Text

Initial Exam 2010-301 Draft SRO Written Exam
	ML101590078
Person / Time
Site:	Brunswick
Issue date:	05/03/2010
From:	; NRC/RGN-II
To:	; Progress Energy Carolinas
References
	50-324/10-301, 50-325/10-301
	Download: ML101590078 (122)
	v • d • e

76. Which

76. Which one one ofof the following correctly the following correctly completes completes the the statement statement below?below?

Technical Specifications Technical Specifications do do NOTNOT require require the the RWCU RWCU isolation isolation from from the the SLC SLC control control switch in Mode switch in Mode due (1) due to (2)

(1) to (2)

(1) 33 A (1)

A!I (2) control rods are not not able to be be withdrawn since the reactor reactor mode mode switch must must be be in the shutdown position in position and and aa control control rod rod block block is is applied (1)3 B. (1) 3 (2) the operability of each individual control rod scram accumulator is required which will ensure that the control rods can be inserted (1)5 C. (1) 5 (2) control rods are not able to be withdrawn since the reactor mode switch must be in the shutdown position and a control rod block is applied (1)5 D. (1) 5 (2) the operability of each individual control rod scram accumulator is required which will ensure that the control rods can be inserted Feedback K/A: 204000 G2.02.25 KIA:

Knowledge of the bases in Technical Specifications for limiting conditions for operations and safety limits.

Reactor Water Cleanup System 41.5 / 41.7 /43.2)

(CFR: 41.5/41.7/43.2)

There are no safety limits associated with RWCU system, so question is written directly to the TS.

ROISRO 3214.2 RO/SRO Rating: 3.2/4.2 Objective:

Reference:

Cog Level: Low Explanation: Are no safety limits for RWCU. SLC is required in mode 33 (RO knowledge) and the bases for the mode 33 requirement is SRO knowledge.

Distractor Analysis:

Choice Choice A: Correct answer A: Correct answer from the the bases bases document.

document.

Choice B:

B: Plausible Plausible becasue becasue this is is the bases bases for Mode Mode 4/5 from the bases bases document.

document.

Choice C:

Choice Choice D:

D: Plausible Plausible because because thethe scram scram accumulators accumulators are are capable capable ofof inserting inserting the the control control rods rods with with low low reactor pressure conditions, reactor pressure but the conditions, but accumulators are the accumulators are not not required required toto be operable in be operable in Mode Mode 3.3.

SRO SRO Basis:

Basis: 1010 CFR CFR 55.43(b)-2, 55.43(b)-2, Facility Facility operating operating limitations limitations in in the the TS TS and and their their bases.

bases.

Knowledge Knowledge of of TS TS bases bases that that is is required required to analyze TS to analyze TS required required actions actions and and terminology.

terminology.

Notes Notes RECTC MODE RE~.CTOR MODE AVERIo.GE AVERA3E REACTOR REACTOR MDDE MODE TTLE TiTLE S1TCH POSfllON SWlTCH POSfIICN COOLANT EMPERTURE COOLANTIEMPERA Th'RE

(=-l-t (F:

11 Pr Operatoo P'))VI,'Er Cperatbn Run Run N."'.

22 S1aiup Slartup Reftle? ,or Refuer"') o- S;,artup;'HD1 SatupHot Nt.!;.

NA Sanby S1aoooji 33 Hot Shutdall.m'1I)

Hot hutdon Sutcii Shutdb~\T1 >2'&2

> 21.2 44 Cold Silub:i'o\'ltn Cold Sutdcn t* , iuchr, Sh,utdbltn 21:2 S;212 5 Reje1g RefueEflg(ll1 Sttubcii or R~ilel Silutdoltn ReieI ..

NI!.

From Bases 3.3.6.1 One channel of the SLC System Initiation Function is available and OPERABLE only in MODES 'I1 and 2, since these are the required to be OPERA.BLE only MODES where the reactor can be critical, and these MODES are consistent with the Applicability for the SLC System (LCO 3.'1.7). 3.1.7).

From bases 3.1.7 APPLICABILITY In MODES 'II and 2, shutdown capability is required. In MODES 3 and 4,

[n control rods are not able to be withdrawn since the reactor mode switch is in the shutdown position and a control rod block is applied. This provides adequate controls to ensure that the reactor remains sub criticaL In subcritical.

MODE 5, 6, only a single control rod can be withdrawn from a core cell containing fuel assemblies. Determination of adequate SDM {LCO 3. '1.'1 ,

(LCD 3.1.1, SHUTDOWN "SHUTDOWN MARGIN (SDM}") (SDM)) ensures that the reactor ...willvill not become critical with the analytically determined strongest control rod withdrawn.

Therefore, the SLC System is not required to be OPERABLE OPERA.BLE when only aa single control rod can be withdrawn.

Categories K/A:

KIA: 204000 204000 G2.02.25 Tier / Group:

Group: T2G2 RO Rating:

RORating: 3.2 3.2 SRO Rating: 4.2 SRORating: 4.2 LP LP Obj:

Obj: 05-11 Source:

Source: NEW NEW Cog Cog Level:

Level: LOW LOW Category Category 8:

8:

77. Unit

77. Unit One One isis operating operating at at full full power power when when the the following following plant plant conditions conditions occur:

occur:

Load Reject

- Load

- Reject Signal Signal received received Line 31 (Whiteville

- Line 31 (Whiteville Line)

- Line) PCBs PCBs redred lights lights are are lit lit Line 31

- Line

- 31 (Whiteville (Whiteville Line)

Line) white white VOLT VOLT lightslights are are not not lit lit All other

- All

- other line line PCBs PCBs green green lights lights are are lit lit 230 KV

- 230

- KV BUS BUS 1A BUS POT IA BUS POT UNDER UNDERVOL VOLTAGETAGE is is in in alarm alarm 230 KV

- 230

- KV BUS BUS 1B lB BUS BUS POT POT UNDERVOLTAGE UNDER VOL TAGE is is in in alarm alarm Which one Which one of the following of the following identifies identifies the the initial initial RPS RPS trip trip signal signal and and the the procedure procedure which which contains the contains the guidance guidance to trip the to trip the Whiteville Whiteville Line Line PCBs?

PCBs?

A'I Control Valve Fast A Control Fast Closure; Closure; OAOP-36.1, Loss OAOP-36.1, Loss ofof Any Any 4160V 4160V Buses Buses or 480V E-Buses.

or 480V E-Buses.

Stop Valve Closure; B. Stop OAOP-36. 1, Loss OAOP-36.1, Loss of Any 4160V41 60V Buses Buses or 480V E-Buses. E-Buses.

C. Control Valve Fast Closure; OAOP-22, Grid Instability.

D. Stop Valve Closure; OAOP-22, Grid Instability.

Feedback Feedback K/A: 212000A2.12 KIA: 212000 A2.12 Ability to Ability to (a)

(a) predict predict thethe impacts impacts ofof the the following following on the REACTOR on the REACTOR PROTECTION PROTECTION SYSTEM;SYSTEM ; andand (b) based (b) based on on those those predictions, predictions, use use procedures procedures toto correct, correct, control, control, oror mitigate mitigate the the consequences consequences of those of those abnormal abnormal conditions conditions or or operations:

operations:

Main turbine Main turbine stop stop control control valve valve closure closure (CFR: 41.5 (CFR: 41.5/ 45.6) 145.6)

RO/SRO Rating:

RO/SRO Rating: 4.0/4.1 4.0/4.1 Objective: CLS-LP-03, Objective: CLS-LP-03, Obj. Obj. 8.

8.

List the List the RPS RPS trip signals, including trip signals, including setpoints setpoints andand how/when how/when each each signal signal is is bypassed.

bypassed.

Reference:

SD-03 Reactor SD-03 Reactor Protection Protection System, System, section section 3.1 3.1 RPS RPS Trips Trips Cog Level Cog Level High High Explanation:

Explanation:

load reject A load reject signal at any reactor power level will cause a turbine control valve fast closure scram. The load reject signal does not input into the turbine stop valve closure scram logic. During During a loss of offsite power, if the grid is lost all PCBs are opened per OAOP-36.1. OAOP-36. 1.

Distractor Analysis:

Choice A: Correct answer, see explanation Choice B: Incorrect Load reject initiates a TCV fast closure scram not a TSV. A misconception of the difference between TCV and TSV scrams may cause a student to select this answer.

Choice C: Incorrect. OAOP-22 does not have an action for loss of grid only for degraded.

Choice D: Incorrect. Load reject initiates a TCV fast closure scram not a TSV. A misconception of the difference between TCV and TSV scrams may cause a student to select this answer. OAOP-22 does not have an action for loss of grid only for degraded.

SRO Basis: 10 10 CFR 55.43(b)-5, Assessment of facility conditions and selection of appropriate procedures during normal, abnormal, and emergency situations.

Notes Notes An example An example of of Turbine Turbine Control Control Valve Valve Fast Fast Closure Closure is is aa load load reject.

reject.

The definition The definition of of aa load load reject reject is is greater greater than than 40%

40% mismatch mismatch betvveen between electrical output and electrical output and mechanical mechanical input input as as sensed sensed by by generator generator stator stator amps and amps and the the Cross Cross Around Piping Piping pressLire.

pressure. ThisThis isis to to prevent prevent excessive overs excessive peed of overspeed of the the Turbine Turbine on on loss loss ofof load.

load.

AA load reject reject signal will energize energize thethe fast fast acting Solenoid Valves on on the control the control valve valve actuators, which removes removes hydraulic hydraulic triptrip fluid fluid pressure. The trip signal pressure. signal comes comes from from pressure switches switches on the the fast Vast acting trip control (FATC) supply to the control valve disc dumps (refer to EHC Hydraulics). Loss of this pressure pressure will causecause a rapidrapid closure of the control valves. Circuitry Circuitry is is designed such that the pressure switch on either control valve 'lor 1 or 3 Will will trip RPS Trip System A. Either system Either control valve 2 or 4 will trip RPS RPS TripTrip System B.

These switches will also provide a Scram signal on loss of hydraulic trip fluid pressure when a load reject signal is not present - loss of -

hydraulic fluid pressure can result in a fast closure of the control valves.

valves 1 SD-03 Rev. 9 Page 2'1 891 21 of 89 7.

7, IF tile the SAT was lost due to loss of power on the Progress Energy System, THEN PERFORM the following:

a. PLACE AUTO RECLOSE

.PLACE RELOSE switches in MANUAL. D LI

b. PLACE transmission line PCB SUPERVISORY D LOCAL/REMOTE LOCAUREMOTE switches in LOCAL. LOCAL c,

c. TRIP all transmission line PCBs. D LI OAOP-36.2 IOAOP-36.2 Rev. 4'1 I Page 44 of 1961961 Categories K/A:

KIA: 212000 A2.12 Tier / Group: T2G1 T2G 1 RO Rating: 4.0 SRO Rating: 4.1 LP Obj: 03-08 Source:

Source: PREY PREV Cog Cog Level:

Level: HIGH Category Category 8:8: Y

78. The Unit

78. The Unit isis at 7% power at 7% power during reactor startup.

during reactor startup.

The operator The operator withdraws withdraws control control rod rod 26-27 26-27 to position 48.

to position 48.

The following The following indications indications are are noted:

noted:

ROD DRIFT

- ROD

- DRIFT alarm alarm seals seals in in ROD OVER

- ROD

- OVER TRAVEL TRAVEL alarm alarm seals seals inin Rod 26-27

- Rod

- 26-27 full full core core display display red red light light out out Which one Which one of the following of the following identifies:

identifies:

(1) the indication (1) the indication thatthat would would be displayed on be displayed on the the four-rod four-rod group group display display and and (2) the required action for the inoperable control rod lAW Technical Specification 3.1.3, Control Rod Operability?

3.1.3, (1)48 A. (1) 48 Fully insert control rod 26-27 and disarm the HCU (2) Fully HCU B. (1)

B. (1)4848 (2) Verify ~12 1 2 control rods are withdrawn and implement GP-11, GP-1 1, Second Operator Rod Sequence Checkoff Sheets C (1) Blank C!I (2) Fully insert control rod 26-27 and disarm the HCU D. (1) Blank (2) Verify ~12 1 2 control rods are withdrawn and implement GP-11, GP-1 1, Second Operator Rod Sequence Checkoff Sheets

Feedback Feedback K/A: 214000 KIA: 214000 A2.03 A2.03 Ability to Ability to (a) predict the (a) predict impacts of the impacts the following of the following on on the the ROD ROD POSITION POSITION INFORMATION INFORMATION SYSTEM; SYSTEM; and (b) and (b) based based on on those those predictions, predictions, use use procedures procedures to to correct, correct, control, control, or or mitigate mitigate the the consequences of consequences of those those abnormal abnormal conditions conditions or or operations:

operations:

Overtravel/in-out Overtravel/in-out (CFR: 41.5 (CFR: 41.5 /45.6)

/ 45.6)

RO/SRO Rating:

RO/SRO Rating: 3.6/3.9 3.6/3.9 Objective: CLS-LP-07 Objective: CLS-LP-07 Obj Obj 5b 5b Given plant Given plant conditions, conditions, determine determine ifif the the following following conditions conditions exist:

exist: b.

b. Indications Indications of of an an uncoupled uncoupled control control rod.

rod.

Reference:

SD-07 page SO-07 page 27 27 TS 3.1.3 TS 3.1.3 Level Cog Level Cog Low Low Explanation:

Explanation:

If the control rod is in the overtravel out position, the corresponding corresponding digital indicator will be blank since the magnet will not be near any of the 00 to 48 reed switches. lAW TS the rod is declared inoperable then magnet inserted to 00 (within 3 hours3.472222e-5 days 8.333333e-4 hours 4.960317e-6 weeks 1.1415e-6 months ) and disarmed (within 4 hours4.62963e-5 days 0.00111 hours 6.613757e-6 weeks 1.522e-6 months ). TS 3.1.6 if the RWM is inoperable then if

>12 control rods are withdrawn GP-11

.=::12 GP-1 1 would be implemented, unless the rod is at 00 and is not intended to be moved.

Distractor Analysis:

Oistractor Choice A: Plausible because the full in and 00 indications are at the same point or the examinee exam inee may think that the rod may settle to the 48 position.

Choice B: Plausible because the full in and 00 indications are at the same point or the examinee may think that the rod may settle to the 48 position. These are TS actions for an inoperable RWM, not control rod.

Choice C: Correct answer, see explanation.

Choice D: 0: Plausible because this is the correct indication but these are TS actions for an inoperable RWM, not control rod.

SRO Basis:

Basis: 1010 CFR CFR 55.43(b)-2, 55.43(b)-2, Facility Facility operating limitations limitations in in the technical specifications and their bases. Application of bases. of required actions statements.

Notes Notes From SD-07 From SD-07

- Coupling integrity Coupling integrity of of aa control control rod rod st"lall shall bebe checked checked anytime anytime aa control rod control rod isis fully fully withdrawn withdrawn by by verifying verifying that that the the rod rod does does not not reach the reacl) the overtravel overtravel position.

position. An An uncoupling uncoupling check check can can bebe performed by performed by maintaining maintaining the the continuous continuous withdrawwithdraw signal signal for for approximately 33 to approximately to 55 seconds seconds whenwhen the the control control rod rod t"las has reached reached position 48 position 48 and and verifying verifying thethe control control rodrod does does notnot retract retract beyOnd beyond position 48.

position 48. IfIf the the rod rod isis uncoupled, uncoupled, then then thethe four four rod rod display display indication will indication will go go out out for for the the uncoupled uncoupled rod rod and and the the Rod Rod Over Over Travel Annunciator Travel Annunciator A-05 4-2 A-05 4-2 will will illuminate.

illuminate.

I SD-07 SD-07 Rev. 6 Rev.S Page Page 2727 of 571 57 C. One or more control rods C. 1 C.*1 ----------NOTE--------

inoperable for reasons other Inoperable control rod may than Condition A or B. 8. be bypassed in the RWM or RWM may may be bypassed bypassed as allowed bv by LCO 3.3.2.1:

3.3.2.1, if required, to allow insertion of inoperable control rod and continued operation.

Fully insert inoperable 3 hours3.472222e-5 days 8.333333e-4 hours 4.960317e-6 weeks 1.1415e-6 months control rod.

AND (continued)

C. (continued) C.2 Disarm the associated 4 hours4.62963e-5 days 0.00111 hours 6.613757e-6 weeks 1.522e-6 months CRD.

From GP-11:

This procedure provides aa method method for aa second second licensed licensed operator or qualified qualified member member of of the plant plant technical staff to to verify control control rod movement movement and and compliance compliance withwith the prescribed BPWS BPWS control rod pattern pattern with thethe rod worth minimizer (RWM) inoperable minimizer (RWM) inoperable in in conformance conformance with with the the requirements of of Technical Technical Specification Specification 332.1.

3.3.2.1. IfIf the the RWM RWM is is inoperable inoperable due due to to bypassed bypassed control control rod(s) rod{s) that will Will not not be be moved moved during during the the startup/shutdow, startup/shutdown, then then this this procedure procedure is is not not required.

required.

Categories Categories K/A:

KIA: 214000 214000 A2.03 A2.03 Tier Tier // Group:

Group: T2G2 T2G2 RO Rating:

RORating: 3.6 3.6 SRO Rating:

SRORating: 3.9 3.9 LP Obj:

LPObj: 07-5B 07-SB Source:

Source: NEW NEW Cog Cog Level:

Level: HIGH HIGH Category Category 8: 8: YY

79. Given the

79. Given the following following ATWS ATWS conditions conditions on Unit Two:

on Unit Two:

2A CRD 2A CRD Pump Pump Overcurrent trip Overcurrent trip 2B CRD 2B CRD PumpPump Shaft uncoupled Shaft uncoupled HPCI System HPCI System Under Clearance Under Clearance SLC SLC Both squib Both squib valves valves failed failed to to fire fire RCIC RCIC Running with Running with an an unisolable unisolable steam steam supply supply leak leak Suppression Pool Suppression Pool Level Level -24 inches

-24 inches Reactor Power Reactor Power 10%

10%

Reactor Water Reactor Water Level Level 160 inches 160 inches Which one Which one of of the the following following identifies identifies the the action action that that would would be be taken taken concerning concerning the the RCIC system based RCIC system based on on the conditions above?

conditions above?

The RCIC The RCIC system system would:

would:

source of the steam leak lAW OAOP-05.0, Radioactive A. be isolated to secure the source Radioactive Spills, High Spills, High Radiation, Radiation, and and Airborne Activity.

B. have the high suppression pool water level transfer defeated and its suction transferred back to the CST lAW SEP-1 SEP-i 0, Circuit Alterations Procedure.

C C~ remain running because it is needed for boron injection lAW LEP-03, Alternate Boron Injection.

D. be terminated and prevented to reduce level to 90 inches lAW LPC.

Feedback Feedback K/A: 217000 KIA: 217000 G2.04.08 G2.04.08 Knowledge of Knowledge how abnormal of how abnormal operating operating procedures procedures are are used used in in conjunction conjunction with with EOPs.

EOPs.

Reactor Core Reactor Core Isolation Isolation Cooling System (RCIC)

Cooling System (RCIC)

(CFR: 41.10/43.5/45.13)

ROISRO Rating:

RO/SRO Rating: 3.8/4.5 3.8/4.5 Objective: CLS-LP-300-J Objective: CLS-LP-300-J Obj Obj 44 Given plant Given conditions and plant conditions and aa copy copy of of the the LEPs, LEPs, determine determine which which method method ofof alternate alternate boron boron injection injection isis appropriate.

appropriate.

Reference:

AOP-50/SCCP/LEP-03/LPC AOP-50 1 SCCP 1 LEP-03 1 LPC Cog Level Cog Level high high Explanation: EOP Explanation: EOP action that supercedes supercedes the AOP AOP action is is what the question is asking.

AOP-5 does have a step to isolate the system that is leaking, but SCCP overrides that if the system is required by EOPs. WithWith the AATWS TWS the RCIC system is required for alternate boron injection. The Suppression Pool level is high but this will only transfer the HPCI suction valves as RCIC only transfers on CST level. LEP-03 would first want to use CRD then RCIC as long as suction is from CST.

Distractor Analysis:

Choice A: Plausible because AOP-5 does have a step to isolate the system that is leaking, but SCCP overrides that if the system is required by EOPs.

Choice B: Plausible because the high suppression pool level would transfer HPCI and SEP-10 has a section for transferring the suction to the CST from the Suppression Pool.

Choice C: Correct answer, see explanation Choice D: Plausible becasue LPC does have a step for terminating and preventing but this does not address RCIC.

SRO Basis: 10 10 CFR 55.43(b)-5, 55.43(b)-5, Assessment of facility conditions and selection of appropriate procedures during normal, abnormal, and emergency emergency situations.

Notes Notes Actions from Actions from AOP-05.0 AOP-05.O 1.

1. INITIATE aa search INITIATE coolant or coolant search to or steam steam leak.

to locate leak.

locate and and isolate isolate the the source source of of any any ofl 2.

2. IF radiography IF radiography isis in danger of danger of abnormal in progress, progress, AND abnormal exposure, exposure, THEN AND personnel personnel are THEN SECURE SECURE are in in oD radiography.

radiography.

3.

3. ENSURE all ENSURE dosimetry and dosimetry all personnel personnel in and report in the report unusual the area area monitor unusual exposure monitor their exposure to their to E&RC.

E&RC.

o OAOP-05.C IOAOP-05.0 Rev. 23 Rev. 23 Page Page 66 of"10 of 10 I From SCCP From SCCP has has thethe actions actions to to leave the system leave the system running:

running:

ISOLATE ALL ISOLATE ALL SYSTEMS SYSTEMS DISCHARGING INTO DISCHARGING INTO THE AREA EXCEPT EXCEPT SYSTEMS SYSTEMS REQUIRED:

- TOEMERGENCY BE OPERATED BYAN TO BE OPERATED BY AN EMERGENCY OPERATING OPERATING PROCEDURE

- DAMAGE CONTROL FOR DAMAGE FOR CONTROL I SCCP-14 From LEP-03 A NOTE: HPCI/RCIC should be used only it HPGIIRCIC if suction is from the CST.

CST A

From LPC, RCIC is not on list to Terminate and prevent (HPCI is):

LOWER REATQR LOWEll; ReACTOR WATER WATER LEVEL LEVEL J!UEsrECTIvE IRlIE$PEC11VE OF OF ANY REACTOR POWER OR REACTOR REACTOR

\VATER LEVELOSCILLATIONS WATER I.EVELOSCILLAllONS ry BY TERMflIAJ1NO TI"~MIIIAliNG AND AND PRlilfliNTING tNJFCTIOU P4flTMflN IN.I~cnOIJ FRCM FRQM TIlE THE FOLLO%1NG FOLlOVlIHG SYSTE1S SYSTEMS UNLESS UHLESSTHE THE SYSTEM IS SEINO

~ YSTEM IS BEING USED USElJ IO INJECT CORON TO NJEC1 IIOItOH;

GQNDENSATEFEECWATER CO.'lDENSATelFESOWATI!R

NPCI IIPCI

RHR IIHR

CORE SPRAY CORESPRAY

ALTERNATE COCL?.NT NJTCtIflN YTPM!

RC)L- 17 Categories Categories KJA:

KIA: 217000 217000 G2.04.08 G2.04.08 Tier!

Tier / Group:

Group: T2G1 T2G1 RO Rating:

RORating: 3.8 3.8 SRO Rating:

SRORating: 4.5 4.5 LP Obj:

LPObj: 300J-4 300J-4 Source:

Source: NEW NEW Cog Level:

Cog Level: HIGH HIGH Category Category 8:8: Y Y

80. Unit Two

80. Unit Two was was operating operating at at rated rated power power with with the following conditions:

the following conditions:

- AA dual

- dual Unit Unit Loss Loss OfOf Offsite Offsite Power Power (LOOP)

(LOOP)

Spent Fuel Pool level

- Spent Fuel Pool level is

- lowering is lowering rapidly rapidly due due toto aa dropped dropped test test weight weight RRCP has

- RRCP

- has been entered due been entered due to to high high rad rad conditions conditions on on the the refuel refuel floor floor Which one Which one of of the the following following isis the first makeup the first makeup source source to to be be used used forfor filling filling the the fuel fuel pool pool and identifies the and identifies the procedure procedure to perform the to perform the action?

action?

A. Emergency Diesel A. Emergency Diesel Makeup Makeup Pump Pump via via hoses hoses lAW lAW OAOP-38.0, OAOP-38.0, Loss Loss of of Fuel Fuel Pool Pool Cooling Cooling B RHR B!'" RHR BB Loop Loop via Fuel Fuel Pool Pool Cooling Cooling System System lAW lAW OAOP-38.0, OAOP-38.0, Loss Loss of of Fuel Fuel Pool Pool Cooling Cooling C. Emergency Diesel Diesel Makeup Makeup Pump via hoses lAW lAW OEDMG-002, OEDMG-002, Spent Fuel Fuel Pool Pool Makeup/Spray and Makeup/Spray and Refuel FloorFloor Enhanced Enhanced Ventilation under Conditions of Extreme Damage Damage D. RHR B Loop via Fuel Pool Cooling System lAW OEDMG-002, OEDMG-002, Spent Fuel Pool Makeup/Spra Makeup/Spray y and Refuel Floor Enhanced Ventilation under Conditions of Extreme Damage

Feedback Feedback K/A: 233000 KIA: 233000 G2.04.06 G2.04.06 Knowledge of Knowledge mitigation strategies.

EOP mitigation of EOP strategies.

Fuel Pool Fuel Pool Cooling Cooling and and Clean-up Clean-up (CFR: 41.10/43.5/45.13)

(CFR: 41.10/43.5/45.13)

There are There are no no direct direct EOP EOP actions actions associated associated with FPC, aa loss with FPC, loss of oflevel level in in the the fuel fuel pool pool will will cause cause entry entty into into RRCP which RRCP which is is an an EOP.

EOP. So these actions So these actions are are mitigation mitigation strategies strategies to RRCP.

to RRCP.

RO/SRO Rating:

RO/SRO Rating: 3.7/4.7 3.7/4.7 Objective:

Objective:

CLS-LP-1 3, Obj.

CLS-LP-13, Obj. 11.

11. State State the the sources sources of of makeup makeup water water for for the the Fuel Fuel Pool Pool inin order order of of preference.

preference.

Reference:

OAOP-38.0 OAOP-38.0 LossLoss of of Fuel Fuel Pool Pool Cooling Cooling Cog Level Cog Level High High Explanation:

Explanation:

sources is from the normal fill, Demin water hose stations, The order of the makeup sources stations, Fire protection hose stations, demin water through RHR keepfill, and then other sources that are not service water. For a high capacity water source and the gates installed RHR Loop B would be used via the FPC system. With a LOOP the demin pumps have no power. If no other sources are availble then the proceudre has injection from the EDMP.

Distractor Analysis:

Choice A: Plausible because although this is a makeup source it is not the preferred source (last resort per the procedure) and is performed per the EDMG procedures. Although upon entering the AOP there is a step to start lining this system up for injection because of the time required to get all of the hoses run in the procedure up to the fuel pool.

Choice B: Correct answer see explanation Choice C: Plausible because although this is a makeup source it is not the preferred source (last resort per the procedure). Although upon entering the AOP there is a step to start lining this system up for injection because of the time required to get get all of the hoses hoses run in the procedure procedure up to the fuel pool.

Choice D:0: Plausible because RHR is the high capacity source that will need to be used, but the EDMG procedure procedure does does not provide this guidance.

SRO SRO Basis: 10 CFR Basis: 10 CFR 55.43(b)-5, 55.43(b)-5, Assessment of of facility conditions and and selection selection of of appropriate procedures during during normal, normal, abnormal, abnormal, and and emergency situations.

situations.

Notes Notes 2.

2. IMMEDIATELY ENTER IMMEDIATELY ENTER OEDMG-002, OEDMG-002, Spent Spent Fuel Fuel Pool Pool Makeup/Spray and Makeup/Spray and Enhanced Enhanced Refuel Refuel Floor Floor Ventilation Ventilation D

LI Under Conditions Under Conditions of of Extreme Extreme Damage, Damage, AND AND make make preparations to preparations to makeup makeup to to the the fuel fuel pool pool using using the the EDfvIP.

EDMP.

3.2.12 3.2:12 IF aa high IF high capacity capacity makeup makeup source source of of water water through through the the RI-IR System RHR System isis required required toto maintain maintain fuelfuel pool pool level level AND AND tt1e fuel pool the fuel gates are pool gates are installed, installed, THEN PERFORM tile THEN PERFORM the following:

following:

1. CONFIRM one CONFIRM of the one of the following flow ow paths paths available available for for use with the Fuel use Pool Cooling Fuel Pool Cooling System:

RHR Loop B only (RHR B only (RHR LoopLoop 6B Shutdown Shutdown Cooling Cooling D must be be secured)

RHR Loop A through RHR Loop Cross-Tie to the D Li RHR Loop 6B discharge.

discharge. (60th (Both RHR Loop Loop A and LoopLoop 6B Shutdown Cooling must be secured).

QAOP-38.O IOAOP-38.0 Rev. 22 Page 1-1 11 of 351 35 Actions for Emergency Diesel Makeup Pump:

3.2.19 3.2:19 IF no actions have been successful, THEN ENTER D LI OEDMG-OD2, Spent Fuel Pool Makeup/Spray OEDMG-002, Makeup/Spray and Refuel Floor Enhanced Ventilation under Conditions of Extreme Damage.

From EMG-002:

3.3 Normal fuel pool makeup methods and the B.5.b 6.S.b requirement for aa diverse internal strategy (using installed plant equipment) eqUipment) are contained in OAOP-38.O, OAOP-38.0, Loss of Fuel Pool Cooling. OEDMG-002 is entered when the methods contained in in OAOP-38.O OAOP-38.0 have proven to be inadequate or cannot be performed.

Categories Categories K/A:

KIA: 233000 233000 G2.04.06 G2.04.06 Tier Tier / Group:

Group: T2G2 T2G2 RO Rating:

RORating: 3.7 3.7 SRO Rating:

SRORating: 4.7

4.7 LPObj

LPObj: 13-11 13-11 Source:

Source: NEW NEW Cog Level:

Cog Level: HIGH HIGH Category Category 8:8:

81. With Unit

81. With Unit Two Two at at rated rated power, power, which which one one of of the the following following identifies:

identifie (1) the (1) required number the reql:lired number of of operable operable SRVs SRVs for for safety safety function function lAW lAW Technical Technical Specification 3.4.3, Safety/Relief Specification 3.4.3, Safety/Relief Valves and Valves and (2) the (2) the bases bases for for this this number number of of operable operable SRVs? SRVs?

A. (1)

A. (1)99 (2) prevent overpressurization (2) prevent overpressurization associated associated with with anan ATWS ATWS event event B (1)

B!'" (1) 10 10 (2) prevent (2) prevent overpressurization overpressurization associated associated with with anan ATWS ATWS event event C. (1)

C. (1)9 9 (2) prevent overpressurization (2) prevent overpressurization associated associated with with anan MSIV MSIV closure closure D. (1)

D. (1)1010 (2) prevent overpressurization (2) prevent overpressurization associated associated with with an an MSIV MSIV closure closure Feedback Feedback K/A: 239002 G2.02.25 KIA: G2.02.25 Knowledge of the bases in Technical Specifications Specifications for limiting conditions for operations and safety limits.

Safety Relief Valves 41.5 / 41.7/43.2)

(CFR: 41.5/41.7/43.2)

ROISRO Rating: 3.2/4.2 RO/SRO Objective: CLS-LP-25, Obj. 10 Given plant conditions and TS, including the Bases, TRM, ODCM, and COLR determine the required actions to be taken in accordance with TS associated with the Reactor Recirculation System. (SRO only)

Reference:

TS 3.4.3 3.4.3 and bases document Cog Level Low Low Explanation:

Explanation:

TS TS 3.4.3 3.4.3 states states 10 10 must must be be operational for the safety function, the the safety the bases bases states states the the reason, reason, ATWS.

ATWS.

Distractor Distractor Analysis:

Analysis:

Choice Choice A: A: Plausible Plausible because because thethe bases bases states states that are required that 99 are required for for the the MSIV MSIV closure.

closure.

Choice Choice B: B: Correct Correct answer, see explanation answer, see explanation Choice Choice C: C: Plausible Plausible because because thethe bases bases states states that are required that 99 are required forfor the the MSIV MSIV closure closure andand the the MSIV MSIV closure not the closure isis not the binding binding failure failure mode.

mode.

Choice Choice D: D: Plausible Plausible because because 10 10 are are required required forfor the the ATWS ATWS and and thethe MSIV MSIV closure closure isis not not the the binding binding failure failure mode.

mode.

