Text

02-2005-Final Operating Exam
	ML060940698
Person / Time
Site:	Diablo Canyon
Issue date:	02/07/2005
From:	; - No Known Affiliation
To:	; Office of Nuclear Reactor Regulation
References
	50-275/05-301, 50-323/05-301
	Download: ML060940698 (423)
	v • d • e

Appendix D Scenario Outline Form ES-D-1 Facility: _DCPP Units 1 & 2___ Scenario No.: __NRCSIM-01__ Op-Test No.: _01_

Examiners: ____________________________ Operators: _____________________________

Initial Conditions:

100% power, equilibrium Xe, 590 ppm, MOL (IC-510). DG 1-1 is tagged out for repairs on starting circuit, OOS for 5 hours5.787037e-5 days 0.00139 hours 8.267196e-6 weeks 1.9025e-6 months , due back in 12 hours1.388889e-4 days 0.00333 hours 1.984127e-5 weeks 4.566e-6 months (Drill File 34)(Place yellow tags on CB). PRA YELLOW. PDP in service (Drill File 90). Diluting 20 gal/hr. STP I-1C was completed one hour ago, due in seven. Last run two days ago.

Turnover:

Swap from PDP to CCP 1-2.

Event Malf. Event Event No. No. Type* Description 1 Drill File N, ALL Secure PDP and place CCP 1-2 in service 6020 2 Xmt I, ALL LT 459 Failure low pzr40 3 N, ALL Restore Letdown 4 Xmt I, RO Loop 4 TC failure rcs138 5 R, ALL CALL - EPOS requests ramp to 900 MW within 30 min. Start ramp in 10 min. (Call 12 minutes before ASW pump trip) 6 Pmp C, BOP Loss of ASW pumps asw1 Pmp asw2 7 Cnv C, RO SG 1-2 FRV failure mfw5 8 Mal M, ALL SG 1-2 safety valve sticks open mss6c 9 Mal M, ALL SGTR from SG 1-2 rcs4c 10 Pmp C, ALL Failure of RHR Pumps 1-1 and 1-2 to AUTO start rhr1/2

(N)ormal, (R)eactivity, (I)nstrument, (C)omponent, (M)ajor

2 drl_file 6020

Appendix D Required Operator Actions Form ES-D-2 Op-Test No.: _01__ Scenario No.: _01__ Event No.: _01__ Page _1_ of _9_

Event

Description:

____Secure PDP and place CCP 1-2 in service.___________________________

Time Position Applicants Actions or Behavior SFM Tailboard the event using OP B-1A:V RO Places FCV-128 in manual to control charging flow.

BOP Verify CCP recirc valves open BOP Start CCP 1-2 RO Reduce PDP speed while opening FCV-128 BOP Secure PDP RO Control charging flow to maintain PZR level and place FCV-128 in Auto RO Adjust seal injection as necessary with HCV-142

Appendix D Required Operator Actions Form ES-D-2 Op-Test No.: _01__ Scenario No.: _01__ Event No.: _02__ Page _2_ of _9_

Event

Description:

____PZR Pressure Transmitter LT-459 Failure Low_______

Time Position Applicants Actions or Behavior BOP Diagnoses LT-459 failing low RO Takes manual control of PZR level control and reduces charging to minimum **

SFM Enters AP-5 RO Selects B/U channel for control ALL Re-establishes letdown (event #3)

SFM Refers to Tech Specs 3.3.1.M SFM Determines bistables in Protection Set 1, Rack 1 to be tripped SFM Directs Asset Team to investigate

Appendix D Required Operator Actions Form ES-D-2 Op-Test No.: __01_ Scenario No.: __01__Event No.: _03_ Page _3_ of _9_

Event

Description:

___Restore Letdown________________________________________

Time Position Applicants Actions or Behavior SFM Tailboard OP B-1A:XII for restoring Letdown, including reactivity control.

BOP Take manual and open TCV-130 and PCV-135 to prescribed position BOP Open Letdown isolation valves RO Increase charging to 87 gpm while maintaining seal injection BOP Adjust PCV-135 and TCV-130 as needed RO/BOP If Letdown Relief Valve lifted, reseat valve RO Return charging flow to auto when level stabilized

Appendix D Required Operator Actions Form ES-D-2 Op-Test No.: _01_ Scenario No.: _01__ Event No.: __4___ Page _4_ of _9_

Event

Description:

________Loop 4 Tcold Failure High______________

Time Position Applicants Actions or Behavior SFM/BOP Diagnoses Loop TAVG failure of TE-441 RO Places rods in manual**

SFM Enters AP-5 RO Defeats Loop 4 for TAVG and T RO Withdraws rods in manual to restore TAVG **

SFM Refers to Tech Specs (3.3.1.E & X and 3.3.2.M)

RO Returns rod control to AUTO (as time allows)

SFM Determines bistables to be tripped in Racks 15 and 16 SFM Notifies Asset Team to investigate

Appendix D Required Operator Actions Form ES-D-2 Op-Test No.: _01_ Scenario No.: _01__ Event No.: __5___ Page _5_ of _9_

Event

Description:

_____Ramp for Path 15 Emergency_________________________________

Time Position Applicants Actions or Behavior Tailboard to include target MW, rate, amount of boration and reactivity oversight SFM to be used.

RO Start a boration per OP B-1A:VII BOP Start the ramp per OP C-3:III RO Verify boration BOP Verify ramp

Appendix D Required Operator Actions Form ES-D-2 Op-Test No.: _01_ Scenario No.: _01__ Event No.: __6___ Page _6_ of _9_

Event

Description:

_____Loss of ASW Pumps_________________

Time Position Applicants Actions or Behavior BOP Acknowledge alarm on PK01-03 and loss of ASW pump 1-1 BOP Attempt to start other train ASW pump 1-2 SFM Enters OP AP-10 RO/BOP Direct Unit 2 to start an ASW pump and open cross-connect valve to supply Unit 1 ASW **

BOP Open Unit 1 cross-connect valve **

RO/BOP Direct Aux Watch to stop any radwaste discharges overboard SFM Enter T.S. 3.0.3 BOP Verify CCW temperatures normal or decreasing Appendix D Required Operator Actions Form ES-D-2

Op-Test No.: _01_ Scenario No.: _01__ Event No.: __7___ Page _7_ of _9_

Event

Description:

_____SG 1-2 FRV Failure________________________

Time Position Applicants Actions or Behavior RO Respond to Steam/Feed Mismatch alarm by checking MFW pump and MFRVs BOP Identifies SG 1-2 level increasing rapidly RO Attempt to take manual control of Feed Reg Valve FCV-520 RO Recognizes no control of SG 1-2 level SFM Directs Reactor Trip RO/BOP Trips the reactor **

SFM Enters E-0 ALL Performs immediate actions from memory

Appendix D Required Operator Actions Form ES-D-2 Op-Test No.: _01_ Scenario No.: _01__ Event No.: __8 & 10___ Page _8_ of _9_

Event

Description:

_____SG 1-2 Safety Valve sticks open and Failure of RHR to Auto Start_______________

Time Position Applicants Actions or Behavior ALL Identifies Faulted SG 1-2 SFM Directs SI if did not occur automatically SFM Conducts E-2 tailboard BOP Isolates S/G 1-2 **

Closes MSIV

Isolates AFW flow SFM Directs transition to E-1.1 (may transition to E-1.0 first)

BOP Performs E-0 Appendix E BOP Recognizes RHR pumps failure to start on SI BOP Manually starts RHR pumps and informs SFM of event SFM If in E-1.0, Foldout Page step 5 directs kickout to E-3 on uncontrolled SG level increase; kickout to E-3 (next event, #9)

Appendix D Required Operator Actions Form ES-D-2 Op-Test No.: _01_ Scenario No.: _01__ Event No.: __9___ Page _9_ of _9_

Event

Description:

_____SGTR on SG 1-2 ________________________________

Time Position Applicants Actions or Behavior BOP Identifies Ruptured SG 1-2 SFM Transitions to E-3 and Conducts tailboard BOP Sets 10% steam dump to 8.67 turns (1040 psig)

BOP Isolates S/G 1-2 **

Verifies MSIV Closed

Verifies AFW isolated

Isolates steam to TDAFP SFM Transitions to ECA-3.1 based on Ruptured SG pressure SFM Conducts tailboard RO Shuts down the RHR pumps RO Cools down the RCS using intact 10% steam dumps if not already cooled down from faulted SG

Appendix D Scenario Outline Form ES-D-1 Facility: _DCPP Units 1 & 2___ Scenario No.: __NRCSIM-02__ Op-Test No.: _01_

Examiners: ____________________________ Operators: _____________________________

Initial Conditions:

100% power. MOL. 590 ppm boron. Diluting 20 gal/hr. Last dilution 15 minutes ago.

Turnover:

Maintain current plant conditions.

Event Malf. Event Event No. No. Type* Description 1 Pmp C, BOP Trip of CCW Pp 1-1 and failure of CCW Pp 1-3 to auto start ccw1 ccw2 2 R, ALL CALL - EPOS: Path 15 emergency. Ramp to 900 MW within 30 min.

Start ramp in next 10 min.

