condition I (HPCI).
SRO Basis: Facility operating limitations in the TS and their bases.(10 CFR 55.43(b)(2)). Application of Required Actions (Section 3) and Surveillance Requirements (SR) (Section 4) in accordance with rules of application requirements (Section 1). PANEL 3A LOCATION: NORMAL SUPPLY: ALTERNATE SUPPLY: Reference Orawng: LL-3C24-6 Con:toi 2uding 4 Ea Swchbcard 1A N!A CKY* LOAD EEFECT Rx .ru,ar_Ic, -M2-5 1. f&s1aTewzce.PaqEI28c3rcL1tI. Pa,I 525 2. e{je 3rrtjncL3jcr 5-f. 2 HPC 9c4Ontr2f 1. OI,?taE cO%recie E4l-FIC-2 (2 OC1 2. Lca*c1iTw 1,ao,.
- 3. Recet? 3rnJcb1cf A-O1 2-f.
& Lc5athCI
- 5. LcF,qchr.
3.4.9 Power Supplies The normal power supply to Panel P601 HPCI ow instruments, flow indicating controller and pressure transmitters is 125 VcIc Bus A 125 \dC power from Panel 3(4)A powers a 24 Vdc power supply, E41-ES-K603, which supplies power to the Johnson-Yokogawa FIC. Panel 3(4)A also powers a 52.5 VcIc power supply, E4i-ES-K600, which supplies power for the vertical board HPCI pressure instrumentation. HPCI is not affected by a loss of the 515 Vdc power supply except that P601 indication is lost. In the event of a 24 Vdc power supply failure, resulUng in the flow controller failing down scale due to the loss of power, Annunciator HPCI FIC POWER LOSS (APP A-01 2-5) alarms. If the HPCI Turbine is operating, turbine speed would run back to below the 2100 rpm minimum and the system would have to be secured. SD-19 Rev.21 Page30ofi08
aat La .Ew.JJf sat tcI-t wLaU .&3JK.ti. ECCSOperating 3.5.1 3.5 EMERGENCY CORE COOLING SYSTEMS (ECCS) AND REACTOR CORE ISOLATION COOLING (RCIC) SYSTEM 3.5.1 ECCSOperating LCO 3.5.1 Each ECCS injection/spray subsystem and the Automatic Depressurization System (ADS) function of six safetqrelief valves shall be OPERABLE. APPLICABILITY: MODE 1, MODES 2 and 3, except high pressure coolant injection (HPCI) and ADS valves are not required to be OPERABLE with reactor steam dome pressure 150 psig. ACTIONS LCO 3.0.4.b is not applicable to HPCI. CONDITION REQUIRED ACTION COMPLETION TIME A. One low pressure ECCS A. 1 Restore low pressure 7 days injection/spray subsystem ECCS injection/spray inoperable, subsystem to OPERABLE status. OR One low pressure coolant injection (LPCI) pump in each subsystem inoperable. B. One LPCI pump inoperabla 6.1 Restore LPCI pump to 72 hours OPERABLE status. AND OR One core spray (CS) subsystem inoperable. B.2 Restore CS subsystem to 72 hours OPERABLE status.
- 7crrr
ZiLIIIL C. Required Action and C.i Be in MODE 3. 12 hours associated Completion Time of Condition A or B not met. ANQ C.2 Be in MODE 4. 36 hours D. HPCI System inoperable. D. 1 Verify by administrative Immediately means RCIC System is OPERABLE. AND D.2 Restore HPCI System to 14 days OPERABLE status. E. HPCI System inoperable. E.1 Restore HPCI System to 72 hours OPERABLE status. AND OR One low pressure ECCS injectionispray subsystem is E.2 Restore low pressure 72 hours inoperable. ECCS injectionfspray subsystem to OPERABLE status. I. Required Action and 1.1 BeinMODE3. 12 hours associated Completion Time of Condition D, E, F. G, or H AND not met. 1.2 Reduce reactor steam 36 hours OR dome pressure to 150 psig. Two or more required ADS valves inoperable. J. Two or more low pressure J.1 Enter LCO 3.0.3. Immediately ECCS injectionfspray subsystems inoperable for reasons other than Condition A or B. OR HPCI System and two or more required ADS valves inoperable.
i!IIiILi4 f !1II11
- 83. S271000 I During performance of OPT-09.2, HPCI System Operability Test, a HPCI steam line leak occurs. The following conditions exist:
Process OG Vent Pipe Rad Hi - In alarm Process Rx Bldg Vent Rad Hi - In alarm Stack radiation levels Rising SBGT IA Operating Under the conditions listed above, which one of the following choices completes the statement below? SBGT I B (I) auto started and the required action lAW RRCP is to enter (2) A. (1) has (2) RSP B. (1) has (2) OGP-05 C. (1) has not (2) RSP D. (I) has not (2) OGP-05 Answer: D K/A: 271000 Offgas System A2.04 Ability to (a) predict the impacts of the following on the OFFGAS SYSTEM ; and (b) based on those predictions, use procedures to correct, control, or mitigate the consequences of those abnormal conditions or operations: Offgas system high radiation (CFR: 41.5 / 45.6) RO/SRO Rating: 3.7/4.1 Objective: LOI-CLS-LP-300-N, Obj. 1gb, Given plant conditions and OEOP-04-RRCP, determine the following: Required actions to be taken.
Reference:
None Cog Level: High Explanation: High stack radiation levels without associated steam jet high radiation levels is indicative of a radioactive leak source other than the reactor vessel. With steam jets high and stack rad rising, fuel failure would be indicated. The stem does not provide steam jet high rad therefore a transition to the RRCP leg for determining fuel failure or resin intrusion is not required. With high stack rad levels and a primary system discharging, RRCP requires lowering power to limit release rate. If steam jet hi hi rad were provided, a hi stack and rising stack rad with main steam line hi hi would require a manual scram. SBGT auto start function is provided to satisfy the Iç/AIor impacts on the off-gas system since SBGT discharges to the stack and is an off-gas treatment path during accident conditions.
ziriir. iiiivu Distractor Analysis: Choice A: Plausible because SBGT would auto start on a Process OG Vent Pipe Rad Hi-Hi alarm. A scram would be required by SCCP if primary system is discharging and before any area reaches its max safe operating value. RRCP only requires reducing power under this condition. Choice B: Plausible because SBGT would auto start on a Process OG Vent Pipe Rad Hi-Hi alarm. Choice C: Plausible because a scram would be required by SCCP if primary system is discharging and before any area reaches its max safe operating value. RRCP only requires reducing power under this condition. Choice D: Correct answer, see explanation. SRO Basis: Assessment of facility conditions and selection of appropriate procedures during normal, abnormal, and emergency situations.[10 CFR 55.43(b)(5)]. iirniiiitr w
I. I.. I 0 C-) uJ 0 0 LU 0
JI1IIIIIUU A1IL
- 84. S295004 I Unit Two is operating at rated power when the following indications occur:
RCIC LOGIC BUS B PWR FAILURE In Alarm Which one of the following choices completes the following statements? There has been a loss of power to I 25V DC Distribution Panel (1) lAW Technical Specifications LCO 3.5.3, RCIC System, the required action isto (2) A. (1) 3B (2) be in Mode 3 within 12 hours B. (1) 4B (2) be in Mode 3 within 12 hours C. (1) 3B (2) restore RCIC to operable within 14 days D. (1) 4B (2) restore RCIC to operable within 14 days Answer: D K/A: 295004 Partial or Complete Loss of D.C. Power G2.01 .23 Ability to perform specific system and integrated plant procedures during all modes of plant operation. (CFR: 41.10 / 43.5/45.2 / 45.6) RO/SRO Rating: 4.3/4.4 Objective: LOl-CLS-LP-051, Obj. 7d Given plant conditions, determine the effect that a loss of DC power will have on the following: Reactor Core Isolation Cooling LOl-CLS-LP-051, Obj. 16- Given plant conditions and Technical Specifications, including the Bases, TRM, ODCM and COLR, determine the required action(s) to be taken in accordance with Technical Specifications associated with the D.C. Distribution system.
Reference:
None Cog Level: High Explanation: Loss of power from Panel 4B results on a loss of RCIC start capabilities causing RCIC to be inoperable. With HPCI operable, RCIC is required to be returned to operable within 14 days lAW T.S. 3.5.3, condition A.2. If HPCI were not operable, RCIC inoperability would require entering mode 3 within 12 hours lAW T.S. 3.5.3, condition 8.1.
JL Distractor Analysis: Choice A: Plausible because mode 3 would be required within 12 hours if HPCI were also mop. Panel 3B supplies Unit 1 The same alarms would result but on the opposite unit. Choice B: Plausible because mode 3 would be required within 12 hours if HPCI were also mop Choice C: Plausible because Panel 3B supplies Unit 1. The same alarms would result but on the opposite unit. Choice D: Correct answer, see explanation. SRO Basis: Facility operating limitations in the technical specifications and their bases [10 CFR 55.43(b)(2)]. 2APP-03 1-4 Unit 2 APP A-03 1-4 Page 1 of 2 RCIC LOGIC BUS B PR PMLURE NOTE: Inc.perability of this annunciator may result in a TEN Rquirad Cc.mprsatc.ry Measure ATJTO ACTIONS NONE CAUSES L Loss of power to 125V DC Distribution Panel 4B OBSERVATIONS 7 The following valves will go open: Torus Suction Vlv E51-P029 Torus Suction Vlv ES1-P031 ACTIONS 1 Isolate RCIC steam supply in accordance with 2OP-16 Section 8.4, Isolating the RCIC Syatem Steam Supply
3.5.3 RCIC System LCO 35.3 The RCIC System shall be OPERABLE APPLICABILITY: MODE 1, MODES 2 and 3 with reactor steam dome pressure> ISO psig. ACTIONS LCO 3.O.4.b is not applicable to RCIC. CONDITION REQUIRED ACTION COMPLETION TiME A. RCIC System inoperable. A.1 Verify by administrative Immediately means High Pressure Coolant Injection System is OPERABLE. AND A.2 Restore RCIC System to 14 days OPERABLE status. B. Required Action and 8.1 Be in MODE 3. 12 hours associated Completion Time not met. AND 8.2 Reduce reactor steam 36 hours dome pressure to
< 150 psig.
E!rZZrE.ZZE
i1IIILIIIIIIIJL
- 85. S295006 I While performing PT 9.2, HPCI OPERABILITY TEST, the HPCI steam supply line ruptured. HPCI failed to automatically isolate and attempts to manually isolate HPCI are unsuccessful.