SRO Basis:

SRO 10 CFR Basis: 10 CFR 55.43(b)-2, 55.43(b)-2, Facility Facility operating operating limitations the technical limitations inin the technical specification speCifications and their s and their bases.

bases. This This isis knowledge knowledge of oftech tech spec spec bases bases to to determine determine the the reason reason 10 10 are are required.

required.

Notes Notes 3.4.3 Safety/Relief 3A.3 Safetv?RehefValves Valves (SRVs)

(SRVs)

LCO 3.4.3 LCO 3.4.3 The safety The safety function of '10 function of 10 SRVs SRVs shall shall be be OPERABLE.

OPERASLE.

APPUCABILITY:

APPLICABILITY: MODES 1,1. 2, MODES 2. and 3.

and 3.

From the From the Bases Bases document:

document:

APPLICABLE APPLICABLE The overpressure The overpressure protection protection system system must must accommodate accommodate the the most most SAFETY ANAL SAFETY ANALYSES severe pressurization YSES severe pressunzattan transient.

transient EvaIatioi aac1eterrnined that the rnoLseere4ransient icioit1flar&fres (MP7 fIod by readorscrt n 1f1hr{ie fiIure of of the the direct scrani i+/-tdith MSIV position) (Ref. 1 For the purpose direct purpose ofof the analyses;*

the 9 SRVs are assumed,tooperate.in,the analyses, 9SR\lsare assumed to operate in the safety mode. The safety mode. The analysis results results demonstrate demonstrate that the the design design SRV SRV capacity capacity isis capable capable ofof maintaining reactor maintaining reactor pressure pressure below below thethe ASME CodeCode limit limit of of 110%

110% of vessel design pressure P'l design pressure (110%

0% xx '1250 psig =1375 1250 psig 1375 psig).

psig). This This LCO LCO helps helps to ensure that the acceptance acceptance limit limit of of '1375 1375 psig is is met met during during the Design Design Basis Event.

Basis Event.

(continued)

(continued APPLICABLE For overpressurization associated with an A ATWS event, 10 SRVs are nNS event.

SAFETY ANALYSES assumed to operate in the safety mode. The analysis (Ref. 2)

(continued)

(continued} results demonstrate that the design capacity is capable of maintaining reactor pressure below the ASME Section III Ill Code Service Serice Level C limits (1500 psig).

From an overpressure standpoint, the design basis events are bounded by the overpressurization associated with the AnNS ATWS event described above. Reference 3 discusses additional events that are expected to actuate the SRVs.

Criteiion 33 of '10 SRVs satisfy Criterion 10 CFR CER 50.36(c){2)(ii) 50.36(cX2)(ii) (Ref. 4).

Categories KJA:

KIA: 239002 G2.02.25 Tier / Group: T2G1 RO Rating:

RORating: 3.2

3.2 Rating

SRORating:

SRO 4.2 LP Obj: 25-10 Source:

Source: NEW Cog Cog Level: LOW LOW Category Category 8:8: Y

82. Unit

82. Unit One One isis operating operating at full power at full power when when the the Main Main Stack Stack Rad Rad Monitor Monitor lost lost its its normal norm power supply.

power supply.

Which one Which one of of the the following following identifies identifies the the procedure procedure that that contains contains thethe steps steps to to transfer transfer the Main the Main Stack Stack Rad Monitor to Rad Monitor its alternate to its alternate powerpower supply?

supply?

A. IOP-52, 120 A. 10P-52, 120 Volt Volt ACAC UPS, UPS, Emergency, Emergency, and and Conventional Conventional Electrical Electrical Systems Systems Operating Procedure Operating Procedure B

B~ 20P-52, 120 20P-52, 120 Volt Volt ACAC UPS, UPS, Emergency, Emergency, and and Conventional Conventional Electrical Electrical Systems Systems Operating Procedure Operating Procedure C. 1APP UA-03 C. 1APP UA-03 6-3, PROCESS SMPL 6-3, PROCESS SMPL OG OG VENTVENT PIPE PIPE DNSCIINOP DNSC/INOP D. 2APP UA-03 D. 2APP UA-03 6-3, PROCESS SMPL 6-3, PROCESS SMPL OG OG VENTVENT PIPE PIPE DNSCIINOP DNSC/INOP Feedback Feedback K/A: 262002 KIA: 262002 G2.01.23 G2.0l.23 Ability to perform specific system and integrated plant procedures during all modes of plant operation.

Uninterruptable Power Supply (A.C'/D.C.)

(A.C.ID.C.)

41.10/43.514 (CFR: 41.10 5.2/45.6) 143.5 1 45.2 145.6)

ROISRO Rating: 4.3/4.4 RO/SRO Objective: CLS-LP-11.0, 15a Given plant conditions and a trip or failure of one of the following Radiation Monitors, determine appropriate plant response and use procedures to determine the actions required to control and/or mitigate the consequences consequences of the event:

a. Main Stack.

Reference:

20P-52, Section 8.7, Stack Radiation Monitor UPS Power Supply Transfer Cog Level High Explanation:

The normal power supply for the Main Main Stack Rad MonitorMonitor is from Unit Unit Two. On aa loss of of power the from the normal normal power supply the operators will need to transfer to the alternate power power supply. This direction direction isis only only in in the the U2 U2 procedure. ThereThere is no directions is no directions to perform this in in the U1 procedure the Ui procedure or or the the APPs APPs for either Unit.

Unit.

Distractor Distractor Analysis:

Analysis:

Choice Choice A: A: Plausible Plausible because because the stem states the stem states this this is is Ui U1 but but the the actions actions areare in in the the U2 U2 procedure.

procedure.

Choice Choice B: B: Correct Correct answer, see explanation.

answer, see explanation.

Choice Choice C: C: Plausible Plausible because because the downscale imp the downscale 1 inop annunciator annunciator will be actuated will be actuated onon aa loss loss of of power power but but the the APP5 APPs do do not address transfer not address transfer of of power power toto backup backup supply.

supply.

Choice Choice D: D: Plausible Plausible because the downscale because the downscale /1mop inop annunciator annunciator will will be be actuated actuated onon aa loss loss of of power power but but the the APPs APPs do do not not address address transfer transfer of of power power to backup supply.

to backup supply. U2 U2 is is the the normal normal power power supply supply to to the the rad rad monitor.

monitor.

SRO Basis:

SRO Basis: 10 10 CFR CFR 55.43(b)-5, 55.43(b)-5, Assessment Assessment of offacility facility conditions conditions andand selection selection of of appropriate appropriate procedures procedures during during normal, normal, abnormal, abnormal, and and emergency emergency situations.

situations.

Notes Notes 8.0 8.0 INFREQUENT OPERATIONS INFREQUENT OPERATIONS..................................................................................

32 32 8.1 8.1 Transferring UPS Transferring UPS Loads Loads From From Alternate Alternate SourceSource to to PrimalY Primary UPS UPS Unit Unit 2A.....

2A 32 32 8.2 8.2 Transferring UPS Transferring UPS Loads Loads From From Alternate Alternate SourceSource to to Standby Standby UPS UPS Unit 28 Unit 28 .... 35 35 8.3 8.3 Transferring UPS Loads Transferring UPS Loads From Standby From Standby UPS UPS Unit Unit 2B2B to to Alternate Alternate SourceSource .... 39 39 8.4 8.4 Transferring UPS Transferring UPS Loads Loads From From Primary Primary UPS UPS Unit Unit 2A2A to to Alternate Alternate SourceSource..... 4'1 41 8.5 8.5 Alignment of Alignment of Standby Standby UPS UPS Unit Unit 2B2B After After aa Loss Loss of of Alternate Alternate SourceSource Power.Power 43 43 8.6 8.6 Returning Standby UPS Unit 2B to Normal Returning Standby UPS Unit 2B to Normal Operating Condition Operating Condition Upon Upon Regaining Altemate Regaining Alternate SourceSource Power Power Supply Supply ............. ..... ........... ...... ........... ....... 45 45 8.7 8.7 Stack Radiation Stack Radiation Monitor Monitor UPS UPS PowerPower Supply Supply Transfer.

Transfer ................................... 47 47 20P-52 120P-52 Rev. 53 Rev. 53 Page Page2of782 of 78 I 8.0 8.0 INFREQUENT OPERATIONS INFREQUENT OPERATIONS .... ............ ................. ...... ................ ........... ............. 34 34 8.1 8.1 Transferring UPS Transferring UPS LoadsLoads From From Alternate Source Source to to PrimalY Primary UPS UPS Unit Unit 1A.....

1A 34 34 8.2 8.2 Transferring UPS Loads From Alternate Source to Standby UPS Unit 18 18 .... 38 8.3 Transferring UPSUPS Loads From From Standby UPS Unit 'I1 B to Alternate Source .... 42 8.4 Transfening UPS Loads From Primary UPS Unit 'lA Transferring IA to Alternate Source ..... 44 8.5 Alignment of Standby UPS Unit 1lB B After a Loss of Alternate Source Power ................................................................................................................

Power. 46 8.6 Returning Standby UPS Unit 11 B to Normal Operating Condition Upon Regaining Alternate Source Power Supply. Supply ........... ...... ..... ......... ................ ..... 48 I OP-52 l'lOP-52 Rev. 35 Page 3 of 74 I Unit22 Unit APP UA-03 6-3 Page 1 of'l1 1 of PROCESS SMPL OG VENT PIPE DNSCllNOP DNSCIINOP (Process Sample Off-Gas Pipe Down-Inoperable)

Down-Inoperable)

AUTO ACTIONS NONE CAUSE CAUSE 1.

'I. Off-gas vent Off-gas vent pipe pipe (stack)

(stack) radiation radiation monitor monitor downscale downscale or out of or out of service.

service.

2.

2. Circuit Circuit malfunction.

malfunction.

3.

3. Change Change inin background background counts, counts, possibly possibly from from unitunit power power reduction.

reduction.

Categories Categories K/A:

KIA: 262002 262002 G2.0 G2.01.23 1.23 Tier Tier // Group:

Group: T2G1 T2Gl RO Rating:

RORating: 4.3 4.3 SRO Rating:

SRORating: 4.2 4.2 LP Obj:

LPObj: I11-15A 1-15A Source:

Source: NEW NEW Cog Level:

Cog Level: HIGH HIGH Category Category 8: 8: YY

83. The

83. The following following conditions conditions exist exist on on Unit Unit Two Two following following aa spurious spurious Main Main Turbine Turbine trip trip at at rated power:

rated power:

SDV HI-HI SDV HI-HI WTR WTR LVL LVL TRIP TRIP BYPASS BYPASS In alarm In alarm OTBD NSSS OTBD VALVES MTR NSSS VALVES MTR OVERLOAD OVERLOAD In alarm In alarm Reactor level Reactor level 185 185 inches inches and and steady Reactor Pressure Reactor Pressure psig with BPVs 900 psig BPVs controlling Rods All Control Rods Fully inserted Fully inserted Scram Being reset reset lAW lAW LEP-02 RWCU System RWCU Isolated Isolated by 2-G31-F001 2-G31 -FOOl The 2-G31-F004 (RWCU Outboard Isol lsol Vlv)

VIv) failed to automatically close on a valid isolation signal due to motor overload.

Which one of the following identifies the Technical Specification requirements when the RSP is exited?

The RSP can be exited to OGP-05, Unit Shutdown, provided an active LCO is implemented for Technical Specification (1) .

The start time of the LCO action completion time is when the (2)

A. (1) 3.3.1.1, Reactor Protection System (RPS) Instrumentation (2) condition occurred B. (1) 3.3.1.1, Reactor Protection System (RPS) Instrumentation (2) RSP is exited C. (1) 3.6.1.3, Primary Containment Isolation Valves (PCIVs) (PCIV5)

(2) condition occurred D (1) 3.6.1.3, Primary Containment Isolation Valves (PCIVs)

D!'" (PCIV5)

(2) RSP is exited

Feedback Feedback K/A: S295006G KIA: S295006G 2.02.22 2.02.22 Knowledge of Knowledge of limiting limiting conditions conditions forfor operations operations andand safety safety limits.

limits.

SCRAM SCRAM (CFR: 41.5 (CFR: 41.5/43.2/45 .2) 1 43.2 1 45.2)

RO/SRO Rating:

RO/SRO Rating: 4.0/4.7 4.0/4.7 CLSLP300C*1 I Objective: CLS-LP-300-C*11 Objective:

11. Given

11. Given plant plant conditions, conditions, the the Unit Shutdown Procedure Unit Shutdown Procedure (GP-05),

(GP-05), and and the the Reactor Reactor Scram Scram Procedure, Procedure, determine ifif conditions determine conditions allow allow exiting exiting the the Reactor Reactor Scram Scram Procedure.

Procedure.

Reference:

1 OCFR5O.36 10CFR50.36 OEOP-01-UG, Revision OEOP-01-UG, Revision 55, Page 31, Section 55, Page Section 3.5 3.5 Cog Level:

Cog Level: High High Explanation:

The EOPs authorize actions outside of technical specifications to mitigate the consequences consequences of an emergency condition. The EOPs EOPs also provide provide for returning returning the system or component to service.

service, IfIf the system or component is is not returned to its standby or operable condition prior to exiting the EOPs, EOP5, then the appropriate limiting condition of operation shall be implemented in accordance with Technical Specifications The starting time for the limiting condition of operation is the time that the EOPs were Specifications.

exited.

In order to exit EOP, compatibilty with GP-05 along with active LCOs need to be implemented. PCIV G31-F004 is inoperable, TS 3.6.1.3 Condition A (A1)requires isolating the affected penetration flow path by use of at least one closed and de-activated automatic valve, closed manual valve, blind flange, or check valve check valve wi1~q9Ll~l~J~q~~ilbiml&~l:Ii~(.~ify witf witj pursQ() jerify the affected penetration flow path is isolated Once per 31 days for isolation devices outside primary containment AND Prior to entering MODE 2 or 3 from MODE 4, if primary containment was de-inerted while in MODE 4, if not performed within the previous 92 days, for isolation devices inside primary containment.

Distractor Analysis:

Choice A: Plausible because OGP-01 would be entered in order to restart the reactor and TS 3.3.1.1 would be correct if the unit was in Mode 1I or 2 - SDV Hi level is not required in Mode 3.

Choice B: Plausible because OGP-05 is correct and TS 3.3.1.1 would be correct if the unit was in Mode 1 1 or 2 - SDV Hi level is not required in Mode 3.

Choice C: Plausible because OGP-01 would be entered in order to restart the reactor and TS 3.6.1.3 is correct.

Choice D: Correct Answer SRO Only Basis: Requires assessment of plant conditions (RPS SDV Hi Hi Level Bypass and Failed open PCIV) and prescribing a procedure procedure with which to proceed (OGP-05).

Notes

3.5 3.5 Technical Specifications Technical Specifications The EOPs The EOPs authorize authorize actions actions outside outside of of technical technical specifications specifications to to mitigate mitigate thethe consequences of consequences of an emergency condition.

an emergency condition. TheThe EOPs EOPs also also provide provide for for returning the returning the system system or or component component to service. IfIf the to service. the system system or or component component isis not returned not returned to to itsits standby standby or or operable operable condition condition prior prior to to exiting exiting the the EOPs, EOPs, then the then the appropriate appropriate limiting limiting condition condition ofof operation operation shall shall be be implemented implemented in in accordance with accordance with Technical Technical Specifications.

Specifications. The The starting starting time time for for the the limiting limiting condition of condition of operation operation is is the time that the time that the the EOPs EOPs were were exited.

exited.

OEOP-01-IJG IOEOP-01-UG Rev. 55 Rev. 55 Page 31 Page of 'IS" 31 of 151 I Completion Completion Times Times 1.3 1.3 1.0 USE AND APPLICATION

'1.0 APPLICATION 1.3

'1.3 Completion Times Completion Times PURPOSE PURPOSE The purpose of this section is to establish the Completion Time convention and to provide guidance for its use.

BACKGROUND BACKGROUND Limiting Conditions for Operation (LCOs) specify minimum requirements for ensuring safe operation of the unit. The ACTIONS associated *with with an LCO state Conditions that typically describe the ways in which the requirements of the LCO can fail to be met. Specified with each stated Condition are Required Action(s) and Completion Times(s).

DESCRIPTI DESCRIPTION ON The Completion Time is the amount of time allowed for completing a Required Action. itIt is referenced to the time of discovery of a situation (e.g., inoperable equipment or variable not within limits) that requires (e.g, entering entering an ACTIONS Condition Condition unless otherwise specified, providing providing the unit is in in a MODE MODE or specified condition stated in the Applicability of of the LCO. Required Required Actions must be completed prior to the the expiration of ofthe the specified specified Completion Completion Time.

Time. An An ACTIONS Condition Condition remains in in effect and and the Required Actions apply until the ConditionCondition no no longer longer exists or or the the unit unit isis not not within the LCO LCO Applicability Applicability..

WHITE 5-5 55 OTBD NSSS OTBD VALVES MTR NSSS VALVES MTR OVERLOAD OVERLOAD Page Page 11 of2 of 2 1.0 OPERATOR 1.0 OPERATOR ACTIONS: ACTIONS:

1.1 OBSERVE 1.1 OBSERVE Automatic Automatic Functions:

Functions:

1.1.1 IF 1.1.1 IF one one of of the affected valves the affected valves waswas being being operated, operated. THEN:

THEN:

1. Valve motion

1. Valve will stop motion will stop

2. Valve will

2. Valve will NOT NOT respond respond to to control control signals signals

3. Valve

3. Valve position position will will still still be indicated be indicated 1.2 PERFORM Corrective 1.2 PERfORM Corrective Actions:

Actions:

NOTE: Resetting NOTE: Resetting valve motor motor overload overload devices devices oror manual manual operation operaon of tripped of tripped motor-operated valves sl10uld motor-operated should onlyonly be be attempted attempted in in emergency Situations situations as directed by directed the Unit by the SCO.

Unit SeQ.

CAUTION During manual During manual operation oPeration of motor-operated motor-operated valves, personnel sh.ould should stand clear clear of the valve while vhiIe either:

1. Resetting the thermal overload

1. 3verload device device Ofor Operating the valve remotely.

2. Operating remo:ely.

1.2.1 IF IF the affected valve is required for operation, THEN PERFORM the following steps:

1. RESET the thermal overload device at the affected valve breaker compartment AND OPERATE the valve again.

themal overload device actuates again,

2. IF the themlal again. THEN MANUALLY OPERATE the valve.

3. WHEN the valve is broken off its closed or open seat, THEN RESET the themial overload device at the affected valve breaker compartment AND themlal OPERATE the valve.

REFERt0TS 1.2.2 REFER to T.S.. 3.6.1.3 and TRM TRMApp App. 0DTable Table 3.6.1.3-2.

24,PP-A-02 12APp-A-02 Rev. 32 Page Page 57 of 57 of 791

PCIVs 3.6.1.3 3.6 CONTAINMENT 3.6 CONTAINMENT SYSTEMS SYSTEMS 3.6.1.3 3.6.'1.3 Primary Containment Primary Containment Isolation Isolation Valves Valves (PCIVs)

(PCIVs)

LCO 3.6.

LCO 16.1.3'1.3 Each PC1V, except Each PCIV, except reactor reactor building-to-suppression building-to-suppression chamber chamber vacuum vacuum breakers, shall breakers, shall bebe OPERABLE.

OPERABLE.

APPLICABILITY:

APPLICABILITY: MODES 1, MODES 1, 2, 2, and and 3, 3, When associated VV'hen associated instrumentation instrumentation is is required required to to be OPERABLE per be OPERABLE per LCO 3.3.6.1, "Primary LCO 3.3.6:1, Primary Containment Containment Isolation Isolation Instrumentation."

Instrumentation.

CONDITION REQUIRED ACTION REQUIRED ACTION COMPLETION COMPLETION TIME A. NOTE-------

NOTE-----------

A.l A:I Isolate the affected Isolate 8B hours Only applicable to penetration flow path by penetration flow paths 'Nithwith use of at least one closed two PCIVs. and de-activated automatic valve, closed manual valve, blind flange, or check valve One or more penetration with flow through the valve flow flo ....' paths with one PCIV secured.

inoperable except for MSIV leakage not within limit. AND

3.3.1 :1 3.3.1.1 3.3 3.3 INSTRUMENTATION INSTRUMENTATION 3.3.1 .1 3.3.*'.'1 Reactor Protection Reactor Protection SystemSystem (RPS) (RPS) Instrumentation Instrumentation LCO 3.3.-l.'I LCO 3.3.1.1 The RPS The RPS instrumentation instrumentation for for each each Function Function in in Table 3.3.l.1I shall Table 3.3.'U-*1 shall be be OPERABLE.

OPERABLE APPLICABILITY:

APPLICABILITY: According to According to Table Table 3.3.13.3.1 ...1-i. 1-*1.

ACTIONS

NOT NOTE--- E ---------------------------------------------------------

Separate Condition Condition entry is is allowed for each channel. channel.

RPS Instrumentation 3.3.1.1 3.3.1.*'

Tt :t Tatfe 3~ 1.1'"~ (~3J~ .:.3 o~

3l.t c 3t R: Pf\:t~,:.~or.

R~3-:'>>r Pn:r Sj':~m. Oy: !n:.~".m~r:~l!:':f:

i:Irc

,%PPLICABLE

.'V'PLICASLE OC:NOION CO~OI7fONS

!MOD

..~ooszc,:(O REOAED RSQIJt:tED REHCE

,:t5FERENCEO OTHER C*"iHER. CHANNEL2 CHANNeLS FO AAOM 2PECWE (lPECIFIE::l i=E?t r,:w=

TA QLLAE FtEQIJ1,"iEO aVR\'EII.LAJIoICE.

2URVELLArCE "'1.L!:JW ALLOWABLE *.r..sLE FL4O2N FUNCttON OCNDI:ON CC'NOI7'ICNS 3OTE3 3'V!l"iE\1 AOT2r 0.1 ACtiON D. RSQUIREAlEma REOLJIREVENTO VAL,VE VALUE 7.

7. Ccr CEcharc Vr ScramOtscharoeVO\\m""e 1,2 1,;: 2 ~ OR ~.l.4.1";

WLHh W3!er Le'o!et-HIgn a.=t OR :;.3:.~.4.:S OR 3.3."L~LH a,;:(

a,=\

OR 3.3:.il.L1':'

CONDITION REQUIRED ACTION COMPLETION TIME A. One or more required Al A.'I Place channel in trip. *12 12 hours1.388889e-4 days 0.00333 hours 1.984127e-5 weeks 4.566e-6 months channels inoperable.

OR A.2 NOTE

NOTE------------- 12 12 hours1.388889e-4 days 0.00333 hours 1.984127e-5 weeks 4.566e-6 months Not applicable for Functions 2.a, 2.b, 2.c, 24, 2.d, or 2.f.

Place associated trip system in trip.

Categories Categories KJA:

KIA: S2950006G S2950006G 2.02.22 2.02.22 Tier / Group:

Tier! Group: T1GI TIG!

RO Rating:

RORating: 4.0 4.0 SRO SRORating: Rating: 4.7 4.7 LP Obj:

LPObj: CLSLP300C*11 CLS-LP-300-C* 11 Source:

Source: NEW NEW Cog Cog Level:

Level: HIGH HIGH Category Category 8: 8:

84. Following

84. Following aa scram scram onon Unit Unit Two, Two, which which one one of of the the following following correctly correctly identifies:

identifies:

(1) the (1) the initial initial response response of of reactor reactor water water level level ifif an an SRV SRV is is opened opened andand (2) the (2) the procedure procedure thatthat contains contains the the guidance guidance to close the to close the MSIVs MSIVs due due toto water water level?

level?

A. (1)

A. Shrink (1) Shrink (2) Reactor Scram Procedure (2) Reactor Procedure B. (1)

B. (1) Shrink Shrink (2) 2APP-A-07, (2) 2APP-A-07, REACTORREACTOR WATER WATER LEVEL LEVEL HIGH/LOW HIGH/LOW C

c~ (1) Swell (1)

(2) Reactor Scram Procedure Procedure (1) Swell D. (1) Swell (2) 2APP-A-07, REACTOR WA WATER TER LEVEL HIGH/LOW Feedback K/A: 295008 A2.05 KIA:

andlor interpret the following as they apply to HIGH REACTOR WATER Ability to determine and/or LEVEL:

Swell 41.10/43.5/45.13)

(CFR: 41.10/43.5/45.13)

ROISRO Rating: 2.9/3.1 RO/SRO Objective: CLS-LP-300-C, 10 Given plant Given conditions and plant conditions and the the RSP, RSP, determine determine the required operator actions.

Reference:

I 001-37.3 RSP 1001-37.3 2APP-A-07, page 12 Cog Level: High Explanation:

Opening of the SRV will cause the reactor water level to swell up due to the reduction in pressure in the vessel and if level reaches the value in figure 1 1 on the RSP then closure of the MSIVs is directed. The MSIVs will close automatically but only on low level 3, not high level.

Distractor Analysis:

Choice A: Plausible if the examinee thinks that opening the SRV would reduce the water volume in in the RPV, The RSP does contain the actions to close the MSIVs. MSIVs.

Choice B: Plausible if the examinee thinks that opening the SRV would reduce the water volume in the RPV, the examinee may think that the the closure is an auto action, which are contained in the APP.

Choice Choice C:C: Correct see explanation Correct see explanation Choice Choice D:0: Plausible because because reactor water level level will will swell, and and the examinee examinee may may think think that that the the closure is is an auto an action, which auto action, are contained which are contained inin the the APP.

APP.

SRO Basis:

Basis: 10 CFR 55.43(b)-5, 10 CFR 55.43(b)-5, Assessment of of facility facility conditions and and selection of of appropriate procedures during during normal, abnormal, and normal, abnormal, and emergency emergency situations.

situations.

Notes Notes

REACTOR WATER REACTOR WATER LEVELLEVEL HIGHlLOW I-IIGHFLOW PaOe 11 of Page 2 012 1.0 OPERATOR ACTIONS:

1.0 OPERATOR ACTIONS:

1 .1 CONFIRM 1.1 CONFIRM by by multiple multiple indications indications actual actual high high or or low low reactor reactor water water level:

level:

1.1.1 Reactor water level indication 1.1.1 Reactor water level indication on on RTGB RTGB Panel Panel P603 P603 may may I)e be used used for ror verification of verification of water level:

water level:

1. Reactor Water 1, Reactor Water Level Level .4" A, C32-Ll-R606.'\.

C32-Ll-R606A.

2. Reactor

2. Reactor Water Water Level Level B, B, C32-Ll-R606B.

C32-Ll-R6066.

3. Reactor

3. Reactor Water LevelLevel C, C, C32-U-R606C.

C32-Ll-R606C.

4. Reactor

4. Reactor Level/Pressure Level/Pressure Recorder, Recorder, C32-R608.

C32-R608.

1.2 OBSERVE Automatic 1.2 Automatic Functions:

Functions:

1.2.1 IF 1.2.1 reactor level IF reactor level decreases decreases to to '166 136 inches, inches, THEN THEN aa reactor reactor Scram Scram results.

results.

1.2.2 IF 1.2.2 IF reactor reactor level level increases increases toto 206 206 inclles, THEN the inches. THEN the Main Main Turbine.

Turbine, RFPTs, RFPTs, RCIC and turiines '.vill and HPCI tUri)ines will trip.

1.2.3 IF either of tile 1.2.3 IFeitller the RFPs have tripped AND reactor water level is less than 182 inches. THEN inches, TI-lEN a Recirculation Pump runback will occur.

2APP-A-07 12APP-A-07 Rev. 32 Page 12 Page 12 of 451 of 45 From the Reactor Scram Procedure:

flWAN J_Yt PL.CLALLMIV lTCHE1OCLO o1 LOWER REAtOR WATER WH RWCU

ATTACHMENT66 ATTACHMENT Page 19 Page of 19 19 of 19 FIGURE 21 FIGURE 21 Reactor Water Reactor Water Level Level at at MSL MSL (Main Steam Line Flood (Main Steam Line Flood Level) Level>

300 (Cl)

J)

LU W

MSL I

J:

()

(U)

--Z

.-J

.J REF lEG LU W LEG 250 TEMP LU W

ABQVEOR ABOVE OR EQUAL TO

..J

-j 200-F 200F Ca .r REF LEG

,. TEMP TEMP W BELOW

~

o 200"F 2COF cz z

200 11111 IIIIII lIIlllII!1lllIJIjI1,15o 100 00 300 300 500 500 700 900 13100 60 60 200 200 400 400 600 600 800 800 1,000 1,000 REACTOR REACTOR PRESSURE PRESSURE (PSIG) (PSIG)

WHEN WHEN REACTOR REACTOR PRESSURE PRESSURE IS IS LESS LESS THAN THAN 60 60 PSIG, USE INDICATED PSIG, USE INDICATED LEVEL.

LEVEL.

MSL MSL IS+250 IS +250 INCHES.

INCHES.

DEOP-Ol-UG OEOP-01-UG I Rev.

Rev. 55 55 Page 106 Page 106 of of 151 151

Categories Categories KJA:

KIA: 295008 A2.05 295008 A2.05 Tier// Group:

Tier Group: TlG2 T1G2 RU Rating:

RORating: 2.9 2.9 SRO Rating:

SRORating: 3.1 3.1 LP Obj:

LP Ubj: 300-C, 10 300-C,1O Source:

Source: NEW NEW Cog Level:

Cog Level: HIGH HIGH Category 8:8:

Category YY

85. Unit Two

85. Unit Two isis operating operating at 74% power at 74% when the power when the FW-V120, FW-V120,FW HTRS 44 && 55 BYP "FW HTRS BYP VLV, VLV, isis inadvertantly opened inadvertantly opened by by mechanics.

mechanics. TheThe valve valve isis bound bound and and can can not not be be reclosed.

reclosed.

Initial Final Initial Final Feedwater Feedwater Temperature Temperature was was 404°F.

404°F.

Conditions are Conditions are now now stable stable with with reactor reactor power power atat 8181%% and and Final Final Feedwater Feedwater Temperature at Temperature at 314°F.

314°F.

(Reference provided)

Which one Which one of of the the following following identifies identifies the the required required action action based based onon the the information information above?

above?

Continued operation:

Av is A'! is not allowed and reactor not allowed reactor shutdown is is required required lAWlAW OGP-05, OGP-05, Unit Unit Shutdown.

Shutdown.

B. is not allowed and a manual reactor scram is required lAW 001-01.01, BNP Conduct of Operations Supplement.

C. is allowed provided the FW Heaters 4 & 5 are isolated lAW 20P-32, Condensate and Feedwater Operating Procedure.

D. is allowed provided reduced thermal limits are established within 4 hours4.62963e-5 days 0.00111 hours 6.613757e-6 weeks 1.522e-6 months as required by Technical Specifications.

Feedback Feedback K/A: 295014 KIA: 295014 G2.01.25 G2.01.25 Ability to Ability to interpret interpret reference reference materials, materials, such such as as graphs, graphs, curves, curves, tables, tables, etc.

etc.

Inadvertent Reactivity Inadvertent Reactivity Addition Addition (CFR: 41.10/43.5/45.12)

(CFR: 41.10 / 43.5 / 45.12)

ROISRO Rating:

RO/SRO Rating: 3.9/4.2 3.9/4.2 Objective: CLS-LP-34, Objective: CLS-LP-34, Obj. 1 Ic Obj. 11c Given plant Given plant conditions, conditions, describe describe the the effect effect aa loss/malfunction loss/malfunction of of the the feedwater feedwater heaters heaters wi" will have have on:

on:

c. Feedwater

c. Feedwater Temperature Temperature

Reference:

20P-32, Attachment 44 (provided) 20P-32, Attachment (provided)

Cog Level Cog Level HIHI Explanation:

Explanation:

From Attachment 4 of 20P-32 operation is outside of the a"owable From allowable range <<11 0.3°F) this wil require a

(<1 10.3°F)

Unit shutdown lAW GP-05.

Distractor Analysis:

Choice A: Correct see explanation Choice B: Plausible because the 01 has a table with the Selected Out-of-Service Equipment Contingencies. In this case the FW heater meets the definition of the heater OOS and operation is permitted.

Choice C: Plausible because operation is allowed if the FW heaters are isolated but not with the final temperature greater than the 110.3 limit.

Choice 0:D: Plausible because if turbine bypass is inoperable with a FW Heater OOS then TS 3.7.6 requires this action.

SRO Basis: Assessment of facility conditions and selection of appropriate procedures during normal, abnormal, and emergency situations (43(b)(5) a. Assessement of the plant conditions and then prescribing the shutdown lAW the GP.