3 Cnh C, ALL Auto RMUW system failure cvc4 4 Xmt I, ALL Failure of LT-549, SG 1-4 level transmitter mfw46 5 Xmt I, RO Failure of FCV-128 auto control cvc4 6 Mal syd3 C, ALL Grid frequency variation (drops to 58 HZ) 7 Mal C, ALL Full load rejection gen4 8 Mal pzr1 M, ALL PZR steam space break 9 Ovr C, ALL Trip of 52-HG-15 (vital bus startup supply) vx4i222o 10 Mal C, ALL Trip of DG1-1 deg1a 11 Mal I, ALL Failure of Phase A Train A to actuate ppl1a 12 M, ALL SBLOCA Response per E-1 and E-1.2

(N)ormal, (R)eactivity, (I)nstrument, (C)omponent, (M)ajor

4 drl_file 6040; drl_file 15 - secure vacuum, SJAE, Gland Seal; drl_file 48 - swap Batt 13 to Batt Chrg 131

Appendix D Required Operator Actions Form ES-D-2 Op-Test No.: __1__ Scenario No.: __2__ Event No.: __1__ Page _1_ of _10_

Event

Description:

_____Trip of CCW Pp 1-1_______________________________

Time Position Applicants Actions or Behavior RO Acknowledge PK01-09, CCW Pumps BOP Diagnose CCW Pp 1-1 trip and 1-3 failure to auto start BOP Manually start CCW Pp 1-3 BOP Verify action by referencing OP AP-11 SFM Reference Tech Spec 3.7.7

Appendix D Required Operator Actions Form ES-D-2 Op-Test No.: __1__ Scenario No.: __2__ Event No.: __2__ Page _2_ of _10_

Event

Description:

______________Path 15 Emergency Ramp____________________

Time Position Applicants Actions or Behavior SFM Tailboard to include target MW, rate, amount of boration and reactivity oversight to be used.

RO Start a boration per procedure OP B-1A:VII (start of failure of boration, event #3)

BOP Start the ramp per procedure OP C-3:III RO Verify boration completes as set in BOP Verify ramp progressing to target at set ramp RO Verify rods step in

Appendix D Required Operator Actions Form ES-D-2 Op-Test No.: __1__ Scenario No.: __2__ Event No.: __3__ Page _3_ of _10_

Event

Description:

________Failure of Reactor Makeup System _____________________

Time Position Applicants Actions or Behavior RO Acknowledge alarm PK05-11 CVCS Makeup Deviation RO Verify 1/MU OFF BOP Verify valves aligned per PK05-11 on VB-2 or AP-6, Emergency Boration SFM May transition to AP-19, Malfunction of Reactor Makeup Control SFM May transition to OP B-1A:VII, CVCS Makeup Control System Operation SFM Tailboards use of Manual for Makeup BOP Verifies control board lineup RO Takes HC-110 to manual and aligns integrators and 43/MU and 1/MU for operation RO Performs boration and verifies proper operation **

IF AP-6 is used:

BOP Start BA Transfer pumps in Hi BOP Close HCV-104 or 105 BOP Open 8104

Appendix D Required Operator Actions Form ES-D-2 Op-Test No.: __1__ Scenario No.: __2__ Event No.: __4__ Page _4_ of _10_

Event

Description:

_____Failure of LT-549, SG 1-4 Level Transmitter______________

Time Position Applicants Actions or Behavior RO/BOP Acknowledge DFWCS alarm and diagnose LT-549 failure RO/BOP Verify control systems controlling in auto SFM Enter AP-5 SFM Contact Maintenance Services SFM Determine failure in Rack 8 SFM Determine Tech Spec and ECG association SFM Determine MDAFW supply to SG 1-2, LCV-113 operability per TS and ECG (3.3.2.D, 3.3.2.J, 3.3.1.E and ECG 4.1) **

BOP Place LCV-113 in Manual and Open (may be completed during E-0 actions) **

Appendix D Required Operator Actions Form ES-D-2 Op-Test No.: __1__ Scenario No.: __2__ Event No.: __5__ Page _5_ of _10_

Event

Description:

______Failure of FCV-128 Auto Control _______

Time Position Applicants Actions or Behavior RO/BOP Recognize increased seal injection flow RO Recognize FI-128 indicates 0 gpm RO Diagnose problem as a failure of FCV-128 to control in auto RO Take manual control of FCV-128 and reduce actual charging flow **

SFM Enter AP-5 RO Verify charging and seal injection are stable BOP Verify Letdown normal SFM Contact Maintenance Services RO Maintain manual control of FCV-128 RO Maintain seal injection 6-12 gpm and control PZR level in band

Appendix D Required Operator Actions Form ES-D-2 Op-Test No.: __1__ Scenario No.: __2__ Event No.: __6__ Page _6_ of _10_

Event

Description:

___Grid Frequency Variation_______________________________________

Time Position Applicants Actions or Behavior SFM Recognize conditions relating to lowering grid frequency RO Respond to reduced seal injection flow ALL Diagnose problem as grid frequency problem ALL Determine plant response appropriate for condition SFM Tailboard event

Appendix D Required Operator Actions Form ES-D-2 Op-Test No.: __1__ Scenario No.: __2__ Event No.: __7__ Page _7_ of _10_

Event

Description:

_____Full Load Rejection/Reactor Trip_______________

Time Position Applicants Actions or Behavior ALL Recognize Full Load rejection RO/BOP Verify plant control systems operating as expected ALL Recognize reactor trip ALL Perform Immediate Actions **

SFM Enter E-0 ALL Verify Immediate Actions BOP Verify SI if it occurs SFM Transition to E-0.1 if no SI occurs; Continue in E-0 is SI occurs BOP Control AFW cooldown **

RO Verify PZR level and pressure controlling

Appendix D Required Operator Actions Form ES-D-2 Op-Test No.: __1__ Scenario No.: __2__ Event No.: __8, 11__ Page _8_ of _10_

Event

Description:

_____PZR Steam space break / Failure of Phase A Train A to Actuate_____________

Time Position Applicants Actions or Behavior RO/BOP Recognize PZR pressure decrease and level increase ALL Determine need for SI RO Manually SI if not already initiated SFM Enter E-0 again if in E-0.1 ALL Perform Immediate Actions BOP Verify Vital Buses F & G energized, H NOT energized BOP Perform Appendix E BOP Recognize Failure of Phase A and align valves accordingly **

RO Verify AFW flow and SG levels RO Verify RCS temperature trending to 547° RO Verify SGs intact RO Verify RCS is NOT intact **

SFM Transition to E-1

Appendix D Required Operator Actions Form ES-D-2 Op-Test No.: __1__ Scenario No.: __2__ Event No.: __9, 10__ Page _9_ of _10_

Event

Description:

_____Trip of 52-HG-15 and DG 1-1______________________________________

Time Position Applicants Actions or Behavior BOP Recognize Vital Bus H deenergized BOP Recognize non-vital buses energized BOP Recognize DG 1-1 tripped BOP Diagnose problem as 52-HG-15 trip and power is available to ALL buses except bus H RO/BOP Identify equipment not available because of Bus H (AFWP 1-2, CCWP 1-3, RHRP 1-2, SIP 1-2, CSP 1-2)

Appendix D Required Operator Actions Form ES-D-2 Op-Test No.: __1__ Scenario No.: __2__ Event No.: _12__ Page _10_ of _10_

Event

Description:

____E-1 and E-1.2______________________________________________________

Time Position Applicants Actions or Behavior SFM Tailboard E-1 BOP Check RCPs already secured BOP Check SGs intact RO/BOP Check PORVS intact RO Check Containment Spray not running SFM Check ECCS flow should NOT be reduced BOP Determines DGs can NOT be secured SFM Transition to E-1.2, Post LOCA Cooldown and Depressurization SFM Tailboard transition RO Reset SI and Phase A **

BOP Establish air to Containment BOP Cooldown to Cold Shutdown using 10% steam dumps < 100°/hr RO Block Low Steam Line and Low PZR Pressure SI at P-11

Appendix D Scenario Outline Form ES-D-1 Facility: _DCPP Units 1 & 2___ Scenario No.: __NRCSIM-03__ Op-Test No.: _01_

Examiners: ____________________________ Operators: _____________________________

Initial Conditions:

100% power(IC-510). MOL with boron at 590 ppm. Diluting 20 gal/hr. Last dilution was 15 minutes ago. It was reported 5 minutes ago that DG 1-1 had an air leak from the turbocharger receiver relief valve. The receiver has been isolated and the compressor secured. Technical Specifications have NOT been reviewed for this event.

Turnover:

Swap Condensate Booster Pumps per procedure for clearance on set 1-1.

Event Malf. Event Event No. No. Type* Description 1 N, BOP Swap Condensate Booster Pump sets 2 Xmt I, ALL VCT Level Indicator LI-114 Fail High cvc20 3 Cnh pzr4 I, RO Failed Auto control of PZR Pressure Controller HC-455 4 Mal C, ALL Increased screen and condenser DP (Call from Intake Watch of sudden cws3a increase in swells breaking over breakwater; kelp buildup.) (Call as cws3b Steve David reporting Environmental report rapidly building storm.