The following Steam Leak Detection NUMAC channels are in alarm: B21-XY-5949A, Channel A3-3, reading 303°F B21-XY-5949B, Channel A3-3, reading 298°F B21-XY-5948A, Channel AS-I, reading 301°F B21-XY-5948B, Channel A5-I, reading 296°F No other channels are in alarm. Which one of the following choices completes the statements below? (Reference provided) The required actions lAW SCCP are to Scram the reactor and (1) The highest EAL classification for this event is (2) A. (1) emergency depressurize (2) a Site Area Emergency B. (1) commence a cooldown at normal rates (2) a Site Area Emergency C. (1) emergency depressurize (2) an Alert D. (1) commence a cooldown at normal rates (2) an Alert Answer: B K/A: 295006 SCRAM G2.01 .25 Ability to interpret reference materials, such as graphs, curves, tables, etc. (cFR: 41.10 /43.5 /45.12) ROISRO Rating: 3.9/4.2 Objective: LOl-CLS-LP-300-M, Obj. 1 3a Given plant conditions and the Secondary Containment Control Procedure, determine the required actions if the following limits are exceeded: Maximum Safe operating values with a primary system discharging into Secondary Containment.
Reference:
OEOP-01-UG, Figure 22; OPEP-02.1 Cog Level: High
wLJIIjg1vjJJivyAL, L Explanation: SCCP requires a reactor scram if a primary system is discharging into the reactor building and before any area exceeds its max safe operating value. A cooldown at normal cooldown rates follows that unless more than one area exceeds its max safe operating value within the same table in which case emergency depressurization is required.. B21-XY-5949A1B Ch. A3-3 and B2i -XY-5948A1B are separate temperatures for RCIC and HPCI steam tunnels but are both in the same area. Both values are exceeding max safe, however since they are in the same area emergency depressurization is not required. EAL is based on the Fission Product Barrier matrix of OPEP-02.i. In the conditions provided, Table F-i C. Isolation for Containment barrier would apply for Containment Barrier Loss, RCS Barrier Loss and RCS Barrier Potential Loss. FU i.i UE would apply for Containment Barrier Loss. FA1.1 (Alert) would apply for RCS Barrier Potential Loss and FS1 .i (Site Area Emergency) applies for Loss or Potential Loss of any two barriers (in this case, containment and RCS) and is the highest classification for this event. Distractor Analysis: Choice A: Plausible because ED is required if primary system is discharging AND two or more areas have exceeded their max safe values. The temperatures provided are related to two separate systems but share a common area. If Figure 22 is not interpreted correctly, the candidate could assume two areas are exceeding and call for ED. Choice B: Correct Answer, see explanation Choice C: Plausible because ED is required if primary system is discharging AND two or more areas have exceeded their max safe values. The temperatures provided are related to two separate systems but share a common area. If OEOP-Oi -UG, Figure 22, is not interpreted correctly, the candidate could assume two area temperatures are exceeded and call for ED. OPEP-02.i, Fission Product Barrier Matrix provides information for the EAL call. A Site Area Emergency is appropriate for the stated conditions based on Loss of potential loss of any two barriers (Table F-i RCS Barrier C.3 and C. 2 and Containment Barrier C.3 and C.5). The highest EAL is Site Area Emergency. If only the RCS barrier is identified, an Alert may be selected. Choice D: Plausible because OPEP-02.1, Fission Product Barrier Matrix provides information for the EAL call. A Site Area Emergency is appropriate for the stated conditions based on Loss of potential loss of any two barriers (Table F-i RCS Barrier C.3 and C. 2 and Containment Barrier C.3 and C.5). If only the RCS barrier is identified, an Alert may be selected. SRO Basis: Assessment of facility conditions and selection of appropriate procedures during normal, abnormal, and emergency situations.[iO CFR 55.43(b)(5)]. Use of EOPs and the appropriate interpretation of EALs is required.
OEOP-01 -UG ATTACHMENT 10 Page 2 of 5 Secondary Containment Temperature and Radiation Limits FIGURE 22 Secondary Containment Area Temperature TABLE I AREA TEMPERATURE LIMITS PLANT PLANT STEAM LEAK INSTRUMENT MAX NORM MAX SAFE AUTO AREA LCCATION DETECTION NUMBER! OPERATING OPERATING GROUP DESCRIPTION CHANNEL! WINDOW VALUE F1 VALUE (°F) ISOL LOCATION (NOTE 1) N CORE N CORE PANEL XU-3 VA-TI-i 603 120 175 NIA SPRAY SPRAY ROOM S CORE S CORE PANEL XU-3 VA-TI-16D4 120 175 WA SPRAY SPRAY ROOM RWCU PUMP B21-XY-5949A G31-TE-NO16A ROCM A B21-XY-5949B G31-TE-N0165 CH. Al-i RWCU PUMP B21-XY-5949A G31-TE-NO16C RWCU 140 225 3 ROOM B B21-XY-5949B G31-TE-NOI6D CN. A2-1 RWCU HX B21-XY-5949A G31-TE-NO16E ROOM B21-XY-5949B G31-TE-NOI6F CH. A3-1 NRHR B21-XY-5946A E11-TE-NOD9A N RNR EQUIP ROOM CN. A5-4 175 295 NIA PANEL XU-3 VA-TI-i 601 S RHR B21-XY-59486 El 1-TE-N0095 EQUIP ROOM CN. A5-4 175 295 WA PANEL XU-3 VA-TI-i 602 S RNR RCIC EQUIP B21-XY-5949A ES1-TE-N023A ROOM 821-XY-5949B E51-TE-N0235 165 295 5 CN. A1-3 NPCI HPCI EQUIP B21-XY-5946A E41-TE-NOSOA ROOM B21-XY-5948B E41-TE-N0305 165 4 CN. Al-i NOTE 1: MAX NORM OPERATING /ALUE IS THE ANNUNCIATOR!GROUP ISOLATION SETPOINT WHERE APPLICABLE ry, EElS
-a W Siitta (it eS WITS as ATTACHMENT 10 Page 3 of 5 Secondary Containment Temperature and Radiation Limits FIGURE 22 Secondary Containment Area Temperature (Cant)
TABLE 1 AREA TEMPERATURE LIMITS PLANT PLANT STEAM LEAK INSTRUMENT MAX NORM MAX SAFE AUTO AREA LOCATION DETECTION NUMBER! OPERATING OPERATING GROUP DESCRIPTION CHANNEL! WINDOW VALUE (F VALUE (°F: ISOL LOCATION (NOTE 1 RCIC STM 521-XY-5949A E51-TE-N025A TUNNEL 521-XY-59495 E51-TE-N0255 190 295 5 STEAM c. AS-S TUNNEL HPCI STM 521-XV-5945A E51-TE-NO2SC TUNNEL 521-XY-59455 E51-TE-N025D 190 295 4 CH. AS-i 20 FT NORTH 521-XY-5945A Chz. Al-a 521-TE-576IA I 20 FT 2C FT SOUTH 521-XY-59485 521-TE-57638 140 200 N/A CH. Al-a 50 FT 50 FT NW 521-XY-5948A 52 1-TE-5762A CH.A2-4 140 200 N:A 50 FT SE 52i-X-5948B 621-TE-S7648 CH. A2-4 REACTOR MULTIPLE AN NUNCIATOR WINDOW ALARM N/A 3.4, BLDG AREAS PANEL A-02 5-7 SETPOINT AND/OR REACTOR MSIV ANNUNCIATOR WINDOW ALARM N/A BLDG PIT PANEL A-OS 6-7 SETPOINT NOTE 1: MAX NORM OPERATING VALUE IS THE ANNUNCIATOR/GROUP ISOLATION SETPOINT WHERE APPLICABLE a
- 86. S295010 1 Unit One CRS has entered 0AOP-14.0, Abnormal Primary Containment Conditions.
Current primary containment conditions are: Drywell pressure 7 psig, slowly rising Suppression Pool Pressure 4 psig, slowly rising Drywell temperature 145° F, slowly rising Suppression Pool Level -28 inches, stable Which one of the following choices completes the statement below? The CRS will (1) and order (2) A. (1) exit 0AOP-14.0 and enter PCCP (2) SEP-02, Drywell Spray Procedure B. (1) exit OAOP-14.0 and enter PCCP (2) SEP-03, Suppression Pool Spray Procedure C. (1) execute OAOP-14.0 and PCCP concurrently (2) SEP-02, Drywell Spray Procedure D. (1) execute OAOP-14.0 and PCCP concurrently (2) SEP-03, Suppression Pool Spray Procedure Answer: B K/A: 295010 High Drywell Pressure G2.01 .20 Ability to interpret and execute procedure steps. (CFR: 41.10 / 43.5 / 45.12) RO/SRO Rating: 4.6/4.6 Objective: LOI-CLS-LP-300-L, Obj. 02- Given plant conditions, determine if the Primary Containment Control Procedure should be entered. LOI-CLS-LP-300-L, Obj. 05b Given the Primary Containment Control Procedure, determine the appropriate operator actions if any of the following limits are approached or exceeded: Suppression Chamber Spray Initiation Pressure Limit
Reference:
None Cog Level: High Explanation: If drywell pressure is greater than 1.7 psig then AOP-14 is exited and PCCP is entered as the definition from the user guide states for GO TO. The operator then has to select the appropriate procedure to be implemented, SEP-02 or 03. Spraying the torus is perfromed before chamber pressure reaches 11.5 psig, while drywell sprays are performed when suppression chamber reaches 11.5 psig. With the given indications the students will be within the DW Spray Initiation Limit Graph.
I& && LJII1JIIL Distractor Analysis: Choice A: Plausible because the conditions are within the DW spray graph but lAW the DW pressure leg of PCCP cannot spray until suppression chamber reaches 11.5 psig. Choice B: Correct Answer, see explanation Choice C: Plausible because AOP-14 is the only AOP that is exited, others are performed concurrently with the EOP. The conditions are within the DW spray graph but lAW the DW pressure leg of PCCP cannot spray until suppression chamber reaches 11.5 psig. Choice D: Plausible because AOP-14 is the only AOP that is exited, others are performed concurrently with the EOP. SRO Basis: Assessment of facility conditions and selection of appropriate procedures during normal, abnormal, and emergency situations (10 CFR 55.43(b)(5)) t7 CAN PSIG PRIMARY CTMT PRESS YES BE MAINTAINED BELOW 1.7 PSIG PCIP-04
/
/ CAUTION REDUCING PRIMARY CONTAINMENT
\\
\ PRESSURE WILLREDUCE NPSK I
\ FOR PUMPS TAKING SUCTION FROM /
\ THESUPPRESSION POOL
/
PC/P-05 115 I PSIG SUPPRESSION CHAMBER PRESS REACHES 11.5 PS1G INITIATE SUPPRESSION POOL SPRAY PER EOP-O1-SEP-03 EXCEPT RHR PUMPS REQUIRED FOR ADEQUATE CORE COOLING BY CONTINUOUS OPERATION IN LPCI MODE
iIZJIUI NO SUPPRESSION CHAMBER PRESS ABOVE 115 PSIG PC/P- (lB
/ CAUTION
(
/ SUCTION OPERATION OF RHR WITH FROM SUPPRESSION )
\ POOLAND PUMP FLOWABOVEJ
\ NPSI-{ OR VORTEX LIMIT MAY!