Notes

LL 0 a (4 -c LL U a) >

From OP-32, Attachment 4, Final Feedwater Temperature Vs Power E

2 (5 C) E a) C (5 a) ci) IE ci) 0 I- Co 0 ci) I RX Ck x - C 0 U

Nominal -

Nominal ) C

1*1O.3°F U :

PWR FVV Temp RlV Temp. c Reduced F'vV cl) ;

ReducedçL) 0 Temp 1)

LL

'10°F in in in in cc cc cc cc r r-100 C C C4 CJ cN C 429.0 4'19.0 328.7

0) CO I- W [C) t C) CI, 99 427.6 4'17.6 327.7 cc Lfl 98 426.5 4'16 . 6 . C 327.0 r cc cc cc in in

00) 0) I-. W [C)

) C1.

97 425.5 4'15.6 326.3

00) CO I-. CO [C) 96 424 . 4 4']4.6 325.7 95 t 1 423.4 4'13.6 325.0 94 C) C 422.4 4'12.6 324.3 c 93 42'1.4 4'I'L7 323.7 9,2 C C4 C C c4 420.4 4'10..7 323.0 91 419.5 t 409.8

322.4 ci - cc L -r c cJ r*

90 4-18.5 408.8 321.7 cc in a c cc 1

j 89 4'17.5 407.9 32'1.'1

t i iZ 88 416.5 406 .. 9 320.4

0) 0) 0) 0)0) 0) 0) 0) 0) 0) CO CO CO CO CO CO CO CO CO CO - I 1 I - I C) C) C C C 87 415.6 406.0 D 319.8 86 4'14.6 405.0 319:1 -

t t J f f C) C C 85 413.6 404.1 318.5 84 4-12.6 403 .. 1 317.8 -

COCCCQCO C CJ c c 83 41'1.7 402.2 317.2 -

82 4-10.7 40'1.2 316.5 -

81 409.7 400.3 315.8 C C CrJ 80 408.7 399.3 315.2 79 407.6 398.3 314.5 rzrr 78 406.6 397.3 313.8 (V) j 77 405.6 396.3 313:1 C C3 Cr) 76 404.5 395.3 312.4 CJ CJ 75 403.5 394.2 31'1.7 74 402.4 393.2 311.0 -

- C) Cr) 73 40'1.3 392.1 310.3

CAUTION CAUTION Unit operation Unit operation outside outside tile the bounds bounds of of the the Loss Loss of of Feedwater Feedwater Heating Heating analysis analysis isis prohibited.

prohibited.

9.

9. IF Step 8.7.2.8.c IF step 8.7.2.8c criteria criteria is NOT met, is NOT met. THENTHEN PERFORM PERFORM the following:

the following:

a.

a. IMMEDIATELY NOTIFY IMMEDIATELY NOTIFY the Unit seo.

the Unit SCO b

b. RESTORE unit RESTORE operation within unit operation within the the bounds bounds of of the cycle the cycle Loss Loss of of Feedvvater Feedwater Heating Heating analysis analysis OR OR c

c. COMMENCE unit shutdown COMMENCE shutdown in in accordance 'ljVith with OGP-05.

20P-32 120P-32 Rev. 165 165 Page 1-17 117 of 300 1 Permitted Condition Condition Operation Operation Comment 005 Single DOS Sin gI (See NOTES)

FWHOOS FWHOOS FWR I,FFTR)

FW,R FFrR Yes Yes Defined as a 10°F ID F or greater reduction in Defined as a 10 'F Delined n nom inal feedwater temperature.

nominal F or greater reduction inn feedwater feecwater temperature.

Defined as a cycle extension strategy.

MSIVOOS MSIIfOOS Yes-base

MSIVOOS permits 1 I MSIV to be inoperable.

MSIVOOS. THEN thermal power shall be limited to 70% of rated.

IF MSIVOOS_

TBPCCS TepOOS Yes

TSPOOS assumes all a! turbine bypass valves (I BV) are inoperable.

(ThV) incoerable.

SLO SLO ODS Combination (See 005 Yes (See NOTES)

- Permitted with a thermal limi: penalty.

ihemlallimii TBPOOS TSPOOS & WH003 TBPOOS TBPOOS

&& FWTR FlNHOOS Yes Yes **

Combined 003 Combined 003 OOS condition is permtted condition s 005 condtion permitted with a thermal limit penalty.

is permitted with a thermal limit penalty.

imit penatty.

FltHR (FFTR)

(FFTR)

Operefinq Operating Power/Flow Power/Flow Map Power Map'" ICF Power Coas:down Coastdown ICF Yes-base Yes-base Permitted Pem,iited operations Permitted Pemlitted operations with thermal operations with operations with thermal limits with thermal limits defined thermal limits defined by limits detned by 003 defined by OOS condition.

by OCS condition.

OOS condition.

condition.

Turbine Control Turbine RCR Control Mode Pump Per RCR Pump Mode Pwr Source Source Yes-base Yes-base Yes-base Yes-base

- Partial Partial arc Power aro cpera:ion Power source operation isis supported source can supported by be protiided can be provided by by safety by the saier{ analysis the UAT UAT or analysis for or SAT SAT or for all for all all 005 all ODS OOS conditions conditions.

OOS conditions.

conditions.

Yes:

Yes: Operations are Operations permitted with are permitted with restrictive thermal limits.

restrictive thermal limits.

Yes-base:

Ye.s-base: Operations Operations are are permitted permitted with with base thermal limits.

base thermal limits. fo No thermal flnit changes thermallirnit changes are are required.

required.

001-01.01 1001-01.01 Rev. 29 Rev. 29 Page 121 Page of '1771 12-1 of 177 Categories Categories K/A:

KIA: 295014 295014 G2.01.25 G2.01.25 Tier / Group: T1G2 Tier/Group: T1 G2 RORating:

RO Rating: 3.9 3.9 SRO SRO Rating:

Rating: 4.2 4.2 LPObj:

LPObj: 34-11C 34-11C Source:

Source: NEW NEW Cog Cog Level:

Level: HIGH HIGH Category 8:8:

Reference:

OAOP-15, Revision 23, Page Page 11, 11, Section 3.2.14 3.2.14 Cog Level:

Cog Level: High High Explanation:

During conditions in which there is no circulation, the reactor vessel water level, as read on B21-LI-R605A(B), should be maintained between 200" 821-Ll-R605A(8), 200 and 220",

220, or as directed by the Shift Superintendent based on plant conditions, until forced circulation is restored. With a LOOP present and no actions taken to restore Off-site power (not provided in the question), the only available means of decay heat removal is alternate shutdown cooling utilizing SRVs.

Distractor Analysis:

Choice A: Plausible because 200 inches is correct and OP-17 contains actions for using FPC and SSFPC, but these would only be available if the reactor head is removed and fuel pool gates removed.

Choice B: Correct Answer Choice C: Plausible because 254 inches is the level of the MSLs and could be confused with Natural Circulation level due to the requirement to be at this level during alternate SDC and OP-17 contains actions for using FPC and SSFPC, but these would only be available if the reactor head is removed and fuel pool gates removed.

Choice D: Plausible because Plausible because 254 inches is the level of the MSLs and could be confused with Natural Circulation level due to the requirement to be at this level during alternate SDC SRO Only Basis: Assessment of facility conditions and selection of appropriate procedures during normal, abnormal, and emergency emergency situations (43(b)(5)a. Requires assessing plant conditions (LOOP, Mode 3, power availability, impact of power losses) and prescribing correct section of a procedure to provide DHR.

Notes

2.0 2.0 AUTOMATIC ACTIONS AUTOMATIC ACTIONS Loop /\(8)

Loop A(B) INBOARD INBOARD INJECTION INJECTION 1I.4L V4LVE VE, 0 E11-FO15.4iB, will E11-F015A(B), will close close (Low (Low Level Level One One Only)

Only)

The RHR The RHR Pump Pump inin service service for for Shutdown Shutdown Cooling will trip Cooling will trip 0 on aa loss on loss of of suction suction path.

path.

3.0 OPERATOR ACTIONS OPERATOR ACTIONS 3,1 3.1 Immediate Actions Immediate None None 3.2 Supplementary Actions ii CAUTION CAUT.lON IfIf reactor coolant temperature ten,pera:ire is greater grea:er tllan 212rF and than 21.2°F and reactor reac:or water level has been l:ieen raised to greater than than 218 212 inches foror 10 minutes minLtes or more, more, a false RPV low level signal could result when the reference leg condensing condensing pot N12A(B)

N12A(B: nozzle n3zzie is uncovered as subsequentl lowered below 218 level is subsequently 21S inches.

inches.

3.2.1 IF Shutdown Cooling has been lost due to a tripped RHR Pump, THEN START an RHR Pump in the loop being o[]

used for Shutdown Cooling.

NOTE: During conditions in which there is no circulation, the reactor vessel water 82f-LI-R6OA8, should be maintained between level, as read on B21-Ll-R605A(B), between 200" 200 and 220, or as directed by the Shift Superintendent based on plant conditions, 220~, conditions.

until rorced unol forced circulation is restored.

3.2.2 IF forced circulation has been lost, AND natural 0 circulation has NOT been established, THEN RESTORE AND MAINTAIN reactor vessel water level.

OAOP-15.0 IOAOP-15.0 Rev. 23 Page Page 33 of 21 of21 I

10 3.0 OPERATOR ACTIONS OPERATOR ACTIONS j.j, IF the IF the reactor reactor coolant coolant temperature temperature isis less less than than 212°F, THEN 212°F, THEN ENSURE ENSURE the the following following valves valves are are open:

open:

- INBOARD RX iNBOARD RX HE.4D HE4D VENT VENT VLV, VLV. B21-F003 821-F003 0

- OUTBOARD RX OUTBOARD PXHEAD HEAD VENT VENT VLV VLY. B21-F004.

B21-F004. 0U k.k. MAINTAIN RHR in MAINTAIN in Shutdown Cooling accordance with 11(2)OP-17.

accordance (2)OP-17.

in Cooling in oU

& IFIF RHR RHR has has NOT 3.2.11.5, THEN 3.2.11.5, NOT been THEN PLACE been restored restored in PLACE the RHR in accordance RHR loop accordance with StepStep loop that was operating oU in Shutdown Cooling in Shutdown back in Cooling back in service in in accordance with (2)OP-1 7 as soon as conditions permit.

1I (2)OP-17 3.2.12 IF necessary to minimize reactor coolant temperature IF rise,, THEN PERFORM one of the following feed and rise oU Not AvaD (LOOP)P) II bleed combinations:

I~ot Not Avail (RPS notJ not reset)

FEED BLEED CONDIFW in accordance with RWCU Reject in accordance CONDJFW 1 (2)OP-32 1(2)OP-32 with 1(2)OP-14 CRD in accordance with Reactor Water level Level Control 11(2)OP-08 (2)OP-08 using Main Steam Lines in accordance with 11 (2)OP-32.

Core Spray in accordance Maintaining RPV Level Using with 1(2}OP-18 1(2)OP-18 LPCI in accordance with the Main Steam Line Drains with 1 (2)OP-25.

1(2)OP-25.

I jvail (LOOP) I 1(2)OP-17 1(2)OP-1 7 3-23 IF NEITHER RHR loop can lJe be placed in Shutdown 0 U

Cooling, THEN PLACE the Condensate System in Condenser Cooling in accordance with "vith 1(2}OP-32.

1(2)OP-32.

OAOP-15.0 IOAOP-15.0 I Rev.23 Rev. 23 Page 10 Page 10 of of21 21 I

3.0 3.0 OPERATOR ACTIONS OPERATOR ACTIONS 3.2.14 3.2.14 IFIF ALL ALL of of the the allove above methods methods cancan NOTNOT maintain maintain reactor reactor vessel coolant vessel coolant temperature temperature below below 212°F, 212°F, THEN THEN INITIATE alternate INITIATE alternate Shutdown Shutdown Cooling Cooling with.

with the the SRVs SRVs asas follows:

follows:

1.

1. ENSURE ALL ENSURE control rods ALL control rods are are funy fully inserted.

inserteth 0 2.

2. CONFIRM reactor CONFIRM reactor vessel vessel head head isis installed installed and and 0 tensioned.

3.

3. IF the Reactor IF Reactor Recirculation Recirculation Pumps Pumps are are running:

running, THEN PERFORM the PERFORM the following:

a.

a. RAISE AND MAINTAIN RAISE MAINTAIN reactor reactor water water level level 0 between 200" 200 and 220" 220 as read on 321-LI-R6O548J. or as directed by Shift B21-U-R605A(8),

Superintendent I)asedbased on plant conditions.

b.

b. running Reactor Recirculation Pumps STOP the running Pumps in 0 accordance with 11(2)OP-02.

(2)OP-02.

4. SHUT DOWN the RHR RI-IR loop that was operating in 0U Cooling in accordance with 1(2)OP-17.

Shutdown Cooltng

5. PLACE one RHR loop in the Suppression Pool Cooling 0U mode in accordance with 11(2)OP-17.

(2)OP-17.

6. IF Suppression Pool temperature rises above 95 F, 95°F.

Q 0U THEN GO TO OEOP-02-PCCP, Primary Containment Control Procedure AND PERFORM CONCURRENTLY with this procedure.

J OAOP-15.O IOAOP-15.0 Rev. 23 Page Page 11 11 of 21 I Categories Categories K/A:

KIA: S295021 S295021 A2.03 A2.03 Tier/Group:

Tier / Group: T1G1T1 G 1 RO RO Rating: 3.5 3.5 SRO SRO Rating: 3.5 3.5 LP LP Obj:

Obj: CLSLP120*06 CLS-LP-120*06 Source:

Source: ~vv NEW Cog Level:

Cog Level: NIGH HIGH Category Category 8:8:

87. While

87. While performing performing refueling refueling activities activities on on Unit Unit Two, Two, aa spent spent fuel fuel bundle bundle was was dropped dropped and and the following the following alarms alarms were were received:

received:

AREA RAD AREA RAD REFUEL REFUEL FLOOR FLOOR HIGH HIGH PROCESS RX PROCESS FV( BLDG BLDG VENT VENT RADRAD HIGHHIGH Which one Which one of the following of the following identifies:

identifies:

(1) the (1) the immediate immediate operator operator action action that that is is required required to to be be performed performed and and (2) the bases for the performance of this (2) the bases for the performance of this action? action?

A. (1)

A. (1) Standby Standby Gas Gas Treatment Treatment (SBGT)

(SBGT)

(2) Ensures control room operators (2) Ensures control room operators will will receive receive .:5. 22 Rem Rem TEDE TEDE B. (1)

B. (1) Standby Standby Gas Gas Treatment Treatment (SBGT)

(SBGT)

(2) Ensures control room operators (2) Ensures control room operators will receive receive .:5. 55 Rem

. Rem TEDE TEDE C. (1) Control Room Emergency Ventilation (CREV)

(2) Ensures control roomroom operators will receive .:5.<2 2 Rem TEDE D

D~ (1) Control Room Emergency Ventilation (CREV)

(2) Ensures control room operators will receive .:5. 5 Rem TEDE

Feedback Feedback K/A: S29S023G KIA: S295023G 2.04.49 2.04.49 Ability to Ability to perform perform without without reference reference to to procedures procedures those those actions actions thatthat require require immediate immediate operation operation of system of system components components and and controls.

controls.

Refueling Accidents Refueling Accidents (CFR: 41.10 (CFR: 41.10 //43.2/45.6) 43.2 /4S.6)

ROISRO Rating:

RO/SRO Rating: 4.6/4.4 4.6/4.4 CLSLP302..J*02 Objective: CLS-LP-302-J*02 Objective:

2. Given plant conditions with

2. Given plant conditions with spent spent fuel fuel damage damage and and aa high high airborne airborne activity activity problem problem in in progress, progress, determine ifif the determine the appropriate appropriate automatic automatic actions actions have have occurred occurred in in accordance accordance with with OAOP-S.O, OAOP-5.0, Radioactive Radioactive Spills, High Spills, High Radiation, Radiation, and and Airborne Airborne Activity.

Activity.

Reference:

OAOP-05, Revision OAOP-OS, Revision 23,23, Page Page 2, 2, Section Section 3.1

3.1 Level

High Cog Level: High Explanation:

OAOP-05 immediate action for a dropped or damaged fuel assembly is to ENSURE CREVS is in OAOP-OS operation.

The dose consequence calculation for the fuel handling accident does not credit the secondary containment or automatic CREVS start, however, it does assume that CREVS is manually initiated within 20 minutes of a dropped/damaged fuel assembly. Based on this analysis, Technical Specifications do not require secondary containment or CREVS automatic initiation instrumentation except during Modes 1, 2, or 3 or during operations with the potential to drain the Reactor vessel. The CREV System is designed to maintain a habitable environment in the CRE for a 30 day continuous occupancy after a DBA without exceeding S 5 rem total effective dose equivalent (TEDE).

Knowledge of DBA analysis initial conditions.

Distractor Analysis:

Choice A: Plausible because PROCESS RX BLDG VENT RAD HIGH annunciator is easily confused with the auto start for SBGT verifing Auto actions can be confused with Immediate Actions. SBGT start is a supplemental action which will reduce control room dose and 2 Rem TEDE is a site administrative dose limit and can be confused with the actual Dose Analysis from FHA of 2.69 rem TEDE.

Choice B: Plausible because PROCESS RX BLDG VENT RAD HIGH annunciator is easily confused with the auto start start for SBGT verifing Auto actions actions can be be confused with Immediate Immediate Actions. SBGT start is is aa supplemental supplemental action action which which will reduce control room dose dose and 5S Rem TEDE is is correct.

Choice Choice C:C: Plausible because because CREV CREV is is correct and and 22 Rem TEDETEDE is is aa site site administrative dose dose limit limit and and can be be confused confused withwith the the actual actual Dose Dose Analysis from from FHA of 2.69 FHA of 2.69 rem TEDE.

TEDE.

Choice Choice D: Correct Answer.

D: Correct Answer.

SRO SRO Only Only Basis:

Basis: Conditions Conditions and and limitations limitations in in the the facility facility license license (43(b)(1)

(43(b)(1)

Notes Notes

Unit 22 Unit APP U.A.-03 APP UA-03 3-73-7 Page 11 of Page of 11 AREA RAD AREA RAD REfUEL REFUEL FLOOR FLOOR HIGHHIGH AUTO ACTIONS AUTO ACTIONS NONE NONE CAUSE

'1.1. High radiation level High radiation level in in the the cask cask wash wash area.

area.

2.

2. Circuit malfunction.

Circuit malfunction.

3.

3. Refueling cavity Refueling cavity water water sealseal failure.

failure.

OBSERVATIONS OBSERV.A.TIONS 1.

'1. ARM indicator ARM indicator and and trip ip unitunit Upscale Upscale light illuminated on light illuminated on Panel Panel HH 12-P600.

12-P600.

ACTIONS ACTIONS 1.

1. Refer to EOP-03-SCCP, EOP-03-SCCP, Table 3; 3; enter EOP-03-SCCP EOP-03-SCCP as appropriate.

appropriate.

2. Refer to Refer to AOP-OS.O, AOP-05.0, Radioactive Radioacte Spills, Spills, High Radiation, and High Radiation, and Airborne Activity.

3. Suspend refueling Suspend refueling operation ifif due to fuel pool pool low low !evel level from refueling refueling cavity cavity water seal leakage.

4. If a circuit malfunction is suspected, ensure that a Trouble Tag is prepared.

DEVICEISETPOINTS DE\iICE/SETPOINTS ARM Channel 29 K2 1<2 rnRPhr 40 mRlhr POSSIBLE PLANT EFFECTS 1.

1. Suspension of refuel floor activities.

REFERENCES REFERENCES 1.

1. LL-9353-39 LL-9353 - 39

2. AOP-05.0

3. EOP-03-SCC EOP-03-SCCP P 2APP-LIA-03 12APP-uA-03 Rev. 46 Rev. 46 Page 34 Page 34 ofof 631

Unit Unit 22 APP APP U.A.-03 UA-03 4-54-S Page'l Page 1 of of 11 PROCESS RX PROCESS BLDG VENT RX BLDG VENT RADRAD HIGHHIGH AUTO ACTIONS AUTO ACTIONS NONE NONE CAUSE 1.

1. High airborne activity High activity in in Reactor Reactor Building Suiding ventilation exhaust exhaust plenum.

p!enum.

2.

2. Circuit malfunction.

Circuit malfunction.

OBSERVATIONS OBSERV.A.TiONS 1.

'1. Reactor Building Reactor Building Vent Rad Rad Recorder Recorder D12-RR-R605 D12-RR-R605 Channel ChannelAA or or B B indicates indicates high high radiation level.

2. Reactor Building Exhaust Plenum Rad Rad Monitor Channel A or B indicates greater rnRlhr on Panel than 3 mRlhr Panel H12-P606.

H12-P6G6.

ACTIONS

1. EOP-03.SCCP. Secondary Containment Control.

Enter EOP-03-SCCP, Conti-o.

2. Refer to AOP-OS.O, Radioacte Spills, AOP-05.O, Radioactive SpiIs, High Radiation, and Airborne Activit'!.

Activity.

3. If a circuit malfunction is suspected, ensure that a Trouble Troube Tag is prepared.

DEVICEISETPOINTS DEVICE/SETPOINTS D12-RR-R605 red or black pen D'12-RR-R605 3 mRlhr mRihr POSSIBLE PLANT EFFECTS

1. Possible release to environs.

2. niRihr. Reactor Building HVAC isolation, a Group If airborne activity increases to 4 mRlhr, 6 isolation, dr'Jwell drjwell purge isolation, and initiation of the Standby Gas Treatment System ifl1

\'Iill occur.

REFERENC REFERENCESES

1. LL-9353 - 35

2. AOP-O5.O AOP-05.0

3. EOP-03-SCCP

4. Plant Modification 85-081 2APP-UA-03 12APP-uA-03 Rev. 46 Page 41 Page 41 ofof 63 631

1.0 1.0 SYMPTOMS SYMPTOMS 1.1 1.1 AREA RAD RAD REFUEL REFUEL FLOOR FLOOR HIGHHIGH (UA-03 (UA-03 3-7) 3-7) is in alarm.

is in alarm.

1.2 1.2 AREA RAD AREA RAD NEWNEW FUELFUEL STORAGE STORAGE HIGH HIGH (UA-03 (UA-03 4-7) 4-7) isis in in alarm.

alarm.

1.3 1.3 PROCESS RX PROCESS BLDG VENT RX BLDG VENT RAD RAD HI i-il (UA-03 (UA-03 4-5) 4-5) isis in in alarm.

alarm.

1.4 1.4 TVRB BLDG TURB BLDG VENTVENT RADRAD H.IGH HIGH (U.4-03 (U.4-03 3-3)3-3 is is in in alarm.

alarm.

1.5 1.5 Area Radiation Area Radiation Monitor Monitor (ARM)

(ARM) is is in alarm.

in alaml.

1 .6 1.6 Continuous ,",ir Continuous Monitor (CAM)

Air Monitor (CAM) is inin alarm.

1.7 1.7 Turbine Building Turbine Building once-through once-through effluent effluent monitor monitor indicates indicates elevated elevated (higher (higher than expected than expected or or an an unanticipated unanticipated increase) increase) activity.

activity.

1.8 Routine surveys indicate high radiation, contamination and/or andlor airi)ome airborne activity.

1.9 spill. leak, Report of spill, leak. or potential damage to ne'.vnew or spent fuel.

2.0 AUTOMATIC ACTIONS 2.1 IF PROCESS RX BLDG VENT RAD HI-HI (UA-03 3-5) is in alam, THEN the following actions occur:

alam1, Reactor Building Ventilation isolation 0 SBGTSautostart SBGTS auto start 0U Group 6 Isolation. 0 3.0 3,0 OPERATOR ACTIONS 3.1 Immediate Actions

)) 3.1.1 IF aa fuel assembly was dropped or damaged, THEN ENSURE the Control Room Emergency Ventilation U0 System (CREVS) is in operation.

OAOP-05.O IOAOP-OS.O Rev. 24 24 Page Page 22 of of 10 10 I

UPDATED UPDA TED FSAR FSAR evision:

Revision: 21 21 Cr&Ll Pmm, ENGInEERED SAFETY ENGINEERED SAFETYFEA FruREs TURES Chapter:

Chapter:

Page:

Page: 108 oi 108 o 121 66 121 6.44.12 Fuel 6.4.4.1.2 Fuel Handling Handling Accident Accident - Control

- Control RoomRoom Dose Dose Section 15.7.1 Section 15.7.1 discusses discusses the the release release ofof aotivity and its activity and its transport transpoi to to the the environment environment following following aa postulated postulated fuel handling fuel handling aoodent accident (FH (FHA).

..o,).

The design The inpiAs utilized design inputs utilized to to ellaluate evalua:e the tne intake & this ntake of ths acjjvjt~* into the acv,ty into the conlrol control room room and and to to assess assess thethe resultant dose resultant dose to the control to the control room room operators operators are are tabulated tabulated in Table 6-28.

in Table 6-22. AA sensitivity sensitivity study study of of unfiltered unuttered outsde air outside air inleakage inleakage into nto the the oontrol control room room 'lias performed ellaluating was performed evaluating ininleakage rates of leakage rates of 10.000 10,030 ofmofm (bouxidirig case).

(bounding case). 30003000 cfrncfrn {control

{confrol rcom room design).

design), and and 00 cfm.

cfm, Accident Acodent XiQ XIQ values values are are developed developed as as discussed in discussed Section 15.9.2.

in Secticn 15..2. SectionSection '15.9.3 descr bes the 1.9.S describes the parameters parameters utilized utilized in in conjunction conjunction with with the the AQTRAO computer RADTRAD computer code Reterence 8-35}

code {Reference to con~'ert 8-35} to convert the Alternative Source the Alternative Source Tern, Temi activity activity drawn drawn into into the contrcl the control room room during during the postulatec accident the postulated accident into into aa total effective dose total effeclive dose equivalent equivalent (TEDE)

TEDE) dose. dose.

Th 30-day The 30-day FHA FHA dose dose to to the the conirol control room room operator operator from from thethe internal internal cloud cloud associated associated with with the FHA is the FHA is calculated to c,3lculated to be 2.69 rem be 2.69 rem TEOE.

TEQE.

The onsile The onsi:e control control room room operator operator dosedose criterion established by criterion established by Reference Reference 8-368-36 for for this this accident accident is is that that the the total control room total contro!! room cperator operator dose dose should should be be less the to than the less than 10 CFR CFR 50.67 50.67 guidelines; guidelines: i.ei.e.,

.* that the total that the total dose dose should be should be less than 5S rem less than rem TEDE.

TEDE.

3.0 3.0 OPERATOR ACTIONS OPERATOR ACTIONS 3.2.3 3.2.3 IFIF new new oror spent fuel damage spent fuel damage isis suspected, suspected, THEN THEN PERFORM the PERFORM following:

the following:

1.

1. PLACE any PLACE any fuel fuel that that is is being moved in being moved in aa safe safe condition.

condition. D 2.

2. SECURE further SECURE further fuel fuel movement.

movement. D 3.

3. EVACUATE personnel trom EVACUATE personnel from the the following following areas:

areas:

- Refueling Floor Refueling Floor D0

-- Drywell, ifif occupied Drywell, D

-- Reactor Building, Reactor Building, -17'-17 Elev.,

Elev., ifif Shutdown Shutdown Cooling Cooling in in D service.

-- ECCS Pipe Tunnel DU

- Any area determined to have the potential for high D radiation.

4. ISOLATE Secondary Containment Containment. D

5. START Standby Gas Trains. D U

3.2.4 NOTIFY E&RC to perform the following as necessary:

- Area radiation survey D U

- Air sampling D U

- Smear survey D U

- Posttheaffect Post the affected edareaasnece area as necessary ssary D U

Control access to reduce exposure and D U

contamination.

DAD P-05.0 IOAOP-05.0 Rev Rev. 23 Page 44 of 10 10 I

4.0 4.0 GENERAL DISCUSSION GENERAL DISCUSSION Liquid radioactive Liquid radioactive spills spills may may be be caused caused by by valve valve packing packing leaks, leaks, leaky leaky fittings, fittings, system leaks, system leaks, or or system system draining draining evolutions.

evolutions. Liquids Liquids spills spills should should bebe covered covered with with an absorbent an absorbent material material to minimize the to minimize the spread spread of of contamination.

contamination. Solid Solid spills spills may may I)e be caused by caused by leaks leaks from from the the containers containers oror process process streams streams which which handle handle radioactive radioactive material or by material or by an an accident accident during during the transport of the transport of new new or or spent spent fuel, radioactive radioactive sources, or sources, or other other solid solid radioactive radioactive materials.

materials. Solid Solid spills should should bebe covered covered by by aa damp material damp material toto minimize minimize the the spread spread ofof airborne airborne contamination.

contamination. A A spill spill of of highly highly radioactive solid radioactive solid materials materials such such as as spent spent resin, resin, filter filter sludge, sludge, neutron neutron sources, sources, oror irradiated reactor internal irradiated internal components may create aa serious personnel personnel exposure problem and should be handled with extreme caution. In addition, addition, high high radiation high airborne acUvity and higl1 activity may accompany aa spill.

High airborne High airborne activity may occur activity may occur from reactor reactor coolant coolant leaks, leaks, coolant coolant spills, radwaste leal,s, radwaste leaks, sampling, grinding, draining, draining, and other maintenance.

maintenance. High High airborne activity in the turbine buildings may require ventilation shutdown or realignment to the recirculation lineup if the ventilation systems are operating in the once-through lineup.

levels may be caused High radiation [evels caused by radiation "streaming,"

streaming, loss of or degraded element damage, high airi)orne shielding, fuel element airborne activity, coolant spillS, spills, or radiography.

New or spent fuel damage may occur within the plant during fuel handling operations. Fuel may be damaged if it is inadvertently dropped oral[owed or allowed to collide with objects. Damage may also be sustained if heavy objects (shipping casks, reactor vessel head, dlY'Nell drywell head, etc.) are allowed to fall on the fuel. These accidents may release a substantial amount of radioactive noble gases, halogens, and other fission products into the secondary containmentcontainment. The secondary containment will be automatically isolated due to high radiation at its ventilation exhaust plenum. Although Standby Gas Treatment (SBGT) System will reduce the activity released to the environs, there is a activir; a chance that technical speCification specification limits may be exceeded.

The dose consequence calculation for the fuel handling accident does not credit the secondary containment or automatic CREVS start, however, it does assume that CREVS is manually initiated within 20 minutes of a droppedfdamageddropped(damaged fuel assembly. Based on this analysis, Technical Specifications do not require secondary containment or CREVS automatic initiation instrumentation except during Modes 1, 1, 2, or 33 or during operations with the potential to drain the Reactor vessel.

OAOP-05.0 IOAOP-05.0 Rev.

Rev. 23 23 Page Page 8B of of 10 10 I

UPDATED UPDA TED FSARFSAR Revision:

Revion: 21 21 CP&L EzIGIIIEEREDSAFETY ENGmEERED SAFETYFEA CHAPTER 66 TABLES FEATURES TABLES TURES Chapter:

Chapter:

Page: 11 of 63 of 1 CHAPTER TABLE 6-28 TABLE 6-28 Control Room Control Room Design Design Inputs Inputs - Design Basis Accidents Design Basis Accidents Control Room

1. Control room habitabi!ity vo ume 2g8,650 flf 2gS.650

2. Assumed unfiltered inIeakage 10.000 crn 10.000 cim' Control Room Ventilation

1. Normal mode operation outside air intake 2,1C0 cfm 2.100 cm

2. Normal mode roughing filter.

e - roughing filter, aerosol aerosol removal removal 0%

0%

3. Normal mode roughing filter, elemental iod:ne removal 0%

0%

4. Normal mode roughing filter, organic iodine removal 0%

0%

5. Time of manual switchover from normal to radiation mode minutes 20 minutes'"

20

0. Radiation mode operation outsde sir ntake I .500 cfm 1.500 cfm

7. Radiation mode HEPA ft ter. aerosol removal 95%

. Radiation mode charcoal filter, elemental iodine removal 90%

. Radiation mode charcoal fiter,* organic iodine removal 90%

10. Radiation train charcoal depth 2inehes 2 inches

11. Radiation mode filtered recrculated airflow 400 40 cfm

12. Radiation mode - aerosol iodine removal 95%

13. Radiation mode - elemental iodine remollal removal 90%

14. Radiation mode organic iodne removal 90%

NOTES NOTES Sensitivity cases using 3.000 3.00 cfm a,nd cThi unfiliered and 0 dm unfiltered outside air inleakage into the control room were also into also evaluated. The 10,000 10.000 cfm cim unfiltered inleakage case is bounding boundi,ng for the LOCA, lOCA. the FHA, FHA. and the CRDA eRDA events.

ellents.