Direct crew to reduce to 50% within 1 hour1.157407e-5 days 2.777778e-4 hours 1.653439e-6 weeks 3.805e-7 months .)

asw2 5 R, ALL Ramp unit to 50%

6 Xmt I, BOP SG 1-1 Pressure Transmitter Fail High (10% Controller Fail Open in mss62 Auto) 7 Mal sei1 M, ALL Earthquake 8 Mal M, ALL LBLOCA rcs3e 9 Loa sis1 C, ALL Loss of RWST (SI-1) 10 Pmp C, BOP Failure of ASW pump 1-2 to Auto Start asw2

(N)ormal, (R)eactivity, (I)nstrument, (C)omponent, (M)ajor

6 drl_file 6060

Appendix D Required Operator Actions Form ES-D-2 Op-Test No.: __01__ Scenario No.: _03_ Event No.: __01__ Page _1__ of _8__

Event

Description:

___DG Inoperability and Swap Condensate Booster Pump_______________

Time Position Applicants Actions or Behavior SFM Evaluate DG condition for inoperability per TS 3.8.3 and 3.8.1 SFM Declare DG 1-1 inoperable SFM Tailboard swapping from set 1-1 to set 1-3 BOP Notify the Polisher Watch of the swap BOP Place set 1-3 in manual and start the pump per OP C-7A:I, section 6.2 BOP Secure set 1-1

Appendix D Required Operator Actions Form ES-D-2 Op-Test No.: __01__ Scenario No.: _03_ Event No.: __02__ Page _2__ of _8__

Event

Description:

______VCT Level Indicator LI-114 Fail High__________

Time Position Applicants Actions or Behavior RO Observe VCT level increase on PPC RO/BOP Observe FCV-112A has diverted from VCT to LHUT RO Channel check VCT level channels (114 on HSP or PPC reading 100%),

determines that LT-114 failed high SFM Refers to AP-19, Appendix A, determines that start-to-open signal to LCV-112A and VCT LVL LO-LO alarm OOS BOP Places FCV-112A in the VCT position SFM Notify TM to troubleshoot and repair LT-114

Appendix D Required Operator Actions Form ES-D-2 Op-Test No.: __01__ Scenario No.: _03_ Event No.: __03__ Page _3__ of _8__

Event

Description:

______PZR Pressure Controller HC-455 fails in Auto Mode ____

Time Position Applicants Actions or Behavior RO Acknowledge PK05-17, Low PZR Pressure RO/BOP Observe pressure indications on CC2 not trending with VB-2 indications RO Take manual control of PZR Pressure and control pressure in band **

SFM Enter OP AP-5 or OP AP-13 RO Diagnose controller failed, not channel RO Maintain PZR Pressure control in manual SFM Contact Maintenance Services

Appendix D Required Operator Actions Form ES-D-2 Op-Test No.: __01__ Scenario No.: _03_ Event No.: __04__ Page _4__ of _8__

Event

Description:

____Increased Screen and Condenser DP________________________________

Time Position Applicants Actions or Behavior RO Receives alarm PK13-01, Screen Hi Diff Auto Start SFM Referencs to AR PK13-01 BOP Has local operator check screens still running continuously, in fast speed and OK SFM Implements OP AP-7, Degraded Condenser, when Screens P cannot be reduced below 8 inches BOP Checks CWPs status normal; verifies screens in high speed.

SFM Monitors Screens Ps on the PPC SFM References OP O-28, Intake Management

Appendix D Required Operator Actions Form ES-D-2 Op-Test No.: __01__ Scenario No.: __03__ Event No.: _05__ Page _5__ of _8__

Event

Description:

____Ramp to 50% and securing of CWP 1-2_____________________________

Time Position Applicants Actions or Behavior SFM Direct a load reduction to 50% and contacts EPOS BOP Set in ramp and commence turbine ramp if time permits RO Calculate and start boration if time permits SFM Monitor screen DP SFM Observes Screens P for CWP 1-2 increasing above 70 SFM Directs trip of CWP 1-2 (may have been done earlier) **

BOP Trips CWP 1-2 (may have been done earlier) **

RO Monitors programmed ramp to 50% from trip of CWP 1-2 ALL Crew stabilizes plant at 50% and continues to monitor screens, CWPs, and ASW

Appendix D Required Operator Actions Form ES-D-2 Op-Test No.: __01__ Scenario No.: __03__ Event No.: _06__ Page _6__ of _8__

Event

Description:

SG 1-1 Pressure Transmitter PT-516 Fail High____________________________

Time Position Applicants Actions or Behavior BOP Identify SG 1-1 10% Atmospheric open BOP Inform SFM, take manual control and close atmospheric **

SFM Enter OP AP-5 RO/BOP Verify plant controls operating in automatic SFM Contact Maintenance Services SFM Reference TS 3.3.2.D to ensure operability

Appendix D Required Operator Actions Form ES-D-2 Op-Test No.: __01__ Scenario No.: __03__ Event No.: _07__ Page _7__ of _8__

Event

Description:

____Earthquake____________________________________________________

Time Position Applicants Actions or Behavior ALL Recognize earthquake RO Place rods in auto if in manual ALL Monitor plant for proper response BOP Determine magnitude of earthquake at 0.4g ALL Recognize reactor trip ALL Perform immediate actions **

SFM Enter E-0 ALL Verify immediate actions BOP Takes control of AFW to minimize cooldown **

SFM Transition to E-0.1 RO Verify PZR level control trending to 22%

BOP Verify PZR pressure control trending to 2235#

Appendix D Required Operator Actions Form ES-D-2 Op-Test No.: __01__ Scenario No.: __03__ Event No.: _08 and 10__ Page _8__ of _8__

Event

Description:

__LBLOCA and Failure of ASW pump 1-2 to Auto Start___________________

Time Position Applicants Actions or Behavior RO Recognize PZR level and pressure decrease SFM Direct manual SI if recognized early enough RO Manually SI if Auto has not occurred SFM Transition to E-0 ALL Verify Immediate Actions **

BOP Perform Appendix E BOP Manually start ASW pump 1-2 on failure to auto start **

RO/BOP Recognize Adverse Containment and notify SFM SFM Transition to E-1 SFM Tailboard transition BOP Verify RCPs tripped BOP Verify SGs intact RO/BOP Recognize CCPs, SIPs, and RHRps cavitating RO/BOP Reset SI and 4kV transfer relays **

RO/BOP Secure CCPs, SIPs, RHRps **

SFM Diagnose Cold Leg Recirc Capability NOT met **

Appendix D Scenario Outline Form ES-D-1 Facility: _DCPP Units 1 & 2___ Scenario No.: __NRCSIM-04__ Op-Test No.: _01_

Examiners: ____________________________ Operators: _____________________________

Initial Conditions:

100% power. MOL (IC-510). 590 ppm Boron. AFW Pp 1-2 OOS for repair last 12 hours1.388889e-4 days 0.00333 hours 1.984127e-5 weeks 4.566e-6 months . RTS in 24 hours2.777778e-4 days 0.00667 hours 3.968254e-5 weeks 9.132e-6 months (Drill File 43). Diluting 20 gal/hr. Last dilution was 15 minutes ago.

Turnover:

Swap CFCUs from 1-1 to 1-2 for clearance.

Event Malf. Event Event No. No. Type* Description 1 N, BOP Swap CFCU from 1-1 to 1-2.

2 R, ALL CALL - EPOS: Path 15 emergency. Commence ramp in 10 minutes to 900MW in following 30 minutes.

3 pmp C, ALL Trip of running CCP and restoring letdown cvc2 4 Xmt tur2 C, ALL PT-505 hangs up at 100%

5 Xmt I, RO PZR pressure controller bias fails to +100# during ramp pzr27 6 Ovr C, BOP TCV-23 Failure vb3079a 7 Mal sei1 M, ALL Earthquake 8 Mal rcs1 M, ALL LBLOCA (50% DBA loop 2) 9 Mal C, BOP Failure of phase A Train B to actuate ppl1b 10 Mal syd1 M, ALL Loss of Offsite power 11 Transfer to Cold Leg Recirc

(N)ormal, (R)eactivity, (I)nstrument, (C)omponent, (M)ajor

7 drl_file 6070

Appendix D Required Operator Actions Form ES-D-2 Op-Test No.: __01__ Scenario No.: __04__ Event No.: __01__ Page _1_ of _9_

Event

Description:

____Swap CFCU_______________________________________________________

Time Position Applicants Actions or Behavior SFM Tailboard swap of CFCU from 1-1 to 1-2 BOP Swap CFCU per OP H-2:I

Appendix D Required Operator Actions Form ES-D-2 Op-Test No.: __01__ Scenario No.: __04__ Event No.: __02__ Page _2_ of _9__

Event

Description:

____Path 15 Emergency Ramp__________________________________________

Time Position Applicants Actions or Behavior SFM Verify EPOS call for Path 15 Ramp SFM Tailboard crew on emergency ramp and reactivity control RO Start a boration per OP B-1A:VII BOP Start the ramp per OP C-3:III RO/BOP Verify ramp progressing to target at set ramp

Appendix D Required Operator Actions Form ES-D-2 Op-Test No.: __01__ Scenario No.: __04__ Event No.: __03__ Page _3_ of _9_

Event

Description:

___Trip of CCP and restoring letdown__________________________________

Time Position Applicants Actions or Behavior RO Acknowledge alarm PK04-17, CCP 1-2 BOP Diagnose CCP 1-2 trip SFM Enter OP AP-17, Loss of Charging BOP Start CCP 1-1 **

RO Control PZR level control to return PZR level to band **

RO Control Seal Injection flow SFM Tailboard placing letdown in service BOP Place TCV-130 and PCV-135 in manual and open RO Increase charging flow BOP Open letdown orfice isolation valve BOP Place TCV-130 and PCV-135 in auto SFM Reference TS

Appendix D Required Operator Actions Form ES-D-2 Op-Test No.: __01__ Scenario No.: __04__ Event No.: __04___ Page _4_ of _9_