\SULT IN EQUIP DAMAGE
/
PCIP-09 IS NO SUPPRESSION POOL WATER LEVEL BELOW
+21 INCHES PC/P-b YES IS DRYWELL NO SPRAY INITIATION LIMIT IN SAFE REGION PC/P-Il YES INITIATE DRYWE LL SPRAYS PER EOP- 01-SEP-02 EXCEPT RHR PUMPS REQUIRED FOR ADEQUATE CORE COOLING BY CONTINUOUS OPERATION IN LPCI MODE From OAOP-14.O:
- 3. IF Drywell pressure is greater than 1.7 psig, THEN GO LI TO OEOP-02-PCCP.
From the EOP Users Guide: GOTO This term is used when branching from one procedure to another. When the term GO TO is encountered, the operator is to exit the current procedure and execute the referenced procedure.
DRYWELL AVERAGE AIR TEMPERATURE (°F) cJ Ui Ui UI UI I
- 87. S295019 1 Unit Two is operating at rated power when a PNS leak results in the following indications:
PNS SYS DIV I HDR PRESS LOW PNS SYS DIV/I HDR PRESS LOW Drywell pneumatics have been transferred lAW OAOP-20.O, Pneumatic (Air/Nitrogen) System Failures. Which one of the following choices completes the statements below? RNA is supplying pneumatics to (1) lAW OAOP-20.O the required action of Technical Specifications LCO 3.6.3.1, Primary Containment Oxygen Concentration, is to (2) (Reference provided) A. (1) SRVs and MSIVs (2) restore the oxygen concentration limit within 24 hours B. (1) SRVs and MSIVs (2) reduce THERMAL POWER to <15% RTP within 8 hours C. (1) MSIVs ONLY (2) restore the oxygen concentration limit within 24 hours D. (1) MSIVs ONLY (2) reduce THERMAL POWER to <15% RTP within 8 hours Answer: A K/A: 295019 Partial or Complete Loss of Instrument Air AA2.02 Ability to determine and/or interpret the following as they apply to PARTIAL OR COMPLETE LOSS OF INSTRUMENT AIR: Status of safety-related instrument air system loads (see AK2.1 AK2.19) (CFR: 41.10 / 43.5 / 45.13) RO/SRO Rating: 3.6/3.7 Objective: LOI-CLS-LP-020, Obj. 1 5a Given plant conditions, predict how ADS/SRVs will be affected by the following: Loss of Non-Interruptible Air to the Drywell.
Reference:
TS 3.6.3.1 Cog Level: High Explanation: PNS is the normal pneumatic source for all the drywell components when in Mode 1 and reactor power is> 15 % power. AOP-20 directs aligning RNA to supply drywell pneumatics. Since instrument air pressure (PNS) has not lowered to the point that backup nitrogen would initiate, RNA would then be supplying all drywell loads including SRVs. AOP-20 also directs entry into tech spec 3.6.3.1 condition B which requires reducing thermal power to 15% power within 8 hours.
ip V!I[Liii Distractor Analysis: Choice A: Correct answer, see explanation. Choice B: Plausible if the student believes that AOP-20 drives them to Condition B of T.S. 3.6.3.1 instead of Condition A. Choice C: Plausible because RNA is aligned to supply all drywell pneumatic components including MSIVs when reactor power is below 15%. T.S. 3.6.3.1, condition A, is referenced in AOP-20. Choice D: Plausible because RNA is aligned to supply all drywell pneumatic components including MSIVs when reactor power is below 15% and if the student believes that AOP-20 drives them to Condition B of T.S. 3.6.3.1 instead of Condition A. SRO Basis: Facility operating limitations in the technical specifications and their bases [10 CFR 55.43(b)(2)]. OAOP20 3.22 IF the PNS System is lost or degrading, THEN PERFORM the following:
- 4. IF thermal power is greater than 15% AND the PNS System is NOT aligned to the drywefl, THEN ENTER the action statement of Technical Specrtication 3.6.3.1 as if the limit has been exceeded. j T.S. 3.6.3.1 3.6 CONTAINMENT SYSTEMS 3.6.31 Primary Containment Oxygen Concentration LCO 16.3.1 The primary containment oxygen concentration shall be < 4.0 volume percent.
APPLICABILITY: MODE 1 during the time period:
- a. From 24 hours after THERMAL POWER is> 15% RTP following startup. to
- b. 24 hours prior to a scheduled reduction of THERMAL POWER to
<15% RTP.
ACTIONS CONDITION REQUIRED ACTION COMPLETION TIME A. Primary containment oxygen A.1 Restore oxygen 24 hours concentration not within concentration to within limit. limit. B. Required Action and B.1 Reduce THERMAL S hours associated ornp[etion Time POWER to 15% RTP. not met. iit? c
- 88. S295025 1 A reactor scram on Unit One has resulted in the following plant conditions:
Reactor pressure 825 psig Reactor pressure band 800 1000 psig Reactor Water Level 120 inches Control Rod position All unknown SRVs One open Drywell pressure: 3 psig Supp. Pool pressure: 2 psig Supp. Pool water temp: 160°F Supp. Pool water level: -3 feet Which one of the following choices completes the statements below? (Reference Provided) The procedure required for pressure control is (1) The CRS will direct the RO to (2) the current reactor pressure band. A. (1) RVCP (2) lower B. (1) LPC (2) lower C. (1) RVCP (2) maintain D. (1) LPC (2) maintain Answer: B K/A: 295025 High Reactor Pressure EA2.03 Ability to determine and/or interpret the following as they apply to HIGH REACTOR PRESSURE: Suppression pool temperature (CFR: 41.10 /43.5 /45.13) RO/SRO Rating: 3.9/4.1 Objective: LOI-CLS-LP-300-C, Obj. 7 Given plant conditions and the Reactor Scram Procedure, determine if branching into the Level/Power Control Procedure is required. LOI-CLS-LP-300-L, Obj. Given Primary Containment Control Procedure, determine the appropriate operator actions if any of the following limits are approached or exceeded: Heat Capacity Temperature Limit.
Reference:
EOP-01-UG, Attachment 5, Figure 3, Heat Capacity Temperature Limit Cog Level: High
AIIZTII!F12J7 az I Explanation: Suppression pool temperature and water level determine the amount of heat energy that can be absorbed from a blowdown of the reactor. Heat Capacity Temperature Limit (HCTL) graphs that relationship. Maintaining parameters within the safe region of the graph ensures adequate capacity for a blow down. For suppression pool level, the graph line irmediately below suppression pool level is selected for use. With a Suppression Pool level of -4 feet, the
-4.25 ft line would be used. For the parameters given, the HCTL+/-s outsidethe safe region.
With control rod positions unavailable, RSP directs performance of LPC. Distractor Analysis: Choice A: Plausible because if APRMs are downscale AND control rod position can be determined RVCP would be entered. However, lAW RSP if control rod position cannot be determined LPC will be entered. Choice B: Correct Answer, see explanation Choice C: Plausible because incorrect suppression pool level line could be used resulting in a safe determination. If APRMs are downscale AND control rod position can be determined RVCP would be entered. However, lAW RSP if control rod position cannot be determined LPC will be entered. Choice D: Plausible because incorrect suppression pool level line could be used resulting in a safe determination. SRO Basis: Assessment of facility conditions and selection of appropriate procedures during normal, abnormal, and emergency situations.[10 CFR 55.43(b)(5)J.
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- Distractor Analysis:
Choice A: Correct answer, see explanation. Choice B: Plausible because RBCCW leak could provide rising room levels, but area Rad and Temperature would not rise. Choice C: Plausible because the SDV is correct and the OP does not provide for bypassing the hi drywelll pressure isolation signal. SCCP directs to use the OP or the SEP. Choice D: Plausible because RBCCW leak could provide rising room levels, but area Rad and Temperature would not rise and the OP does not provide for bypassing the hi drywell pressure isolation signal. SCCP directs to use the OPor the SEP. SRO Basis: Assessment of facility conditions and selection of appropriate procedures during normal, abnormal, and emergency situations.[10 CFR 55.43(b)(5)]. 001-37.9 STEPS SCCPaII and SCCP-12 ( SOCI- 12 If no plant conditions currently exist which would automatically isolate the Reactor Building [IVAC System, the appropriate procedure to execute is OP-37.1. In the event OP-37. 1 is executed and the HVAC System isolates, an assessment of plant conditions should be made before any further attempts are made to restart the H/AC System. Defeating high drywell pressure and low reactor water level isolation interlocks is appropriate, if needed, since application of these isolations to Reactor Building HVAC is for the sole purpose of limiting radioactivity release to the environment. Once assurance is provided that an excessive release of radioactivity will not occur, these Iwo isolation interlocks become dispensable. The Reactor Building [IVAC Restart Procedure (EOP-01-SEP-04) provides detailed instructions on bypassing these interlocks. The reactor building ventilation exhaust high temperature isolation is not bypassed since the radiation detectors are not qualified for operation at high temperatures which could exist following a high energy line break in the Reactor Building. If the high temperature isolation was defeated, the radiation monitors could not be relied upon to isolate the building to secure a subsequent release.
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- 89. S295033 I Following a Reactor Scram on Unit Two due to a loss of off-site power, the following plant conditions exist:
Area Rad R)( Bldg High In alarm South RHR RM Flood Level Hi In alarm South CS RM Flood Level Hi In alarm Reactor Building 20 Rad Level Rising Reactor Building 20 Temperature Rising Reactor Water Level 150 inches and stable Drywell Pressure 3.1 psig Which one of the following choices completes the statements below? The source of the leak is from (1) and the CRS will direct the restart of RB HVAC lAW (2) A. (1) SDV (2) SEP-04, Reactor Building HVAC Restart Procedure B. (1) RBCCW (2) SEP-04, Reactor Building HVAC Restart Procedure C. (1) SDV (2) OP-37.l, Reactor Building Heating and Ventilation System Operating Procedure D. (1) RBCCW (2) OP-37.1, Reactor Building Heating and Ventilation System Operating Procedure Answer: A K/A: 295033 High Secondary Containment Area Radiation Levels EA2.03 Ability to determine and/or interpret the following as they apply to HIGH SECONDARY CONTAINMENT AREA RADIATION LEVELS: tCause of high area radiation (CFR:41.10/43.5/45.13) RO/SRO Rating: 3.7/4.2 Objective: LOI-CLS-LP-300-M, Obj. 11 Given plant conditions involving Reactor Building HVAC system isolation and the Secondary Containment Control Procedure, determine if the Reactor Building HVAC system should be restarted.
Reference:
None Cog Level: High Explanation: Reactor scram due to LOOP providing indications of SDV rupture. The Maximum Normal Operating Values are the highest radiation levels expected to occur during normal plant operating conditions with all directly associated support and control systems functioning properly. Separate radiation levels are provided for each Secondary Containment area. Flood level Hi is MNOWL entry condition to SCCP. LOOP automatically provides groups 1,2,3,6,8,& 10 isolations. No RWCU (Grp 3) isolation failure provided in stem) Based on flood level status, along with rising 20 temperature and radiationids to SDV rupture. RB HVAC restart based on hi drywell pressure will have to use the SEP vs the OP in order to place the jumpers to bypass this signal.