For the MSLB MSLB event, ellent. 0 cfm unfiltered outside air inleakage nleakage represents the bounding vaiue.

value.

For For the MSLB MSLB event, ellent. aa 5ensitlvity sensitivity study study was performed, isolating the control performed. isolating contrel room room at at various various times between between 5.05.5 seconds seconds and and 30 30 days.

days.

CREV CREV System System BB 3.7.3 37.3 BASES BASES BACKGROUND BACKGROUND The CREV The CREV System System is is designed designed to to maintain maintain aa habitable habitable environment environment in in the the (connued}

(continued} CRE for CRE for aa 30 30 day connuous occupancy day continuous occupancy after after aa DBA DBA without without exceeding exceeding rem total 55 rem total effecti effective dose equivalent

.... e dose equivalent {TED(TEDE), E). A A single single CREV CREV subsystemsubsystem operating at operating at aa f10Vl flow raterate of of s: 2200 2200 cfmcfm ',viii will sligh!ly slightly pressurize pressurize the the CRE CRE relative to relative outside atmosphere to outside atmosphere to to minimize minimize infiltration infiltration of of air air from from surrounding areas surrounding areas adjacent adjacent to to the the CRE CRE boundary.

bounday. CREV CREV System System operation in operation maintaining CRE in maintaining CRE habitability habitability is is discussed discussed in in the the UFSAR, UFSAR, Sections 6.4 Sections 6.4 am!

and 9.4,9.4, (Refs.

(Refs. *11 and and 2,2. respectively).

respectively).

APPLICABLE APPLICABLE ability of The ability of the the CREV CREV System to maintain maintain the habitability habitability ofof the CRE CRE SAFETY .ANAL SAFETY ANALYSES is an YSES is an explicit explicit assumption assumption ror the design for the design basis basis accident accident presented presented in in the the UFSAR (Ref.

UfSAR (Ref. 3}.3). The The radiation/smoke radiation!smoke protection protection. modemode ofthe of the CREV CREV System is System is assumed assumed (explicitly (explicitly or or implicitly) implicitly) to to operate operate fol!owing following aa DBA. DBA.

The radiological doses to the CRE occupants as aa result of a DBA DBA are summarized in ReferenceReference 3. Postulated single single active failures that may may cause the loss of outside or recirculated air from the CRE CRE are bounded by BNP radiological dose calculations for CRE occupants.

BNP Brunswick Unit 2 S 3.7.3-2 B 3.7,3-2 Re .... ision No. 61 Revision Categories K/A:

KIA: S295023G2.

S295023G 04.49 2.04.49 Tier/Group:

Tier / Group: T1G1 TIGl RO Rating:

RORating: 4.6 SRO Rating: 4.4 SRORating:

LP Obj:

LP CLSLP.3O2J*O2 CLS-LP-302-J*02 Source: NEW Cog Level: HIGH Category 8:

88. An event

88. An event on on Unit Unit One One has has resulted resulted in in the the following following plant plant conditions:

conditions:

Reactor pressure Reactor pressure 1000 psig 1000 psig Reactor Water Level Reactor Water Level 120 inches 120 inches Control Rod Control Rod Positions Positions All unknown All unknown APRMs APRMs Downscale Downscale Drywell pressure Drywell pressure 33 psig psig Supp. Pool pressure Supp. Pool pressure 22 psig psig Supp. Pool Supp. Pool water water temp temp 150°0 FF 150 Supp. Pool Supp. Pool water level water level -4 feet

-4 feet (Reference provided)

(Reference provided)

Which one of the following one of following identifies identifies the the status status ofof the the Heat Heat Capacity Capacity Temperature Temperature Limit Limit (HCTL)

(HCTL) and the required procedure required procedure for reactor reactor pressure control?

control?

HCTL Pressure Control Leg of Procedure A. has been exceeded RVCP B has been exceeded By LPC C. has NOT been exceeded RVCP D. has NOT been exceeded LPC

Feedback Feedback K/A: S295026 KIA: A2.03 S295026 A2.03 Ability to Ability to determine determine and/or andlor interpret interpret the the following following as as they they apply apply to to SUPPRESSION SUPPRESSION POOL POOL HIGH HIGH WATER TEMPERATURE:

WATER TEMPERATURE:

Reactor pressure Reactor pressure (CFR: 41.10/43.5/4 5.13)

(CFR: 41.10 143.5 145.13)

RO/SRO Rating:

RO/SRO Rating: 3.9/4.0 3.9/4.0 Objective: CLSLP300L*05a Objective: ClS-lP-300-l *05a

05. Given the PCCP, determine

05. Given the PCCP, determine the the appropriate appropriate actions actions ifif any any of of the the following following limits limits are are approached approached or or exceeded:

exceeded:

a. Heat

a. Heat Capacity Capacity Temperature Temperature Limit.

Limit.

Reference:

Heat Capacity Heat Capacity Temperature Temperature Graph Graph only only isis given given to to examinee examinee PCCP.

PCCP.

Cog level:

Cog Level: High High Explanation:

Explanation:

HCTL has been exceeded. With rods unknown the operator would be in HCTl in lPC.

LPC.

Distractor Analysis:

Distractor Choice A: Plausible because rods are Choice are unknown, would be in lPC. LPC.

Choice B: Correct Answer Choice C: Plausible because HCTl Choice HCTL has been exceeded. rods are unknown, would be in lPC LPC Choice D: Plausible because HCTl HCTL has been exceeded.

SRO Only Basis: Assessment of facility conditions and selection of appropriate procedures during normal, abnormal, and emergency situations (43(b)(5)

Notes Notes

ATTACHMENT 5

.rngm I Page '18 of 27 FIGURE 3 T1

- Heat Capacity Temperature Limit 0 .. r SUPPRESSION POOL WATER TEMPERATURE (°F)

L1.

~ 220 W

$ 210 ~ UNSAFE ABOVE ~

r- W m.,..rn z<

!;;c =f::l=Il

=~I SELECTED LINE ~

rn mm rmi n:: 200

~ 190 5j 180 l-n:: 170

- -4

(-) 0.25 FT W

1flfll1 Cl UI

~ 160

(-) 1.25 FT s: 150

(-) 2.50 FT 0

---4

..J

~~~ (-) 3.25 FT o

Cl

S:i~

o : (-)

I I 1 tit-r tW titt llttVtt 140 D.. =f:: f:: SAFE BELOW =t=~ 4.25 FT LII fl cnr> om mm IE[ ;=I=~

-I

=~~

It Z rn-n 11[ TEl St=~

o

( I)

(I) 130 120 SELECTED LINE I m ,-p.,~

=I::~ (-) 5.50 FT

=~~

I

=~I=

Cl C -

~ 110 - 0 liiii-rn

1=1

=I=~

D.. :=I=~

D.. 100 ,-I-f-

J - - -1,150 F

0 (I) C o

100 200 300 0

400 500 UI 600 700 800 900 1,000 I

1,100 0

rn 0

COO REACTOR PRESSURE (PSIG)

C) - -o CO o U SUPPRESSION POOL WATER TEr. .1PERATURE IS DETERMINED BY:

c j) Z 0 0 rn m -

Cl) rn D m ID CAC-TR-4426-*1A, POINT WTR AVG OR 8 8

--9 CAC-TR-4426-2A, POINTWTRAVG OR 00 I COrvlPUTER POINT G050 OR

ç)ç)QQ COMPUTER POINT G051 OR CAC-TY-4426-1 OR CAC-TY -4426-2 SELECT GRAPH LINE IMMEDIATELY BELOW SUPPRESSION POOL WATER CJ C) Z rnti

-< C - z - o 0 -

mm

-Ui LEVEL AS THE LIMIT. rn I

OEOP-01-UG I,,

C a ti C) Rev. 55 Page 78 of 151 CD g -

I

YES NO MAiNTAIN REACTOR MAINTAIN IIEACTORPRE5S PRESS BELOW THE BELOW TIlE HEAT CAPACITY CAPACfY TEMP Ur.tIT UM1 IRRESPECTIVE 1RRESPECTPJE OF THE RESULTING OFTHERESULTING tOOLDOWN RATE COOLOOWN RCWI1 BNP VOLV1 IEOP-Ot- LPC REldlStONNO 9 UNIT i ONLY

INITIATEAREACTOR INITIATEA REACTORSCRAM SCRAM AND ENTER AND ENTER EOP-01 EOP-9i

~ ______-L______ ~

SPIT-99 SP/T-09 REDUCEREACTOR REDUCE PERTHE PER REACTOR PRESS THE RCIP RCIP SECTION PRE SECTION OFOF I

EOP- 01 EOP- 91 AS AS NECESSARY NECESSARY TO TO REMAIN IN REMAIN IN SAFE SAFE REGION REGION OF HEAT OF HEAT CAPACITY CAPACITf TEMP UMIT TEMP UMIT SPIT-b SPIT-10

/~~~~'~~'~~1iiii£"'~~~~~"\

CONSIDERANTICIRATIONOF I/ CONSIDER ANTICIPATION OF \

EMERGENCY DEPRESSURlZAll0N

( EMERGENCY DEPRESSURIZA11ON \

\\ PERRCIPSECTIONOF PER RCIP SECTION OF /

\

\ ~REACTOR REACTOR VESSEL VESSEL CONTROL/

CONTROL

\

\~~~,:~~~~~:,,~~~~

PROCEDURE{EOP.01. RVCP)

.1. SPIT-11 HCTL HCTL ~/".r(;AN~,

~// THE HEAT ---..........

</" CAPACITY TEMP UMIT '-.,.-!Ii~L.".,

-.,...,. BE MAINTAINED IN THE /"-~

-.,...,., SAFEREGION /""'"

---'::e~ NO SPIT- 12 SPIT-12 f

';EN;-~EP;;~~IZE (MERGENCY DEPRESSURI THE REACTOR PER THE RCIP SECTION OF EOP.01 EOP-91

~~~~~~~~.

SPIT-IS SPIT-13 BNPVOL-VI BNP VOL- VI OEOP-02-PCCP OEOP PCCP REVISION NO NO: 10 10 Categories K/A:

KIA: S295026 A2.03 Tier / Group: TIG!

T1G1 RO Rating:

RORating: 3.9 SRO Rating:

SRORating: 4.0 LP LP Obj:

Obj: CLSLP300L*05A CLS-LP-300-L *05A Source:

Source: PREV PREY Cog Cog Level:

Level: HIGH HIGH Category Category 8:

8: Y

89. Unit

89. Two isis operating Unit Two operating at at rated rated power when half power when half of of the the Orywell Drywell (OW)(DW) Coolers Coolers are are lost.

lost.

Which one Which one of the following of the following correctly correctly completes completes the the statements statements below?

below?

(Assume (Assume initialinitial OW DW andand Suppression Suppression Pool Pool pressures pressures are are equal) equal)

As OW As DW temperature temperature rises, rises, Suppression Suppression Pool Pool pressure pressure willwill rise rise at (1) OW at (1) DW pressure.

pressure.

If DW If OW Air Air Temperature Temperature is is not not restored restored to to within within the the LCO LCO limit limit in (2) hours, in (2) hours, the the Unit Unit is required is required to to be be in Mode 33 within in Mode within the the following following 1212 hours0.014 days 0.337 hours 0.002 weeks 4.61166e-4 months hours per per TS TS 3.6.1.4 3.6.1.4 (Orywell (Drywell Air Air Temperature).

Temperature).

A. (1 A. (1)) the the same same rate rate as as (2)

(2) 8 8 B. (1 B. (1)) the the same same rate rate as as (2)

(2) 1212 C (1) a slower rate Cy rate than (2) 8 D. (1 O. (1)) a slower rate than (2) 12

Feedback Feedback K/A: S295028 KIA: S295028 A2.05 A2.05 Ability to Ability to determine determine and/or andlor interpret interpret thethe following following as they apply as they apply toto HIGH HIGH ORYWELL DRYWELL TEMPERATURE:

TEMPERATURE:

Torus/suppression chamber Torus/suppression chamber pressure:

pressure: Plant-Specific Plant-Specific (CFR: 41.10/43.5/45.13)

(CFR: 41.10 /43.5 /45.13)

RO/SRO Rating:

RO/SRO Rating: 3.6/3.8 3.6/3.8 Objective: CLSLP004A*1 Objective: CLS-LP-004-A 5a

15a

15. Given plant conditions, determine

15. Given plant conditions, determine thethe effects effects that that the following will the following will have have on on the the Primary Primary Containment, Containment, Primary Containment Primary Containment Ventilation Ventilation and Primary Primary Containment Containment Monitoring:

Monitoring:

Loss of

a. Loss Drywell cooling.

of Drywell

Reference:

SD-04, Revision SD-04, Revision 5,5, Page Page 25 25 TS TS Cog Level: High Explanation:

Explanation:

Reduced OW Reduced DW cooling or rising DW DW temperature results in in DW pressure increases increases whose severity is dependent upon plant conditions. OAOP-14.0, Abnormal Primary Containment Conditions provides guidance on indications to be monitored and actions to be taken which include verification of cooling system lineups and reductions in power to maintain average temperature below 150°F. Failure to accomplish this may require entry into the OEOP-02-PCCP Primary Containment Control.

Elevated DW temperature causes DW pressure to rise. As DW pressure rises, SP water level rises causing a rise in SP pressure. Due to the downcomers extending 3 feet below the surface of the SP water level a differential pressure will always exist. Temperature response is different from LOCA response due to steam AND non-condensibles being forced into the SP - steam condensing and-non-condensibles collecting in SP air space.

TS 3.6.1.4 (DW Air Temperature) limit of ~ < 150°F, CONDITION A - Drywell average air temperature not within limit, REQUIRED ACTION A.1 Restore drywell average air temperature to within limit has a COMPLETION TIME of 8 hours9.259259e-5 days 0.00222 hours 1.322751e-5 weeks 3.044e-6 months . If temperature is not restored to ~ .

150°F, CONDITION B, REQUIRED ACTION 8.1 Be in MODE 3 has a COMPLETION TIME of 12 hours1.388889e-4 days 0.00333 hours 1.984127e-5 weeks 4.566e-6 months .

Distractor Analysis:

Choice A: Plausible because SP pressure is changed by the change in SP level only vs pressure, steam, and non-condensibles during a LOCA. The SP air space temperature is in equilibrium with SP water temperature (95°Fmax during normal operations) rising DW pressure would have a direct impact on SP level. However during temperature only (no steam), the DW pressure increase is cushioned by SP water, small changes in SP water level provides small change in SP pressure.

88 hours0.00102 days 0.0244 hours 1.455026e-4 weeks 3.3484e-5 months to restore temperature is correct.

Choice Choice B: Plausible because SP pressure is changed by the change in SP level only vs pressure, steam, in SP and non-condensibles non-condensibles duringduring aa LOCA.

LOCA. The SP air space temperature is in in equilibrium with SP water temperature (95°Fmax during normal operations) rising DW during normal DW pressure pressure would have have a direct direct impact on SP level. However during SP level. during temperature only (no (no steam),

steam), the DW pressure increase is cushioned by by SP water, small changes SP water, changes in SP water in SP water level level provides provides small small change change in in SP SP pressure.

pressure.

12 hours is 12 hours1.388889e-4 days 0.00333 hours 1.984127e-5 weeks 4.566e-6 months is the time time required required to getget to MODE MODE 33 ifif not not restored within within the required Completion Completion Time.

Time.

Choice Choice C: Correct Answer C: Correct Answer Choice Choice D:D: Plausible Plausible because because rising at aa slower rising at slower rate is correct rate is correct and and 12 12 hours1.388889e-4 days 0.00333 hours 1.984127e-5 weeks 4.566e-6 months is the hours is the time time required required to to get get to to MODE 33 ifif not MODE restored within not restored within the the required required Completion Completion Time.

Time.

SRO SRO Only Only Basis:

Basis: Application Application ofof required required actions actions (Section (Section 3)3) and and surveillance surveillance requirements requirements (Section (Section 4)

4) in in accordance accordance withwith rules rules ofof application application requirements requirements (Section (Section 1).

1). (43(b)(2)

(43(b)(2)

Notes Notes Drywell Air Temperature Drvwell Air Temperature 3.6.1.4 3.6.1.4 3.6 CONTAINMENT 3.6 CONTAINMENT SYSTEMS SYSTEMS 3.6.1.4 3.6.'1.4 Drvwetl Air Drywell Temperature Air Temperature LCO 3.6.

LCO 3.6.1.4

'1.4 Drywell average air Drywell average air temperature temperature shall be s '150°F.

shall be 150°F.

APPLICABIUTY:

APPLICABILITY: MODES '1,2, MODES 1,2, and and 3.3.

ACTIONS ACTIONS CONDITION CONDITION REQUIRED ACTION REQUIRED COMPLETION TIME COMPLETION TIME A.

A Drywell average air DrY'Nell A. 1 A.-I Restore drywell average air 8 hours9.259259e-5 days 0.00222 hours 1.322751e-5 weeks 3.044e-6 months Restore hours temperature not within limit.

limit, temperature to within limitlimit.

B.

B. Required Action and B.l B.-I Be in MODE 3. 12 hours1.388889e-4 days 0.00333 hours 1.984127e-5 weeks 4.566e-6 months associated Completion Time not met not met. AND B.2 Be in MODE 4. 36 hours4.166667e-4 days 0.01 hours 5.952381e-5 weeks 1.3698e-5 months

4. Drywell Orywell Temperature A loss of RBCCW to the drywell results in drywell temperature and pressure increases whose severity is dependent upon plant conditions. OAOP-1 OAOP-14.0, 4.0, Abnormal Primary Containment Conditions provides guidance on indications to be monitored and actions to be taken which include verification of cooling system lineups and reductions in power to maintain average temperature below 150°F. '150°F.

Failure to accomplish this may require entry into the OEOP-02-PC OEOP-02-PCCP CP Primary Containment Containment Control.

Loss Loss of of RBCCW to the Drywell due to all Orywell due all RBCCW pumps tripping 1 SD-21 SO-21 Rev.

Rev. 55 Page Page 25 25 of 421 of 42 Categories Categories KJA:

KIA: S295028 S295028 A2.05 A2.05 Tier Tier // Group:

Group: T1G1 T1 G 1 RO Rating:

RORating: 3.6 3.6 SRO SRO Rating:

Rating: 3.8 3.8 LP Obj:

LP Obj: CLSLPOO4A* I 5A CLS-LP-004-A*15A Source:

Source: NEW NEW Cog Level:

Cog Level: HIGH mGH Category Category 8:8: YF YF

90. The

90. The following following plant plant conditions conditions existexist on on Unit Two:

Unit Two:

An ATWS

- An

- ATWS with with aa spurious spurious GroupGroup II Isolation Isolation has occurred has occurred HPCI isis injecting

- HPCI

- injecting to to the RPV to the RPV maintain RPV to maintain level RPV level SUPPRESSION CHAMBER

- SUPPRESSION

- CHAMBER LLVL HI-HI isis inin alarm VL HI-HI alarm Which one Which one of the following of the following identifies identifies the the action action required required for long term for long term HPCI HPCI system system operation and operation and thethe reason reason forfor this this action?

action?

When suppression When suppression pool pool temperature temperature reaches reaches 140°F, 140°F, (1)

(1) to to prevent prevent (2)

(2) .

A. (1)

A. lower HPCI (1) lower HPCI flow flow to to less less than than 2000 2000 gpmgpm lAW lAW LPC LPC (2) pump (2) pump bearing bearing damage damage B. (1)

B. lower HPCI (1) lower HPCI flow to to less less than than 2000 2000 gpm lAW lAW LPC LPC (2) a loss of NPSH C

C~ (1) defeat the automatic suction transfer logic logic and transfer HPCI HPCI suction suction to the CST lAW SEP-10 (2) pump bearing damage D. (1) defeat the automatic suction transfer logic and transfer HPCI suction to the CST lAW SEP-10 (2) a loss of NPSH

Feedback Feedback K/A: 295029 KIA: 295029 G2.01.07 G2.01.07 Ability to evaluate plant performance Ability to evaluate plant performance and and make make operational operational judgments judgments based based on on operating operating characteristics, reactor characteristics, reactor behavior, behavior, and and instrument instrument interpretation.

interpretation.

High Suppression High Suppression PoolPool Water Water Level Level (CFR: 41.5/43.5/45.12/45.13)

(CFR: 41.5/43.5/45.12/45.13)

ROISRO Rating:

RO/SRO Rating: 4.4/4.7 4.4/4.7 Objective:

Objective:

LOl-CLS-LP-0l 9-A, 26g:

LOI-CLS-LP-019-A, Given plant 26g: Given conditions and plant conditions and one one of of the the following following events, events, use use plant plant procedures procedures toto determine the determine the actions required to actions required control and/or to control mitigate the and/or mitigate the consequences consequences of of the the event:

event:

High Suppression High Suppression Pool Pool water water level.

level.

Reference:

001-37.5 001-37.5 SUPPRESSION CHAMBER SUPPRESSION CHAMBER LVL HI-HI APP LVL HI-HI APP Cog Level Cog Level - High

- High Explanation: HPCI Explanation: HPCI system system isis normally normally aligned aligned to the CST, to the CST, with with the the torus high water torus high water level level this this transfers transfers to to meets the KA by having the torus. this meets having to evaluate the suction path has transferred to the torus and the operational implications of the high torus tempeature on continued operation of the HPCI system. this requires the suction to be transferred back to the CST lAW SEP-10. SEP-I 0.

From : The lube oil and control oil for both HPCI and RCIC are cooled by the water being From:

pumped. Very high lube oil temperatures can result in loss of lubricating qualities in the oil and thus cause damage to the bearings. Suction for HPCI and RCIC is aligned to the Condensate Sto(age Storage Tank (CST) if it is available. The HPCI automatic suction transfer logic can be defeated to allow this lineup if necessary provided suppression pool temperature is approaching 140°F.

Distractor Analysis:

Choice A: Plausible because reducing flow would be a correct action if HPCI NPSH was the concern. At high temperatures with low level in the torus this could be a correct action. Pump bearing damage is a correct statement.

Choice B: Plausible because reducing flow would be a correct action if HPCI NPSH was the concern. At high temperatures with low level in the torus this could be a correct action. a loss of NPSH would be correct for the reason to reduce flow.

Choice C: Correct answer, see explanation Choice D:0: Plausible because transferring the suctionsuction isis correct but the concern is is for pump pump bearing bearing damage.

damage.

SRO SRO Basis:

Basis: 10 10 CFR CFR 55.43(b)-S 55.43(b )-5 Assessment Assessment of of facility facility conditions conditions and and selection selection ofof appropriate procedures procedures during during normal, abnormal, abnormal, and emergency emergency situations.

Notes Notes TABLE 1I M1MUM SYSTEM MAXIMUM SYSTEMINJECTlD.'I INJECTION PRESSURES PRESSURES MAXIM MAXIMUMLIM

.,VSTEM INJECTION INJECTION SYSTEM PRESSURE PRESSURE (P5I)

(PSIG)

CONDENSATE1FEEDWATER CONDENSATEIFEEDWA.TER 1250 1250 CRD FLOW NAY CRt FI.OW OE NAY DIl: 1490 141)0 MAXIMIZED PER MAXIMIZED PER EOP0i. SEP-00 eOP.Of. SEP. 09 RCIC WITH RCiC SUCTION FROM WITH SUCTION FROM CST IF IF AVAILABLE.

AVAILABLE. DEFEAT DEFEAT LOW REACTOR LOW REA.CTOR 1190 1190 CST PRESS AND PRESS AND HIGH HIGHAREA TEMPERATURE AREA TEMPERATURE ISOLATION LOGIC ISOLATION LOGIC IF IF NECESSARY NECESSARY PER *CIRCUIT PER CIRCUW ALTERATION ALTERATION PROC EDURE PROCEDURE" leop* Ot* sep* 10) c,.oISEP- io HPCI WITH HPCI WITH SUCTION SUCTION FROM FROM 1250 12110 OST II' CST IF AVAILABLe.

AVAILAbLE, OeFeAT OFIAT HP(;I 1IPCI HI NI SUPPRESSION POOt.

SUI"PRI!$$ION POOL I.I!Yf!1.

I.EVEL SUCflO1 TAANSFSR SUCTION TRANSFER ANI) AND HIGH I-IGt-I AREA TIOMPERATURS ARIOA TEMPERATURE ISOLATION ISOLATION LOGIC IF NECESSARY PER lOGIC If: NeCeSSARY PER *CIRCUIT CIRCUIt ALTERATION PROCEDURe" ALTeRATION PROCEDURE EOP.0I-SEP -

~EOP* 01* SEP* 10) 10)

IPUI- ESTABLISH lPCI* ESTALI5H RHR RHR SERVICE SERVICE WATER FI.OW 200 WATER FLOW AS 500HSOON AS POSSIBLE CAUTION CAUTION OPERATION ERATION OF HPCI HPGI OR OR RCI ROIG WITH SUCTION TEMPERAT H SUCTION TEMPERATURE URES ABOVE 140? MAY OVE 140*' MAY RESULT IN EQUIPMEN EQUIPMENT T DAMAGE RC1L- 23 Distractor Distractor plausibility:

plausibility:

CAUTION CAUTION I HPCI HPCI FLOW WITH FLOW ABOVE WITH SUCTION ABOVE 2000 SUCTION FROM FROM CST 2000 GPM GPM CST AND AND CST LEVEL GST LEVEL BELOWBELOW 55 FEET FEeT MAY MAY RESULT RESULT IN IN VORTEXNG VORTEX<<NG AND AND EQUIPMEN EQUIPMENT T DAMAGE DAMAGE Categories I RCFL Categories K/A:

KIA: 295029 295029 G2.01.07 G2.01.07 Tier/Group:

Tier / Group: T1G2 T1 G2 RO Rating:

RORating: 4.4 4.4 SRO SRO Rating:

Rating: 4.7 4.7 LP Obj:

LPObj: 19-A 19-A 26G26G Source:

Source: BANK BANK Cog Cog Level:

Level: HIGH HIGH Category Category 8:8:

91. Which

91. Which one one of of the the following following identifies identifies the the controlling controlling document document and and the the required required action action to be to be taken taken ifif SJAE SJAE Offgas Offgas Radiation Radiation monitor monitor readings readings increase increase 50%

50% during during steady steady state rated power state rated power operation?

operation?

Notify E&RC Notify E&RC to to perform perform the the Surveillance Surveillance II Test Test Requirement Requirement (SRITR)

(SRITR) required required byby (1) ,which (1) , which confirms confirms the the SJAE release rate SJAE release rate isis within within limits limits within within (2) (2) following following the monitor the monitor reading reading increase.

increase.

A. (1)

A. (1) ODCM ODCM 7.3.2,7.3.2, Radioactive Radioactive Gaseous Gaseous Effluent Effluent Monitoring Monitoring Instrumentation Instrumentation (2) 4 hours (2) 4 hours4.62963e-5 days 0.00111 hours 6.613757e-6 weeks 1.522e-6 months B. (1)

B. ODCM 7.3.2, (1) ODCM 7.3.2, Radioactive Radioactive Gaseous Gaseous Effluent Effluent Monitoring Monitoring Instrumentation Instrumentation (2) 12 (2) 12 hours1.388889e-4 days 0.00333 hours 1.984127e-5 weeks 4.566e-6 months hours C

C~ (1) T.S. 3.7.5, (1) T.S. 3.7.5, Main Main Condenser Condenser Offgas Offgas (2) 4 hours4.62963e-5 days 0.00111 hours 6.613757e-6 weeks 1.522e-6 months D. (1) T.S. 3.7.5, Main Condenser Offgas (2) 12 hours1.388889e-4 days 0.00333 hours 1.984127e-5 weeks 4.566e-6 months

Feedback Feedback K/A: S29503BG KIA: S295038G 2.02.42 2.02.42 Ability to Ability to recognize recognize system system parameters parameters thatthat are are entry-level entry-level conditions conditions for for Technical Technical Specifications.

Specifications.

High Off-Site High Off-Site Release Release Rate Rate (CFR: 41.7/41.10 / 43.2/43.3/45.3)

(CFR: 41.7/41.10/43.2/43.3/45.3)

RO/SRO Rating:

RO/SRO Rating: 3.9/4.6 3.9/4.6 CLSLP30*08 Objective: CLS-LP-30*OB Objective:

08. Given plant conditions and OB. Given plant conditions and Technical Technical Specifications, Specifications, including including the the Bases, Bases, TRM, TRM, ODCM, ODCM, andand COLR, COLR, determine whether determine whether given given plant plant conditions conditions meet meet minimum minimum Technical Technical Specifications, Specifications, TRM, TRM, or or ODCM ODCM requirements associated requirements associated with with the the Condenser Condenser AirAir Removal/Augmented Removal/Augmented Offgas Offgas System.

System.

Reference:

101-03.1, Revision 101-03.1, Revision 10, 10, Page Page 44, 44, Item #57 (CODSR)

Item #57 (CODSR)

Cog Level: High Cog Level: High Explanation:

Explanation:

NOTIFY E&RC NOTIFY E&RC to confirm release rate rate is within limits within 4 hours4.62963e-5 days 0.00111 hours 6.613757e-6 weeks 1.522e-6 months hours following a monitor monitor reading reading increase of greater than or equal to 50% without an accompanying accompanying increase in thermal power. SR SR 3.7.5.1 Distractor Analysis:

Distractor Choice A: Plausible because the SJAE Rad Monitor operability is required by ODCM 7.3.2 and 4 hours4.62963e-5 days 0.00111 hours 6.613757e-6 weeks 1.522e-6 months is correct.

Choice B: Plausible because the SJAE Rad Monitor operability is required by ODCM 00CM 7.3.2 and 12 hours1.388889e-4 days 0.00333 hours 1.984127e-5 weeks 4.566e-6 months is a timeframe for another Required Action in this spec.

Choice C: Correct Answer Choice D: Plausible because TS 3.5.7 is correct and 12 hours1.388889e-4 days 0.00333 hours 1.984127e-5 weeks 4.566e-6 months is a timeframe for another Required Action in this spec.

SRO Only Basis: Application of Surveillance Requirement Requirements s and timeframe greater than 11 hour1.273148e-4 days 0.00306 hours 1.818783e-5 weeks 4.1855e-6 months .

Notes

ATTACHMENT ATTACHMENT 'I1 Page Page 3939 of6?

of 67 ITEM ITEM SI-il FCHECK SHIFT CHECKLIST LIST NOTES NOlES OPER OPER FREQ FREQ TIME':

TIME TSiOPER TSOPER NO.

NO. MODE MODE LIMITS LIMITS RECORD SJAE RECORD SJAEOFFGASOFF(4SRADMONITOR RAD MONiTOR 57 DD DD 1. 2. 3' 1,2' 3 bb 07-13 07-13 012-RM-K60tA. NOTIFY D12-RM-K6Cl"IA. NOTIFYE&RC E&RCto toconfilm confirm release rate rel~ase rate isis within withinlimits lmits,...within hours ijhin 44 hrurs oIIowing aamonitor follO\lotng monitorreading reading increase increase o*f of greaterthan greater than or equal to orequal 50%without to 50% withoutan an 13-19 l3-1 accompanying increase accompanying increase inin thermal thermal pow~r.

pcwer.

SR 3.7.5.1 SR3.7.5.1 RECORD SJAE RECORD S4E OFFGASOFFGASRAD R40 MONITOR MON.TOR 58 CD DD 1. 2. 3 1,2',3' b 07-13 07-12 D12-RM-K6018. NOTIFY D12-RM-K601B. NOTIFY E&RC E&RC to confirm to confirm reLease rate re!~ase rate isis within within limits limits within within 44 hours5.092593e-4 days 0.0122 hours 7.275132e-5 weeks 1.6742e-5 months hours following aa monitor follo',\;ng nionitor reading reading increase increase of of greater than greater than or or equal equal to to 50% without an 50% without an 13-19 accompanying increase accompanying increase in thermal pow~r.

in thermal pcf.ver.