Event

Description:

____PT-505 Failure High (stuck in position)_______________________

Time Position Applicants Actions or Behavior RO Recognize not enough rod motion for plant condition RO Recognize TAVG - TREF difference not normal for condition BOP Diagnose PT-505 failed as is at 100%

RO Place rods in manual to control TAVG - TREF difference **

SFM Enter AP-5 RO/BOP Verify other plant controllers working normally SFM Identify affected bistables and TS 3.3.1 requirements SFM Contact Maintenance Services

Appendix D Required Operator Actions Form ES-D-2 Op-Test No.: __01__ Scenario No.: __04__ Event No.: __05___ Page _5_ of _9_

Event

Description:

___PZR pressure controller bias fails to +100 psig___________________________

Time Position Applicants Actions or Behavior RO Acknowledge PZR low pressure alarm PK05-17 RO Diagnose PZR pressure indication is about 100# over actual RO Take manual control of PZR master pressure controller **

RO Increase PZR pressure to 2235 psig SFM Enter AP-13 BOP Stop ramp if ordered RO/BOP Verify Safeties, PORVS and sprays closed RO Verify heaters energized RO Restore pressure to normal band SFM Exit AP-13 and enter AP-5 RO Loop out PT-455 SFM Contact maintenance services ALL Recommence ramp is secured

Appendix D Required Operator Actions Form ES-D-2 Op-Test No.: __01__ Scenario No.: __04__ Event No.: __06___ Page _6_ of _9_

Event

Description:

___TCV-23 Failure______________________________________________________

Time Position Applicants Actions or Behavior RO Acknowledge alarm PK14-16, Turbine/Generator Trouble SFM Transition to AP-30, Main Generator Malfunctions BOP Diagnose TCV-23 not functioning properly BOP Take manual control of TCV-23 and restore temperature SFM Contact Maintenance Services

Appendix D Required Operator Actions Form ES-D-2 Op-Test No.: __01__ Scenario No.: __04__ Event No.: __07___ Page _7_ of _9_

Event

Description:

____Earthquake___________________________________________________

Time Position Applicants Actions or Behavior ALL Recognize Earthquake ALL Monitor plant response BOP Determine magnitude of earthquake ALL Verify reactor/turbine trip **

ALL Perform immediate actions **

SFM Enter E-0

Appendix D Required Operator Actions Form ES-D-2 Op-Test No.: __01__ Scenario No.: __04__ Event No.: __08 & 09___ Page _8_ of _9_

Event

Description:

___LBLOCA and failure of Phase A Train B________________________________

Time Position Applicants Actions or Behavior RO Acknowledge alarms for SI Actuation ALL Verify SI actuation **

BOP Verify AC Vital Buses energized **

BOP Implement Appendix E BOP Diagnoses Phase A Train B has not actuated BOP Manually position valves required to complete Phase A **

RO Control AFW flow to minimize cooldown **

RO Verify RCPs stopped **

RO Verify SGs intact ALL Diagnose LBLOCA SFM Transition to E-1 SFM Tailboard E-1 RO/BOP Verify Containment Spray required RO/BOP Verify ECCS pumps running RO Check RWST level at 33% **

SFM Transition to E-1.3, Transfer to Cold Leg Recirc **

SFM May enter FR-P.1 SFM Exit FR-P.1 after verifying RHR flow >100 gpm

Appendix D Required Operator Actions Form ES-D-2 Op-Test No.: __01__ Scenario No.: __04__ Event No.: __11___ Page _9_ of _9_

Event

Description:

_____Transfer to Cold Leg Recirc________________________________

Time Position Applicants Actions or Behavior SFM Direct actions of E-1.3 RO/BOP Reset SI, Phase A and Phase B **

RO/BOP Check ECCS lineup RO/BOP Isolate RHR train discharge headers **

RO/BOP Place RHR in service to SIP **

RO/BOP Crosstie CCP suction to SIP suction**

RO/BOP Place RHR in service to CCP **

RO/BOP Close RWST valves **

RO/BOP Align Containment Spray RO/BOP Reduce RHR flow as necessary RO/BOP Monitor sumps

ES-301 Administrative Topics Outline Form ES-301-1 Facility: ___DCPP_______________________ Date of Examination: __02/07/2005__

Examination Level (circle one): RO / SRO Operating Test Number: ___01______

Administrative Topic Type Describe activity to be performed (see Note) Code*

Conduct of Operations SRO - Review RCS Water Inventory Balance RO - Perform RCS Water Inventory Balance NRCADM-01SRO N CFR 43.2/43.3/45.3 RO-3.4 SRO-4.0 NRCADM-01RO M 2.1.33 Ability to recognize indications for system operating parameters which are entry level conditions for TS SRO - Review Outage Safety Checklist Conduct of Operations RO - Perform Outage Safety Checklist NRCADM-02SRO N CFR 41.10/43.2/45.12 RO-3.4 SRO-3.8 NRCADM-02RO M 2.1.32 Ability to explain and apply all system limits and precautions Equipment Control SRO - Safety Function Determination CFR 43.2/45.13 SRO-3.8 NRCADM-03SRO N 2.2.24 Ability to analyze the affect of maintenance activities on LCO status NRCADM-03RO N RO - Determine Clearance Points RO-3.6 2.2.13 Knowledge of tagging and clearance procedures Radiation Control SRO - Approve Emergency Exposure RO - Determine Posting NRCADM-04SRO N CFR 43.4/45.10 RO-2.5 SRO-3.1 NRCADM-04RO 2.3.4 Knowledge of radiation exposure limits and contamination N

control, including permissible levels in excess of those authorized Emergency Plan SRO - GDT Rupture Release and EAL M

NRCADM-05SRO CFR 43.5.45.11 SRO-4.1 2.4.41 Knowledge of SRO responsibilities in emergency plan implementation NOTE: All items (5 total) are required for SROs. RO applicants require only 4 items unless they are retaking only the administrative topics, when all 5 are required.

Type Codes & Criteria: (C)ontrol room (D)irect from bank ( 3 for ROs; 4 for SROs & RO retakes)

(N)ew or (M)odified from bank ( 1)

(P)revious 2 exams ( 1; randomly selected)

(S)imulator

NUCLEAR POWER GENERATION DIABLO CANYON POWER PLANT JOB PERFORMANCE MEASURE Number: NRCADM01RO

Title:

PERFORM RCS WATER INVENTORY BALANCE Examinee:

Evaluator:

Print Signature Date Results: Sat Unsat Total Time: minutes Comments:

References:

STP R-10C, RCS Water Inventory Balance, Rev. 25 Technical Specifications 3.4.14, RCS Leakage Alternate Path: Yes No X Time Critical: Yes No X Time Allotment: 30 minutes Critical Steps: 4, 15 Job Designation: RO/SRO Task Number: G2.1.33 Rating: 3.4/4.0 AUTHOR: JACK BLACKWELL DATE: 01/18/2005 REVIEWED BY: N/A DATE:

JPM COORDINATOR APPROVED BY: N/A DATE:

TRAINING LEADER REV. 0

JPM TITLE: REVIEW RCS WATER INVENTORY BALANCE JPM NO.: NRCADM01RO INSTRUCTOR WORKSHEET Directions: No plant controls or equipment are to be operated during the performance of this Job Performance Measure. All actions taken by the examinee should be clearly demonstrated and verbalized to the evaluator. The student will be given the initial conditions, initiating cue, and task standard. The examiner will then ask if any clarifications are needed. After identifying the appropriate procedure for the task, the examinee may be given the procedure and told the step with which to begin.

Required Materials: STP R-10C, RCS Water Inventory Balance, Rev. 25 Technical Specifications 3.4.14 Initial Conditions: Unit 1 is at BOL, 100% power with the PPC out of service. No other equipment out of service and Zinc Injection is not in service. The SFM has directed a manual RCS leak evaluation per STP R-10C. The prior shift has logged the initial readings, and has taken the final data set, but has not entered the data, nor completed the STP.

The final set of readings, taken one hour ago, are as follows:

o YIC-110 - 40 gal.

o YIC-111 - 0 gal.

o LI-461 - 52.5%

o LI-112 - 25%

o TI-412 - 572.5° o LI-470 - 83%

o LI-188 - 52%

o FI 0 gal o LI-950 - 66%

o LI-952 - 67%

o LI-954 - 65%

o LI-956 - 66%

o RCS to Secondary leak rate - 0.051 gpm o RCS to CCW leak rate - 0 gpm o Other IDENTIFIED leak rates - 0 gpm.

Initiating Cue: The SFM has directed you to complete STP R-10C through Data Reduction and Evaluation, and have it ready for his evaluation.

Task Standard: The procedure is completed and ready for SFM review.

NRCADM01RO.DOC Page 2 of 8 REV. 0

JPM TITLE: REVIEW RCS WATER INVENTORY BALANCE JPM NO.: NRCADM01RO INSTRUCTOR WORKSHEET Start Time:

Step Expected Operator Actions

1. Verify Start Data Section filled 1.1 Confirms Unit 1, Mode 1and out properly. Date/Time correct.

1.2 Confirms Precaution and Limitations (Section 10) meet plant conditions and are all initialed.

1.3 Confirms Prerequisites (Section 11) meet plant conditions and initialed.

Step was: Sat: ______ Unsat _______*

2. Verify Procedure steps 2.1 Confirms Section 12.1.1 data is completed properly. recorded for Rx Power, Pressure and Temperature.