- 90. S295037 1 Unit Two is operating at rated power when a complete loss of UPS occurs.
The CRS has directed the RO to insert a manual reactor scram due to rising drywell pressure. Plant conditions are: Manual Scram pushbuttons Depressed Mode Switch Shutdown position RPS Lights NOT lit ARI Initiated Drywell pressure 2.1 psig APRM Downscales NOT lit Scram Valve P11 Air Hdr Press Hi/Lo In alarm Which one of the following choices completes the statement below? The CRS will direct the RO to perform Section of LEP-02, Alternate Control Rod Insertion. A. 2, De-energize the Scram Pilot Valve Solenoids B. 3, Reset RPS and Initiate a Manual Scram C. 4, SCRAM individual rods D. 5, Insert Control Rods with the Reactor Manual Control System Answer: B K/A: 295037 SCRAM Condition Present and Reactor Power Above APRM Downscale or Unknown EA2.05 Ability to determine and/or interpret the following as they apply to SCRAM CONDITION PRESENT AND REACTOR POWER ABOVE APRM DOWNSCALE OR UNKNOWN: Control rod position (CFR: 41.10/43.5/4513) RO/SRO Rating: 4.2/4.3 Objective: LOl-CLS-LP-300-J Obj. 5, Given plant conditions determine which sections of the Alternate Control Rod Insertion Procedure should be utilized for Control Rod Insertion in accordance with EOP-01 -LEP-02.
Reference:
None Cog Level: High Explanation: With UPS lost there is no power for rod selection so the operator cannot insert control rods using RMCS. Rods will have to be attempted to be inserted using LEP-02 section 3 (Scram reset Scram). APRM recorders do not have power so a power reading is not available. The APRM downscale lights (on the apron) still have power. With the RPS lights out and Air header pressure alarm in Section 2 would not be performed. \Nith hi drywell pressure the scram cannot be reset without jumpers so section 7 cannot be performed.
Distractor Analysis: Choice A: Plausible because if the RPS lights were still lit or air header pressure alarm was not in (electrical ATWS) this would be correct. Choice B: Correct Answer, see explanation. Choice C: Plausible because this is would be correct if DW pressure was not high allowing the scram to be reset without the use of jumpers, Scram cannot be reset directs going back to another section. Choice D: Plausible because this is would be correct if rod select power was available. SRO Basis: Assessment of facility conditions and selection of appropriate procedures during normal, abnormal, and emergency situations (43(b)(5) Knowledge of diagnostic steps and decision points in the EOPs that involve transitions to emeency contingency procedures PERFORM ALTERNATE CONTROL ROD INSERTION (EOP. 01- LEP- 02)
- RCIQ-15
!IMII7iA IIIIIIIIIIiI4
- 91. S500000 I Unit Two is operating at rated power when a LOCA and a LOOP occur.
The following conditions exist: Reactor water level 100 inches rising Drywell Pressure 32 psig Suppression Pool Pressure 29 psig Suppression Pool Level 18 inches DG3 Tripped on differential overcurrent DG4 Running loaded ERFIS Available lAW PCCP, for these conditions, which one of the following choices completes the statements below? Primary containment hydrogen/oxygen concentration must be determined by (1) When venting of primary containment has been established the CRS will direct the purging of the (2) lAW SEP-05, Primary Containment Purging. A. (1) placing CAC-AT-4410 in service (2) Drywell or Suppression Chamber B. (1) placing CAC-AT-4410 in service (2) DrywellONLY C. (1) E&RC sample ONLY, CAC-AT-4410 cannot be used (2) Drywell or Suppression Chamber D. (I) E&RC sample ONLY, CAC-AT-44I0 cannot be used (2) DrywellONLY Answer: D K/A: 500000 High Containment Hydrogen Concentration EA2.01 Ability to determine and / or interpret the following as they apply to HIGH PRIMARY CONTAINMENT HYDROGEN CONCENTRATIONS: Hydrogen monitoring system availability (CFR: 41.10/43.5/45.13) RO/SRO Rating: 3.1/3.5 Objective: LOI-CLS-LP-300-L, Obj. 11 Given Primary Containment Control Procedure, which steps have been completed and plant parameters, determine the required operator actions.
Reference:
None Cog Level: High rI7Iw IIIi1iII
3L I Explanation: Unit two H2/02 monitors CAC-AT-4409 and 4410 are powered from 120V emergency distribution panels 32A and 2B respectively. 32A is powered from Div I AC. On a LOOP with subsequent failure of DG3, power is lost to the 4409. Additionally, with drywell pressure
> 30 psig, PCCP directs isolation of both the 4409 and 4410 (ref PCIP-17). So, while the 4410 has power available, it cannot be used by procedure. PCCP PC/H-05 asks if 4409 or 4410 are available and if not, E&RC is notified to obtain containment samples. Use of Attachment 12 of EOP-01-UG is not required (ref PCCP PC/H-04) because ERFIS is available but primarily because PCCP does not require operators to perform this if E&RC sampling is obtained. It torus level is greater than 1 foot purging is not allowed. With water in the purge line the possibility exists that the purge containment isolation valves may fail to shut when required to secure the purge Distractor Analysis:
Choice A: Plausible because CAC-AT-441 0 has power available but cannot be used because a drywell pressure of 32 psig would require the CAC-AT-4409 and 4410 to be isolated. Suppression Chamber can not be purged with torus level at >1 foot. Choice B: Plausible because CAC-AT-4410 has power available but cannot be used because a drywell pressure of 32 psig would require the CAC-AT-4409 and 4410 to be isolated. Choice C: Plausible because E&RC sampling is required. Suppression Chamber cannot be purged with torus level at >1 foot. Choice D: Correct answer, see explanation. SRO Basis: Assessment of facility conditions and selection of appropriate procedures during normal, abnormal, and emergency situations.[10 CFR 55.43(b)(5)]. This question requires knowledge of the decision points in EOP that involve transition to EOP contingency procedures. PCCP 30 1 F DRYWELLPRESSIS A8OVEOPSIG P SIG ThEN SECIJREAND ISOLATE CAC-AT-4409 AND 4410 pc!p-1r
3UVWL L 4 III1ZIIh PCIH
/
PCIH- 01 PLACE 2 ANALYZERS CACAT-4400 AND 4410 IN SERVICE (OP24) PCiN- 02 MONFTORAND CONTROL \ PRIMARY CTMT HYDROGEN AND OXYGEN CONCEIbIIRATIONS 1 PCH-03 H2ANDO2READINGSARE COMPENSATED BY ERFIS. H2ANDO2 READINGS ON CAC-AT- 4400 AND 4410 MUST BE COMPENSATED FOR PRIMARY CTMT CONDITIONS USING ATTACHMEIIT 12 OF
/
USERSGUIDE(EOP-01-UG) / PCJH-04 H2K2ANALES
-L CAC-AT-4409 CR4410 PCIH- 00 NO NOTIFY E&I TO SAMPLE THE DRYWELLAND SUPPRESSION CHAMBER FOR HYDROGEN AND OXYGEN CONCENTRATIONS PCIH- 06 PCIIl. 19 Suppression chamber purge is not allowed with the water level above the elevation of the purge penetration (÷1 foot). With water in the purge line the possibility exists that the purge containment isolation valves may fail to shut wtlen required to secure the purge. The drywell is purged instead.
001-37.8 Rev. 4 Page 55 of 56
- 92. S600000 1 Surveillance testing of fire protection components on Unit Two have yielded the following results:
North RHR room sprinkler system Inoperable South RHR room sprinkler system Operable
-17 RHR Pump areas Smoke Detectors 1-4, 1-9, 3-6, and 3-9 are Inoperable Fire Barriers All operable Which one of the following choices identifies the required fire watches lAW OPLP-1 .2, Fire Protection System Operability, Action, and Surveillance Requirements?
(Reference provided) A. Establish an hourly fire watch in North RHR and a continuous firewatch in South RHR. B. Establish a continuous firewatch in North RHR and an hourly firewatch in South RHR. C. Establish a continuous fire watch in both North and South RHR. D. Establish an hourly fire watch in both North and South RHR. Answer: B K/A: 600000 Plant Fire On Site AA2.15 Ability to determine and interpret the following as they apply to PLANT FIRE ON SITE: Requirements for establishing a fire watch (CFR: 41.10 /43.5 /45.13) RO/SRO Rating: 3.0/3.6 Objective:
Reference:
OPLP-01.2, Fire Protection System Operability, Action, and Surveillance Requirements: Section 6.2, Spray and/or Sprinkler Systems; Section 6.8, Fire Detection Instrumentation; Attachment 1, Required Water Spray/Sprinkler Systems; Attachments 4, Required Fire Detection Instruments. Cog Level: High Explanation: Sprinkler systems are designated in OPLP-1 .2 as Low or High Safety Significance (LSS or HSS) (ref. OPLP-1 .2, Attachment 1). For HSS systems, a continuous fire watch is required if the system is inoperable. For LSS systems, an hourly fire watch is required IF fire barriers and fire detection in the same area are operable. OPLP-1 .2 requires a minimum of four detectors in both the North and South RHR pump areas. There are a total of five instruments in each area. Detectors 1-4 and 1-9 are in NRHR and their inoperability results in inoperability of the detector system for that area. With the NRHR sprinkler system inoperable, a continuous firewatch is required lAW OPLP-1 .2, 6.2.2.b. Detectors 3-6 and 3-9 are in SRHR and their inoperability results in inoperability of that area system (ref. OPLP-1 .2, Attachment 4). Since no other components are inoperable in that area, an hourly fire watch is required lAW OPLP-1.2, 6.8.2.1.
LiFiUIZI[I1IC Distractor Analysis: Choice A: Plausible because an hourly fire watch in NRHR would be the correct answer if only the detection instrumentation were mop in that area. A continuous fire watch in SRHR would be the correct answer if the sprinkler system were also mop in that area. Choice B: Correct Answer, see explanation Choice C: Plausible because if fire detection instrumentation failures are not correctly a determination could be made for the continuous watch in both areas. Choice D: Plausible because if fire detection instrumentation failures are not correctly a determination could be made for the one hour watch in both areas. SRO Basis: Conditions and limitations in the facility license. [10 CFR 55.43(b)(1)}. Administration of fire protection program requirements such as compensatory actions associated with inoperable sprinkler systems, fire doors, etc. Fire watch requirements are contained in OPLP-01 .2, Fire Protection System Operability, Action, and Surveillance Requirements, which is the Brunswick site technical specification equivalent for Fire Protection Systems. This procedure outlines the operability, action, and surveillance requirements for fire protection systems required for safe shutdown in the event of a fire. This procedure implements license renewal commitments and supports the License Renewal Aging Management Program. OPLP-1.2 1.0 PURPOSE The purpose of this procedure is to outline the operability, action, and surveillance requirements for fire protection systems required for safe shutdown in the event of a tire. It also addresses the surveillance requirements for these systems which fall under the requirements of the Nuclear Electric Insurance Limited (NEIL) Loss Prevention Program. I RI 51 This procedure is required to implement license renewal commitments and supports the License Renewal Aging Management Program. 2.15 BSEP 04-0006, Letter from BSEP to US Nuclear Regulatory Commission, I RI 51 dated October 18, 2004, Application for Renewal of Operating License
I 6.2.2 Actions
- 1. Verify the operability of the tire barriers and fire detection in the area and determine if the spray and/or sprinkler is classified as High Safety Significant (HSS) or Low Safety Significant (LSS)
- a. With one or more of the HSS spray and/or sprinkler systems shown in Attachment 1 inoperable, establish a continuous fire watch ith backup fire suppression equipment for the unprotected area(s) within 1 hour and restore the system to OPERABLE status Within 14 days.