3.7.5.1 SR 3.7.5.1 SR PERFORM ohannel PERFORM channel check utilizing the check utilizing the 59 '6 0 07-13 07-13 Refer~nce Reference calculabon on calculation Table I SJAE on Table SJAE OFF-GAS OFF-GAS RAD RAE) calculation calcufaton onon MONITORS D'12-RM-K601A MONITORS 012-RM-keoIA and and BB ODCM 00CM Table Table 1I TR 7.3.2-1 Function TR 7.3.2-1 FunctionS, 6, rn TR 7.3.2.1 7.3.2.1 PERFORM channel PERFORM channel cheokcheck on on SERVICE SERViCE 00 R R 6 ec 07-13 07-13 channel channel WATER EFFLUENT

~\-:41ER EFFLLJENTRAD MONITOR RAD MONITOR operable operabe 012-RM-K605. ODCM D12-RM-K605, 00CM TableTabie 7.3.1-1, 7,3. 1-i Function 3, Function 3, TR TR 7.3.1.1 7.3.1.1 i PERFORM ohannel PERFORM channel check on on P.ADI*lIASTE RAE)WASTE 61 8 6 ec 07-13 channel EFFLUENT RAD EFFLUENT RAD MOM MONITOR D12-RM-K6C4 TOR D12-RM-K604 operable operabe on Control en Control Room Room Pan~12-H12-Pa04 Panel 2-H12-P804 'Qi1h with recorder D'12-ROOt recorder D12-ROO1 on XU-3, XU-2. o!)CM CCCM Tab!eTable 7.3.1-i, !t~m 7.3.1-1, Item I, 1. TR7.3.l.I TR 7.3.1.1 operation of During operation

'During of the mainmain oondenser condenser air ejector. ejector.

SHIFT SHIFT Dayshift Davshift BRUNSWIC BRUNSWICK PL4.NT K STEAM ELECTRIC PL4NT DAILY SURVE1LL DAilY SURVEIllANCE ANCE REPORT CONTROL OPERATOR CONTROL OPERATORS S 101-03.1 1'101-03.1 Rev. 101 101

Main Condenser Main Condenser Offgas Offgas 3.7.5 3.7.5 3.7 PLANT 3.7 PLANT SYSTEMS SYSTEMS 3.7.5 Main 3.7.5 Main Condenser Condenser Offgas Offgas LCO 3.7.5 LCO 3.7.5 The gross The gross gamma gamma activity activity mte rate of the noble of the noble gases gases measured measured at at the the main main condenser air condenser air ejector ejector shall shall be 243,600 jJCilsecond be ::; 243,600 pCiisecond after after decay decay of of minutes.

30 minutes.

30 APPLICABILITY:

APPLICABILITY: MODE 1, MODEl, MODES 22 and MODES and 33 with with any any main steam line main steam line not not isolated isolated and and steam steam jetjet air air ejector (SJAE) ejector (SJAE) in in operation.

operation.

ACTIONS ACTIONS CONDITION CONDITION REQUIRED ACTION REQUIRED COMPLETION TIME COMPLETION TIME A. Gross gamma activity rate of .1l A.1

.* 1 Restore gross gamma 72 hours8.333333e-4 days 0.02 hours 1.190476e-4 weeks 2.7396e-5 months gases not within the noble gases activity mte activity rate of the noble nobie limit, limit. gases to within limitlimit.

B. Required Action and B.1 Isolate all main steam lines. 2 12 hours1.388889e-4 days 0.00333 hours 1.984127e-5 weeks 4.566e-6 months associated Completion Time no met.

not OR B.2 Isolate SJAE. 12 hours1.388889e-4 days 0.00333 hours 1.984127e-5 weeks 4.566e-6 months OR B.3.1 Be In in MODE 3. 12 12 hours1.388889e-4 days 0.00333 hours 1.984127e-5 weeks 4.566e-6 months AND B.3.2 B.3.2 Be in in MODE MODE4. 4. 313 hours0.00362 days 0.0869 hours 5.175265e-4 weeks 1.190965e-4 months 36 Brunswick Unit 11 Brunswick Unit 3.7-18 3.7-18 Amendment No.

Amendment No. 203 203

Main Condenser Main Condenser OffgasOffgas 3.7.5 3.7.5 SURVEILLANCE REQUIREMENTS SUR'.,.'EILLANCE REQUIREMENTS SURVEILLANCE SURVEILLANCE FREQUENCY FREQUENCY SR 3.7.5.1 SR 3.7.5.1, NOTE

NOTE----------------------------

Not required Not required to to be be performed performed until until 31 31 days days after after any any main steam main steam line line not not isolated isolated and and SJAE SJAE in in operation.

operation.

Verify the Verify the gross gamma activity gross gamma activity rate rate of of the the noble noble 31 days 31 days gases is gases is ::; 243,600 243600 ~Cilseoond pCiisecond after after decay decay of of 30 minutes.

30 minutes. AND Once within 4 hours4.62963e-5 days 0.00111 hours 6.613757e-6 weeks 1.522e-6 months Once hours after a :::50%

afteraa5O%

increase in increase in the the nominal steady nominal steady state fission gas slate release after factoring oul out increases due to changes in THERMAL POWER level 8runswick Brunswick Unit Unit 11 3.7-19 3.7-19 Amendment Amendment No. 1\10. 203 2113

Radioactive Gaseous Radioactive Gaseous Effluent Effluent Monitoring Instrumentation Monitoring Instrumentatlon 7.3.2 7.3.2 Table 7.:?.2-1 T<ible 7.12- ipage of4)4) aage 22 off Radactie Radioacti **-e C-,aSECUS Effluent~,'itocn!!

I3asecusEffluent Mitang1nslrurr,entat'crI reburrentatcn FJNC.'"

FUNCTlml PJCA6_E

,.P?LlCA8:'E aL:iEt REaLiI.~EO ccrilyr1cr4 CCNDlTICNS TEST TE&T Al.ARWiRiP ftp tLC0S 3ft fb:::O:SOR C%NLS CH>.NNELS 9EFEfNCO REfERIENCEO aECuIRE,lETS REQUIREMENTS S:=TPO!NT 3TPCNT OThER OThER ZER

?ER fROM REJIRD ft REQUIRED V,.LUE S,CIFIED S"=:CIFIEO FJt2TDt F;Jr..;:::TiO:-I CCt.FEriSArOv CCI/PEN!l"'TO.RY CCt4DrIC?3 CCN:)liIC/,"S tE4.SLL9E3 Al MEASU~ES" *. l 2.

2. Re:actor Buildng Ventilation Rea,tsorBuild\!1g Ventilatcn Ffonhtorir Systsm Monitoring System (continueo}

(rontinuec}

e.e. £airplerFlcwRaie Sarrpler Row Rate at fmas At all At tme; 1 D0 TR 7.3.2.1 TR 7.3.2.1 (0;1 srernent Device; Measurerr,ent Device TR 7.3.2.6 TR 7.12.6 TR 7.3.2.10 TR 7.3.2. C 3.

3. Turbine Buildillg TLiltine Buildng Venlilaton Venblatcn Monito,ir Systam Monitorir" System

3. a. NobieOasMtWiy Notle Gas Acti'.i.ly Atalmes

.<\t all tim;;s 1 3 TR 7.32.;

TR 7.a2. (b)

(b Atnikr M::fIitcr TR 7.3.2.3 TR 7.3.2.3 7.3.15 TR 7.3.2.5 TR TR 7.3.2.10 TR 7.3.2.10 b.b. bineSanpIer looine Sarrpler Ata[tmes At 'II! limss 1 C C TR 7.3.22 TR 7.3.12 NA NA Cartroge Cmr:dge c.o. Parclate S<1mj:ler PartiCIIlate Sampler Ata[ lim;;s AtaN bmes 1 C C TR 7.3.2.2 TR 7.3.2.2 NA NA Ftter Fiiter d.

d. System Effluent Flem System Effiuem Fbiw At alla lim;;s tines 1 00 TR 7.3.2 7.3.2i .* NA NA Rate Measurement Rate TR 7.3.2.0 7.3.2.

Device De-~ice TR 7.3.2.10 e.

e. low Rar,e Sarrpler Lcv Range Lairpler At aU tines al lim;;s 1 00 7.3.21 TR 7.3.2.1 (e)

(Cl F Rate FbwRate TR 7.3.2.0 7.3.2.5 easarernent Device; Measurement Desice TR 7.3.2.10 f.f. Md:High Rang;;

MeiHigh Rante (rn:

(m) 1 0 TR 77.3.2.10

..3.2.10 NA Sampler ReI'.'

Sarrpler F1cv Rate Measurenrent Device; ME-<1SlJrement 0eioe 4.

4. ManCondenserOtf-Gas Main Condenser Off-Gas (e)

(ei 1 BB TR 1.3.2.1 7.32.1 (bJ Treatment Treatment Systam System Ncble Noble 7.3..2.3 TR 7.3.2.3 Gsa Gas Acbiity Mcnhtor Activity Monitor '" 7.12.6 TR 7.3..2.6 (0onseam (Dovmstream of AOG R 7.3.2.10 TR Treatment Treafment Systam)System)

(CQntinued)

(continued)

(a)

(a) Specif Speci/a a inatn.anentatii instrumentatioo dentlicaban idE!1~licalion nuntere numbers are <ire provided provided in in Appendix E.

AppEIldix 3.

b)

(b) Alarnscrp AlarJl1l~rip setpcints setpoinfs ahat shall be be detemiine determined in in accocdaie acoocdance teth\";th 00CM ODCM methodacgy melhodo:cgy and and set 10 ensure set ia ensure thethe limta lim(s of oi 000MS ODCMS 7.3.7, 'Dose RateGaseous 7.3.7. Dose Rate-Gaseous Effluents, Effluents; are are nt no! eteeeded.

elQOeeaed.

Ic)

(c} Alarmr AJarll1/tr,p seoints setpoints shall shall be be deterntined detemlined in oY..o::roance eith in aconntiartce \\;th asscctsed assccia:;ed desi9n specitcatiDn(5) and desig\!1specifoation{s) and setto set to ensure anS!!re the limits at the ltnit.s of COCMS COCMS 7.3.7,7.:'..7. Dose

'Dose RateGaseous Effluents; are Rate---Gaseous Effluents, are na not exceeded.

exceec'ed.

d)

(d) Provides Provides alarm.

alaml.

ie)

(e) During Main Doring Main Condenser Coodenser 0ff-Gas Oft-Gas Tresurent Treatment System Systemoperatcn operation irn)

(m) During During MWHig MdiHigll RareSymaac Rang;; System opeaIion n Brunswidc Brunswick Units Units 11 and and 22 7.3.2-18 7.3.2-10 Rev. 32 Rev. 32 I1 Categories Categories K/A:

KIA: S295038G2.

S295038G 2.02.42 02.42 Tier/Group:

Tier / Group: TIGl T1G1 RO Rating:

RORating: 3.9 3.9 SRO Rating:

SRORating: 4.6 4.6 LP Obj:

LPObj: CLSLP3O*O CLS-LP-30*08 8 Source:

Source: NEW NEW Cog Cog Level:

Level: HIGH HIGH Category Category 8:8: YY

92. The

92. The following following plant plant conditions conditions exist exist on on Unit Unit Two Two due due to to aa malfunction malfunction of of the the Air Air Dryer:

Dryer:

SERVICE AIR

- SERVICE

- AIR PRESS-LOW PRESS-LOW isis in in alarm alarm

- RB

- INSTR AIR RB INSTR AIR RECEIVER RECEIVER 2A 2A PRESS PRESS LOW LOWis in alarm is in alarm RB INSTR

- RB

- INSTR AIR RECEIVER 2B AIR RECEIVER PRESS LOW 2B PRESS LOW isis inin alarm alarm Instrument Air

- Instrument

- Air pressure pressure isis 93 93 psig psig and and recovering recovering Based on Based on the the above above indications, indications, which which one one of of the the following following correctly correctly identifies:

identifies:

(1) the status of the Service (1) the status of the Service Air Air Dryer Dryer Bypass Bypass Valve, Valve, SA-PV-5067, SA-PV-5067, and and (2) the (2) the procedure procedure that that contains contains the the steps steps to to close close the the Reactor Reactor Building Building Inboard Inboard and and Outboard Isolation Outboard Isolation Valves Valves (BFIVs)?

(BFIVs)?

A'I (1) open A (1) open (2) OAOP-20.O, Pneumatic (2) OAOP-20.0, Pneumatic (Air/Nitrogen)

(Air/Nitrogen) System System Failures Failures B. (1) open (2) 2APP-UA-01, 2APP-UA-O1, Service Air Press-Low C. (1) closed (2) OAOP-20.0, OAOP-20.O, Pneumatic (Air/Nitrogen) System Failures D. (1) closed (2) 2APP-UA-01, 2APP-UA-O1, Service Air Press-Low

Feedback Feedback K/A: 300000 KiA: 300000 A2.01A2.01 Ability to Ability to (a) predict the (a) predict impacts of the impacts the following of the following on on the the INSTRUMENT INSTRUMENT AIR SYSTEM and AIR SYSTEM and (b) based (b) based on on those those predictions, predictions, use use procedures procedures to to correct, correct, control, control, or or mitigate mitigate the the consequences consequences ofof those those abnormal abnormal operation:

operation:

Air dryer Air dryer and and filter filter malfunctions malfunctions (CFR: 41.5 (CFR: 41.5/45.6)

/45.6)

RO/SRO Rating:

RO/SRO Rating: 2.9/2.8 2.9/2.8 Objective:

Objective:

CLS-LP-46, 07i:

CLS-LP-46, 07i: Given Given plant plant conditions, conditions, determine determine ifif the the following following automatic automatic actions actions should should occur:

occur: Air Air Dryer is Dryer is bypassed.

bypassed.

CLS-LP-037.1, 8b:

CLS-LP-037.1, 8b: State State how how the the RBHVAC RBHVAC is is affected affected by by the the following:

following: Loss Loss of of Instrument Instrument Air.

Air.

Reference:

RB INSTR RB INSTR AIR RECEIVER 2B AIR RECEIVER 28 PRESS PRESS LOW LOW (UA-01 (UA-01 1-2)1-2)

SERVICE AIR PRESS SERVICE PRESS LOW LOW (UA-01 (UA-01 5-4) 5-4)

OAOP-20, Pneumatic OAOP-20, Pneumatic (Air/Nitrogen)

(Air/Nitrogen) System Failures System Failures Cog Level:

Cog Level: High High Explanation:

The air dryer malfunction has caused air pressure to lower. The Service Air low pressure alarms comes in at 107 psig. At 105# decreasing the Service Air system isolates, thus the 0 psig indication on Service Air.

The alarms for the receivers low pressure come in at 95# and are located in the Reactor Building. With these alarms in the operators are required to close the BFIVs while there is still sufficient air pressure remaining to make the secondary containment isolation valves close in accordance with the AOP supplemental actions.

supplemental Distractor Analysis:

Choice A: Correct answer; The air dryer bypass valve opens at 98# and dropping and the steps are in the AOP for closing the BFIVs.

Choice B: Plausible because the air dryer bypass valve is open, but the guidance for closure of the BFIVs is contained in the AOP or RB INSTR INS TRAIRAIR RECEIVER 2A(B) PRESS LOW LOWAPP.

APP.

Choice C: Plausible because the AOP is the correct procedure for closure of the BFIVs, but the air dryer bypass valve would be open (requires system knowledge to know the setpoint for the bypass opening).

Choice D: Plausible because because the student may not know the setpoint of the bypass valve opening and the guidance guidance for closure of the BFIVs is contained in the AOP or RB INSTR AIR RECEIVER 2A(B) PRESS LOWAPP.

SRO SRO Basis:Basis: 10 10 CFR CFR 55.43(b)-5 55.43(b)-5 Assessment of of facility conditions conditions andand selection selection of of appropriate procedures procedures during during normal, normal, abnormal, and and emergency emergency situations.

situations.

The first The part of first part of the the question question isis RO RO knowledge knowledge (setpoint (setpoint forfor the the auto auto opening opening of of the the air air dryer dryer bypass bypass valve valve the second part the second part is is Assessing Assessing plant plant conditions conditions (normal, (normal, abnormal, abnormal, or or emergency) emergency) andand then then prescribing prescribing aa procedure procedure to to mitigate, mitigate, recover, recover, oror with with which which toto proceed.

proceed.

Notes Notes

4.

4. IF RB IF INSTR AIR RB INSTR AiR RECEIVER RECEIVER 1A(2A) IA(2A) PRESS PRESS LOWLOW (UA-O1 1-1)

(UA-01 1-1) OR OR RB RB INSTR IWSTR AIR AIR RECElVER RECEiVER 1B(2B) IB(2B,)

PRESS LOW PRESS LOW (UA-01 (UA-Ol 1-2)1-2) alarm alarm isis received, received, THEN PERFORM THEN PERFORM the following:

the following:

NOTE:

NOTE: Isolation or Isolation of the the Reactor Reactor Building Building supply supply andand exhaust exhaust dampers dampers will will render render the the building ventilation building ventilation system system inoperable.

inoperable. Consideration Consideration should should be be given given ror for starting the starting the Standby Standby Gas Gas Treatment Treatment System System to to ensure ensure methe Reactor Reactor Building Building differential pressure differential pressure remains remains negative.

negative.

a.

a. IFIF necessary, necessary. THEN Treatment System.

Treatment THEN START System.

START the the Standby Standby Gas Gas o0 NOTE:

NOTE: Local "Tee Local Tee Handles" Handles may may be be used used to close the to close the Reactor Reactor Building Building Isolation Isolation Dampers ifif insufficient Dampers insufficient control control airair is is available.

available. 11 (2)OP-37.1 (2)OP-37.1 provide provide L instructions for manual instructions manual operation operation of of Reactor Reactor Building Building Isolation Isolation Valves.

b.

b. CLOSE the CLOSE following dampers:

the following dampers:

- ,~B VENT 1NBO VALVES, RB VENTJNBD VALVES, 0 1A2A)-BFiV-RB and tC(2C)-BFIV-RB 1A(2A)-BFlV-RB iCi2C)-BF1 V-RB

.~B RB VENTOUTBD VENT OUTBD VALVES, 0 1B(2B)-BF1V-RB and 1B(2B)-BFlV-RB and 1D(2D)-BFIV-RB ID(2D)-BF1 V-RB

) OAOP-20.O IOAOP-20.0 Rev. 35 Page 5 of 18 Page50f18 I Unit 2 APP UA-01 UA-O1 5-3 Page*lof2 Page 1 of 2 AIR DRYER .2A 2A TROUBLE AUTO ACTIONS 1.

1. Service air dryer bypass valve SA-PV-5067 Service SA-PV-5067 will begin to open ifif service air header pressure decreases to 98 98 psig.

2.

2. IfIf control power is control power is lost lost oror interrupted interrupted the dryer dryer will fail safe, providing providing continued continued air air flow through through one one tower.

3.

3. IfIf aa dryer dryer tower tower moisture moisture sensing probe probe related fault or or malfunction occurs, occurs, the the dryer dryer control system will control system will default default to to aa 44 hour5.092593e-4 days 0.0122 hours 7.275132e-5 weeks 1.6742e-5 months hour drying drying cycle.

cycle.

RB INSTR RB INSTR AIR RECEIVER 2B AIR RECEIVER PRESS LOW 2B PRESS LOW AUTO ACTIONS AUTO ACTIONS 1.

'I. Standby Instrument standby Instrument Air Compressor Compressor 2B 2B starts starts and and loads.

loads.

2.

2. High Pressure BoUIe High Pressure Bottle Rack Rack Isolation Isolation Valve, RNA-SV-548'1 RNA-SV-5481 opens, supplying SRV's supplying SRVs and CAC-V17 AC-V17 with aa pneumatic pneumatic source.

source.

CAUSE CAUSE 1.

'I. Low air pressure Low pressure (95 psig) psig) in in instrument instrument air receiver receiver 2B.

2. Loss of plant Loss plant air compressors.

3. Instrument air pipe Instrument pipe rupture or air leak.

4. Circuit malfunction.

OBSERVATIONS 1.

'1. Standby compressor starts automatically and loads (it will unload at '105 105 psig).

2. Service air header may have isolated.

3. Pressure Indicator 2-RNA-PI-5268 2-RNA-Pl-5268 (XU-51) indicates approximately 100 psig.

ACTIONS 1.

'1. that standby compressor is running.

Check tllat

2. Check to see if instrument air pressure is maintaining or increasing above 95 psig.

3. Check plant compressors.

4. Check for instrument air ruptures.

Cl1eck

5. Isolate any instrument air piping leaks or ruptures.

6. Isolate nonessential air supplies in order to maintain more than 95 psig on instrument header.

heade[

7. Ensure the High Pressure Bottle Rack Isolation Valve, RNA-SV-548'1RNA-SV-5481 (XU-51) opens.

8. If aa circuit malfunction is suspected, ensure that a WRIJO WRfJO is prepared.

9. If secondary containment isolation is required, close secondary containment isolation valves 2B-BFI V-RB and 2D-BFI 2B-BFIV-RB 20-BFIV-RB V-RB prior to accumulator air pressure bleedoff.

2APP-UA-0i

!2APP-UA-O'l Rev. 64 Page 66 of 102 1021 Categories Categories K/A:

KIA: 300000 A2.01 300000 A2.0 1 Tier / Group:

Tier/Group: T2G 1 T2G1 RO RO Rating: 2.9 2.9 SRO Rating:

SRO 2.8 2.8 LP Obj:

Obj: 46-71 46-71 Source:

Source: ~vv NEW Cog Cog Level:

Level: HIGH HIGH Category Category 8:

8: Y Y

93. Unit

93. Two isis operating Unit Two operating at at power power withwith Reactor Reactor Recirculation Recirculation Loop Loop AA isolated isolated due due to to abnormal seal abnormal seal leakage.

leakage. AA fire fire inin the reactor building the reactor building occurs occurs and and the the Site Site Incident Incident Commander has Commander has requested requested thatthat MCCMCC 2XA-2 2XA-2 bebe de-energized de-energized forfor fire fire suppression.

suppression.

Which one Which one ofof the following identifies the following identifies thethe impact impact that that deenergizing deenergizing MCC MCC 2XA-2 2XA-2 hashas on on RHR Loop A availability and the procedure RHR Loop A availability and the procedure which which provides provides this this guidance guidance underunder the the above plant above conditions?

plant conditions?

Deenergizing MCC Deenergizing MCC 2XA-2 2XA-2 will will make make RHR RHR Loop Loop A A Inoperable Inoperable but but Available Available provided provided that the that the 2-E11-F015A, 2-El l-FO15A, InboardInboard Injection Injection Vlv, Vlv, (1)

(1) to support LPCIIAW to support LPCI lAW (2) (2)

A. (1)

A. is maintained (1) is maintained (de-energized)

(de-energized) openedopened (2) OAP-025, BNP (2) OAP-025, BNP Integrated Integrated Scheduling Scheduling B. (1)

B. is maintained (I) is maintained (de-energized)

(de-energized) opened opened (2) 001-01.08, Control of Equipment and System Status (2)

C C~ (1) has a Dedicated Operator is assigned for manual operation (2) OAP-025, BNP Integrated Scheduling D. (1) has a Dedicated Operator is assigned for manual operation (2) 001-01.08, Control of Equipment and System Status

Feedback Feedback K/A: S600000G KIA: S600000G 2.02.37 2.02.37 Ability to Ability determine operability to determine operability and/or andlor availability availability of of safety safety related related equipment.

equipment.

Plant Fire Plant Fire OnOn Site Site (CFR: 41.7/43.5/45.12)

(CFR: 41.7 /43.5 / 45.12)

ROISRO Rating:

RO/SRO Rating: 3.6/4.6 3.6/4.6 Objective: CLS-LP-Objective: CLS-LP

Reference:

OAP-025, Revision OAP-025, Revision 39, 39, Page Page 9, 9, Section Section 3.1 3.1 Cog Level: High Cog Level: High Explanation:

Explanation:

Requires knowledge Requires knowledge of of equipment equipment powered powered from from MCC MCC 2XA-2 2XA-2 (opposite (opposite UnitUnit power power E5).

E5). With With the the RR RR Loop A isolated Loop isolated (RR (RR Discharge Discharge and and Disch Disch Bypass Bypass valves valves will be be close close - required

- required for for LPCI)

LPCI) and and F015A FO15A (located in (located in ECCS ECCS Pipe Pipe Tunnel Tunnel - RB

- RB 20')

20) can can be be manually manually opened opened by dedicated operator.

by aa dedicated operator. 001-01.08 001-01.08 has has recently been recently revised to support been revised support implementation implementation ofof OPS-NGGC-1000, OPS-NGGC-1 000, Fleet Fleet Conduct of of Operations.

Operations. Risk Risk assessment and assessment and equipment equipment removal removal from from service service guidance guidance has has been been removed removed from from 01-01.08.

01-01.08.

Distractor Analysis:

Choice A: Plausible because if the valve is de-energized in its intended state (such as a PC IV in the PCIV closed dierection) this could be considered correct although for this case the open direction for this normally closed valve would have to have interlocks defeated to have both the F015 and F017 both open (potential to pressurize low pressure piping). OAP-025 is correct.

Choice B: Plausible because if the valve is de-energized in its intended state (such as a PCIV in the closed dierection) this could be considered correct although for this case the open direction for this normally closed valve would have to have interlocks defeated to have both the F015 and F017 both open (potential to pressurize low pressure piping). 001-01.08 001-01 .08 no longer provides guidance for evaluating MRIPSA MR/PSA system availability.

Choice C: Correct Answer Choice D: Plausible because RHR Loop is available and 001-01.08 no longer provides guidance for evaluating MRIPSA system availability.

SRO Only Basis: Knowledge of administrative procedures that specify implementation, implementation, and/or coordination of plant normal procedures.

Notes

3.0 3.0 DEFINITIONS DEFINITIONS 3.1 3.1 Available (Availability)

Available (Availability)

The status The status of system, structure of aa system, structure or or component component (SSC) (SSC) that that isis OPERABLE, OPERABLE, inin service or service or can can be be placed placed inin aa FUNCTIONAL FUNCTIONAL state state within within aa reasonably reasonably short short period of period of time time consistent consistent with its intended with its intended need.

need. The The SSC SSC must must be be capable capable of meeting of meeting all of its all of its most most limiting limiting requirements requirements for for the the plant plant mode mode under under consideration Using consideration. Using aa manual manual means means for for placing placing an an SSC SSC in in service service requires aa dedicated requires dedicated operator operator assigned assigned to to bebe cognizant cognizant of of the the SSC SSC along along with aa written with procedure tor written procedure for its restoration. A its restoration. A "dedicated" dedicated operator operator for for the the purpose of purpose of this definition is this definition is one one whowho is is specifically specifically assigned assigned the the task task and and 2vailable, as available, necessary, to as necessary, to perform perform the the required required actions.

actions.

3.2 3.1 Backbone Schedule Backbone Schedule AA preliminary preliminary schedule schedule consisting consisting of of work work items items that that are are either either required required to to be he performed or performed have been or have been deSignated designated by by management management as as high high priority priority items.

items.

The following items items would nom1ally nomally comprise the backbone schedule:

- Implementing Supervisor recommendations Implementing

- KeyI(a)(1 Equipment priority Key/(a)(1) priority action items

SurveillancesiPMs Required Surveillances/PMs

- System Outages

- Committed Items - Priority 11 & 2 C.4.PR's, CAPRs, CORR's, CORRs. and regulatory committed items

Modification EC's ECs determined a priority by Engineering repre.sentative representative or Scheduler (must be ready to work with work orders in ready or approved status)

Engineering recommendations

Reactivity Management flagged Work Orders 13 3.3 Compensatory Actions Measures that are used to mitigate niitigate the impact and minimize the duration of an ELEVATED RISK activity. These measures may include CONTINGENCY PLANS or procedural controls.

3.4 Contingency Contingency Planning A look ahead process whereby potential problems are systematically identified, assessed, assessed, and addressed by adding plans or mitigating mitigating actions.

actions.

The necessity necessity for aa contingency contingency plan plan is is based based on the the potential potential consequences as as well well as the probability as the probabllity of of aa problem occurring.

DAP-025 IOAP-025 Rev Rev. 39 39 I Page 99 of Page of 121 121 I

ATTACHMENT 33 ATTACHMENT Page 15 Page 15 of 19 of 19 480V Substation 4BOV Substation E5/MCC/Panei E5?MCCIPaneI Load Load Summary Summary 480V Motor Control Load: 480V Motor Control Center Load: Center2-2XA-222XA-2 Locadc9: Unit Location: Ur t ReaotorBuilding 2 Reactor 3T. ldin 20' NE ZY NE Draw ng

Reference:

Drawing

Reference:

F-03D49

-2 3c-9 Upstream Power Upstream PowerSource:

Source: 4S0V dBOV SUbstation Substation E5 E5 COMPT COMPT LOAD DESCRIPTION LOAD DESCRIPTION EFFECTS ON EFfECTS ON LOSS LOSS OF OF POWER POWER DF5 DFS F?lR Outboard RHR Cuibaard Injection Valve lniection Valve Loss of Less ofload load 2-El1-F3l7ATS 3.5.1.

2-E11-F017A(TS 3.5.1, 3.6.1.3, 36.1.3, 3.5.2, 3.5.2, 3.3.2.1:

3.3.3.1}

D3 DF3 RHR Inboard Injecticn RHR Inboard lnection Val'.e Valve 2-E11-FOt5A 2-El -FO15A Loss of Loss of load load

{IS 3.51, 3.8.1.3, ITS 3.5.1. 3.8.1.3, 3.5.2, 3.5.2,3.3.3.1) 3.3.3.1}

DGO DGO RHR Tor,s Spray HR Torus Spray Valve Valve 2-E i-F028A 2-El11-1"028.11, Loss of Loss of load load tTS3.8.1,3.8.

(IS 1.3.3.&2.3.

3.5.1. 3.8.1.3, 3.3.3.1) 3.6.2.3, 3.3.3. t) 0D7 007 Rx Recirculation Rx ecrculation Pump Pump 2A2A Discharge Discharge Valve Valve Losa of Loss of load load 2-B22-FO31ATS 2-B;?2-F031A 3.4.1, 3.S.1}

iTS 3.4.1. 3.5.1) 008 008 Rx Recrculation Pump Rx Recirculation Pump 2A Discharge 2A Discharge Loss of Loss of load load Sypass Valve Bypass Valve 2-B32-F032A 2-232-F032A (is 3.4.1 (TS 3.4.1, 3.5.11 3.5.1)

\ 001-50.1 001-50.1 Rev.

Rev. 42 42 Page Page 24 24 of 55\

of 55

ATTACHMENT2.4-ATTACHMENT 2A Page 22 of Page of 33 Residual Heat Residual Hear Removal Removal System System Loop Loop AA Panel Panel Lineup Lineup Number Numb:r Description Description Positicn Positionl Checked Checked Verified Venfied Indication Indication Loop AA Control Loop Control Room Panel H12-P801 Room - Panel H12 P801 El lFcOeC El1-FOOeC Pump CC Shulx:lc'IIn Pump Shuown Cooling Cooling CLOSED CLOSED Suction V SuctionV'N El l-FCOeA Ell-FOOM Pump AA Shutdown Pump Shutdovn Cooling Cooling CLOSED CLOSED Suction V SuclionV'N El l-V32 Ell-V32 Check Va'Ne Check Vae 8}'Jlass Bypass VIII VIv CLOSED CLOSED El l-FO17A Ell-F017.4 Ouoard Injection Outboard lrec:ion VIII Vlv OPEN OPEN El l-F0le4 El1-F01M Drywell Spray Drywell Spray Otbd Otbd IsolIsol VIII Vlv CLOSED CLOSED Ell-F1O4A Ell-Fl04A HX 2A

!-IX Inboard Vent 2A Inboard Vent V'NVlv CLOSED CLOSED Ell-FOI5A Ell-FOl5A Inboard InjeciionVlv Inboard Injection V'N CLOSED CLOSED El l-FO2IA El1-F021A Drywell Spray lnbd Orywell Spray Inbd (501 led Iflv Vlv LOSEO CLOSED Ell-FIO3A El1-Fl03A HX 2.4

!-IX 2A Outboard Outboard Vent Vlv /1v CLOSES CLOSED El 1-F024A El1-F024A Cooling 1501 Torus Cooling Isol VIII Vhi CLOSED CLOSED Ell-FC4A Ell-F048A HX24Bypass1v

!-IX 2A Bypass 'v1v OPEN El l-F027A Ell-F027A led V'N Torus Spray lsol Vi CLOSES CLOSED El 1-FOl IA Ell-FOllA HX 2A

!-IX 2 Drain To Torus Vl Vlv CLOSES CLOSED El 1-FcO4C Ell-FOO4C Pump C Torus Suction V'N Viv OPEN Eil-F028A Ell-F028A Torus Discharge lsol IsolVlv V'N CLOSED El I-F026A E11-F02M -IX 2A Drain To RCIC VI"

!-IX VIv CLOSES CLOSED El I-FcO4A El1-FOO4A Pump A Torus Suction Sucbon VIII VIv OPEN E1l-FcO2A E1l-FOO3.4 HX 2A Outlet

!-IX Cutlev Vlv OPEN El I-FC7A E11-FOO7A Mn Flow Bypass Min Sypass V IJI~' CLOSED El 1-FO2GA Ell-F020A Pump A&C Torus Suction Vlv Ifl... CPEN OPEN El i-Fc47A E11-F047A HX Z 2A InletV Inlet V'N OPEN El l-FE8OA E11-FOOOA Manual lnection ilv M.anuallnieciion Vlv OPEN OPEN El l-PDV-Ffl38A E11-PDV-F068A HX 24 SW Disch

!-IX Z Disch VIv VI" CLOSED CS-517A CS-S17A Containment Containment Spray Spray Valve Va'Ne Control Control OFF OFF Think Think Swilch Switch 20P-17 1 20P-17 Rev.