2.2 Confirms Pressure and Temperature above minimum required on steps 12.1.2 and 12.1.3.

2.3 Confirms Step 12.1.4 marked N/A.

2.4 Confirms Step 12.1.5 marked No and initialed.

2.5 Confirms Step 12.1.6 and 12.1.7 marked yes and initialed, and 12.1.8 initialed.

2.6 Confirms Step 12.2 marked N/A.

2.7 Confirms Step 12.3.1 and 12.3.4 marked N/A, steps 12.3.2 and 12.3.3 initialed.

Step was: Sat: ______ Unsat _______*

Denotes an entry required on the JPM cover sheet.

- Denotes a Critical Step.

NRCADM01RO.DOC Page 3 of 8 REV. 0

JPM TITLE: REVIEW RCS WATER INVENTORY BALANCE JPM NO.: NRCADM01RO INSTRUCTOR WORKSHEET Step Expected Operator Actions

3. Verify Table 1 and 2 data. 3.1 Confirms Table 1 data readings appear accurate.

3.2 Enters Table 2 data for Integrator, PZR Level, VCT level and TAVG.

Step was: Sat: ______ Unsat _______*

- 4. Calculates Table 3 data. 4.1 Calculates BA as 40 gal.

- 4.2 Calculates PW as 0 gal.

4.3 Calculates PZR as -217 gal.

4.4 Calculates VCT as -172.8 gal.

4.5 Calculates TAVG as 39.25 gal.

Step was: Sat: ______ Unsat _______*

5. Calculates T step 12.3.5. 5.1 Calculates T as 240 minutes.

Step was: Sat: ______ Unsat _______*

6. Calculates RCS leak rate step 6.1 Calculates RCS leak rate 1.954 gpm.

12.3.6.

Step was: Sat: ______ Unsat _______*

7. Calculates Leak Error Factor step 7.1 Calculates Leak Error Factor 0.317 12.3.7.b. gpm.

Step was: Sat: ______ Unsat _______*

8. Caluclates Gross Leak Rate step 8.1 Calculates Gross Leak Rate 2.271 12.3.8. gpm.

Step was: Sat: ______ Unsat _______*

Denotes an entry required on the JPM cover sheet.

- Denotes a Critical Step.

NRCADM01RO.DOC Page 4 of 8 REV. 0

JPM TITLE: REVIEW RCS WATER INVENTORY BALANCE JPM NO.: NRCADM01RO INSTRUCTOR WORKSHEET Step Expected Operator Actions

9. Determine Gross Leak Rate is >1 9.1 Marks 12.3.9 No.

gpm and must continue procedure at Table 4.

9.2 Marks 12.3.10 N/A.

9.3 Initials 12.3.11.

Step was: Sat: ______ Unsat _______*

10. Verify Table 4 data. 10.1 Confirms initial data readings appear accurate and enters final readings.

10.2 Caluclates PRT is 123 gal.

10.3 Caluclates RCDT is 10.25 gal.

10.4 Caluclates RCDT totalizer is 0 gal.

10.5 Caluclates Accumulators is 0 gal.

Step was: Sat: ______ Unsat _______*

11. Calculates RCS Identified Leak 11.1 Calculates Identified Leak Rate at Rate step 12.3.12. (Table 4). 0.606 gpm.

Step was: Sat: ______ Unsat _______*

12. Calculate Identified Leak Rate 12.1 Calculates Identified Leak Rate Error Error Factor step 12.3.13.b. Factor at 0.458 gpm.

Step was: Sat: ______ Unsat _______*

13. Calculates Identified Leakage 13.1 Calculates Identified Leakage at step 12.3.14. 1.064 gpm.

Step was: Sat: ______ Unsat _______*

14. Signs test performer signature 14.1 Signs test performer.

step 12.5.

Step was: Sat: ______ Unsat _______*

Denotes an entry required on the JPM cover sheet.

- Denotes a Critical Step.

NRCADM01RO.DOC Page 5 of 8 REV. 0

JPM TITLE: REVIEW RCS WATER INVENTORY BALANCE JPM NO.: NRCADM01RO INSTRUCTOR WORKSHEET Step Expected Operator Actions

- 15. Perform step 13, Data Reduction 15.1 Checks N/A on step 13.1.

and Evaluation.

15.2 Calculates Total RCS leak rate error factor from Aux Board and Control Boards, step 13.2.1.b. at 0.557 gpm.

and initials step.

15.3 Calculates Unidentified Leakage at 1.918 gpm. and initials step.**

NOTE: Critical Task met if Unidentified Leakage Rate calculated > 1gpm.

All other calculations in this JPM may vary as long as the final calculation identifies Unidentified Leakage as > 1 gpm.

Step was: Sat: ______ Unsat _______*

Stop Time:

Total Time: (Enter total time on the cover page)

Denotes an entry required on the JPM cover sheet.

- Denotes a Critical Step.

NRCADM01RO.DOC Page 6 of 8 REV. 0

JPM NO.: NRCADM01RO EXAMINEE CUE SHEET Initial Conditions: Unit 1 is at BOL, 100% power with the PPC out of service. No other equipment out of service and Zinc Injection is not in service. The SFM has directed a manual RCS leak evaluation per STP R-10C. The prior shift has logged the initial readings, and has taken the final data set, but has not entered the data, nor completed the STP.

The final set of readings, taken one hour ago, are as follows:

o YIC-110 - 40 gal.

o YIC-111 - 0 gal.

o LI-461 - 52.5%

o LI-112 - 25%

o TI-412 - 572.5° o LI-470 - 83%

o LI-188 - 52%

o FI 0 gal o LI-950 - 66%

o LI-952 - 67%

o LI-954 - 65%

o LI-956 - 66%

o RCS to Secondary leak rate - 0.051 gpm o RCS to CCW leak rate - 0 gpm o Other IDENTIFIED leak rates - 0 gpm.

Initiating Cue: The SFM has directed you to complete STP R-10C through Data Reduction and Evaluation, and have it ready for his evaluation.

Task Standard: The procedure is completed and ready for SFM review.

NRCADM01RO.DOC Page 7 of 8 REV. 0

JPM TITLE: REVIEW RCS WATER INVENTORY BALANCE JPM NO.: NRCADM01RO ATTACHMENT 1, SIMULATOR SETUP The simulator is not needed for the performance of this JPM.

NRCADM01RO.DOC Page 8 of 8 REV. 0

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PACIFIC GAS AND ELECTRIC COMPANY NUMBER STP R-10C NUCLEAR POWER GENERATION REVISION 25 DIABLO CANYON POWER PLANT PAGE 1 OF 18 SURVEILLANCE TEST PROCEDURE UNITS TITLE: Reactor Coolant System Water Inventory Balance 1 2 AND 05/20/04 EFFECTIVE DATE PROCEDURE CLASSIFICATION: QUALITY RELATED

1. SCOPE 1.1 Determine the gross leak rate or IDENTIFIED and UNIDENTIFIED LEAKAGE from the reactor coolant system (RCS) by taking the difference in RCS and "chemical and volume control system" (CVCS) inventory change over a reasonable period of time without inventory makeup.

2. DISCUSSION 2.1 Tracking the RCS inventory in a consistent manner provides an effective means of quantifying overall system leakages. Non-RCS sources of water added to the RCDT and PRT are eliminated or quantified when utilizing Table 4 to determine IDENTIFIED LEAKAGE.

2.2 If the gross RCS leak rate measured in STP R-10C exceeds 1 gpm (or 0.965 gpm if zinc injection is in service), this procedure will consider the following leakage parameters to allow the SFM to classify the leakage:

2.2.1 PRT Level 2.2.2 RCDT Level 2.2.3 RCDT Flow Totalizer 2.2.4 Accumulator Leakage to the RCDT 2.2.5 RCS Leakage to Secondary (Stm. Gen.)

2.2.6 RCS Leakage to CCW 2.2.7 Other IDENTIFIED LEAKAGE, which may be tracked as necessary with Volume 9 2.2.8 Initiate an Action Request (AR) to document actions taken. Create a PIMS evaluation screen (EVAL) to be routed to the maintenance rule program (PTMR).

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PACIFIC GAS AND ELECTRIC COMPANY NUMBER STP R-10C DIABLO CANYON POWER PLANT REVISION 25 PAGE 2 OF 18 TITLE: Reactor Coolant System Water Inventory Balance UNITS 1 AND 2 2.3 RCS leaks to closed system (steam generators, CCW system, ECCS systems, etc.) are not directly identified in this procedure.

These leaks are evaluated by other means. Examples are as follows:

2.3.1 RCS to steam generators - determine from activity analysis of secondary coolant.

2.3.2 RCS to CCW - determine from CCW activity analysis and increasing level in surge tank.

2.3.3 Letdown/Charging - determine from increasing auxiliary building area radiation monitors and airborne activity.

2.3.4 RCS to accumulators are not identified in this procedure.

3. RESPONSIBILITIES 3.1 Shift foreman (SFM), for operation of the equipment as required, for obtaining test data, for data reduction as required by this procedure and for evaluation of reactor coolant system leakage.

3.2 Chemistry engineer, for determining primary system leakage to the secondary system and the component cooling water system.

4. FREQUENCY 4.1 This test shall be performed when required by STP I-1B or as directed by the shift foreman.

4.2 RCS water inventory balance whether performed in STP I-1B or STP R-10C shall be current when operating in MODES 1 through 4 and may be performed in MODES 1 through 5.