- b. With one or more of the LSS spray and/or sprinkler system inoperable concurrent vyith a fire detection system or fire barrier in the area inoperable, establish a continuous fire watch with backup fire suppression equipment for the unprotected area(s) within 1 hours and restore the system to OPERABLE status within 14 days.
- c. With one or more of the LSS spray and/or sprinkler systems shown in Attachment 1 inoperable with the fire barriers and fire detection in the area OPERABLE, establish an hourly fire watch with backup fire suppression equipment for the unprotected area(s) within 1 hour and restore the equipment to OPERABLE status within 14 days.
ATTACHMENT 1 Page 2 of 2 Required Water SpraylSprinkler Systems REACTOR BUILDING 2 North and South Core Spray Pump Rooms (LSS) North (LSS) and South (HSS) RHR Rooms 5-O Separation Zone Water Curtain (HSS) 20 Southwest Sprinkler System (LSS) 20 East Separation Zone Water Curtain (HSS) 20 Railroad Bay Sprinkler System (LSS) 20 ECCS Room Sprinkler Head (HSS) 38 Separation Zone Water Curtain (HSS) 50 Sprinkler System (Near Elevator) (LSS) 50 Separation Zone Water Curtain (HSS) 80 Sprinkler System (LSS) Two Standby Gas Treatment Train 2A Deluge Systems (LSS) Two Standby Gas Treatment Train 26 Deluge Systems (LSS) r wn
4 L sS&.JSk JEL m&...aL LsFle aJtWLwar 6.8.2 Actions With one or more of the required lire detection instrument(s) shown in Attachment 4 inoperable:
- a. Within one hour verity the operability of tire barriers and cable coatings in the area covered by the affected zone and the operability of spray and/or sprinkler systems in the area covered by the affected zone and establish an hourly fire watch patrol.
- b. Re-evaluate Steps 6.2.2.1.b, 6.6.2.1.a and 6.10.2.1 for each inoperable spray and/or sprinkler system, fire barrier and cable coating in the area and implement revised lire watch requirements to include consideration of the inoperable tire detection instrument(s).
ATTACHMENT 4 Page 3ot7 Required Fire Detection Instruments MINIMUM INSTRUMENTS OPERABLE DETECTOR ZONE! BUILDING AREA FLAME HEAT SMOKE
- 2. Reactor Building #2
(-)17 N Core Spray Pump Room 0 0 Detectors 1-1, 1-2 (-y17 S Core Spray Pump Room 0 0 1 Detectors 3-1, 3-2 (-)17 N RHRICRD Pump Area 0 0 4 Detectors 1-4, 1-5, 1-6, 1-8, 1-9 (-)17 S RHR Pump Area 0 0 4 Detectors 3-4, 3-5, 3-6, 3-8, 3-9 L j
6.2 Spray and/or Sprinkler Systems 6.2.1 Operability The spray and/or sprinkler systems shown in Attachment I shall be OPERABLE whenever the equipment in the areas protected by the spray and/or sprinkler system is required to be OPERABLE, and is capable of performing its intended function. 6.2.2 Actions Verify the operability of the fire barriers and fire detection in the area and determine if the spray and/or sprinkler is classified as High Safety Significant (HSS) or Low Safety Significant (LSS)
- a. With one or more of the HSS spray and/or sprinkler systems shown in Attachment I inoperable, establish a continuous fire watch with backup fire suppression equipment for the unprotected area(s) within 1 hour and restore the system to OPERABLE status within 14 days.
- b. With one or more of the LSS spray and/or sprinkler system inoperable concurrent with a fire detection system or fire barrier in the area inoperable, establish a continuous fire watch with backup fire suppression equipment for the unprotected area(s) within I hours and restore the system to OPERABLE status within 14 days.
- c. With one or more of the LSS spray and/or sprinkler systems shown in Attachment I inoperable with the fire barriers and fire detection in the area OPERABLE, establish an hourly fire watch with backup fire suppression equipment for the unprotected area(s) within 1 hour and restore the equipment to OPERABLE status within 14 days.
- d. Re-evaluate Steps 6.6.2.1 .a, 6.8.2.1 .a and 6.10.2.1 for each inoperable fire barrier, fire detection and cable coating in the area and implement revised fire watch requirements to include consideration of the inoperable spray and/or sprinkler system.
- 2. Place signs at the backup fire suppression equipment to identify the proper hose to be used.
OPLP-01 .2 Rev. 38 Page 12 of 50
ATTACHMENT I Page 2 of 2 Required Water Spray/Sprinkler Systems REACTOR BUILDING 2 North and South Core Spray Pump Rooms (LSS) North (LSS) and South (HSS) RHR Rooms 5-O Separation Zone Water Curtain (HSS) 20 Southwest Sprinkler System (LSS) 20 East Separation Zone Water Curtain (H SS) 20 Railroad Bay Sprinkler System (LSS) 20 ECCS Room Sprinkler Head (HSS) 38 Separation Zone Water Curtain (HSS) 50 Sprinkler System (Near Elevator) (LSS) 50 Separation Zone Water Curtain (HSS) 80 Sprinkler System (LSS) Two Standby Gas Treatment Train 2A Deluge Systems (LSS) Two Standby Gas Treatment Train 2B Deluge Systems (LSS) SERVICE WATER BUILDING Service Water Pump Area Sprinkler System (HSS) Service Water Cable Spread Area Sprinkler System (HSS) RADWASTE BUILDING Drumming Room Sprinkler System (LSS) MAKEUP WATER TREATMENT BUILDING Makeup Water Treatment Area Sprinkler System (LSS) OPLP-01 .2 Rev. 38 Page 29 of 50
ATTACHMENT I Page 1 of 2 Required Water S praylSpri nkler Systems DIESEL GENERATOR BUILDING Diesel Generator #1 Sprinkler System (HSS) Diesel Generator #2 Sprinkler System (HSS) Diesel Generator #3 Sprinkler System (HSS) Diesel Generator #4 Sprinkler System (HSS) South Basement (2) Cable Spread Area Sprinkler System (H SS) North Basement (2) Cable Spread Area Sprinkler System (HSS) CONTROL BUILDING Unit I Cable Spread Area Sprinkler System (LSS) Unit 2 Cable Spread Area Sprinkler System (LSS) REACTOR BUILDING I North and South Core Spray Pump Rooms (LSS) North (LSS) and South (HSS) RHR Rooms 5-O Separation Zone Water Curtain (HSS) 20 Southwest Sprinkler System (LSS) 20 Southwest Separation Zone Water Curtain (HSS) 20 East Separation Zone Water Curtain (H SS) 20 Railroad Bay Sprinkler System (LSS) 20 ECCS Room Sprinkler Head (HSS) 38 Separation Zone Water Curtain (HSS) 50 Sprinkler System (Near Elevator) (LSS) 50 Separation Zone Water Curtain (HSS) 80 Sprinkler System (LSS) Two Standby Gas Treatment Train IA Deluge Systems (LSS) Two Standby Gas Treatment Train I B Deluge Systems (LSS) OPLP-0I .2 Rev. 38 Page 28 of 50
ATTACHMENT 4 Page 3 of 8 Required Fire Detection Instruments MINIMUM INSTRUMENTS OPERABLE DETECTOR ZONEI BUILDING AREA FLAME HEAT SMOKE
- 2. Reactor Building #2
(-)17 N Core Spray Pump Room 0 0 1 Detectors 1-1, 1-2 (-)17 S Core Spray Pump Room 0 0 1 Detectors 3-1, 3-2 (-)17 N RH RD Pump Area 0 0 Detectors 1-4, 1-5, 1-6, 1-8, (-)17 S RHR Pump Area 0 0 /43 Detectors 3-4, 3-5, 3-83 20NRHRRoom 0 0 1 Detectors 1-10, 1-11 20SRHRRoom 0 0 1 Detectors 3-10, 3-11 20 South 0 0 10 Detectors 2-1, 2-5, 2-6, 2-7, 2-8, 2-9, 2-10, 2-11, 2-12, 2-13, 2-14, 2-15, 2-16 20North 0 0 6 Detectors 2-19, 2-20, 2-21, 2-22, 2-23, 2-24, 2-26, 2-27, 2-28 20East 0 0 8 Detectors 2-31, 2-32, 2-33, 2-34, 2-35, 2-36, 2-37, 2-38, 2-39, 2-40, 2-41 20 Personnel Airlock 0 0 1 Detectors 2-2, 2-3 20 Pipe Tunnel North Half 0 0 1 Detectors 2-29, 2-30 20 Pipe Tunnel South Half 0 0 1 Detector 2-17 OPLP-01 .2 Rev. 38 Page 36 of 50
ATTACHMENT 4 Page 1 of 8 Required Fire Detection Instruments MINIMUM INSTRUMENTS OPERABLE DETECTOR ZONE! BUILDING AREA FLAME HEAT SMOKE
- 1. Reactor Building #1
(-)17 N Core Spray Pump Room 0 0 1 Detectors 1-1, 1-2 (-)17 S Core Spray Pump Room 0 0 1 Detectors 3-1, 3-2 (-)17 N RHRICRD Pump Area 0 0 4 Detectors 1-4, 1-5, 1-6, 1-8, 1-9 (-)17 S RHR Pump Area 0 0 4 Detectors 3-4, 3-5, 3-6, 3-7, 3-9, 3-10 20 N RHR HX Room 0 0 1 Detectors 1-10, 1-11 20 S RHR HX Room 0 0 1 Detectors 3-11, 3-12 20 South 0 0 10 Detectors 2-7, 2-11, 2-12, 1-13, 2-14, 2-15, 2-16, 2-17, 2-18, 2-19, 2-20, 2-21, 2-22, 2-23, 2-24 20North 0 0 6 Detectors 2-27, 2-28, 2-29, 2-30, 2-31, 2-32, 2-33, 2-35, 2-36 20East 0 0 8 Detectors 2-39, 2-40, 2-41, 2-42, 2-43, 2-44, 2-45, 2-46, 2-47, 2-48, 2-50, 2-51 20 Personnel Airlock 0 0 1 Detectors 2-8, 2-9 20 Pipe Tunnel North Half 0 0 1 Detectors 2-37, 2-38 20 Pipe Tunnel South Half 0 0 1 Detectors 2-25 20 Personnel Decon Room 0 0 1 Detectors 2-53, 2-54 OPLP-01 .2 Rev. 38 Page 34 of 50
1111 WIIIIIIIth a L aJidZI1P!I
- 93. S700000 I Which one of the following choices completes the statement?