Rev. 155 155 Page Page 244 244 of of 297 2971 Categories Categories K/A:

KIA: S600000G2.02.37 S600000G 2.02.37 Tier/Group:

Tier / Group: T1G1 T1 G 1 RO Rating:

RORating: 3.6 3.6 SRO Rating:

SRO Rating: 4.6 4.6 LP LP Obj:

Obj: Source:

Source: NEW NEW Cog Cog Level:

Level: HIGH HIGH Category Category 8:8:

94. What

94. What action action isis required required to to be taken ifif Alternate be taken Alternate Safe Safe Shutdown Shutdown (ASSD)

(ASSD) Staffing Staffing drops drops below minimum below minimum complement complement due due to an emergent to an emergent on-shift on-shift AO AC illness illness and and what what procedure provides procedure provides the the guidance guidance for for this this action?

action?

The guidance The guidance for for establishing establishing an an (1)

(1) ifif ASSD ASSD staffing staffing composition composition isis less less than than the the minimum required minimum required isis provided provided by by (2)

(2)

(1) ASSD A (1)

A'I ASSD Impairment Impairment (2) OASSD-OO, (2) OASSD-00, User User Guide Guide B. (1)

B. (1) Active Active LCOLCO forfor T.S.

T.S. 5.2.2, 5.2.2, Facility Facility Staff Staff Organization, Organization, (2) OASSD-00, User (2) OASSD-OO, User Guide Guide C. (1)

C. ASSD Impairment (1) ASSD Impairment (2) 001-01.01, (2) 001-01.01, BNP BNP Conduct Conduct of of Operations Operations Supplement Supplement D. (1) Active LCO for forT.S. 5.2.2, Facility Staff T.S. 5.2.2, Staff Organization, Organization, (2) 001-01.01, (2) 001-01 .01 , BNP BNP Conduct of Operations Supplement Operations Supplement

Feedback Feedback K/A: SG2.01.05 KIA: SG2.O1 .05 Conduct of Conduct Operations of Operations Ability to Ability to use procedures related use procedures related to shift staffing, to shift staffing, such such asas minimum minimum crew crew complement, complement, overtime overtime limitations, etc.

limitations, etc.

(CFR: 41.10/43.5/45.12)

(CFR: 41.10 / 43.5/45.12)

ROISRO Rating:

RO/SRO Rating: 2.9/3.9 2.9/3.9 CLSLP304M*1 3m Objective: CLS-LP-304-M*13m Objective:

13. Given

13. Given ASSD ASSD procedures procedures and and plant plant conditions conditions that that require require use use of of ASSD ASSD procedures, procedures, determine determine the the following:

following:

m. The

m. The manpower manpower required required to support the to support the ASSD ASSD actions.

actions.

Reference:

OASSD-00, Revision OASSD-OO, Revision 37, 37, Page Page 30, 30, Section Section 5.3.3 5.3.3 Cog Level:

Cog High Level: High Explanation:

Explanation:

The ASSD The ASSD staffing staffing composition composition may may be less than be less than the the minimum minimum requirements requirements for for a period period of time time not not to to accommodate unexpected absence of on-duty shift crew members provided exceed two hours in order to accommodate immediate action is taken to restore requirements to within the minimum requirements of the shift ASSD staffing. If the ASSD staffing composition is less than the minimum minimum required, establish an Alternative Safe Shutdown Impairment in accordance with OPLP-1.5, OPLP-1 .5, Alternative Shutdown Capability Controls, and OFPP-020, Impairment Notification.

Distractor Analysis:

Choice A: Correct Answer Choice B: Plausible because an impairment is the same as LCO (001-01.01) but impairments are not established against TS 5.2.2 and ASSD User Guide is correct.

Choice C: Plausible because ASSD impairment is correct and 001-01.01 provides staffing requirements OASSD-00 procedure use for required staffing.

for TS 5.2.2 but directs use of OASSD-OO Choice D: Plausible because Plausible because an impairment is the same as LCO (001-01.01) but impairments are not established against TS 5.2.2 and 001-01.01 provides staffing requirements for TS 5.2.2 but directs use of OASSD-00 OASSD-OO procedure use for required staffing.

SRO Only Basis: Requires knowledge of TS 5.2.2 Facility Staff Staff Organization - and prescribes the procedure procedure required for guidance guidance during during periods of ASSD minimum minimum complement complement notnot maintained.

Notes

5.0 5.0 INSTRUCTIONS INSTRUCTIONS 5.3 General 5.3 General GUidelines Guidelines for ASSD Staff for ASSD Staff 6.31 5.3.1 All ASSD All ASSD Staffing Staffing Roster Roster members members must must be capable of be capable of prompt prompt response when response when events events are are inin progress progress that that may may require entry into require entrl into ASSO procedures.

ASSD procedures.

5.12 5.3.2 All ASSD All ASSD members members shall shall obtain obtain aa deSignated designated radio radio at at the the beginning beginning of of shift and shift ensure that and ensure that itit is is charged.

charged.

rNol-E:

NOTE: Planned reduction Planned reduction of of ASSn ASSD personnel personnel below below the the minimum minimum number number required required permitted.

NOT permitted.

is NOT is

5.3.3 The ASSD staffing composition composition may be less may be less than the minimum minimum requirements for aa period requirements period of of time not toto exceed exceed twotwo hours hours in order order toto accommodate unexpected accommodate unexpected absenceabsence of of on-duty on-duty shift shift crew crew members members provided immediate provided immediate action is is taken to restore restore requirements requirements to within the minimum requirements of the shift ASSn ASSD staffing.

5.3.4 ASSD staffing composition is less than the minimum required, If the ASSn required.

establish an Alternative Safe Shutdown Impairment in accordance '.'1ith with OPLP-1.5, Alternative Shutdown Capability Controls, and OFPP-020, Impairment Notification.

5.3.5 If an impairment exceeds two hours, initiate a Condition Report.

5.3.6 With both units in Mode 4 or 5, ASSD staffing is not required.

DASSO-QO 10ASSD-OO Rev. 37 Rev. 37 Page Page 30 of 53 30 of 531

5.0 5.0 INSTRUCTIONS INSTRUCTIONS 5.4 Minimum 5.4 Minimum ASSD ASSD Nuclear Nuclear Shift Shift Staffing/Assignments Staffing/Assignments 5.4.1 5.4.*1 Senior Reactor Senior Reactor Operators:

Operators:

1 Unit 11 seo:

Unit SCO: Unit Unit 11 Remote Remote Shutdown Shutdown Panel Panel 11 Unit 22 seo:

Unit SCO: Unit Unit 22 Remote Remote Shutdown Shutdown Panel Panel 5.4.2 5.4.2 Auxiliary Operators:

Auxiliary 1 Unit 11 Reactor Unit Reactor Building/MCe BuildinglMCC Operator Operator or as directed by as directed by the Unit seo Unit SCO 11 Unit 22 Reactor Unit Reactor BuildingJMCe BuildinglMCC Operator Operator or or as directed by as directed by the the Unit seo Unit SCO 11 Diesel Generator Operator or as directed by the Unit seoSCO 11 Switchgear Operator Emergency Switchgear Operator or as directed by the as directed Unit seo Unit SCO 11 Service Water Building Operator or as directed by the Unit SCO UnitSeO

\ QASSD-00 OASSD-OO Rev. 37 Page Page 3131 ofj of 53\

9.4 9.4 Operations Leadership Operations Leadership RoleRole in in Station Station Activities (continued)

Activities (continued) 5.

5. Operators work Operators work closely closely with with station station support support personnel personnel to to establish establish appropriate priorities appropriate priorities for resoMng plant for resolving plant equipment equipment and station program and station program deficiencies. Being deficiencies. Being aware aware of of the the integrated integrated effect effect of of equipment equipment out out of of service and service and establishing establishing priorities priorities for for equipment equipment return-to-service return-to-service consistent with consistent with plant impact are plant impact are key key components components of of this this philosophy.

philosophy.

6.

6. Operations pursues OperatiOns pursues the root cause(s}

the root cause(s) of of problems; problems; provides provides direction direction to implement to implement corrective corrective actions actions and hold department and hold department and and station station personnel accountable personnel accountable for for achieving achieving expected expected levels levels ofof perfornlance.

perfomance.

9.5.

9.5. Operations Shift Operations Shift Slaffing StaflinO Standards Standards Operations ensures Operations ensures that the Control Room Room is is adequately adequately staffed for for plant plant operations with appropriately with appropriately qualified qualified individuals.

individuals. Additionally, Additionally, Operations Operations ensures ensures staffing staffing is is adequate to meet adequate meet regulatory regulatory andand programmatic programmatic requirements.

requirements.

Expectations 1.

1. General

a. The CRS and Shift Manager are responsible for ensuring that watchstanders hold required positions. Personnel only qualified watchstanders should verify they are qualified for the position to be held prior to assuming the watch.

b. Individual qualifications qualifications for specific specifc positions can be found in REG-NGGC-0012, Confirmation of Personnel Qualifications REG-NGGC-0012, Qualifications Associated with Commitments Commitments to Regulatory Guide 1.8.

c. The shift complement may be one less than the minimum requirement for a a period of time not to exceed 2 hours2.314815e-5 days 5.555556e-4 hours 3.306878e-6 weeks 7.61e-7 months in order to accommodate accommodate unexpected absence of on-duty onduty shift meml)ers members provided immediate action is taken to restore the shift complement to within the minimum requirements. requirements. This provision does not permit any shift member position to be unmanned upon shift change due to an oncoming shift member meml)er beingI)eing late or or absent.

absent.

d.

d. Shift staffing shall shall meet the the requirements of the indMdual individual plant license/Tech licenseffech Specs Specs and and other regulatory regulatory andand programmati programmatic c required positions at required positions at all all times. Required Required staff staff numbers numbers and and positions positions can be be found in in Attachment I1 "Shift Shift Staffing.

Staffing".

I OPS-NGGC -1000 OPS-NGGC-1000 Rev.2 Rev. 2 I 1491 Page 46 of 1491 Page46of

Attachment 1I - Shift Attachment - Shift Staffing Staffing Sheet 11 of2 Sheet of 2 Shift Manning Shift Manning BNP BNP Position Position Minimum Minimum Note Note staffing staffing SM SM 1I CRS CRS 22 SRO/STA SRO/STA 11 RO RO 33 AO AO 99 9.5 9.5 Operations Shift Operations Shift Staffing Staffing In addition In addition to to the the requirements requirements of of OPS-NGGC-1000, OPS-NGGC-1000. the the following following requirements apply:

requirements apply:

9.5.1 General The following table outlines the administrative guideline for the normal Operations normal Operations shift complement A.flY Any deviation from the normal shift complement must remain in accordance with Section 5.2.2 of Technical Specifications, and applicable apphcable sections of OASSO-OO, IJASSD-OO.

OFPP-031. and Attachment 13. (Attachment 13 contains a listing of OFPP-031, required ERO Watch Stations and qualifications for each and ASSD position& This attachment may be used as a tool to support positions.

determining shift staffing requirements.)

Watchstations BNP Watchstations BNP Shift Com~lement Complement License Shift Manager (SM) 1 Shift Manager 1 SRO Control Room Supervisor (CRS) 2 CRSs (1 for each unit) SRO Reactor Operator (RO) 4 Reactor Operators (typically, rtypically. 2 ROISRO RO/SRO for each unit)

Auxiliarj Auxiliary Operator (AO) 9 (includes 2 in Radwaste) N/A Operations Center/Field SRO 11 Operations Center/Field Center/Freid SRO SRO STA STA' 11 STA STA STA Qualified STA Qualified The

'The STASTA may may stand stand watch watch as as aa CRS CRS or or Reactor Reactor Operator Operator provided the the following requirements are are met:

met:

At At least least 44 SROs SROs are are available available on on shift shift (this (thiS includes includes thethe STA STA hut but does NOT does .NOT include include the the Fire Fire Brigade Brigade Advisor Advisor which which may may bebe filled by an filled by an RO RO licensed licensed individual).

individual).

Another Another Licensed Licensed Operator Operator is is designated designated to to relieve relieve the STA as the STA as Unit Unit CRS CRS or or RO.

RO.

(Relief as (Relief as Reactor Reactor Operator Operator is is required only one required ifif only one operator operator isis assigned assigned to to aa unit unit Relief Relief as as CR5 CRS shall shall he be filled filled from from thethe CR5 CRS position position on on the the shift shift staffing staffing roster.)

roster.)

The The designated deSignated relief relief must must NOTNOT be be assigned assigned as the Fire as the Fire Brigade Brigade Advisor.

Advisor.

- The The designated deSignated relief relief has taken turnover has taken turnover on on the the affected affected unit.

unit.

The designated The deSignated relief relief must must he able to be able to relieve relieve the the STA within 10 STA within 10 minutes.

minutes.

001-Ui .131 1001-01.01 Rev.

Rev. 29 Page Page 1414 of 1771 of 177

ATTACHMENT 13 ATTACHMENT 13 of 22 Page 11 of Page Operations Staffing Operations Staffing Roster Roster Date: _ _ _ _ _ __

Date:__________________ Shift _______

Shift:________________

STA P613 STA SRO PB11 Unit 1 CRS/U-1 RSD Panel SRO PB1I Unit 2 CRS1U-2 RSD Panel ARO PB12 Unit 1 RB MCC Operator ARC P812 Unit 2 RB MCC Operator ARC PB12 *FB Advisor CREC PBI7 CREC AC PB14 SW Operator AC P614 DC Operator AC P814 Emergency Switchgear Operator FB (SIC) FBO2FBD3 FB (SIC)

FB FBO2 F8 F8 FBO2 FB FB FGO2 FB FB F802 FB Security Contact E&RC Contact Maintenance Contact A

May hold an RO OR SRO license.

Security Key Accountabili Accountability ty Unit I RB AC iiiui; ii*;

Unit2 RBAO Outside AC Unit 1 CRS Un1t2CRS Shift Manager 001-01.01 1001-01.01 Rev. 29 Rev 29 Page Page 97 97 of 1771 of 177

5.2.2 5.2.2 Facility Staff Facility Staff The facility The facility staff staff organization organization shall shall include include the the fol[owing:

following:

a.

a. AA total total of three non-licensed of three non-licensed operators operators shall shall bebe assigned assigned for for Brunswick Brunswick Units 1I and Units at all and 22 at all times.

times.

( continued)

Organization Organization 5.2 5.2 5.2 Organization 5.2 Organization 5.2.2 5.2.2 Facility Staff Facility Staff (continued) continued) b.

b. east one At least one licensed licensed Reactor Reactor Operator Operator (RO)

(RO) shall shall be be present present in in the control room control room when fuel is is in in the reactor.

reactor. InIn addition, when eithereither unit unit is is in in MODE 1, 2, or MODEl, or 3. at at least least one licensed licensed Senior Senior Reactor Reactor Operator Operator (SRO)(SRO) shall be present in be present in the control room. With one unit room. V'I'ith unit in in MODEl, 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

c. Shift crew composition may be lee.s less than the minimum requirement of 10 50.54(m)(2)(i) and Specifications 5.2.2.a and 5.2.2.g for a peliod CFR SO.54(m){2)(i) period of time not to exceed 2 hours2.314815e-5 days 5.555556e-4 hours 3.306878e-6 weeks 7.61e-7 months 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 hours2.314815e-5 days 5.555556e-4 hours 3.306878e-6 weeks 7.61e-7 months , in order to provide for unexpected absence: absence, provided immediate action is taken to fill the required pOSition. position.

e. Deleted.

f. The The operations manager manager or or assistant assistant operations operations manager manager shall shall hold hold an.

an SRO SRO license.

license.

g.

g. The The shift shift technical technical advisor advisor shall serve in shall serve in an an advisory advisory capacity capacity to to the the shift shift superintendent superintendent on on matters pertaining to matters pertaining to the the engineering engineering aspects aspects assuring assuring safe safe operation operation of of the the unit unit when either unit when either unit is is in in MODE MODE 1.2, 1, 2, or or 3.

3.

Eiunswick Brunswick Unit Unit 1'1 5.0-3 5.0-3 Amendment No.

Amendment No. 253 253 II

Categories Categories K/A:

KIA: SG2.0L05 SG2.01.05 Tier/Group:

Tier / Group: T3 T3 RO Rating:

RORating: 2.9 2.9 SRO Rating:

SRORating: 3.9 3.9 LP Obj:

LP Obj: CLSLP3O4M*13M CLS-LP-304-M* 13M Source:

Source: NEW NEW Cog Level:

Cog Level: HIGH HIGH Category 8:8:

Category YY

95. OFH-11,

95. OFH-1 1, Refueling, Refueling, prohibits prohibits control control rod rod withdrawal withdrawal during during the the core core load load sequence sequence untiluntil neutronic bridge aa neutronic bridge isis established.

established.

Which one Which one of of the the following following core core loading loading sequences sequences establishes establishes aa neutronic neutronic bridge bridge as as described inin OFH-11?

described OFH-1 1?

Four fuel Four bundles are fuel bundles loaded around are loaded around (1)

(1) ,then fuel

,then fuel isis loaded loaded inin all all fuel fuel cells cells in in aa line between line between SRMs SRMs (2) (2) .

A. (1) SRMs A. (1) SRMsAandA and D Donly only (2) A (2) AandD and D B. (1)

B. (1) SRMs SRMs BandB and D D only only (2) Band (2) B and D D c.

C. (1) each of the four SRMs and D (2) A and D

D~ (1) each of the four SRMs B and D (2) Band

Feedback Feedback K/A: SG2.01.42 KIA: SG2.01.42 Conduct of Conduct of Operations Operations Knowledge of Knowledge new and of new and spent spent fuel fuel movement movement procedures.

procedures.

(CFR: 41.10/43.7/45.13)

RO/SRO Rating:

RO/SRO Rating: 2.5/3.4 2.5/3.4 CLSLP305C*

Objective: CLS-LP-305-C*

Objective:

Reference:

Revision 93, Page OFH-1, Revision Page 9, Section 4.37 High Cog Level: High Explanation:

Explanation:

Provide ENP-24-12, Provide ENP-24-12, Figure Figure 11 as a reference reference From FH-11, 4.37 To help ensure that an unmonitored criticality will not occur, control rod withdrawal is not allowed during the core reload sequence until a neutronic bridge is established. The neutronic bridge ensures that two SRMs are neutronically coupled, thus monitoring the loaded area of the core. The reload sequence has three basic steps. Four fuel bundles are loaded around each of the four SRMs, the neutronic bridge is established and a spiral reload of the other fuel bundles completes the sequence. The neutronic bridge is established by loading fuel in all fuel cells in a line between two SRMs. These SRMs must be on opposite sides of the core and the line of loaded fuel cells must intersect the center of the core (A to D would not intersect the center, B to D would).

Distractor Analysis:

Choice A: Plausible because loading fuel around 2 SRMs and a line between them would establish a neutron bridge (between those 2 SRMs)SRM5) but not lAW PAW OFH-11 OFH-1 1 and A&D are adjacent.

Choice B: Plausible because loading fuel around 2 SRMs and a line between them would establish a neutron bridge (between those 2 SRM5) JAW OFH-1 SRMs) but not lAW OFH-11I and A&D are adjacent.

Choice C: Plausible because loading fuel around all SRMs is correct but A&D are adjacent.

Choice D: Correct Answer SRO Only Basis: IOCFR55.43.6 10CFR55.43.6 Procedures and limitations involved in initial core loading, alterations in core configuration, control rod programming, and determination of various internal and external effects on core reactivity.

IOCFR55.43.7 Fuel handling 10CFR55.43.7 handling facilities and procedures.

Notes

4.0 4.0 PRECAUTIONS AND PRECAUTIONS AND LIMITATIONS LIMITATIONS 4.34 4.34 RPS shorting RPS shorting links links SHALL SHALL be be removed removed for for control control rod rod withdrawal withdrawal (except (except forfor control rods control rods removed removed inin accordance accordance withwith Technical Technical Specifications)

Specifications) inin thethe refuel mode refuel mode whenwhen corecore verification verification AND AND subsequent subsequent strongest strongest rod rod out out verification have verification have NOT NOT been been performed.

performed. Control Control rods rods may may be be withdrawn withdrawn with with the shorting the shorting links links installed, installed, provided provided core core verification verification (OENP-24.13),

(QENP-24.13),

subsequent strongest subsequent strongest rodrod out out verification verification (single (single control control rod rod sub criticality test suhcriticality test in accordance in accordance with with OFH-11)

OFH-1 1) have have IJeen been performed, performed, andand the the one-rod-out one-rod-out refuel interlocks refuel interlocks havehave been been demonstrated demonstrated to to IJe be operable.

operable.

4.35 4.35 An SRO An SRO withwith nono other other concurrent concurrent duties duties shall shall directly directly super"ise supervise all all core core alterations.

alterations.

4.36 4.36 Members of Members of fuel fuel handling handling crew, scheduled for crew, scheduled consecutive dail~'

for consecutive daily duty, duty, should should NOT nom13l1y NOT normally work work more more than than 12 12 hours1.388889e-4 days 0.00333 hours 1.984127e-5 weeks 4.566e-6 months hours outout of of each each 2424 hours0.0281 days 0.673 hours 0.00401 weeks 9.22332e-4 months .

hours.

4.37 help ensure that an unmonitored To help unmonitored criticality will NOT NOT occur, control rod rod withdrawal is is NOT allowed allowed during the the core core reload reload sequence until until aa neutronic neutronic bridge is bridge established. The is established. The neutronic neutronic bridge bridge ensures ensures that two two SRMs SRM5 are are neutronically coupled, neutronically coupled, thus monitoring monitoring thethe loaded loaded area area of the the core.

core. The reload seQuence reload sequence has three basic steps. Four Four fuel bundles are loaded loaded around each ofthe of the four SRMs, the neutronicneutronic bridge is established and a spiral reload of the other fuel bundles completes the sequence. The neutronic bridge is established by loading fuel in all fuel cells in a line between two SRM& These SRMs must be on OPPOsite SRMs. opposite sides of the core and the line of loaded fuel cells must intersect the center of the core.

4.38 With fuel removed, if a control rod is withdrawn without blade guides capability shall be removed for the control rod.

installed, the insertion capabiliPf 4.39 The Bridge Operator should immediately push the STOP button if the bridge fails to respond to Operator commands, such as speed changes or jogs. The failS STOP button will prevent all an bridge movement.

movement 4.40 If attaching tools, such as a ajet jet pump grapple or controll)lade control blade latching tool, tOOl, to either the monorail or frame mounted hoist, verify proper thread engagement/

engagement/size size by ensuring there is no play in the connection prior to thread engagement of three (3) full tum& turns. The correct tool and coupling thread size is 7116-14 UNC. AdditionallyAdditionally,, aa 112-13 112-13 UNC bolt IlOlt will wiI[ not fit into a a proper size tool (7/16-14

{7/16-14 U NC); thus, this check may be performed if UNC);

practical.

practical. Failure Failure to detect detect mis-matched mis-matched thread sizes will significantly significantly reduce the the load load capacity callacity of of the the tool/hoist.

tOOl/hoist.

4.41 4.41 Indication Indication of criticality observed observed on the SRM SRM indicators indicators during during functional, subcritical, subcritical, or or shutdown shutdown marginmargin rod rod checks checks shall shall be be reason reason to to teminate temlinate fuel loading loading until until aa complete evaluation evaluation of of the the cause of oftM criticality indication the criticality indication is is determined.

determined.

OFH-1 1 IOFH-11 Rev.

Rev. 93 93 Page90f55 Page 9 of 55 I

FIGURE 09.1-FIGURE 09.1-22 IN-Core Instrumentation IN-Core Instrurnenation Location Location Diagram Diagram 00 PLANT ~

PLANT NORTH NORTH ,..".-

r n r h r!, /'"

~

r LA..J

l/ B,-l --41

(, ~

~~ ~ ffr -33 4:f.~ '( -:; t:l.f; ~ -25

~~ -11 L ~ :J w rHl L I L.~

w --09 L W I

12 20 28 D

LI IRM DETECTOR LOCATION ZSIO SRM DETECTOR LOCATION CORE CORE CORE SRM LOCATION RM IRM LOCATION IRM LOCATION A 12.33 A 12.41 12*41 E 28.25 28*25 a

B 2641 28-41 BB 36-41 FF 20-26 20*25 C 36-26 36-25 C 20-33 20*33 G 36-09 36*09 0

0 20.17 20*11 0 283.3 28.33 H 12-09 12*09 ISD-091 SD-09.1 Rev.

Rev. 66 Page 49 of 61 Page490f61 I Categories K/A:

KIA: SG2.O1.42 SG2.01.42 Tier/Group:

Tier / Group: T3 T3 RORating:

RO Rating: 2.5 2.5 SRO SRORating:

Rating: 3.4 3.4 LP Obj:

LPObj: Source:

Source: BANK BANK Cog Cog Level:

Level: HIGH HIGH Category 8:

Category 8: YY

96. With Unit

96. With Unit Two Two operating operating atat power, power, Annunciator Annunciator RCIC RCIC TURBINE TURBINE STMSTM LINE LINE DRN DRN POT POT LEVEL HI LEVEL alarms and HI alarms and the the RO RO observes observes the the E51-F054, E51-F054, F025, F025, && F026 F026 indicate indicate closed closed on Panel on Panel P601.

P601.

Which one Which one ofof the the following following identifies identifies the the cause cause of of the the above above indications indications and and the the operability status operability status ofof RCIC?

RCIC?

(Reference provided)

These valves These valves areare closed closed due due to to loss loss of of (1) (1) and after and taking the after taking the appropriate appropriate actions in actions in the the annunciator annunciator procedure procedure the the system system would would bebe declared declared Inoperable Inoperable and(but) and (but) (2) .

(2)

(1) pneumatics A. (1) pneumatics (2) Unavailable (2) Unavailable B (1) pneumatics B:-"

(2) Available C. (1) DC Power (2) Unavailable D. (1) DC Power (2) Available

Feedback Feedback K/A: SG2.02.15 KIA: SG2.02.15 EQUIPMENT CONTROL EQUIPMENT CONTROL Ability to Ability to determine determine the the expected expected plantplant configuration configuration using using design design andand configuration configuration control control documentation, such documentation, such as as drawings, drawings, line-ups, line-ups, tag-outs, tag-outs, etc.

etc.

(CFR: 41.10/43.3/45.13)

(CFR: 41.10 143.3/45.13)

ROISRO Rating:

RO/SRO Rating: 3.9/4.3 3.9/4.3 CLSLP016*15e Objective: CLS-LP-016*15e Objective:

15. Given

15. Given plant plant conditions, conditions, predict predict the the RCIC System response RCIC System response toto the the following following conditions:

conditions:

s. Loss

s. Loss ofof instrument instrument air.

air.

e. DC power

e. DC power failure.

failure.

Reference:

2APP A-03 2APP A-03 3-5, 3-5, Revision Revision 49, 49, Page Page 44 44 Cog Level: High Explanation:

Explanation:

Valves fail closed on loss of DC DC power or Pneumatics, however with a loss loss of power, position indication on P601 will also be lost. Per APP A-03, 3-5 -IfIf either E51-F025 or E51-F026 has been failed closed for more than 5 minutes, perform the following:

a. Close Turbine Trip and Throttle Valve, E51-V8, E51 -V8, to prevent water hammer damage from a RCIC auto start.

b. If RCIC must be started, proceed to OP-16.

this would still make RCIC available for use per the procedure but it is inoperable because it will not auto start as required.

This will make the RCIC system inoperable but available to be restarted per the procedure.

Distractor Analysis:

Choice A: Plausible because loss of pneumatics only is correct and the system will not start in auto when required, but could be manually started.

Choice B: Correct Answer Choice C: Plausible because a loss of power will cause valves to fail closed, but with loss of power position indication will be lost and the system will not start in auto when required.

Choice D: Plausible because pneumatics and power power will cause valves to to fail closed, but with loss loss of of power power position position indication indication will be be lost and itit is is available available to start per the procedure procedure which makes makes itit available.

available.

SRO SRO Only Basis: Assessment Only Basis: Assessment of of facility facility conditions conditions and and selection of appropriate procedures selection of procedures during during normal, normal, abnormal, abnormal, emergency emergency conditions.

conditions.

Notes Notes

RCIC STEAM POT Partial P&ID QRAIN!

~I

.f.£lI .

F04S

Unit Unit 22 APP APP A-03 A-03 3-5 3-S Page Page 22 ofof 22 (Cont a)

ACTIONS (Cont'd)

ACTIONS CAUTION CAUTION If Main Steam If r~in Steam Line Line Drain Drain v~v, Vlv, MVD-F021, NVD-F021, fails fails toto close, close, then then the the Main Main Steam Steam Line Drain Line Drain Inboard Inboard and and Outboard Outboard Isolation Isolation valves valves must must bebe closed.

closed.

6. If required,

6. If reguii-ed, then then close close [*lain Main Stearn Steam Line Line Drain Drain Inbd Inbd Iso1 lad V1v, Vlv, B21-F0l6, and B21-F016, Main Steam and Main Steam LineLine Drain Drain Otbd Otbd Iso1 Isol V1v, Vlv, B21-F019.

B2l-FOl.

7. If

7. If alarm alarm fails fails to to clear clear within within five five minutes minutes after after completion completion of of actions 11 2,3, actions 2,3, 5,5, oror 6, 6, then then dispatch dispatch an an Auxiliary Auxiliary Operator Operator to to thethe Drwell access Drywe11 access roofroof to to determine determine if if the the HPCI/RCIC HPCI/RCIC Condensate Condensate Drain Drain Line Back Line Back Pressure Pressure Orifice Orifice is is plugged plugged oror the the drain drain line line isolated.

isolated.

NOTE NOTB: Greater than 500 psig on Greater on HPCI/RCIC BackBack Pressure Orifice Inlet Inlet Pressure Gauge, 2-1*lYD-PI-7146 Pressure Gauge, 2-MVD-PI-7146 would would be be an an indication indication of aa plugged orifice.

orifice.

8. IF back pressure orifice is plugged, a,

a. Open HPCI/RCIC Cond (ond Drn Line Back Press Orifice Bx~ss Bypass Valve, 2-MVD-VSOO2.

2-['lVD-V5002.

b. BPCI/RCIC Cond Drn Line Back Press Orifice Inlet Iso1 Close HPCI/RCIC Isol Valve, 2-t-1VD-V5000.

2-MVD-VS000.

c. Close HPCI/RCIC Cond Drn Line Back Press Orifice Outlet Isol Valve, 2-MVD-VSO 2-t-r.rD-V5001.

Ol.

d. Place valves val,,'es under proper adrninistrat administrativ ive e control.

9. If NPCI/RCIC HPCI/RCIC Cond Drain Line is isolateth isolated:

a. Open HPCI/RIC HPCI/RCIC (ondCond Drn Line Back Press Orifice Inlet Isol Valve, 2-MVD-VS0 2-M'~~-V5000. 00.

b. Open HPCI/RCIC HPCI/RCIC Cond Drn Line Back Press Orifice Orifice Outlet Isol Valve, 2-MVD-VSO 2-M'.rD-V5001.Ol.

10.

10. IF IF a circuit circuit malfunction malfm1ction is is suspected, suspected, ensure ensure thatthat a WE/JO WR/JO is is prepared.

prepared.

DEVICE/SE TPOINTS D~vlCE/SETPOINTS Level Level Switch Switch ES1-LSH-NO E51-LSH-NOIO-l 1O-l Instrument failure in Instrument in the the Switch Switch Point Pc,int tl

1 dry condition/19 dry condition/1980 mV.

80 my.

Level Level Switch Switch ESl-LSH-NO E51-LSH-NOIO-l l0-l .~lso detects Also detects instrument instrument failure fa.ilure in ill the Switch the Switch Point #2/0" +/-+/- 2 Point #2/0 2" wet condition.

wet Inc01~orates condition. Incox-porate s water.

water. 100 sec 100 sec time time delay delay in ~1nunciator in annunciator circuitry.

circuitry.

POSSIBLE POSSIBLE PLANT PLANT EFFECTS EFFECTS Damage Damage to to the the RCIC RCIC turbine turbine due due to to high high moisture moisture carryover carryover on on the the steam.

steam.

REFERENC REFERENCESES 1.