5. TECHNICAL SPECIFICATIONS 5.1 This test is performed to satisfy Technical Specification SR 3.4.13.1.

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PACIFIC GAS AND ELECTRIC COMPANY NUMBER STP R-10C DIABLO CANYON POWER PLANT REVISION 25 PAGE 3 OF 18 TITLE: Reactor Coolant System Water Inventory Balance UNITS 1 AND 2

6. ACCEPTANCE CRITERIA 6.1 The terms used herein are defined in the Technical Specifications. Reactor coolant system leakage shall be limited to:

6.1.1 NO PRESSURE BOUNDARY LEAKAGE 6.1.2 1 GPM UNIDENTIFIED LEAKAGE (or 0.965 gpm if zinc injection is in service).

6.1.3 10 GPM IDENTIFIED LEAKAGE 6.2 If these limits are exceeded, comply with the appropriate Technical Specification ACTION requirements.

6.3 If the gross RCS leak rate exceeds 1 GPM (or 0.965 gpm if zinc injection is in service),

further evaluation must be performed to determine the source of the leakage and to differentiate between IDENTIFIED and UNIDENTIFIED LEAKAGE as defined in the Technical Specifications.

6.4 UNIDENTIFIED LEAKAGE, and IDENTIFIED LEAKAGE used at the decision points in this test include an error factor for readability.

7. REFERENCES 7.1 STP R-10, "Reactor Coolant System Leakage Evaluation."

7.2 Acceptance Criteria Basis AC R-10C.

7.3 NRC Information Notice 94-46, "Nonconservative Reactor Coolant System Leakage Calculation."

8. APPENDICES None

9. ATTACHMENTS 9.1 "Pressurizer Level and RCS TAVG Adjustment Factor Curves," 03/20/03 9.2 "Pressurizer Level Correction Curves for Pressurizer Pressures," 03/20/03 STP R-10C.Doc 06 0119.0817

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PACIFIC GAS AND ELECTRIC COMPANY NUMBER STP R-10C DIABLO CANYON POWER PLANT REVISION 25 PAGE 9 OF 18 TITLE: Reactor Coolant System Water Inventory Balance UNITS 1 AND 2 PERF

f. During zinc acetate injection and if CALCULATED LEAK RATE, AVG LEAK RATE or 95% UCL LEAK RATE recorded above is greater than or equal to 0.965 gpm, perform the manual calculation per step 12.3. If all the leak rates are less than 0.965 gpm, N/A step 12.3 and go to step 12.4.

N/A [ ] ______

OR

g. With NO zinc acetate injection and if CALCULATED LEAK RATE, AVG LEAK RATE or 95% UCL LEAK RATE recorded above is greater than or equal to 1 gpm, perform the manual calculation per step 12.3. If all the leak rates are less than 1 gpm, N/A step 12.3 and go to step 12.4.

N/A [ ] ______

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PACIFIC GAS AND ELECTRIC COMPANY NUMBER STP R-10C DIABLO CANYON POWER PLANT REVISION 25 PAGE 12 OF 18 TITLE: Reactor Coolant System Water Inventory Balance UNITS 1 AND 2 NOTE: The duration of the manual leak rate evaluation should be 4 hours4.62963e-5 days 0.00111 hours 6.613757e-6 weeks 1.522e-6 months or longer. Minimum T is 2 hours2.314815e-5 days 5.555556e-4 hours 3.306878e-6 weeks 7.61e-7 months .

TABLE 3 PARAMETER AVG FINAL AVG INITIAL DIFFERENCE & CONVERSION from Table 2 from Table 1 (FINAL-INITIAL)1 X FACTOR5

a. B.A. Integrator 7 gal gal galx1.0 = gal (BA)

( ) ( ) ( )x1.0 =( ) gal

b. P.W. Integrator 8 gal gal galx1.0 = gal (PW)

( ) ( ) ( )x1.0 =( ) gal 3

c. Pzr Level % % %xF1x62.0 gal/% = gal (PZ)

( ) ( ) ( )x( )x62 =( ) gal

d. VCT Level % % %x19.2 gal/% = gal (V)

( ) ( ) ( )x19.2 =( ) gal

e. RCS Tavg 4 °F °F °FxF2x78.5 gal/°F = gal (TAVG)

( ) ( ) ( )x( )x78.5=( ) gal NOTES:

1 Sign convention is: If Final >Initial = Positive If Final < Initial = Negative 2

Computer values are preferred :

3 Pzr level channels are (LI-461 OR L0482A), (LI-459A or L0480A), and (LI-460 or L0481A). If the pressurizer pressure is below 2185 psig, determine actual Pzr level per Attachment 9.2. Refer to precautions and limitations, step 10.5.

4 When the RCS temperature is less than 530°F, determine TAVG by averaging the available computer points for loop temperature RTDs T0406A, T0419A, T0426A, T0439A, T0446A, T0459A, T0466A, and T0479A, or using U0491.

5 If pressurizer pressure is below 2200 psig or TAVG is below 530°F, refer to precautions and limitations, step 10.4.

6 VCT level channels are LI-112 or L0112A.

7 B.A. Integrator YIC-110 or PPC point F0110D.

8 P.W. Integrator YIC-111 or PPC point F0111D.

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PACIFIC GAS AND ELECTRIC COMPANY NUMBER STP R-10C DIABLO CANYON POWER PLANT REVISION 25 PAGE 13 OF 18 TITLE: Reactor Coolant System Water Inventory Balance UNITS 1 AND 2 PERF 12.3.5 T Calculation = End Time - Start Time T = = min Table 2 Time Table 1 Time 12.3.6 BA + PW PZ V + TAVG Calculated RCS Leak Rate =

T

=

( )+ ( ) ( ) ( )+ ( )

( )

Calculated RCS Leak Rate = ____________ GPM NOTE: If the PW is not zero, the RCS leak rate is invalid because the primary water integrator is not a qualified PME device.

12.3.7 Calculated RCS Leak Rate Error Factor (EFg)

a. When readings are taken from the PPC, 14.48 gal 14.48 gal EFg = = = gpm Tmin min N/A [ ] ______

b. When readings are taken from the control boards, 76.12 gal 76.12 gal EF g = = = gpm Tmin min N/A [ ] ______

NOTE: T may be increased beyond 4 hours4.62963e-5 days 0.00111 hours 6.613757e-6 weeks 1.522e-6 months to lower the effects of EFg on the leak rates.

12.3.8 Gross RCS Leak Rate Calculated Leak Rate + EFg = Gross Leak Rate

____________ + ___________ = ___________ gpm ______

Step 12.3.6 Step 12.3.7a or 12.3.7b STP R-10C.Doc 06 0119.0817

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PACIFIC GAS AND ELECTRIC COMPANY NUMBER STP R-10C DIABLO CANYON POWER PLANT REVISION 25 PAGE 14 OF 18 TITLE: Reactor Coolant System Water Inventory Balance UNITS 1 AND 2 PERF 12.3.9 Zinc injection in service? YES NO

[ ] [ ] ______

12.3.10 If the gross leak rate calculated in step 12.3.8 is

< 1 gpm (or 0.965 gpm if zinc injection was in service) go to step 12.4.

N/A [ ] ______

12.3.11 If the gross leak rate calculated in step 12.3.8 is N 1 gpm (or 0.965 gpm if zinc injection was in service), fill out below to determine the IDENTIFIED LEAKAGE from the RCS system. Initiate an AR and route Eval to PTMR.

N/A [ ] ______

NOTE: Do NOT run ECCS pumps that take suction from the RWST while performing the IDENTIFIED LEAKAGE portion of this test. This will preclude possible leakage into the RWST or PRT from the ECCS system.

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PACIFIC GAS AND ELECTRIC COMPANY NUMBER STP R-10C DIABLO CANYON POWER PLANT REVISION 25 PAGE 16 OF 18 TITLE: Reactor Coolant System Water Inventory Balance UNITS 1 AND 2 PERF 12.3.13 Calculated IDENTIFIED LEAKAGE Error Factor (EFID)

a. When readings are taken from the Aux Board +PPC, 22.17gal 22.17 gal EFID = = = gpm Tmin min N/A [ ] ______

b. When readings are taken from the Aux Board + Control Boards, 109.8gal 109.8 gal EFID = = = gpm Tmin min N/A [ ] ______

NOTE 1: T may be increased beyond 4 hours4.62963e-5 days 0.00111 hours 6.613757e-6 weeks 1.522e-6 months to lower the effects if EFID on the leak rates.

NOTE 2: If any control board indicator is used in Table 4, use 12.3.13b.

12.3.14 IDENTIFIED LEAKAGE = RCS Calculated Identified Leak Rate + EFID.

____________ + ___________ = ___________ gpm ______

Step 12.3.12 Step 12.3.13a or 12.3.13b 12.3.15 If IDENTIFIED LEAKAGE is greater than 10 gpm, refer to Technical Specification 3.4.13 for LCO.

AR # _________________

N/A [ ] ______

12.4 REMARKS:

12.5 Test performers and verifiers:

Name Signature Date/Time Init

/

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PACIFIC GAS AND ELECTRIC COMPANY NUMBER STP R-10C DIABLO CANYON POWER PLANT REVISION 25 PAGE 17 OF 18 TITLE: Reactor Coolant System Water Inventory Balance UNITS 1 AND 2 PERF

13. DATA REDUCTION AND EVALUATION NOTE: If the manual RCS LEAK calculation is performed, use the leak rate data from step 12.3.8.