If switchyard (1) with full time load shedding in effect, the crew will enter Technical Specifications 3.8.1, AC Sources Operating, and perform
- (2)
A. (1) voltage drops to <225 kV (2) OAOP-22.O, Grid Disturbances B. (1) voltagedropsto<225kV (2) OAOP-36.1, Loss of Any 4160V Buses or48OV E-Buses C. (1) frequency drops to <59.5 Hz (2) OAOP-22.O, Grid Disturbances D. (1) frequencydropsto<59.5Hz (2) OAOP-36.1, Loss of Any 41 60V Buses or 480V E-Buses Answer: A K/A: 700000 Generator Voltage and Electric Grid Disturbances G2.02.42 Ability to recognize system parameters that are entry-level conditions for Technical Specifications. (CFR: 41.7 / 41.10 /43.2/43.3/45.3) RO/SRO Rating: 3.9/4.6 Objective: LOI-CLS-LP-050-A, Obj. 16 Given plant conditions, determine whether plant conditions meet minimum Technical Specifications requirements associated with the 230kV Electrical Distribution system.
Reference:
Tech Spec 3.8.1, AC Sources Operating Cog Level: High Explanation: lAW AOP-22.0, The Minimum Required Switchyard Voltage is 225kV with full time load shedding (normal configuration) or 223 kVwith part time load shedding. If switchyard voltage is low, the offsite transmission network must be declared inoperable and applicable sections of Tech Spec 3.8.1 followed. This requirement is in OAOP-22, Grid instability, step 3.2.2. AOP-22 also defines time limits and scram actions for off-normal system frequencies but declaration of inoperability is not required for off-normal frequencies. Declaring offsite power inoperable does not require entry into AOP-36.1 although the condition could result in a loss of power.
J t 1 ilf I JI!I1I Distractor Analysis: Choice A: Correct answer, see explanation. Choice B: Plausible because AOP-36.1 is entered upon loss of power or specific loss of buses. If loss of operability is considered a loss of power, the procedure might be selected. Choice C: Plausible because low frequency alarms at 59.8 Hz and is an AOP-22 entry condition. There is no requirement to declare offsite power inoperable based on frequency. Choice D: Plausible because low frequency alarms at 59.8 Hz and is an AOP-22 entry condition. There is no requirement to declare offsite power inoperable based on frequency. SRO Basis: Facility operating limitations in the technical specifications and their bases.[1 0 CFR 55.43(b)(2)]. OAOP-22 a.o OPERATOR ACTIONS NOTE: The Minimum Required Switchyard Voltage is the voltage necessary in the switchyard to ensure proper operation of nuclear plant emergency loads following a Loss of Coolant Accident (LOCA) or a unit trip. 3.Z2 IF Grid Voltage on the 230kV Buses is less than the Minimum Required Switchyard Voltages (225kV with Full Time Load Shedding or 223kV with Part Time Load Shedding implemented), DECLARE the offsite transmission network inoperable AND follow the applicable actions in Tech Spec 18.1 or Tech Spec 3.8.2.
- 94. SG2.O1.42 1 Which one of the following choices completes the statement below lAW FH-1 IA, Refueling Platform Operations?
To support a fuel shuffle in the core, a minimum of (1) are to be assigned to the fuel handling team on the refueling floor!bridgeand minimum of (2) of the fuel movement sheets is(are) to be utilized by the fuel handling team. A. (1) 2 workers - SRO/FH Lead and Bridge Operator (2) one copy B. (1) 2 workers SRO/FH Lead and Bridge Operator (2) two copies C. (1) 3 workers - SRO/FH Lead, Bridge Operator, and Spotter (2) one copy D. (1) 3 workers SRO/FH Lead, Bridge Operator, and Spotter (2) two copies Answer: D K/A: G2.O1 .42 - Knowledge of new and spent fuel movement procedures. (CFR. 41.10 / 43.7 I 45.13) ROISRO Rating: 2.5/3.4 Objective: LOl-CLS-LP-305, Obj. 15 Summarize the responsibilities for the following individuals when unloading or reloading the reactor core:
- a. Refuel Floor SRO
- b. Grapple Operator
- c. Spotter
Reference:
None Cog Level: Low Explanation: While performing core alterations all three personnel are required with the SRO and spotter placekeeping in the procedure. When only performing moves in the fuel pool, not to the core, then the SRO can be replaced with a Fuel Handling Lead ajiththe-spotterisnpjgj,jiecLA recent change has been to require two copies of the procedure for both actions. Distractor Analysis: Choice A: Plausible because this is the number of workers required for non core moves. One copy was the previous requirement. Choice B: Plausible because this is the number of workers required for non core moves. Two copies is correct. Choice C: Plausible because this is the number of workers required but two copies are now required. Choice D: Correct Answer, see explanation SRO Basis: Fuel handling facilities and procedures. (10 CFR 55.43(b)(7))
J JJ J ATTACHMENT 4 Page 1 of 1 Refuel Platform Manpower Requirements Cask New Handling ***ln..vessel Core Loading) Fuel Spent Fuel Maintenance (NOT Alterations unloading Receipt in SFP Core Alterations) 3RD YES *NQ *NQ *NQ NO Fuel Handling **YES **YES NO YES NO Lead Bridge YES YES YES YES YES Operator Spotter YES NO NO NO YES Dedicated YES NO NO NO NO CR RD An SRO may substitute for the Fuel Handling Lead. Fuel Handling Lead is responsible for the Spotter duties. In-Vessel Maintenance such as blade guide moves and control rod replacement that are NOT core alterations do NOT require an SRO OR Fuel Handling Lead, however, a second person to fulfill the duties of a Spotter is expected to be used. NOTE: During core alterations, the Spotter shall maintain the Bridge Operators copy of OENP-24.12-3. 3.1.7 Maintains an approved copy of Form DENP-24.12-3 independent of the Refuel Floor SRO/Fuel Handling Leads copy and uses it for verification when a Spotter is NOT assigned. cIFH-11A Rev.70 Page4of42
- 95. SG2.02.191 Unit Two is at rated power with Core Spray Loop 2A is under clearance for pump repairs.
RHR North Room Cooler develops a leak and must be isolated for repairs. Which one of the following choices completes the statements below? The RHR North Room Cooler work order priority for this condition is Priority (1) lAW WCP-NGGC-0300, Work Request Initiation, Screening, Prioritization and Classification. Maintenance can begin (2) A.(1) I (2) ONLY following planning of the Work Ordera1d-eonsider-if-wofkaround-the clock coverage is warranted. B. (1) 2 (2) ONLY following planning of the Work Order and consider if work around the clock coverage is warranted. C.(l) I (2) in parallel with planning and risk assessment per Shift Manager direction. D.(I) 2 (2) in parallel with planning and risk assessment per Shift Manager direction. Answer: C K/A: G2.02.19 - Knowledge of maintenance work order requirements. (CFR: 41.10/43.5 /45.13) RO/SRO Rating: 2.3/3.4 Objective:
Reference:
None Cog Level: High Explanation: Two or more LP ECCS injection/spray subsystems mop requires LCO 3.0.3 entry. Per WCP-NGGC-0300, Attachment 3: PRIORITY 1 classification is an emergency classification and includes activities required to support mitigation of plant conditions requiring plant shutdown as defined in Technical Specifications 3.0.3 or more limiting. Shift Manager can authorize the maintenance to begin immediately in parallel with planning and risk assessments. PRIORITY 2 classification is an urgent classification and includes activities such as failure of equipment that has a direct impact on NPDES requirements (reference Attachment 3). Candidate must determine correct prioritization, and definition.
LIIL ¶ tAL .II!I. Distractor Analysis: Choice A: Plausible because Priority I is correct classification however definition provided is for Priority 2. Choice B: Plausible because Priority 2 is incorrect for given conditions. Choice C: Correct answer, see explanation. Choice D: Plausible because Priority 2 is incorrect for given conditions. Additionally, definition is incorrect. SRO Basis: Facility licensee procedures required to obtain authority for design and operating changes in the facility. [10 CFR 55.43(b)(3)]. PRIORITY I Emergency Shift Manager can authorize the maintenance to begin immediately in parallel with planning and risk assessments. Definition Those items that have an immediate and direct impact on the heaith and safe1 of the general public, pose a significant industria hazard, or require immediate attention to prevent the deterioration of plant conditions to a possibe unsafe or unstable level. CRITERIA
. Mitigation of plant conditions requiring plant shutdown as defined in Technical Specifications 3.0.3 or more limiting . The plant Emergency Plan . Emergency Operating Procedures a Emergency activities associated with Abnormal Operating Procedures PRIORITY 2 Urgent Begin immediately folowing planning of the Work Order and consider if work around the clock coverage is warranted.
Definition Those items that have a near-term direct impact on the health and safety of the general public, the reiability of power generation, or industrial safety.
L A AIIIIIIIIt ECCSOperatng 3.5.1 ACTIONS (continued) CONDITION REQUIRED ACTION COMPLETION TIME J Two or more low pressure J.l Enter LCO 3.0.1 Immediately ECCS injection?spray subsystems inoperable for reasons other than Condition A or B. OR HPCI System and two or more required ADS valves inoperable.
- 4. ECCS Room Coolers NOTE: The following step is NOT required to be performed if the ECCS Room Cooler is inoperable due to the loss of a 4160V OR 480V E-Bus. E Bus inoperability impacts the operability of ECCS subsystems.
Technical Specifications AND the SFDP will provide Required Actions to be taken for the loss of the E-Bus. NOTE: In Mode 4 AND Mode 5, ECCS Room Coolers are NOT required to be operable to support operability of the associated ECCS Systems. a WHEN any ECCS Room Cooler is determined to be inoperable, THEN the ECCS equipment associated with that room cooler is to be declared inoperable per the applicable Technical Specifications. EXAMPLE: The RHR Room Coolers are to be considered redundant components required to support the operation of RHR. Therefore, should a room cooler be found OR made inoperable, a 7 day Active LCO is required to be established on the RHR system. Likewise, should both room coolers be found inoperable, the action required is the same as if both RHR loops AND HPCI were inoperable. Should it be identified that one RHR Room Cooler is inoperable AND one RHR Loop is also inoperable (specific combinations do NOT matter), the action is as if only one RHR Loop is inoperable (7 days). 001-01.01 Rev. 45 Page 34 of 167
- 96. SG2.02.38 I Which one of the following choices completes the statements below concerning Shift Staffing requirements?