1. LL-9364 LL-9364 - 50 50 2.

2. OP-iS, OP-16, RCIC RCIC System System Operating Opera.t ing Procedure Procedure 2APP-A-03 I2APP-A-03 Rev.

Rev. 49 49 Page 45 Page 45 ofof 102

'1021

Unit Unit 22 APP A-03 3-S APP "~.-03 3-S Page Page 11 of of 22 RCIC RCIC TIJREINE STI-l TURBINE STh LINE LINE DRN DRN POT POT LEVEL LEVEL HI HI RCIC iRCIC Turbine Steam Line Turbine Steam Line Water Water Drain Drain Pot Pot High High L=vel)

Level)

AUTO

"~UTO ACTIONS ACTIONS 1.

1. Supply Drain Supply Drain Pot Pot Drain Drain Bypass Bypass Valve, Valve, ESI-F054, ES1-F054, opens.

opens.

CCATJSE

....USE 1.

1. Heavy condensate Heavy condensate load load during dul-ing steam steam line line warmup.

warmup.

2.

2. Normal orifioe Normal orifice clogged.

clogged.

3,

3. HPCIJRCIC Cond.

HPCI!RCIC ond. Drain Drain Line Line Baok Back Pressure Pressure Orifioe Orifice is is plugged.

plugged.

4. Drain line Drain isolation valves line isolation valves to main oondenser condenser olosed.

closed.

5.

S. Drain pot level Drain level instrument instrument failure failure oror loss loss of instrument instrument, po;;er.

power.

6.

6. Circuit malfunction.

Cirouit OBSERVATIONS 1.

1. RCIC Supply Drain Pot Drain Byp Valve, ES1-F054, ESl-F054, opened.

NOTE:

NOTE: If alarm ooours If occurs and the E51-FOS4 ES1-F054 valve does not automatioally automatically open, the the most probable cause is instrument failure or loss of instrument oause power (Panel 2B-Rx "HIO" PH1O1T CKT 14/.

14),

NOTE: indications are available inside the level element Additional LED indioations control box device H5E HSE (P£ (RB 20' 20 elevation) as follo;;s:

follows:

Normal Normal status No annunciator No LEDs illuminated High Water level Green LED on Instrument failure Red LED on011.

ACTIONS ACTIONS 1.

1. Ensure Supply Drain Pot Drain Byp Vlv, "lIlv, ESl-F054, ESI-FOS4, is open.

2.

2. Ensure Ensure RCIC RCIC Supp Supp Pot Pot Inbd IOOd Isolation Isolation Valve, Valve, ESi-F025, ESI-F025, is is open.

open.

3.

3. Ensure Thlsure RCIC RCIC Supp Supp Pot Pot Outbd Outbd Isolation Isolation Valve, Valve, ES1-F026, ESI-F026, is is open.

open.

NOTE:

NOTE: Valves Valves E51-F025 E51-F025 andand E51-F026 ESI-F026 will will close close on on loss loss ofof instrument instrument airair and will also and will also close close if if E5l-F045 ESI-F045 is is not not fully fully closed.

olosed. Valves Valves ES1-F025 ESI-F025 and and E51-F026 E51-F026 cannot oannot bebe opened opened in in either either of of these these conditions.

conditions.

4.

4. If If either either ESl-Ft25 ESI-F025 oror ESl-F026 ES1-F026 ha has been been failed failed closed olosed forfor more more than than 55 minutes, minutes, perform perform the the following:

following:

a.

a. Close Close Turbine Turbine Trip Trip and and Throttle Throttle Valve, Valve, ESl-V8, ESI-V8, to to prevent prevent water water hammer hammer damage damage frcm from aa RCIC RCIC auto auto start.

start.

b.

b. If If RCIC RCIC must must hebe started, started, proceed proceed toto OP-16.

OP-16.

S.

S. Ensure Ensure Main Main Steam Steam Drain Drain Lme Line Vlv, Vlv, MVD-F02l, MVD-F02l, is is cload.

olosed.

2APP-A-03 I2APp-A-03 Rev.

Rev. 4949 Page 44 Page of '1021 44 of 102

8.4 8.4 Isolating the Isolating the RCIC RCIC System System Stearn Steam Supply Supply RR Rijlefijr R3r9rr Use Uae 8.4.1 8.4.1 Initial Conditions Initial Conditions 1.

1. All applicable All applicable prerequisites prerequisites listed in listed in Section Section 4.0 4.0 are are met.

met. o 8.4.2 8.4.2 Procedural Steps Procedural Steps 1.

1. IF rapid IF rapid isolation isolation of of RCIC steam line RCIC steam line is is desired, desired, THEN THEN PERFORM the PERFORM following:

the following:

a.

a. CLOSE STEAM CLOSE E1-FOO7.

E51-F007.

STEAM SUPPL SUPPLYIWBOARD Y INBOARD /SOL ISOL VL V.

VLV, o b.

b. CLOSE STEAM CLOSE E1-FOO8.

E5'1-FOOB.

STEAMSUPPLYOUTBOARDISOL SUPPL Y OUTBOARD ISOL VL 11, VLV, o CAUTION Opening the TURBINE STEAM SUPPL Y VL V. E51-F045, to de-pressurize the RCIC steam line will roU the RCIC turbrne, I

2. IF rapid isolation is NOT desired, THEN PERFORM the following to isolate and de-pressurize the RCIC RCIC steam supply line:

a. CLOSE STEAM SUPPL SUPPLY Y INBOARD ISOL 1SOL VL V.

VLV, 0 E51-F007.

E51-FOO7,

b. OPEN HPCLRCIC HPCllf?CIC COND DRN LINE SACK BACK 0 ORiFiCE BYPASS VALVE, PRESS ORIFICE VALVE MVD-V5002.

c. TURBiNE STEAM SUPPLY OPEN TURBINE SUPPL Y VL V, 0 E1-FO45, E51-F045, AND MONITOR turbine response.

d. CLOSE SUPPLY SUPPLYDRAIN DRAIN POT1N8D POT lNBD DRAIN VLV; VLV, 0 E1-FO25.

E51-F025.

e.

e. CLOSE SUPPLY DRAiN DRAIN POT OTBD OTaD DRAIN VLV, VL1I, 0 E51-FO26.

E51-F026.

20P-16 12oP-16 Rev.

Rev. 107 107 Page oi1 34 of Page 34 891

8.12 8.12 Controlled Manual Controlled Manual Start Start of of the the RCIC RCIC System System With With Turbine Turbine Steam Steam Line Line RR Reference Refrence Drain Pot Drain Pot High High Level Level oror RCIC RCIC Pump Pump LowLow Discharge Discharge Pressure Pressure Use Indicated Indicated 8.12.1 8.12.1 Initial Conditions Initial Conditions

1. IF RCIC IF RCIC is being operated for aa planned is being planned evolution evolution (non-(non o emergency operation), THEN THEN Health Health Physics Physics (HPs)

(HPs) shall shall be notified be notified to attend the pre~ob attend the pre-job briefing AND aa log briefing AND log entry entry made to identify the individual made individual contacted.

2. One of the follol;1,1ng following conditions exist exist:

a. The RCIC turbine has been shutdown or tripped and annunciator RCICRC!C TURBINE STM LINE DRN o

POT LEVEL HI (A-03 3-5) sealed in.

b. The RCIC turbine has been shutdown or tripped and the RCIC PUMP DISCH DISH PRESS LOW o

annunciator (A-02, 1-6) is sealed in.

3. A controlled manual start of RCIC is desired. o 8.12.2 Procedural Steps CAUTION The RCIC turbine has the potential for failures that could cause personnel injuries. The potential is most significant when the system is initially started after control system maintenance, or after an extended period of being idle. Announcing turbine starts and clearing of all personnel from the RCIC area are required during this period of risk.

Permission to access this area during initial RCIC roll requires the approval of the Unit SCO.

sco.

1. EVACUATE all personnel ffrom 0111 the he RCIC turbine area. 0 20P-16 120p--.16 Rev. '107 Rev. 107 Page 52 Page 52 of 891 of 89

ATTACHMENT ATTACHMENT2A 2A Page Page5of 5 of 30 30 PANEL 4A LOCATION NORMAL NORMAL SUPPLY SUPPLY Switchboard 21\

Control Building 49 ft East Reference Reference Drawin Drawing LL-3024-6 Switchboard 2A CIRCUIT CIRCUIT LOAD lOAD EFFECT EFFECT 1 Rx.Annunciator R>:. Logic, 2-H12-P630 Annunciator logic, 2-H12-P630 i.1. Auto transfers toto alternate Auto transiers alternatesoursource, ce, Panel4B PaneI4B circuit circuit 11.

Panels 60 Panels 80t/503

!fe03 2. Receive

2. Receive annunciator annunciator Ae-5-S.

A5-&.

22 HPCI Flow HPCI Flow controller controler [11. Controller Controller fails fails downsca.ie.

downscale.

E41-FIC-K800 (24 E41-FIC-K600 (24\lDC)

VDC;. 2.

2. Loss of loss offlo'l/

f;ow indication.

ndicatcn.

3.

3. Receive annunciator Reoeive annunciator AA1-2-5. 1-2-5.

4.

4. Loss of loss of HPCr HPCI 5.

5. Loss of loss of ASSD ASSD function.

function.

HPC1 Super~'isory HPCI Supervisory Lights Lights 1.1. Loss of loss of E41-1/S E41-VS and E4l-V indication and E41-V9 indicaticn 2.

2. Loss HPCI loss HPCI oiloil lank tank le ... el HillO level HilLO alam *.

alarm.

HPCI Vertical Board HPCI Vertical Board meters meters [1.1. Loss of loss pressure transmitters/meters of pressure transmitters/meters Re01, R32, Re~3, R6OI R602. R503, R60!

R80(

VDC)

(52.5 VDC)

(52.5 HPCI Turbine HPCl Turbine Speed Speed Control Control 1. loss

i. Loss ofof speed speed control, control, EGM 2GM and and speed soeed sensor.

sensor.

2. Loss of

2. loss of speed speed indication indicaon on on vertioal veitcal board.

board.

E41-F053, E41-F054.

E41-F053, E41-FOE4, E41-F026 E41-F026 1. Faif

1. Fail closed.

closed.

2. E41-F054 and

2. E41-F054 and E41-F026 241-F026 loss loss ofindication.

of indication.

251-F006,E51-F025 E51-F005, E51-F025 1. Failclosed.

1. Fail clcsed.

2. loss

2. Loss of indication, of indication.

001-50 1001-50 Rev.

Rev. 45 I

ATTACHMENT ATTACHMENT 29 25 Page Page 9 of 32 90132 PANEl:4B PANEL: 4B LOCATION:

LOCATION: NORMAL SUt>,P:L NOBMAt SUPPLY: Y: J Reference Reference Drawing:

Drawing: lL..,3024-7 LL-3024-7 Contra!

Control Building Building 49 49 ftSouth ft South .. Switchboard Switchboard 2B 2B Cku:,

Ckt# LOAD LOAD EFFECT EFFECT 5E Recrc Pump Recirc Auxiliary Equipment Pump 83 AuxillaI'I EqtLpment 1.1. Loss Loss ofof alternate aJternate control power to:

control power to:

Aternate Control Alternate Control PeweI Power ** Recilc Recirc B B Gen.

Gen. FieldFielo Breaker, Breaker, oontre!,

control, trip frp and and indication.

indicaior.

- Recilc Recirc BB Scoop Scoop Tube Thbe Power Power Failure Failure Lock Look && Reset Reset

- Recife Recirc LubeLube Oil Oil Pumps Pumps 8-1 ano B-2.

S-i and 8-2, control control andand indicatien.

ndication.

- Recifc Reciro BB lockLock out out ITrip lTrp logicLogic

- AATWS TWS Trip Trip logic Logic 8S

2. Normal power

2. Normal power is is irom from Panel Panel lOA, 1OA, ckt ckt 3.2.

t) Backup Scram Backup valve. 2-C12-i=11 Scram valve, 2-C12-1 lflS08 1.

1. Backup Backup ScramScram val'Jevalve fails fails dosed; closed; Div Div II Backup Backup ScramScram valve valve can can still still fun, fun Div IIII Backup Div Scram Logic Backup Scram Logic 1. Sc-ram Discharge

1. Sc*ram Discharge Volume Volume Vent Vent and and Drain 2ran Va.lves Vaves will wil nOl no receive receive aa close close valves wilf lIallies will still sti I function function wilhwith DillDiv I.I.

2. DFWLCS w~1

2. DFWLCS wD not rot receive receive auto, auto set set down down from Div rI.

from Dill II. Digital eedwater w Digital Feedwater w

2. Ten-second

3. Ten-second time tme delav delay prior por to scram reset, to scram reset, will will not not function function for for B B RPS RPS !

77 Spare Spare Spare Spare 8 RCIC Flo'll RCIC Flow contrcller controller 1. Controller fails

1. Conlroller cownecale.

sails downscale.

E51 -FlC-KDO (24 E5l-FIC-K800 (24 VDq VOC) Loss of

2. Loss of flow flow indication.

indication.

3. Receive

3. Receive annunciatcr annunciator />'3-6-5.

A3-6-5.

RCIC Supervisory RCtC Supervisory lights Lights Loss of

1. Loss

1. of E51-V8 E51-V8 and E51-V9 ndic-ation.

251-VQ indicatien.

RCIC Vertical Board meters RCIC meters 1. Loss

1. Loss of of pressure :ransmittersmeters ReOl, pressure transmitters/meters ROOl, R602,RO2, R60S.

RD3, R804 RCIJ4 on thethe R P i2.5 VDC1 (52.5VDCi E51-F026.

E51-F02& E51-F004.

E51-FCD4. E51-F054 1. Fai.!closed.

Fail closed.

2. Loss of of indication.

indicaon.

HPCI E41-F025 HPCr E41-FC25 1. Fail closed.

2. Loss of indication.

indicaon.

It RCIC 2GM RC1CEGM 1. Loss of speed control.

2. Loss of speed indication on RTG8 RIGS Initiation and Contrel RCIC Initiation Control Logic 1. will noi RCIC 'liill not auto iniliate.

initiate. Cannot be manuati'l manually operated.

2. Receive annunciator />'3-14. A3-1-4.

2,_

2. Mm flow Min flc'II valve will not auto open. open.

4. Barometric condenser lIacuum vacuum tank auto level control oontrol inop.

nop.