13.1 If the Gross RCS leak rate from step 12.2.2 or 12.3.8 is less than 0.965 gpm if zinc injection is in progress or less than 1 gpm if zinc injection is NOT in progress, assume the following:

N/A [ ] ______

UNIDENTIFIED LEAKAGE < 1 gpm, and IDENTIFIED LEAKAGE < 10 gpm 13.2 If the Gross RCS leak rate from step 12.3.8 is greater than or equal to 1 gpm, record.

N/A [ ] ______

13.2.1 Calculated Total RCS Leak Rate Error Factor (EFTTL).

a. When readings are taken from the Aux Board + PPC, 26.48 gal 26.48 gal EFTTL = = = gpm Tmin min N/A [ ] ______

b. When readings are taken from the Aux Board + Control Boards, 133.6 gal 133.6 gal EFTTL = = = gpm Tmin min N/A [ ] ______

NOTE 1: T should be the same for Table 1 through Table 4. If not use the shortest T.

NOTE 2: If any control board indicator is used in Table 1 through Table 4, use 13.2.1b.

13.2.2 UNIDENTIFIED LEAKAGE = Calculated RCS Leak Rate - RCS Calculated Identified Leak Rate + EFTTL.

____________ - ___________ + ___________ =

Step 12.3.6 Step 12.3.12 EFTTL 13.2.1a or 13.2.1b UNIDENTIFIED LEAKAGE = ______________ ______

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PACIFIC GAS AND ELECTRIC COMPANY NUMBER STP R-10C DIABLO CANYON POWER PLANT REVISION 25 PAGE 18 OF 18 TITLE: Reactor Coolant System Water Inventory Balance UNITS 1 AND 2 PERF

14. PRIMARY REVIEW 14.1 Verify the acceptance criteria have been satisfied for the reactor coolant system leak rate.

UNIDENTIFIED LEAKAGE is less than 0.965 gpm if zinc injection is in progress or less than 1 gpm if zinc injection is NOT in progress. (Steps 13.1 or 13.2.2.) ______

IDENTIFIED LEAKAGE is less than 10 gpm.

(Steps 13.1 or 12.3.14) ______

14.2 REMARKS: Describe any malfunctions, explain any NO or N/A entries in any of the data and list any discrepancies.

14.3 Review the completed procedure.

If the acceptance criteria has not been satisfied, notify management promptly, write an Action Request and refer to applicable Technical Specifications limiting conditions for operations.

AR # ___________________________

Signature: Date/Time /

Shift Foreman

15. SECONDARY REVIEW 15.1 Review procedure for completeness and acceptability. ______

15.2 REMARKS:

Reviewed By: Date Second Reviewer STP R-10C.Doc 06 0119.0817

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03/20/03 Page 1 of 2 DIABLO CANYON POWER PLANT TITLE:

STP R-10C ATTACHMENT 9.1 Pressurizer Level and RCS TAVG Adjustment Factor Curves 1 2 AND STP R-10C.Doc 06 0119.0817

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03/20/03 Page 2 of 2 STP R-10C (UNITS 1 AND 2)

ATTACHMENT 9.1 TITLE: Pressurizer Level and RCS TAVG Adjustment Factor Curves RCS TAVG CONVERSION FACTOR ADJUSTMENT FACTOR F2 1.15 1.10 1.05 1.00

.95

.90

.85

.80

.75 F2

.70

.65

.60

.55

.50

.45

.40

.35

.30

.25 100 150 200 250 300 350 400 450 500 550 600 Tavg 00982802 STP R-10C.Doc 06 0119.0817

- - UNCONTROLLED PROCEDURE - DO NOT USE TO PERFORM WORK or ISSUE FOR USE ***

03/20/03 Page 1 of 1 DIABLO CANYON POWER PLANT TITLE:

STP R-10C ATTACHMENT 9.2 Pressurizer Level Correction Curves for Pressurizer Pressures 1 2 AND STP R-10C.Doc 06 0119.0817

NUCLEAR POWER GENERATION DIABLO CANYON POWER PLANT JOB PERFORMANCE MEASURE Number: NRCADM01SRO

Title:

Review RCS WATER INVENTORY BALANCE Examinee:

Evaluator:

Print Signature Date Results: Sat Unsat Total Time: minutes Comments:

References:

STP R-10C, RCS Water Inventory Balance, Rev. 25 Technical Specifications 3.4.14, RCS Leakage Alternate Path: Yes No X Time Critical: Yes No X Time Allotment: 30 minutes Critical Steps: 4, 15, 16 Job Designation: RO/SRO Task Number: G2.1.33 Rating: 3.4/4.0 AUTHOR: JACK BLACKWELL DATE: 01/18/2005 REVIEWED BY: N/A DATE:

JPM COORDINATOR APPROVED BY: N/A DATE:

TRAINING LEADER REV. 0

JPM TITLE: REVIEW RCS WATER INVENTORY BALANCE JPM NO.: NRCADM01SRO INSTRUCTOR WORKSHEET Directions: No plant controls or equipment are to be operated during the performance of this Job Performance Measure. All actions taken by the examinee should be clearly demonstrated and verbalized to the evaluator. The student will be given the initial conditions, initiating cue, and task standard. The examiner will then ask if any clarifications are needed. After identifying the appropriate procedure for the task, the examinee may be given the procedure and told the step with which to begin.

Required Materials: STP R-10C, RCS Water Inventory Balance, Rev. 25 Technical Specifications 3.4.14 Initial Conditions: Unit 1 is at BOL, 100% power with the PPC out of service. No other equipment out of service and Zinc Injection is not in service. As the SFM, you have requested the RO to conduct a manual RCS leak evaluation per STP R-10C. The RO has completed the procedure and has returned it for the SFM to complete.

Initiating Cue: As the SFM, review and complete the STP R-10C and determine appropriate actions as needed.

Task Standard: The procedure is reviewed, data reduction and evaluation completed, and the primary review comleted and signed.

NRCADM01SRO.DOC Page 2 of 8 REV. 0

JPM TITLE: REVIEW RCS WATER INVENTORY BALANCE JPM NO.: NRCADM01SRO INSTRUCTOR WORKSHEET Start Time:

Step Expected Operator Actions

1. Verify Start Data Section filled 1.1 Confirms Unit 1, Mode 1and out properly. Date/Time correct.

1.2 Confirms Precaution and Limitations (Section 10) meet plant conditions and are all initialed.

1.3 Confirms Prerequisites (Section 11) meet plant conditions and initialed.

Step was: Sat: ______ Unsat _______*

2. Verify Procedure steps 2.1 Confirms Section 12.1.1 data is completed properly. recorded for Rx Power, Pressure and Temperature.

2.2 Confirms Pressure and Temperature above minimum required on steps 12.1.2 and 12.1.3.

2.3 Confirms Step 12.1.4 marked N/A.

2.4 Confirms Step 12.1.5 marked No.

2.5 Confirms Step 12.1.6 and 12.1.7 marked Yes and initialed, and 12.1.8 initialed.

2.6 Confirms Step 12.2 marked N/A.

2.7 Confirms Step 12.3.1 and 12.3.4 marked N/A, steps 12.3.2 and 12.3.3 initialed.

Step was: Sat: ______ Unsat _______*

Denotes an entry required on the JPM cover sheet.

- Denotes a Critical Step.

NRCADM01SRO.DOC Page 3 of 8 REV. 0

JPM TITLE: REVIEW RCS WATER INVENTORY BALANCE JPM NO.: NRCADM01SRO INSTRUCTOR WORKSHEET Step Expected Operator Actions

3. Verify Table 1 and Table 2 data 3.1 Confirms Table 1 and Table 2 data for accurate. B.A and P.W. integrator readings appear accurate.

3.2 Confirms Table 1 and Table 2 data for PZR Level, VCT level and TAVG appear accurate.

3.3 Confirms start and stop times greater than 2 hours2.314815e-5 days 5.555556e-4 hours 3.306878e-6 weeks 7.61e-7 months .

Step was: Sat: ______ Unsat _______*

- 4. Verifies Table 3 data accurate. 4.1 Confirms BA is 40 gal.

- 4.2 Confirms PW is 0 gal.

4.3 Confirms PZR is -217 gal.

4.4 Confirms VCT is -172.8 gal.

4.5 Confirms TAVG is 39.25 gal.

Step was: Sat: ______ Unsat _______*

5. Verify T calculation step 5.1 Confirms T is 240 minutes.

12.3.5.

Step was: Sat: ______ Unsat _______*

6. Verify Calculated RCS leak rate 6.1 Confirms Calculated RCS leak rate is calculation step 12.3.6.b. 1.954 gpm.

Step was: Sat: ______ Unsat _______*

7. Verify calculated Leak Error 7.1 Confirms Calculated Leak Error Factor calculation step 12.3.7.b. Factor at 0.317 gpm.

Step was: Sat: ______ Unsat _______*

8. Verify Gross Leak Rate 8.1 Confirms Gross Leak Rate calculated calculation step 12.3.8. at 2.271 gpm.

Step was: Sat: ______ Unsat _______*

Denotes an entry required on the JPM cover sheet.

- Denotes a Critical Step.