The minimum required staffing-numbers for the AO position at BNP is (1) lAW OPS-NGGC-I000, Fleet Conduct of Operations. lAW Technical Specifications 5.2.2, Facility Staff, the shift complement may be one less than the minimum requirement for a period of time not to exceed (2) in order to accommodate unexpected absence of on-duty shift members provided immediate action is taken to restore the shift complement to within the minimum requirements. A. (I) ,three (2) one hour B. (1) nine (2) one hour C. (I) tthree (2) / two hours D. (I) nine (2) two hours Answer: D K/A: G2.02.38 Knowledge of conditions and limitations in the facility license. (CFR: 41.7 141.10 /43.1 /45.13) ROISRO Rating: 3.6/4.5 Objective:
Reference:
None Cog Level: Low Explanation: lAW the procedure 9 AC makeup the minimum shift staffing and two hours is the time to find a replacement. One hour is the time on stepping out limitation of the control room personnel. The tech Specs 5.2 only address the number of AOs for the Units which is 3, this does not take into account ASSD and Fire Brigade.
______ isdaA. e Aaa tahcSIS Distractor Analysis: Choice A: Plausible because TS 5.2.2 requires 3 AOs for both Units which does not take into account ASSD and Fire Brigade requirements. One hour is the stepping out time limit for control room personnel. Choice B: Plausible because nine is correct but one hour is the stepping out time limit for control room personnel. Choice C: Plausible because TS 5.2.2 requires 3 AOs for both Units which does not take into account ASSD and Fire Brigade requirements. Choice D: Correct Answer, see explanation SRO Basis: Conditions and limitations in the facility license. (10 CFR 55.43(b)(1)) 9.5. Operations Shift Staffing Standards
- 1. Operations ensures that the Control Room is adequately staffed for plant operations with appropriately qualified individuals. Additionally, Operations ensures staffing is adequate to meet regulatory and programmatic requirements.
Expectations
- 1. General
- a. The CR5 and Shift Manager are responsible for ensuring that only qualified watchstanders hold required positions. Personnel should verify they are qualified for the position to be held prior to assuming the watch.
- b. Individual qualifications for specific positions can be found in REG-NGGC-0012, Confirmation of Personnel Qualifications Associated with Commitments to Regulatory Guide 1.8.
- c. The shift complement may be one less than the minimum requirement for a period of time not to exceed 2 hours in order to accommodate unexpected absence of on-duty shift members provided immediate action is taken to restore the shift complement to within the minimum requirements. This provision does not permit any shift member position to be unmanned upon shift change due to an oncoming shift member being late or absent.
- d. Shift staffing shall meet the requirements of the individual plant license/Tech Specs and other regulatory and programmatic required positions at all times. Required staff numbers and positions can be found in Attachment 1 Shift Staffing.
OPS-NGGC-i000 Rev. 6 Page 47 of 147
K Ia £LJ.LC £SUiLWL Attachment I Shift Staffing Sheet 1 of 1 Shift Manning BNP Minimum Position Note staffing SM 1 CR5 2 SRO?STA 1 RO 3 AC 9 SS.2. Stepping Out Standards
- 1. Temporary reliefs and turnovers are conducted in a manner such that the relieving watchstander has the knowledge of current station status.
current station conditions and is prepared to continue safe and efficient operation of the station.
- 2. It is understood that for certain situations (e.g. plant evolutions, training) it may be desirable that CR operators leave the Control Room. The time spent outside the CR shall be minimized to maintain normal CR crew complement.
Expectations
- 1. Stepping Out by a Reactor Operator from the Control Room is intended to be for a short duration of 1 hour or less. For longer reliefs, use of a replacement watchstander will be required.
- 2. IF a Reactor Operator is required to leave the Control Room, THEN the following shall occur
- a. A briefing shall take place, to discuss the following (as a minimum):
(1) Plant status (2) Workfevolutions in progress (3) Pertinent plant operations information I OPS-NGGC-1000 Rev. 6 Page 57 of 147 V cewr r
IILLJIL1 1& J 6.2.2 Facility Staff The faciHty staff organization shall include the following:
- a. A total of three non-Icensed operators shaH be assigned for Brunswick Units 1 and 2 at all times (continued)
Brunswick Unit 2 6.0-2 Amendment No 233
Organization 5.2 5.2 Organization 5.2.2 Facility Staff (continued)
- b. At least one licensed Reactor Operator (RO) shall be present in the control room when fuel is in the reactor. In addition, when either unit is in MODE 1, 2, or 3, at least one licensed Senior Reactor Operator (SRO) shall be present in the control room. With one unit in MODE 1, 2, or 3 and the other unit defueled, the minimum shift crew shall include a total of two SROs and two ROs.
- c. Shift crew composition may be less than the minimum requirement of 10 CFR 50.54(m)(2)(i) and Specifications 5.2.2.a and 5.2.2.g for a period of time not to exceed 2 hours in order to accommodate unexpected absence of on-duty shift crew members provided immediate action is taken to restore the shift crew composition to within the minimum requirements.
- d. An individual qualified in radiation protection procedures shall be on site when fuel is in the reactor. The position may be vacant for not more than 2 hours, in order to provide for unexpected absence, provided immediate action is taken to fill the required position.
- e. Deleted.
- f. The operations manager or assistant operations manager shall hold an SRO license.
- g. The shift technical advisor shall serve in an advisory capacity to the shift superintendent on matters pertaining to the engineering aspects assuring safe operation of the unit when either unit is in MODE 1, 2, or 3.
Brunswick Unit 2 5.0-3 Amendment No. 281 I
- 97. SG2.03.05 1 Unit Two is in an ATWS condition with the given Drywell Monitor indications.
Which one of the following identifies the Loss or Potential Loss of the Fission Product Barrier(s) lAW OPEP-02.1, Brunswick Nuclear Plant Initial Emergency Action Level Matrix? (Reference provided) A. Fuel Clad Barrier ONLY. B. Reactor Coolant System Barrier ONLY. C. Both the Fuel Clad and Containment Barriers. D: Both the Containment and Reactor Coolant System Barriers. Answer: A ii&iiiiiiii wc
J1i1IIL K/A: G2.03.05 Ability to use radiation monitoring systems, such as fixed radiation monitors and alarms, portable survey instruments, personnel monitoring equipment, etc. (CFR: 41.11 /41.12/43.4/45.9) ROISRO Rating: 2.9/2.9 Objective: LOI-CLS-LP-1 1.1, Obj. 3a Describe the function/operation of the following: Drywell High Range Radiation Monitors Reference; OPEP-02.1 Cog Level: High Explanation: Drywell high range area monitors provide indications of gross fuel failure and are used to determine emergency plan emergency action level associated with abnormal core conditions. With the function switch in the E2-E5, meter readings are taken from the lower scale between 100 10,000 R/h. This reading would be
- R/h: Under the old EALs the reading had to be greater than 5000 R to have fuel cladding failure:
Distractor Analysis: Choice A: Correct Answer, see explanation Choice B: Plausible because if the wrong scale is applied to the DW Rad Monitor or if the scale selector switch was in a different position and the reactor was shutdown this would be correct. Choice C: Plausible because if the wrong scale is applied to the DW Rad Monitor then this would be correct Choice D: Plausible because if the reactor was shutdown this would be correct. SRO Basis: Radiation hazards that may arise during normal and abnormal situations, including maintenance activities and various contamination conditions. 10 CFR 55.43(b)(4) (Analysis and interpretation of radiation readings including comparison to emergency plan criteria.) FIGURE 11.1-5 DRYWELL HIGH RANGE RADIATION MONITOR CONTROL/TRIP UNIT 1o-1o ici-icY ID IOU I s TEST RIhr SLACK HIGH SAFE CIIANNEI RLET TEST
.JIIIIIE1I Fuel Clad Barrier Loss Potential Loss A. RPV Level 1. Prnary Ctainmeri Flccc ii 1. RPV eeI cannot e required due to any of te ollcwirg: restcetI a ain:ained
- RPV eve! cant be resIce > AF cr cannot cc and aintaned abe -7.5 de:emlned inches (et Pump Sucticn; vh at least one ore spray crrp inject into the reactor vessel
- RPV eve! carnot be restceo and nsinta ned above LL-4 (fSCRWLr
- RPV water eve! c.ar,t be de:enrined and RPV boding crdit one cannot be maintained B. PC Pressure (Temperature to,e 1o,re C. Isolation oe N:re
- 0. Rad 2. Dryve radation > 2,CD Rihr
- 3. Prirr.ary coolant act vit?
> 2O LCgm I-131c0!e equvalen:
E. Judgment 4. Any cnditon the opinion of the 2. Any cond don in the spin Cf Site 2merency Cosrdinator that the Site En-ergency Coordina:c indicates o:he Fe! Clad that indicates potent ci ccc c barrier the Fue C ad .arrier
7!WflaaSfrs a ,EL& IW C. Reactor Coolant System Barrier Loss Potential Loss
- 1. RFV eve! cannot be restored and maintaned TAP or cannot be detern med c1e
- 2. PC pessure> .7 ps p due to RCS leakage P1tre
- 3. Release pathway easts outside 1. RCS leakage > 50 ppm inside the primary contanment resulting from cywell isolaton allure in any of the followrg (excucing norma 2. Unisolable pnrrary system process system flowpaths frcm an c scharge outside prirrary unisolable system): containment as ncicaed by
- Main steam line Seccncary Contaiment area
- 1°C! steaii line rac atoit cn terrperatu-e above any
- RCIC steam line Maximum Ncna! Operating Limit
- RWCU ;CEOP-03-SCCP Tables 3,:)
- Feecwater
- 4. Emergency Cepressurizaton is requirec
- 5. Drywell rac alion > 27 R(hr with reactor shutcwn
( Ntre C Any condition in the opinicn of the 3. Any :cnc iticn in the op n on of the Site Emergency Coordinator that Ste Emergency Ccocinator that indicates loss of the RCS barrier inc rates potential oss of the RCS barrier t SktE7Ih2ISitflWIIa
- 98. SG2.03.06 I The Unit One Salt Water Release Tank (U/i SWRT) is nearing capacity and is scheduled to be released. The tank has been recirculated per OOP-06.4, Discharging Radioactive Liquid Effluents to the Discharge Canal.
A radioactive liquid release permit has been prepared with the foUowirg data: Tank Level 83% Tank Volume 30,088.3 gallons Recirc Start Date/Time 10/23/08, 2230 hours Sample Valve Opened Date/Time 10/24/08, 0410 hours Sample Taken Date/Time 10/24/08, 0420 hours Which one of the following identifies whether or not the time requirements for recirculation and sampling of the U/i SWRT have been met IAV\/ OOP-06.4? (Reference provided) The requirements of OOP-06.4 for recirculation and sampling of the U/i SWRT: A. have been met. The recirculation and sample times are satisfied. B. have NOT been met. The recirculation time was incorrectly calculated. C. have NOT been met. The recirculation time was calculated correctly; however, the tank was not recirculated long enough. D. have NOT been met. The recirculation time was calculated correctly; however, the sample valve was not open long enough before obtaining the sample. Answer: A K/A: G2.03.06 Ability to approve release permits. (CFR: 41.13 /43.4 I 45.10) ROISRO Rating: 2.0/3.8 Objective: CLS-LP-6.3, Obj. 5. Given a level in one of the Radwaste Release Tanks, calculate the minimum time required for recirculation.