001-50 1001-50 I Rev. 45 r ,

PIPING HY tLJRB.VNDQR -

REv)SL PLR C 6t164 56 - QLlo PUi tC 1i4 13 h1 55 55 R-Vt5fD PIR rc 64l2

~~~'-L-------------~---~~~----------------;A A C3 PROGRESS ENERGY li- (MQRMATIOtI ON HS oNAw:1o CDMPI IFS WllI Ctk %l tJtT 2 5i S/A 2-FP46 (l PO 729E48B SH & 2) REACIOR BL.i1DING

- RLACJOR CORE ISOLATION NOTE RLVISONS TO THIS ORAWING MUST ALSO 1E COOLiNG SYSII M P4CORPORATO ON lE CORRESPONDING ORAW14S:

C) 1476, C) 0427, 0 04219, I) 04220 & 1) 04221 PIPING I)IAG AM Dø229

Q 1 R NIT S:

I QUPPJ

NS RUM1S & PIPING AR REIXII IY UNlf JNII &:

SSIM LMiS 2 th NLSS OTHWS N0ED UEFNCE LRWIIN(S SEC D Ø1q.

. 511 IMEiAT ION

.), IINSmUI,lENtAl lfl\ PENElRt'\nONS

- N RA1 ICNS AREAf* MULTI-LINES MJI I; L,IHFS THROUGH IR)JH ONl: QN S Iy 4.

4. ALL INSRUAENT RACKS A...L INSTRUMENT R.CKS M~E AFE PREnXED PREIXED H21" 2H2l.

S. Ki REFER

5. *FR O OG1C INTERLOCK.[NrR1QC<,

&A.L

6. ANUNCLA0 ALARMS ALL ANNUNCIATOR ALAQS N~E PRE-IXED 2 HI2**P60 R[ PI~EFlXED W2 Pfl XX,

,**xX".

7.

7. - . OENDTS W"LVE DENOTES VALVE LEAKOFF LEAKOFT 'NHICl-I WHICH VaLL E NORMAllY WILL BE NORM&L OPEN OPEN

& WIll WILL BE3E PIPED PIPED TO TO C.R.W.

CRW, UNDER CASS 160",

UNDER CLASS 6e.

8. VENDOR B. FURNIS ED.

VENDOR fURN1StIED.

9 < MASIR EQU!PMENI

)oio f,S MA$rEI~ EQUIPMN LIsr NUAt*R LISt NUMBER,

. SC- CS24 IS USt) fl Sl Ci filHfR IC RkI O FtC 3S? AI[) FO NSR OWR JIY SOUCf 10 1 NiL.

U, x=:s: CLASS -

=Qc)..LTY Cj.,S (1,2.3.M.- )

EE 0- ø2i9 FQF AOantONAl ADDI IIQNA NOTES.

NOTES t2 HtGH

12. HIGH POiNTPOThT VENT UTNT CAP CAE !S NORMALY R(MOVEO S NORMAlt,Y MQVEO ANO ORAIN tOSF INSTALLED AND DRAIN FOR iS1FM VENrtNG.

NSTAI D r.OR VENTING ii SE ECHNICAL *PoRr PI52 F"OR R(PORf FOR APPLICABLE APPUCADLE ASPE SCrION XI RFQUIRMN7S.

XI LF Categories Categories KJA:

KIA: SG2.02.15 Tier!/ Group:

Tier T3 RO Rating:

RORating: 3.9 3.9 SRO Rating:

SRORating: 4.3 LP Obj:

LPObj: CLSLPO16*15E CLS-LP-016*15E Source: NEW Cog Level: HIGH Category 8: YF

97. The

97. The following following conditions conditions exist exist on on Unit Unit One One after after aa transient:

transient:

Jet Pump Flow Jet Pump Flow Loop Loop AA 22 22 Mlbs/hr Mlbs/hr Jet Pump Jet Pump FlowFlow Loop Loop BB 33 Mlbs/hr 33 Mlbs/hr Recirc Pump Recirc Pump AA Percent Percent Speed Speed 47%

47%

Recirc Pump Recirc Pump BB Percent Percent Speed Speed 66%

66%

Total Core Total Core Flow Flow (U1CPWTCF)

(UICPWTCF) 55 55 Mlbs/hr MIbs/hr Which one Which one of the following of the following identifies identifies the the Required Required Action Action lAW lAW T.S.

T.S. 3.4.1, 3.4.1, Recirculation Recirculation Loops Operating, Loops Operating, andand the the bases bases forfor this this action?

action?

Recirculation Recirculation (1) mismatch is (1) mismatch is exceeded exceeded requiring requiring Recirculation Recirculation Loop Loop AA to be be considered out of considered of service service (2)

(2)

(1) Loop A (1)

A'! Loop Flow Flow (2) to ensure that assumptions of the LOCA analysis are satisfied B. (1) Loop FlowFlow (2) due to the inability to detect significant degradation in jet pump performance C. (1) Pump Speed (2) to ensure that assumptions of the LOCA analysis are satisfied D. (1) Pump Speed (2) due to the inability to detect significant degradation in jet pump performance

Feedback Feedback K/A: SG2.02.22 KIA: SG2.02.22 Equipment Control Equipment Control Knowledge of Knowledge limiting conditions of limiting conditions for for operations operations andand safety safety limits.

limits.

(CFR: 41.5 / 43.2 (CFR: 41.5 143.2 1 45.2) I 45.2)

RO/SRO Rating:

RO/SRO 4.0/4.7 Rating: 4.0/4.7 CLSLP002*34 Objective: CLS-LP-002*34 Objective:

27. Explain

27. Explain why there isis aa limit why there limit for mismatch between for mismatch between total total Jet Jet Pump Pump LoopLoop flows flows

34. Given

34. Given plant plant conditions conditions and and Technical Technical Specifications, Specifications, including including the the Bases, Bases, TRM, TRM, ODCM, ODCM, and and COLR COLR determine the determine the required required action(s) action(s) to to be taken in be taken in accordance accordance with with Technical Technical Specifications Specifications associated associated with with the Reactor the Reactor Recirculation Recirculation System.

System. (SRO/STA (SROISTA only)only)

Reference:

Unit 11 Technical Unit Technical Specification Specification 3.4.13.4.1 andand BASES BASES Cog Level:

Cog Level: High High Explanation:

Explanation:

Two recirculation recirculation loops are are normally normally required required to be operation with their flows matched within the limits be in operation specified inin SR 3.4.1.1 to ensure that that during a LOCA LOCA caused by break of the piping of one by a break recirculation loop the assumptions of the LOCA LOCA analysis are satisfied.

Jet pump loop flow mismatch should be maintained within the following limits:

- jet pump loop flows within 10% (maximum indicated difference 7.5 x1 xl 006 Ibs/hr) lbslhr) with total core flow less than 58 X102 lbs/hr x10 Ibs/hr

- jet pump loop flows within 5% (maximum indicated difference 3.5 x10 xl 066 Ibs/hr) lbslhr) with total core flow greater than or equal to 58 x10 xl 066 Ibs/hr lbs/hr Distractor Analysis:

Choice A: Correct Answer Choice B: Plausible because Loop flow mismatch is correct and vibrations would be a result of low or reverse flow.

Choice C: Plausible because Pump Speed used to be the indication utilized and LOCA analysis is correct.

Choice D: Plausible because Pump Speed used to be the indication utilized and vibrations would be a result of low or reverse flow.

SRO Only Basis:

Basis: Application of Required Actions and Knowledge of TS Bases.

Notes

Recirculation Loops Recirculation Loops Operating Operating 3.4."1 3.4.1 3.4 3.4 REACTOR COOL.A.NT RE.A.CTOR COOLANT SYSTEM SYSTEM (RCS)(RCS) 3.4.1 3.4 ..1 Recirculation Loops Recirculation Loops Operating Operating LCD 3.4.1 LCO 3.4.1 Two recirculation loops Two recirculation loops with with matched matched lIows flows shall shall be be inin operation.

operaon, OR One recirculalion loop One recirculation may be oop may be in operation provided in operation provided thethe following following limits limits are applied when applied when thethe associated associated LCO LCD is applicable:

is applicable:

a.

a. LCO 3.2.1, LCO 3.2.1. "AVERAGE AVERAGE PLANAR PLANAR LINEAR LINEAR HEAT HEAT GENERATION GENERATION (APLHGR), single loop operation limits specified in the COLR; RATE (APLHGR),

b. MINIMUM CRITICAL POWER RATIO {MCPR},

LCO 3.2.2, "MINIMUM (MCPR), single loop operation limits specified in the COLR;COLR; c.

c. 3.2.3. "LINEAR LCO 3.2.3, LINEAR HEAT HEAT GENERATION GENERATION RATE RATE (LHGR),"

(LHGR) single specified in the COLR; loop operation limits specified COLR: and

d. Reactor Protection System (RPS)

LCO 3.3.1.1, "Reactor RPS) Instrumentation, Insfrumentation, Function 2.b (Average Power Range Monitors Simulated Thermal PowerHigh), Allowable Value Power-High), Value of Table Tabe 3.3.

3.3.1.1-1

'1.1-1 is resetror reset for single loop operation.

APPLICABILITY: MODES 1 I and 2.

ACTIONS COMPLETION CONDITION REQUIRED ACTION TIME A. Requirements of the LCO LCD A.1 Satist the requirements of SatisPj 6 hours6.944444e-5 days 0.00167 hours 9.920635e-6 weeks 2.283e-6 months not met, nolmet the LCO.

LCD.

( continued)

(continued)

Brunswick Unit Siunswick Unit 11 3.4-1 3.4-1 Amendment Amendment No. 246 No. 246 I

Recirculation Loops Recirculation Loops Operating Operating 3A:l 3.4.1 ACTIONS (continued)

ACTIONS continuedi COMPLETION COMPLETION CONDITION CONDITION REQUIRED .A.CTlON REQUIRED ACTION TIME TIME 5.

B. Required Action Required Action and and 8.1 B.1 Be n MODE Se in MODE 3.3. 12 12 hours1.388889e-4 days 0.00333 hours 1.984127e-5 weeks 4.566e-6 months hours associated Completion associated Completion Time Time of Condition of Condition .A.

A not met.

not met.

OR No recirculation No recirculation loops loops in in operation.

operation.

SURVEILLANCE_REQU SURVEILLANCE IREMENTS REQUIREMENTS SURVEILLANCE SUR\lEILLANCE FREQUENCY FREQUENCY SR 3.4:1.1 SR 34.i.1 ------------------NOTE----

NOTE--------------------------

Not required to be performed until 24 hours2.777778e-4 days 0.00667 hours 3.968254e-5 weeks 9.132e-6 months after both recrcuation lOops recirculation toops are in operation.

Verify recirculation loop jet pump flow mismatch with 24 hours2.777778e-4 days 0.00667 hours 3.968254e-5 weeks 9.132e-6 months both recirculation loops in operation:

a. 5: -t10%

0% of rated core flow when operating at

<: 75% of rated core flow;

< flow; and

b. 5: 5% of rated core flow flO'A' when operating at

75% of rated core flow.

Snjnswjck Brunswick Unit Unit 11 3.4-2 3.4-2 Amendment Amendment No.No. 244 244

Recirculation Loops Recirculation Loops Operating Operating BB 3.4.1 3.4.1 BASES B.A.SES APPLICABLE APPLICABLE For AREVA For AREVA fuel, ftiel, the the COLR COLR presents presents single single loop loop operation operation APLHGR APLHGR limits limits SAFETY AN.A.L SAFETY ANALYSES YSES in the in the form form of of aa multiplier multiplier that that is is applied applied to to the the two two loop loop operation operation (continued)

(continued} APLHGR limits.

APLHGR limits.

The transient The transient analyses analyses of of Chapter Chapter 15 15 of of the the UFSAR UFSAR have have also also been been evaluated for evaluated for single recirculation loop single recirculation loop operation.

operation. The The evaluation evaluation concludes that concludes that results results ofof the the transient transient analyses analyses are are not not significantly significantly affected by affected by the the single recirculation loop single recirculation loop operation.

operation. There There is, however, an is, however, an impact on impact on the fuel cladding the fuel cladding integrity integritj SL SL since since some some of of the the uncertainties uncertainties forfor the parameters the parameters used used inin the the critical critical power power determination determination are are higher higher in in single loop single loop operation.

operation. The The net result is net result is an an increase increase in in the the MCPR MCPR operating limit.

operating limit.

During single During recircutation loop single recirculation loop operation, operation, modification to to the Reactor the Reactor Pratection System Protection System (RPS)(RPS) average average power power range range monitor monitor (APRM)

APRM)

Simulated Themlal Simulated Themial Power-High PowerHigh Allowable 'Value Value isis required required to to account for the different analyzed limits bet'A'een between two-recirculation two-recirculation drive tlow now loop loop operation and operation with only one loop. loop. The APRM channel subtracts W value from the measured the l!.W recirculation drive flow to effective~1 measured recirculation effectivej the limits shift the limits and uses the the adjusted recirculation drive flow value to determine the APRM Simulated Themlal Themial Power-High PowerHigh Function trip setpoint.

Recirculation loops operating satisfies Criterion 2 of Recirculation S0.36(cX2)(ii) (Ref. 4}.

10 CFR 50.36(c)(2)(ii) 4).

LCO lCO Two recirculation loops are normally required to be in operation with their flows matched within the limits specified in SR 3.4.1.1 3.4.1,1 to ensure !hat that dunng a LOCA caused by a break of the piping of one recirculation loop during the assumptions of the LOCA analysis are satisfied. Alternately, with only

!he one recirculation loop in operation, modifications to the required APLHGR LCO 3.2.1, "AVERAGE limits (LCO AVERAGE PLANAR LINEAR L1NEAR HEAT GENERATION (APLHGRy), MCPR limits (LCO RATE (APLHGR)"), 3.2.2. "MINIMUM (LeO 3.2.2, MINIMUM CRITICAL POWER RATIO MCPRfl, POWER (MCPR)"), LHGR limits (LCO 3.2.3, LINEAR "LINEAR HEAT HEAT GENERATION RATE (LHGR)), (LHGR)"), and APRM Simulated Simulated Thermal Power Power-High Allowable Value (LCO 3.3.1.1), as applicable, must be applied to allo\'/ continued operation. The COLR defines adjustments or allow modifications modifications required for the APLHGR, APLHGR, MCPR, MCPR, and LHGR LHGR limits limits for the current operating cycle.

current (continued) continued)

Brunswick Brunswick Unit Unit 11 6B 3.4.1-3 3.4.1-3 Revision No.58 Revision No. 58

Recirculation Loops Recirculation Loops Operating Operating B 3.4.1 3.4.1 BASES BASES APPLICABILITY APPLICABILITY InIn MODES MODES 1I and and 2,2, requirements requirements ior for operation operation ofthe of the Reactor Reactor Coolant Coolant Recirculation System Recirculation System are are necessary' necessary since since there there isis considerable considerable energy energy in in the reactor core the reactor and the core and the limiting liniiting design design basis basis transients transients andand accidents accidents are are assumed to assumed to occur.

occur.

InIn MODES MODES 3, 3,4, and 5, 4, and 5, the the consequences consequences of of an accident are an accident are reduced reduced and and the coastdown the coastdown characteristics characteristics of of the recircuation loops the recirculation loops areare not not important.

important.

ACTIONS ACTIONS Al With the requirements With the requirements of of the LCO not the LCO met, the not met, recirculation loops the recirculation loops mustmust be be restored to restored to operation operation with with matched flows within matched flows within 66 hours7.638889e-4 days 0.0183 hours 1.09127e-4 weeks 2.5113e-5 months .

hours. AA recirculation recirculation loop is loop is considered considered not not inin operation operation when the pump when the pump in in that that loop loop is idle or is idle or when the when the mismatch mismatch between between total total jet pump flows of pump flows of the the two two loops loops is is greater than greater than thethe required required limits.

limits. The The loop loop with the lower lower flow must must be be considered not considered not in operation. Should in operation. Should aa LOCA LOCA occur occur with with one one recirculation recirculation loop no!

loop not inin operation, operation, the the core core flow flow coastdown coastdown and resultant core and resultant core response may not not be bounded bounded by the LOCA analyses. Therefore, only a limited time limited time is allowed allowed to restore the to restore the inoperable loop loop toto operating operating status.

status.

Alternatively, ifif the single Altematively, single loop requirements of the lCO loop requirements LCO are applied applied to operating limits and RPS setpoints, as applicable, operation with only one recirculation loop would satisfl/ satisfy the requirements of the LCO and the initial conditions of the accident sequence.

The 6 hour6.944444e-5 days 0.00167 hours 9.920635e-6 weeks 2.283e-6 months Completion Time is bae.ed based on the the low probability of an accident occurring during this time period, on a reae.onable reasonable time to complete the Required Action (I.e., (i.e.. reset the applicable limits or setpoints for single recirculation loop operation), and on frequent core care monitoring by operators allowing abrupt changes in core flow conditions to be quickiy quickly detected.

This Required Action does not require tripping the recirculation pump in the lowest flow loop when the mismatch between the total jet pump flaws flows of the two loops is greater than the required limits. However, in cases where large flow mismatches occur, low flow or reverse flow can occur in the low flaw flow loop jet pumps.

pumps, causing vibration of the jet pumps. If causingllibralion If zero or reverse flawflow is detected, the condition condition should should be alleviated by changing pump speeds to re-establish forward flow.

pump I(continued) continued)

Brunswick Brunswick Unit Unit 11 BB 3.4.1-4 3.4.1-4 Revision No.

Revision No. SB58

Recirculation Loops Recirculation Loops Operating Operating BB 3.4:1 3.4.1 BASES BASES ACTIONS ACTIONS .!U.

(continued)

With no With no recirculation recirculation loopsloops in in operation operation or or the the Required Required Action Action and and associated Completion associated Completion Time Time of oi Condition Condition AA not not met, met, the the plant plant must must be be brought to brought to aa MODE MODE in in which which the the LCO LCO does does not not apply_

apply. ToTo achieve achieve this this status, th.e status, the plant pant must must be be brought brought to to MODE MODE 33 within within 12 hour1.388889e-4 days 0.00333 hours 1.984127e-5 weeks 4.566e-6 months s_

hours. In In this this condition, the condition, the recirculation recirculation loopsloops areare not not required required to to be be operating operating because because of the of the reduced reduced severity severity of of DBAs DBAs and and minimal minimal dependence dependence on on the the recirculation loop recirculation loop coastdown coastdown characteristics.

characteristics. The The allowed allowed Completion Completion Time of Time of 12 12 hours1.388889e-4 days 0.00333 hours 1.984127e-5 weeks 4.566e-6 months hours is is reasonable, reasonable, based based on on operating operating experience, experience, to to reach reach MODE 33 from MODE from full full power power conditions conditions in in an orderly manner an orderly manner and and *...without

'ithout challenging piant challenging plant systems.

systems.

SURVEILLANCE SURVEILLANCE SR 3.4.U SR 3.4.1.1 REQUIREMENTS REQUIREMENTS This SRSR ensures the recirculation loops loops are within the allowable limits for mismatch. At low mismatch_ low core t/ow flow (Le.,

(i.e.. -<< 75%

76% of rated rated core flow), the MCPR requirements provide larger larger margins to the fuel cladding integrity Safety Limit such that the potential Limit potential adverse effect effect ofof early boiling transition during a LOCA is reduced_reduced. A larger flow mismatch can, therefore, be allowed when core tlow aI/owed flow is < 75% of rated core flow. The recirculation loop jet pump flow, as used in this Surveillance, is the summation of the flows from all of the jet pumps associated with a single recirculation loop_ loop.

The mismatch is measured in terms of the percent of rated core flow. If the flow mismatch exceeds the specified limits, the loop wiltl with the lower flow is considered not in operation. The SR is not required when both loops are not in operation since the mismatch limits are meaningless during single looploop or natural circulation operation. The Surveillance must be performed within 24 hours2.777778e-4 days 0.00667 hours 3.968254e-5 weeks 9.132e-6 months after both loops are in operation. operation_ The 24 hour2.777778e-4 days 0.00667 hours 3.968254e-5 weeks 9.132e-6 months frequency Frequency is consistent with the Surveillance Frequency for jet pump OPERABILITY OPERA.BILITY verification and has been shown by operating experience to be adequate to to detect off normal jet jet pump pump loop loop flows in a timely manner.

manner.

REFERENCES REFERENCES 1.

1. UFSAR.

UFSAR, Section Section 5.4.1.3.

SA 1.3.

2.

2. UFSAR, UFSAR, Chapter Chapter 15. 15.

3.

3. NEDC-31776P, NEDC-31776P, Brunswick Brunswick Steam Steam Electric Electric Plant Units 1 Plan! Units I andand 22 Single Single Loop Loop Operation, Operation, February February 1990. 1990.

4.

4. 10 10 CFR CFR E0.36(c)t2)(ii).

SD_36(c)(2)(ii)_

Brunswick Unit 11 Brunswick Unit B63A1-5 3.4.1-6 Revision No.58 Revision No. 58 Categories Categories KJA:

KIA: SG2.02.22 SG2.02.22 Tier Tier // Group:

Group: T3 T3 RO Rating:

RORating: 4.0 4_0 SRO Rating:

SRORating: 4.7 4.7 LP LP Obj:

Obj: CLSLP0O2*34 CLS-LP-002*34 Source:

Source: NEW NEW Cog Cog Level:

Level: HIGH mGH Category Category 8:8: YY

98. Which

98. Which one,Qf the following oneQf the following identifies identifies the the procedure procedure required required to to control control drywell drywell pressure pressure within PCPL-A lAW PCCP withinPCPL-A:IAW PCCP and the release and the release rate rate restrictions, restrictions, ifif any, any, inin effect effect during during the the venting?

venting?

A A~ SEP-01 Section SEP-01 Section 11 'j Venting Venting Primary Primary Containment Containment irrespective irrespective of of Off Off Site Site Release Release rate rate B. SEP-01, B. SEP-01, Section Section 2.,.

2 Venting Venting Primary Primary Containment Containment via via the the Suppression Suppression Chamber Chamber within Site Release Rate within Site Release Rate Limit Limit C. SEP-01, Section C. SEP-01, Section 3,3, Venting Venting Primary Primary Containment Containment via via the the Drywell Drywell within within Site Site Release Rate Release Rate LimitLimit D. OEDMG-003, D. OLDMG-003, Containment Containment Venting Under Under Conditions Conditions of of Extreme Extreme DamageDamage irrespective of Off Site Release irrespective of Off Site Release RatesRates

Feedback Feedback SG2.03.11 K/A: SG2.03.11 KIA:

Radiation Control Radiation Control Ability to Ability to control control radiation radiation releases.

releases.

(CFR: 41.11 /43.4/45.10)

(CFR: 41.11/43.4/45.10)

ROISRO Rating:

RO/SRO Rating: 3.8/4.3 3.8/4.3 Objective: CLSLP300L*08d Objective: CLS-LP-300-L *08d

8. Given the

8. Given the Primary Primary Containment Containment Control Control Procedure Procedure and and plant plant conditions, conditions, determine determine ifif thethe following following actions are actions are required:

required:

c. Venting

c. Venting thethe primary primary containment containment while while staying staying within within radioactivity radioactivity release release rate rate limits limits

d. Venting

d. Venting thethe primary primary containment containment IRRESPECTIVE IRRESPECTIVE of radioactivity release of radioactivity release rate rate limits limits

Reference:

001-37.8, Revision 001-37.8, Revision 4,4, Page Page 33, 33, Step Step PC/P-18 PC/P-18 Cog Level:

Cog Level: High High Explanation:

Explanation:

Action to Action to vent the the primary primary containment is is taken before before drywell pressure pressure rises rises to Primary Containment Containment Pressure Limit Pressure Limit AA to assure that the integrity of the primary containment is maintained maintained and to prevent core damage that might be caused by the inability to vent the reactor, as necessary, to permit injection of damage of water cool the core. Venting of the primary containment is performed irrespective of the off-site radioactivity to cool release rate that will occur, and defeating isolation interlocks if necessary, because the consequences release consequences of not doing so may be either severe core damage or loss of primary containment integrity and uncontrolled not radioactive release much greater than might otherwise occur. Note that primary containment venting is performed performed only, as necessary, to restore and then maintain pressure below the limit.

Distractor Analysis:

Distractor Choice A: Correct Answer.

Choice Plausible because within ODCM limits is utilized during SEP-01 section 11 when venting the Choice B: Plausible torus due to containment Hydrogen/Oxygen concentration concerns and before exceeding PCPL-A PCPL-A Choice Choice C: Plausible because within ODCM limits is utilized during SEP-01 section 1 1 when venting the drywell due to containment Hydrogen/Oxygen concentration concerns and before exceeding PCPL-A.

Choice D: Plausible becausebecause irrespective is correct and after exceeding PCPL-A is is wrong SRO OnlyOnly Basis: Detailed knowledge of of diagnostic steps and decision decision points in the EOPs that involve transitions transitions to emergency emergency contingency procedures.

Notes Notes PRIMARY PRIMARY CONTAINMENT CONTAINMENT PRESSUREPRESSURE LIMIT-ALlMIT-A The lesser The lesser of the pressure of the pressure capability capability of of the primary containment, the primary containment, pressure pressure atat which containment which containment vent vent valves valves sized sized toto reject reject all all decay decay heat heat from the containment from the containment can be can be opened opened and and closed, or pressure closed, or pressure at at which which SRVs SRVs can be opened can be opened and and will will remain remain open open (Figure (Figure 2).

2).

DE0P-01-UG IOEOP-01-UG Rev.

Rev. 5555 Page 71 Page of 151 71 of 151 I

ATTACHMENT 5 G)CDQ Page 17 of 27 4

.mgm 01 i k) N) Z D C -.x FIGURE 2 Primary Containment Pressure Limit-A

= -H I 100 V I

1'10 H

DRYWELL PRESSURE (PSIG)

( !)

80 (j)

Il.

III .11 70 W

a:

I 60 0)

CJ)

U) w 50 ai t:t: iii

a. 40

...I

..J ii.

W 30 W

~

il2 C

20 N)

Hi H

10 0

10 0 10 20 30 40 50 GO 70 8Q PRIMARY CONTAINMENT WATER LEVEL (FEET) -4 ma I

nl m IF USING THE FOLLOWING INSTRUMENT:

-rT C

Cl) z H C) 0

-1 C m PCPl-A IS:

-o o

-w r CAC-PI-1230 70 PSIG

)OO C.,

-tJ-U-o cci CAC-PI-4176 .-

0 Cl)cl)

USE THE GRAPH

t;x mmo CAC-PR-1257-1 USE THE GR.f\PH HH IOEOP-01-UG m

0

-o C) Rev. 55 C,,

0, Page 77 of 151

-v 1 -b 01

-s I

PCPL-A PCPL-A j

DRYWELL PRESS rmVWllll. PHESS R~AClfeS PCPL.A PCPL*A VENT THE VENT ThE PRIMARY PRMARV elMT CTT RRSPCTIV& OF IRRESPECTIVE OF OFFSITE OFcsIrE NZUASE RAtE "IELEASe RAXF PER

$ETtON 1I OF SECTION OF 1001"*01* SEP. 01 pcjp. i "CiP*HI STEP BASES:

Action to vent the primary containment containment isis taken before drywell pressure pressure rises rises to Primary Primary Containment Pressure Containment Pressure Limit Limit A,defined A, defined toto be be tile the lesser lesser of either:

a. The pressure capability of the containment, or

b. The maximum containment pressure at whicll which vent valves sized to reject all decay heat from tile the containment can be opened and closed, or c.

C. The maximum containment pressure at which SRVs can be opened and will remain open, or

d. The maximum containment pressure at which reactor vent valves can be opened and closed.

This action is taken to assure that the integrity of the primary containment is maintained and to prevent core damage that might be caused by tile the inability to vent the reactor, as necessary, to permit injection of water to cool tile the core.

Lo vent "before The directions to before drywell pressure reaches pePl-A" PCPL-A allows, but does not require, venting at significantly lower pressures.

pressures Early or extended venting can permit primary containment pressure reductions before significant fuel damage occurs, occllrs, thereby increasing the capacity of the containment to retain lission fission products and reducing the radioactivity released to the environment.

environment, If the primary containment has failed, venting may also reduce the offsile offsite dose by directing fission prodllcts products throllgh through an elevated release point.

001-37.8 1001-37.8 Rev. 4 I 33 of 58 Page 33 581

STEP PC/P*18 STEP PCIP.18 (continued)

(continued)

Venting of venting of the the primary primary containment containment isis petiormed performed irrespective irrespective of of the the off-site off-site radioactivity radioactivity release rate that release rate that will occur, and will occur, defeating isolation and defeating isolation interlocks interlocks ifif necessary, necessary, because because the the consequences of consequences of not not doing doing so so may may bebe either either severe severe core core damage damage or or loss loss of primary of primary containment integrity containment integrity and and uncontrolled uncontrolled radioactive radioactive release release much much greater greater than than might might otherwise occur.

otherwise occur. NoteNote that that primary primary containment containment venting venting isis petiormed performed only,only, as as necessary, to necessary, to restore restore andand then then maintain maintain pressure pressure below below the the limit limiL Primary containment Primary containment venting venting is performed using is performed using Primary Primary Containment Containment Venting, Venting, EDP-O1 -SEP-01.

EOP-01-SEP-0'I.

001-37.8 1001-37.8 Rev. 44 Rev. Pane 34 Page 34 of 581 of 58

PRIMARY CONTAINMENT PRIMARY CONTAINMENT VENTINGVENTING 1.0 1.0 ENTRY CONDITIONS ENTRY CONDITIONS

- dhrected by As directed by the PC/P PC/P section section ofof Primary Primary Containment Contaniment Control Control Procedure, Procedure, EOP-O2-PCCP EOP-02-PCCP OR OR

- As directed directed by by the the PC/H section of of Primary Containment Control Procedure, EOP-02-PCCP 2.0 OPERATOR OPERA TOR ACTIONS CO: 2.1 IF while executing this procedure, IF procedure, it is recognized the actions 0 can NOT be performed, OR will NOT be effective, THEN GO TO Containment Venting Under Conditions of Extreme Damage, OEDMG-003, if dlrected directed by the Unit SCO.

CO: 2.2 IF venting for pressure control, THEN PERFORM Section oJ, 1, 0 on page 3.

CO: 2.3 IF venting for H2/02 control control,. THEN PERFORM section of 0 procedure directed by SCO.

IIOEOP-ool-SEP-O'l OEOP-Oi-SEP-Oi Rev. 24 Rev. 24 Page 22 of Page of 22 22 1 Categories Categories KJA:

KIA: SG2.03.I SG2.03.111 Tier Tier // Group:

Group: T3 T3 RU RO Rating:

Rating: 3.8 3.8 SRO Rating:

SRU Rating: 4.3 4.3 LP Ubj:

LP Obj: CLSLP3OOL*O8D CLS-LP-300-L*08D Source:

Source: ~vv NEW Cog Cog Level:

Level: HIGH HIGH Category 8:

Category 8:

99. During non-ATWS

99. During non-ATWS emergency emergency conditions conditions on on Unit Unit Two, Two, Emergency Emergency Depressurization Depressurization isis required with required with reactor reactor pressure pressure at at 1100 1100 psig.

psig.

Which one Which of the one of the following following identifies identifies the the bases bases for for the the Minimum Minimum Number Number of of SRVs SRV5 Required for Required for Emergency Emergency Depressurization Depressurization and and the the required required procedure procedure utilized utilized ifif this this number of number SRVs open of SRVs open cannot cannot be be achieved?

achieved?

The Minimum The Minimum Number Number of of SRVs SRVs Required Required forfor Emergency Emergency Depressurization Depressurization is is based based on on the low pressure ECCS system with the the low pressure ECCS system with the lowest head lowest head being being capable capable of of making making upup the the SRV steam SRV steam flow flow at at the the Minimum Minimum (1) (1)

(2)

(2) Procedure is Procedure is required required ifif the the minimum minimum number number of of SRVs SRVs cannot cannot be be opened.

opened.

(1) Reactor A. (1)

A. Reactor Flooding Flooding Pressure Pressure (2) Primary Containment (2) Primary Containment Flooding Flooding B. (1) Reactor Flooding Pressure (2) Alternate Emergency (2) Emergency Depressurization Depressurization C. (1) Alternate Reactor Flooding Pressure (2) Primary Containment Flooding D (1) Alternate Reactor Flooding Pressure D:'

(2) Alternate Emergency Depressurization Depressurization

Feedback Feedback KJA: SG2.04.17 KJA: SG2.04.17 Emergency Procedures Emergency Procedures II PlanPlan Knowledge of Knowledge of EOP EOP terms terms andand definitions.

definitions.

(CFR: 41.10 /45.13)

(CFR: 41.10/45.13)

ROISRO Rating:

RO/SRO Rating: 3.9/4.3 3.9/4.3 CLSLP300H*002 Objective: CLS-LP-300-H*002 Objective:

2. Given

2. Given plant plant conditions conditions and and the the Emergency Emergency Operating Operating Procedures, Procedures, determine determine ifif execution execution of of the the Alternate Emergency Alternate Emergency Depressurization Depressurization Procedure Procedure isis required.

required.

Reference:

OEOP-01-UG, Revision 55, OEOP-01-UG, Revision 55, Page Page 70, 70, Attachment Attachment 55 (Definitions)

(Definitions)

RVCP RVCP Cog Level:

Cog High Level: High Explanation:

Explanation:

The Minimum Number of Minimum Number of SRVs SRVs Required Required forfor Emergency Emergency Depressurization Depressurization (5)

(5) is is defined defined toto be be the the least least number of number SRVs which of SRVs which correspond correspond to to aa Minimum Minimum Alternate Alternate Reactor Reactor Flooding Flooding Pressure Pressure sufficiently low low that the ECCS with the lowest head will be capable of making up the SRV steam flow at the corresponding corresponding Minimum Alternate Reactor Flooding Pressure. If the number of SRVs specified cannot be opened, the be depressurized by other means. A list of alternate systems that can be used reactor must be used for depressurizing the reactor is included in the Alternate Emergency Depressurization depressurizing Depressurization Procedure, EOP-01 -AEDP.

EOP-01-AEDP.

Distractor Analysis:

Choice A: Plausible because Minimum Reactor Flooding Pressure is easily confused with Minimum Alternate Reactor Flooding Pressure and Primary Containment Flooding requires exiting all EOPs which is wrong for the given conditions.

Choice B: Plausible because Minimum Reactor Flooding Pressure is easily confused with Minimum Alternate Reactor Flooding Pressure and AEDP is correct.

Choice C: Plausible because Minimum Alternate Reactor Flooding Pressure is correct and Primary Containment Flooding requires exiting all EOPs which is wrong for the given conditions.

Choice D: Correct Answer SRO Only Basis: Detailed knowledge of diagnosticdiagnostic steps and decision decision points in in the EOPs that involve involve transitions to emergency contingency procedures.

Notes Notes MINIMUM MINIMUM ALTERNATE ALTERNATE FLOODING FLOODING PRESSURE PRESSURE The lowest reactor The lowest reactor pressure pressure atat which steam 110w which steam through open flow through open SRVs SRVs is is sufficient sufficient to to preclude preclude anyany clad clad temperature temperature from from exceeding exceeding 1500F 1500°F even even ifif the the reactor reactor core core is is not not completely completely covered covered IOEOP-Oi-UG OEOP-O*!-UG Rev.

Rev. 55 55 Page 69 Page 69 of of *15"1 151 I

ATTACHMENT ATTACHMENT 55 Page Page 1010 ofof 2727 Definitions Definitions MINI MUM CORE MINIMUM CORE FLOODING FLOODING INTERVAL INTERVAL The greatest The greatest amount amount of of time time required required to to flood flood the the reactor reactor toto the the top top of of the the active active Tuel with reactor pressure fuel with reactor pressure at at the the minimum minimum reactor reactor flooding flooding pressure pressure and and atat least least the minimum the minimum number number of SRVs required of SRVs required for for emergency emergency depressurization depressurization open. open.

MINIMUM INDICATED MINIMUM INDICATED LEVEL LEVEL The highest The highest reactor reactor water level level instrument instrument indication indication which which results results from from off-calibration instrument off-calibration instrument run temperature conditions conditions when reactor water level is is actually at at the elevation of of the instrument instrument variable leg leg tap.

M1NIMUM NUMBER OF SRVS REQUIRED FOR EMERGENCY MINIMUM EMERGENCY DEPRESSURIZATION DEPRESSURIZATION The least number of SRVs which correspond to a minimum altemate alternate reactor flooding pressure sufficiently low that tile the ECCS witll with the 101,vest lowest head will be capable of making up the SRV steam flow at the corresponding corresponding minimum alternate reactor flooding pressure.

MINIMUM REACTOR FLOODING PRESSURE The minimum SRV reopening pressure; 50 psid witl) with 5 5 SRVs open. This pressure is utilized to assure sufficient liquid injection into the reactor to maintain SRVs open and to flood the reactor to the elevation of the main steam lines during the flooding evolution when the reactor is shutdown.

MINIMUM SRV REOPENING PRESSURE The lowest reactor pressure at which an SRV will lufly fully open and remain fully opened when its control switch is placed in the OPEN position.

MINIMUM STEAM MINIMUM STEAM COOLING COOLING REACTOR REACTOR WATER LEVEL The lowest reactor water level at which the covered covered portion of of the reactor core will generate generate sufficient sufficient steam steam to to preclude preclude any any clad clad temperature in in the the uncovered uncovered portion portion of the core of the core from exceeding exceeding 150DF. 1500°F. This This limit limit is is used used during during anan ATWS ATWS event event to to prevent prevent fuel fuel damage damage when when level level isis lowered lowered below below TAFTAF (Unit (Unit 1'I only:

only:

Figure Figure 18;18; Unit Unit 22 only:

only: Figure Figure'18A).18A).

IOEOP-01-UG OEOP-O'l-UG Rev. 55 Rev. 55 Page 70 Page 70 of of 151

'15'1 I

STEPS RC/P*23 STEPS RCIP-23 through through RC/P*25 RC/P-25 PERFORM "ALTERNATE PERFORM ALTERNATE EFi.ERGE.NCY EMERGENCY DEPRESSURI2.ATION DEPRESSURIZATION PROCEDURE (EOP.

PROCEDURE" (EOP. 01.

01. AEDPI AEDP IRRSPECTNE OF IRRESPECTIVE OFFSITE OI OPPSIlE RADIOACTIVITY REll:

RADIOACTIVITY RELEASE 1 RATE

.... SE RAT!:

RCIP-ZS STEP BASES:

The Minimum Number of SRVs Required for Emergency Depressurization Depressurizatiori (5) is defined to be the least number of SRVs which correspond to a Minimum Alternate Reactor Flooding Pressure sufficiently low that the ECCS with the lowest !lead head will be capable of making up the SRV steam flow at the correspondin corresponding g Minimum Alternate Reactor Flooding Pressure.

The Minimum SRV Re-opening Pressure is the lowest reactor pressure at which an an SRV will remain fully open with its control switch in the open position. The accuracy of the re-opening pressure and the indication available to determine detelmine reactor pressure result in conditions such that the SRVs are not always open when the pressure indicated indicated is 50 psig. One hundred hundred psig has has been selected as aa value which can be used used to determine determine the SRV5 SRVs have failed to function. When reactor pressure is below below this value, depressurizat depressurization ion is considered considered complete and reactor pressure reduction need need notnot be be augmented augmented by by use use of of additional additional systems systems even even ifif less less than than the the minimum minimum number number ofof SRVs SRVs are are open.

open. IfIf the number number of of SRVs SRVs specified cannot be be opened, the reactor mustmust be depressurizec i by other be depressurized by other means. A list means. list of of alternate alternate systems systems that that can be be used used for depressurizin depressurizing the reactor is g the is included included in in the the Alternate Emergency Depressuriza Depressurization tion Procedure, Procedure, EOP-O1-AE EOP-01-AEDP. DP. However,However, since since event event independenc independence must be e must be maintained maintained and specific plant and specific plant conditions conditions cannot cannot be be presumed, presumed, no no priority priority regarding regarding system system use use is is indicated. This approach indicated. This approach provides provides an an operator operator the the flexibility flexibility of of being being able able to to use use whatever whatever system(s) system(s) may may bebe most most appropriate appropriate underunder current current plant plant conditions.

conditions.

001-37.4 1001-37.4 Rev. 88 Rev. Page 59 Page of 78 59 of 781

ALTERNATE EMERGENCY ALTERNATE EMERGENCY DEPRESSURIZATION DEPRESSURIZATION PROCEDURE PROCEDURE 1.0 1.0 ENTRY CONDITIONS ENTRY CONDITIONS As directed by As directed by the the RC/P RCIP section section ofof Reactor Reactor Vessel Vessel Control Control Procedure, Procedure, EOP-Ol-RVCP EOP-01-RVCP OR

- As directed As directed byby the the RC/P RC/P section section ofof Level/Power Level/Power Control, Control, EOP-01-LPC EOP-O1-LPC OR OR As directed by As directed SAMG Primary by SAMG Primary Containment Containment Flooding, Flooding, SAMG-O'l SAMG-Ol 2.0 OPERATOR ACTIONS NOTE: Manpower: '1I Control Operator

'1I Auxiliary Operator

'II Independent Verifier equipment:

Special eqUipment: 4 jumpers (32, 33, 34, and 35} 35:

11 flathead Ilathead screwdriver 1I locking screwdriver tape NOTE: Performance of this procedure will affect any main steam line leakage control pathways established by EOP-O'l-SEP-'!'L EOP-Oi -SEP-i t 2.1 2,1 EVACUATE tile the Unit 1 1 and 2 Turbine Buildings using the following actions:

CO: 2.11 2.1:1 SOUND the Unit 1 1 and Unit 22 Turbine Building evacuation alarms AND ANNOUNCE the o

evacuation.

CO: 2.1.2 REQUEST the SCO to notify the TSC that the Turbine Building Building is being evacuated due due to to potential o

high high radiation conditions conditions during during the the alternate alternate emergency clepressuriza depressurization.

tion.

CO:

CO: 2.2 2.2 IF either IF either Unit service service in Unit 11 or or Unit Unit 22 Turbine the once-through in the Turbine Building once-througlliineup, Building ventilation lineup, THEN ventilation is THEN SECURE SECURE that is in that in o units units' turbine turbine building building ventilation ventilation (OP-37.3).

(OP-37.3).

IOEOP-Oi-AE OEOP-O'I-AEDP DP Rev. 18 Rev. '18 Page 22 of Page 161 of 16 Categories Categories K/A:

KIA: SG2.04.17 SG2.04.17 Tier/Group:

Tier / Group: T3 T3 RO RO Rating:

Rating: 3.9 3.9 SRO SRO Rating:

Rating: 4.3 4.3 LP Obj:

LP Obj: CLSLP3OOH *OO2 CLS-LP-300-H*002 Source:

Source: ~VV NEW Cog Cog Level:

Level: HIGH mGH Category Category 8:8: YY

100. An 100. ATWS has An ATWS has occurred occurred on on Unit Unit Two:

Two:

ARI has ARI has been been actuated.

actuated.

No blue lights are No blue lights are lit lit on on the Full Core the Full Core Display.

Display.

Suppression Pool Suppression Pool Temperature Temperature isis 112 112°0 F.

F.

The 2A The 2A SLC SLC pump pump has has aa red red light light indication.

indication.

The 2B The 2B SLC SLC pump pump has has aa green green light indication light indication The SLC The SLC AA Squib Squib Valve Valve Continuity Continuity white white light light isis lit lit The SLC The SLC BB Squib Squib Valve Valve Continuity Continuity white white light light isis extinguished.

extinguished.

Which one Which one ofof the the following following identifies identifies the the procedure procedure that that an an AO AC would would bebe directed directed to to perform based on the above conditions and perform based on the above conditions and the resultant the resultant effect effect of of those those actions?

actions?

A'! Perform LEP-02, A Perform LEP-02, Section Section 22 toto insert insert control control rods rods in in order order to to shutdown shutdown thethe reactor reactor by by venting the Scram Air venting the Scram Air Header. Header.

B. Perform LEP-02, B. Perform LEP-02, Section Section 66 toto insert insert control control rods rods in order to in order to shutdown shutdown thethe reactor reactor by by venting the overpiston area of the control rods.

C. Perform LEP-03, Section 2 to inject inject boron to shutdown the reactor using RCIC.

D. Perform LEP~03, D. LEP03, Section 3 to inject boron to shutdown the reactor using RWCU via the SLC tank.

Feedback Feedback K/A: SG2.04.35 KIA: SG2.04.35 Emergency Procedures Emergency Procedures II PlanPlan Knowledge of Knowledge of local local auxiliary auxiliary operator operator tasks tasks during during an an emergency emergency and and thethe resultant resultant operational operational effects effects.

(CFR: 41.10/43.5/45.13)

(CFR: 41.10 /43.5/45.13)

RO/SRO Rating:

RO/SRO Rating: 3.8/4.0 3.8/4.0 CLSLP300J*005 Objective: CLS-LP-300-J*005 Objective:

5. Given plant

5. Given conditions and plant conditions and the the Local Local Emergency Emergency Procedures, Procedures, determine determine which which sections sections ofof the the Alternate Control Alternate Control Rod Rod Insertion Insertion Procedure Procedure should should be be utilized utilized for for Control Control RodRod Insertion Insertion (EOP-01 -LEP-02).

(EOP-01-LEP-02).

4. Given

4. Given plant plant conditions conditions andand the the Local Local Emergency Emergency Procedures, Procedures, determine determine which which method method of of the the Alternate Alternate Boron Injection Boron Injection is appropriate (EOP-01-LEP-03) is appropriate (EOP-01-LEP-03)

Reference:

OEOP-01 -LEP-02 OEOP-01-LEP-02 Cog Level: High Cog Explanation:

Explanation:

Based on Based on the conditions given, determines that scram valves have not opened (no blue lights on full core core display) and that Boron is injecting with A pump running (red light on) and B squib valve opened (white display) and light extinguished) light extinguished) so so LEP-03 is not required. The pumps discharge into a c9ffi..rnQ!lJl~~der c9rflmoIheader before going to the squib to the squib valves.

valves. Requires assessment of alternate control rod insertion seCtions seOtions and"determines anddetermines venting the scram air header is appropriate.

venting Distractor Analysis:

Distractor Analysis:

Choice A: Correct Answer Choice Choice Choice B: Plausible because venting of the over piston area will insert the control rods but would be the inappropriate decision for rod insertion given the conditins. The operational effect is reactor shutdown with control rod insertion.

Choice Choice C: Plausible because because suppression suppression pool pool temperature is greater than 1100 is greater 110° FF and boron injection injection isis required.

required. With A pump running but the A squib valve not open and not and no B B pump a common misconceptio misconception n is is that SLC flow will not not occur to the Reactor. this would be be correct under under different different conditions conditions in in the the stem. The operational stem. The operational effect effect is is reactor shutdown shutdown with with boron boron injection.

injection.

Choice Choice D: D: Plausible Plausible because because suppression suppression pool pool temperature temperature is is greater than 1100 greater than 110° FF and and boron boron injection injection isis required.

required. With With A pump pump running but the running but the A squib squib valve valve not not open open and and no no BB pump pump aa common common misconceptio misconception is that n is that SLC SLC flow flow will will not not occur occur to to the the Reactor.

Reactor. this this would would bebe correct correct under under different different conditions conditions in in the the stem. The operational stem. The operational effect effect is is reactor reactor shutdown shutdown with with boron boron injection.

injection.

SRO SRO OnlyOnly Basis:

Basis: Assessing Assessing plant plant conditions conditions and and prescribing prescribing aa section section ofof aa procedure procedure with with which which to to proceed.

proceed.

Notes Notes

21 2.7 INSERTcontrol INSERT controlrods rodsbybyoneoneorormore moreofofthe thefollowing following metho ds:

methods:

2.71 2.7.1 DE-ENERGIZEthe DE-ENERGIZE VENT thescram thescram scramair scrampilot airheader, pilotvalve headerSection valvesolenoids Section22on solenoidsAND AND o VENT the Page9.9 on Page 2.7.2 2.7.2 RESETRPS RESET 33on on Page RPSAND Page '14.

AND INITIATE 14.

INITIATEaamanual manual scram,scram. Section Section o 2.7.3 2.7.3 SCRAM individual SCRAM Sectio n 4 indMdual rods on Page Section 4 on Page H.

rods with 17.

with the the scram scram testtest switches, switches, o 2.7.4 2.7.4 INSERT control INSERT System control rods System,, Section rods with Section 55 on with the on Page the Reactor Page 21.

21.

Reactor ManualManual Control Control oLI 2.7.5 2.7.5 VENT the VENT on Page on the over Page 22.

over piston 22.

piston area area of of control control rods, rods, Section Section 66 o IOEOP-01-LEP-02 OEOP-O1-LEP-02 Rev. 26 Rev. 26 Page 3 of Page of 29 I 2.2 2.2 INJECT boron INJECT boron with one or more of the following following systems:

systems:

ENOTE:

NOTE: System (s) should System(s) should be selecte selected d in order listed and based upon system I availab ility and access availability and ibility.

accessibility.

CO:

CO: - CRD, CRD, Sectio Section n 11 on on page page 33 o LI NOTE NOTE:: HPCI/ RCIC should HPCIIRCIC should bebe used used only only ifjf suction suction is is from from the the CST.

CST.

CO:

CO: - HPCl/ RCICS HPCIIRCIC, ection2 Section onpage 2 on page14 14 o LI CO:

CO: - RWCU RWCU via via SLC SLC tank, tank, Section Section 33 on on page page 21 21 o LI CO:

CO: - RWCU RWCU with with borax, borax, Section Section 44 on on page page31 31 o LI IOEOR -Di-LEP-03 OEOP-01-LEP-03 Rev.

Rev. 27 27 Page22of Page 4-1 of41 I

f002A SQUIB F004A PUMP COO1A F002B PUMP

~026 C001B SQUIB F004B Categories Categories K/A:

KIA: SG2.04.35 SG2.04.35 Tier / Group: T3 RORating:

RORating: 3.8 3.8 SRO Rating: 4.0 SRORating:

LP Obj:

LPObj: CLSLP3OOJ *OO5 CLS-LP-300-J*005 Source: NEW Cog Cog Level:

Level: NIGH HIGH Category 8:

ML101590078

Text

Reference:

Reference:

Reference:

Reference:

Reference:

Reference:

Reference:

Reference:

Reference:

4.7 LPObj

Reference:

Reference:

3.2 Rating

Reference:

Reference:

Reference:

Reference:

Reference:

Reference:

Reference:

Reference:

Reference:

3.1 Level

Reference:

Reference:

1=1

Reference:

Reference:

Reference:

Reference:

Reference:

Reference:

Reference:

Reference:

Reference:

Reference:

Reference:

Reference:

Reference:

Reference:

Reference:

Reference:

Reference:

Reference:

Reference:

Reference:

Reference:

Reference:

Reference:

Reference:

Reference:

Reference:

Navigation menu

Search