Reference:
OOP-6.4 Discharging Radioactive Liquid Effluents the Discharge Canal, (only section 5.7 provided to examinee) Cog Level: High Explanation: The student must evaluate the data and determine that all of the criteria are met. First determines that 83 times 4 doeequal 332 rninufs, second that 332 minutes correlates to the start of recirc until the sample is taknn that the sample vave V\ as opei for greater than 5 minutes before the sample was taken.
Distractor Analysis: CHOICE A: Correct answer, see explanation CHOICE B: Plausible because may determine to be correct if miscalculation of recirculation time is performed. CHOICE C: Plausible because may determine to be correct if miscalculation of recirculation time is performed. CHOICE D: Plausible because may determine to be correct if sample valve open duration is miscalculated. SRO Basis: Radiation hazards that may arise during normal and abnormal situations, including maintenance activities and various contamination conditions. [10 CFR 55.43(b)(4)] Process for gaseous/liquid release approvals, i.e. release permits.
- 6. SAMPLE Unit I Saltwater Release Tank by performing the foIloving:
- a. ALLOW Saltwater Release Tank to recirculate for 4 minutes for each percent of indicated tank voume.
b IF necessary, THEN CONNECT a vent and drain rig to UI SALTWATER RELEASE SYSTEM FILTER INLET SAMPLE STATION VALVE 1-SWR-VI 6.
- c. OPEN UI SALTWATER RELEASE SYSTEM FILTER INLET SAMPLE STATION VALVE I-SWR-VI 6.
- d. ALLOW sample to run for at least 5 minutes to Li ensure a representative sample is obtained.
- 99. SG2.04.03 1 Which one of the following choices completes the statements below lAW Technical Specifications LCD 3.3.3.1, Post Accident Monitoring (PAM) Instrumentation?
PAM instrumentation for (1) shall be operable. If one required channel for this function is inoperable, the required channel must be restored to operable status within (2) days. A. (1) Drywell Area Radiation Monitor (2) 7 B. (1) Drywell Area Radiation Monitor (2) 30 C. (1) Drywell Airborne Radioactivity Monitor- CAC-AR-1260 (2) 7 D. (1) Drywell Airborne Radioactivity Monitor CAC-AR-1 260 (2) 30 Answer: B K/A: G2.04.03 Ability to identify post-accident instrumentation. (CFR: 41.6 / 45.4) ROISRO Rating: 3.7/3.9 Objective:
Reference:
None Cog Level: High Explanation: PAM instrumentation includes the drywell high range radiation monitors but does not include the CAC-1 260 radiation monitors. With one required instrument channel inoperable, tech spec 3.3.3.1 requires restoration of the inoperable channel within 30 days. If more than one channels were inoperable, restoration of one required channel within 7 days is required. Distractor Analysis: Choice A: Plausible because restoration within seven days is required if more than one required channel is inoperable Choice B: Correct answer, see explanation. Choice C: Plausible because CAC-1 260 is required accident monitoring instrumentation (TRM 3.4) but is not PAM instrumentation as covered in TS 3.3.3.1. A single required PAM instrument failure would require a 30 day restoration time. Choice D: Plausible because CAC-1 260 is required accident monitoring instrumentation (TRM 3.4) but is not PAM instrumentation as covered in TS 3.3.3.1. SRO Basis: Facility operating limitations in the technical specifications and their bases [10 CFR 55.43(b)(2)].
WJJfr gJ 7 saJ ad a a ea T.S. 3.3.3.1 PAM Instrumentation 3.3.3.1 3.3 I NSTRUMENTATION 3.3.3.1 Post Accident Monitoring (PAM) Instrumentation LCO 3.3.3.1 The PAM instrumentation for each Function in Table 3.3.3.1-i shall be OPERABLE. APPLICABILiTY: MODES 1 and 2. ACTIONS Separate Condition entry is aIIowd for each Function. CONDITION REQUIRED ACTION COMPLETION TIME A. One or more Functions with A. i Restore required channel to 30 days one required channel OPERABLE status. inoperable. B. Required Action and 6.1 Initiate action in accordance Immediately associated Completion Time with Specification 5.6.6. of Condition A not met. C. One or more Functions with C. 1 Restore one required 7 days two required channels channel to OPERABLE inoperable, status. raraasav
t 11 VWz, 4 PAM instrumentation 3.32.1 Th,e 3.3.3 1-1 ,&c I Ii c: *c crI Vcrtrrç ITI5rt:rn OcNDmcts REFSP.E\C5D PEOJ Ec FF0 V VIP.E2 FLhCON OkAMNELS *r ON Di 2,cIr cs:I :re2:vre C
- 2. FcIr ;csei .ta2er ..est
. -IC I:ht: b lEC ln*t 2 5
,. Chcfit: b.2lCIn:tt 2 5
- . SD rntsI5C rch 2 5
- 2. .s_rpreVr CnrherLteI 2 5
- 4. 9.rre:Vc CIiF0er hler cw,ntr 2 5 S. 3tçrv:br CF3ntcrFrer4re 2 5 S. DyMelI FTCWVS 2 5
. o-jweIrrenhn 2 5 S. 0V Fctb 2rcrpcrtaltr B cw S. Ml :+/-: . D-mellhnaps3aer 2 F TRM TRM Instrument List Appen&x C RM Tabe 3.4-2 page 1 o Ii Accident Moni:on.ng In strt. merta:iori TRM APFLICABLE FUNCTION INSTRUMENT NJMBER(S:
- 1. Suppression Chamber Atmosphere Terpera1ure CAC-TE-12e8-I 7 or e CAC-r-42e-1 CAC-TR-442-1A-5 or 1A-O CAC-TE-12E8-19 or2C CAC-TV-4426-2 CAC-TR-4426-2A-5 or 2A-3
- 2. DryweI RaaiationF CAC-AR-12C0 CAC-AQH-120-I, 2,3 CAC-AR-12e2 CAC-AQH-1262-I. 2.3
100. SG2.04.40 I A General Emergency has been declared. Onsite Emergency Response facilities are being staffed, but are NOT yet activated. Weather conditions are: Temperature 92°F Upper wind speed 9.8 mph Lower wind speed 7.3 mph Upper wind direction 318.9° Lower wind direction 314.3° Which one of the following choices completes the statements below lAW OPEP-02.6.28, Offsite Protective Action Recommendations? (Reference provided) S Recommend to the offsite agencies to evacuate (1) and shelter the remaning zones. The (2) is responsible for making this PAR. A. (1) zonesA,B,M,andN (2) Site Emergency Coordinator B. (1) zonesA,B,M,andN (2) Emergency Response Manager C. (1) zones A, B, L, M, and N (2) Site Emergency Coordinator D. (1) zonesA,B,L,M,andN (2) *Emergency Response Manager Answer: C K/A: G2.04.40 Knowledge of SRO responsibilities in emergency plan implementation. (CFR: 41.10 / 43.5 / 45.11) RO/SRO Rating: 2.7/4.5 Objective: LOl-CLS-LP-301-A, Obj. 07 Given plant conditions, determine protective action recommendations in accordance with PEP-02.6.28, Offsite Protective Action Recommendations (PAR). (SRO Only)
Reference:
Attachment 1 and 2 from PEP-02.6.28 Cog Level: High
Explanation: Prior to activation of the Emergency Response Facilities, the Site Emergency Coordinator in the Control Room is responsible for recommendations of PARS. The lower wind direction is used for determination of wind direction. Based on Att. 2, evacuate zones A, B, L, M, N. Distractor Analysis: Choice A: Plausible because if the upper wind direction was used this would be correct. Choice B: Plausible because if the upper wind direction was used and the facilities were manned this would be correct. Choice C: Correct Answer, see explanation Choice D: Plausible because if the facilities were manned this would be correct. SRO Basis: Radiation hazards that may arise during normal and abnormal situations, including maintenance activities and various contamination conditions. [10 CFR 55.43(b)(4)} Objective is an SRO ONLY objective. The question is asking for the responsibilities of the Site Emergency Coordinator in the Control Room which is an SRO positiQn (usu3ll the SM prior to the emergency facilities being activated. The SEC has to make PARs to offsite facilities when declaring a GE on the ENF.
V L AVIIIt From PEP-2.6.21: 9 METEOROLOGICAL DATAS It using WebEOC and importing Met Data you must first complete Line 6, 7. and 11, then select 9niport Plant/MET Data. Imported Met Data is current data. Enter lower wind direction and wind speed if completing hard copy ENF. CAUTIOtI: Met Data entered on Line 9 must match Met Data used for PAR determination. Met Data on Line 9 may need to be changed to match data used for PAR determination. From PEP-02.6.28 ATTACHMENT 2 Page 1 of 2 Evacuation ZoneslTime EstimatesllO Mile EPZ Map WIND FROM EVACUATE ZONES SHELTER MAXIMUM ZONES EVACUATION TIMES (hours) SUMMERIWINTER 180°- 195° A.B.G.H,J,K C.D.E.F.L,MN 80 4:00 196° 236°
- A,E,H.J.K.L C.D,E,F,G.M.N 10:00 4:00 237° 271°
- A.6.J.KL,M C.D,E.F.G.H.N 10:00 4:00 272° 268°
- A,BJ.L.M C,D.E,F.G.H.K.N 8:40 4:00 N
269°-316° A.8.L.M.N C,DE.FG.H.J.K 8:40 3:50 317°-327° A,8,M,N C.D,E.F,G,H,J.K,L 8:40 3:50 328° - 009° A.5C.M.N D.E.F.G.H.J,K.L 16:30 4:30 010°-021° A.B.C.D.M.N E.F.GJ,,J,I<.L 17:00 7:20 022° 038°
- A5.C.D.E.M,N F.G.H.J.K.L 17:30 7:40 039°-0S1° A.5.C.D.E FGM.J.K,LM.N 17:30 7:40 052° 090°
- A.5.C,D.E,F GH.J.KL.M.N 17:30 7:40 091°-112° A.B.D.E.F CG.H,J.K.L.M.N 17:20 7:30 113°- 179° A. E.F.G,H.J C.DX.L,MN 9:50 6:30 ALL ZONES IN AC.D,E.F,G,H.JKLM.N 17:30 7:40 10 MILE_EPZ 51 The Emergency Response Manager (ERM) shall:
5.1.1 Use the guidance in this procedure to ensure Protective Action Recommendations (PARs) are communicated within 15 minutes of the declaration of a General Emergency. 5.2 The Site Emergency Coordinator (SEC) shall: Prior to EOF activation: 5.2.1 Use the guidance in this procedure to ensure Protective Action Recommendations (PARS) are made within 15 minutes of the declaration of a General Emergency.}}