NRCADM01SRO.DOC Page 4 of 8 REV. 0

JPM TITLE: REVIEW RCS WATER INVENTORY BALANCE JPM NO.: NRCADM01SRO INSTRUCTOR WORKSHEET Step Expected Operator Actions

9. Determine Gross Leak Rate is >1 9.1 Confirms 12.3.9 marked No.

gpm and must continue procedure at Table 4.

9.2 Confirms 12.3.10 N/A.

9.3 Confirms 12.3.11 initialed.

Step was: Sat: ______ Unsat _______*

10. Verify Table 4 data accurate. 10.1 PRT is 123 gal.

10.2 RCDT is 10.25 gal.

10.3 RCDT totalizer is 0 gal.

10.4 Accumulators is 0 gal.

Step was: Sat: ______ Unsat _______*

11. Verify RCS Identified Leak Rate 11.1 Confirms Calculated Identified Leak step 12.3.12 (Table 4). Rate at 0.606 gpm.

Step was: Sat: ______ Unsat _______*

12. Verify Identified Leak Rate Error 12.1 Confirms Calculated Identified Leak Factor step 12.3.13. Rate Error Factor at 0.458 gpm.

Step was: Sat: ______ Unsat _______*

13. Verify Identified Leakage step 13.1 Confirms Identified Leakage at 1.064 12.3.14. gpm.

Step was: Sat: ______ Unsat _______*

14. Verify test performer signature 14.1 Confirms test performer printed and step 12.5. signed name with date, time and initial.

Step was: Sat: ______ Unsat _______*

Denotes an entry required on the JPM cover sheet.

- Denotes a Critical Step.

NRCADM01SRO.DOC Page 5 of 8 REV. 0

JPM TITLE: REVIEW RCS WATER INVENTORY BALANCE JPM NO.: NRCADM01SRO INSTRUCTOR WORKSHEET Step Expected Operator Actions

- 15. Perform step 13, Data Reduction 15.1 Checks N/A on step 13.1.

and Evaluation.

15.2 Calculates Total RCS leak rate error factor from Aux Board and Control Boards, step 13.2.1.b. at 0.557 gpm.

and initials step.

- 15.3 Calculates Unidentified Leakage at 1.918 gpm. and initials step.

NOTE: Critical Task met if Unidentified Leakage Rate calculated > 1gpm.

Step was: Sat: ______ Unsat _______*

- 16. Perform step 14, Primary 16.1 Does NOT initial step 14.1.

Review.

16.2 Makes entry in Remarks for Step 14.2 for any N/A or No entries.

NOTE: Anything is acceptable as long as it relates to the data collected.

- 16.3 Recognizes Acceptance Criteria NOT accepted and LCO T.S 3.4.13 requires leak reduction in 4 hours4.62963e-5 days 0.00111 hours 6.613757e-6 weeks 1.522e-6 months or Mode 3 in 6 hours6.944444e-5 days 0.00167 hours 9.920635e-6 weeks 2.283e-6 months .

16.4 Recognizes need to write AR.

Cue: The SM has initiated AR A0762222.

Step was: Sat: ______ Unsat _______*

Stop Time:

Total Time: (Enter total time on the cover page)

Denotes an entry required on the JPM cover sheet.

- Denotes a Critical Step.

NRCADM01SRO.DOC Page 6 of 8 REV. 0

JPM NO.: NRCADM01SRO EXAMINEE CUE SHEET Initial Conditions: Unit 1 is at BOL, 100% power with the PPC out of service. No other equipment out of service and Zinc Injection is not in service. As the SFM, you have requested the RO to conduct a manual RCS leak evaluation per STP R-10C. The RO has completed the procedure and has returned it for the SFM to complete.

Initiating Cue: As the SFM, review and complete the STP R-10C and determine appropriate actions as needed.

Task Standard: The procedure is reviewed, data reduction and evaluation completed, and the primary review comleted and signed.

NRCADM01SRO.DOC Page 7 of 8 REV. 0

JPM TITLE: REVIEW RCS WATER INVENTORY BALANCE JPM NO.: NRCADM01SRO ATTACHMENT 1, SIMULATOR SETUP The simulator is not needed for the performance of this JPM.

NRCADM01SRO.DOC Page 8 of 8 REV. 0

- - UNCONTROLLED PROCEDURE - DO NOT USE TO PERFORM WORK or ISSUE FOR USE ***

PACIFIC GAS AND ELECTRIC COMPANY NUMBER STP R-10C DIABLO CANYON POWER PLANT REVISION 25 PAGE 17 OF 18 TITLE: Reactor Coolant System Water Inventory Balance UNITS 1 AND 2 PERF

13. DATA REDUCTION AND EVALUATION NOTE: If the manual RCS LEAK calculation is performed, use the leak rate data from step 12.3.8.

13.1 If the Gross RCS leak rate from step 12.2.2 or 12.3.8 is less than 0.965 gpm if zinc injection is in progress or less than 1 gpm if zinc injection is NOT in progress, assume the following:

N/A [ ] ______

UNIDENTIFIED LEAKAGE < 1 gpm, and IDENTIFIED LEAKAGE < 10 gpm 13.2 If the Gross RCS leak rate from step 12.3.8 is greater than or equal to 1 gpm, record.

N/A [ ] ______

13.2.1 Calculated Total RCS Leak Rate Error Factor (EFTTL).

a. When readings are taken from the Aux Board + PPC, 26.48 gal 26.48 gal EFTTL = = = gpm Tmin min N/A [ ] ______

b. When readings are taken from the Aux Board + Control Boards, 133.6 gal 133.6 gal EFTTL = = = gpm Tmin min N/A [ ] ______

NOTE 1: T should be the same for Table 1 through Table 4. If not use the shortest T.

NOTE 2: If any control board indicator is used in Table 1 through Table 4, use 13.2.1b.

13.2.2 UNIDENTIFIED LEAKAGE = Calculated RCS Leak Rate - RCS Calculated Identified Leak Rate + EFTTL.

____________ - ___________ + ___________ =

Step 12.3.6 Step 12.3.12 EFTTL 13.2.1a or 13.2.1b UNIDENTIFIED LEAKAGE = ______________ ______

STP R-10C.Doc 06 0119.0817

- - UNCONTROLLED PROCEDURE - DO NOT USE TO PERFORM WORK or ISSUE FOR USE ***

PACIFIC GAS AND ELECTRIC COMPANY NUMBER STP R-10C DIABLO CANYON POWER PLANT REVISION 25 PAGE 18 OF 18 TITLE: Reactor Coolant System Water Inventory Balance UNITS 1 AND 2 PERF

14. PRIMARY REVIEW 14.1 Verify the acceptance criteria have been satisfied for the reactor coolant system leak rate.

UNIDENTIFIED LEAKAGE is less than 0.965 gpm if zinc injection is in progress or less than 1 gpm if zinc injection is NOT in progress. (Steps 13.1 or 13.2.2.) ______

IDENTIFIED LEAKAGE is less than 10 gpm.

(Steps 13.1 or 12.3.14) ______

14.2 REMARKS: Describe any malfunctions, explain any NO or N/A entries in any of the data and list any discrepancies.

14.3 Review the completed procedure.

If the acceptance criteria has not been satisfied, notify management promptly, write an Action Request and refer to applicable Technical Specifications limiting conditions for operations.

AR # ___________________________

Signature: Date/Time /

Shift Foreman

15. SECONDARY REVIEW 15.1 Review procedure for completeness and acceptability. ______

15.2 REMARKS:

Reviewed By: Date Second Reviewer STP R-10C.Doc 06 0119.0817

NUCLEAR POWER GENERATION DIABLO CANYON POWER PLANT JOB PERFORMANCE MEASURE Number: NRCADM02RO

Title:

PERFORM OUTAGE SAFETY CHECKLIST Examinee:

Evaluator:

Print Signature Date Results: Sat Unsat Total Time: minutes Comments:

References:

AD8.DC55, Outage Safety Scheduling, Rev. 19 OP AP SD-0, Loss of, or Inadequate Deca Heat Removal, Rev. 8 Alternate Path: Yes X No Time Critical: Yes No X Time Allotment: 10 minutes Critical Steps: 1 Job Designation: RO Task Number: 2.1.32 Rating: 3.4 AUTHOR: JACK BLACKWELL DATE: 01/18/2005 APPROVED BY: N/A DATE:

TRAINING LEADER APPROVED BY: N/A DATE:

LINE MANAGER REV. 1

JPM TITLE: PERFORM OUTAGE SAFETY CHECKLIST JPM NUMBER: NRCADM02RO INSTRUCTOR WORKSHEET Directions: No plant controls or equipment are to be operated during the performance of this Job Performance Measure. All actions taken by the examinee should be clearly demonstrated and verbalized to the evaluator. The student will be given the initial conditions, initiating cue, and task standard. The examiner will then ask if any clarifications are needed. After identifying the appropriate procedure for the task, the examinee may be given the procedure and told the step with which to begin.

o Student Handout and blank Outage Safety Checklist for Mode 6 Required Materials:

RCS Level Greater Than or Equal to 111 Initial Conditions: Unit 1 was in Mode 6 when a loss of off site power occurred. All three diesels started, but a fault on Bus H occurred, leaving that bus deenergized. Power was restored within 5 minutes and the plant was stabilized, with the exception of Bus H. Plant Conditions are as follows:

o MDAFW Pump 1-3 was cleared.

o S/G 1-1 and 1-4 were drained for SG cleaning related work.

o S/G 1-2 and 1-3 are at 35%.

o CFCUs 1-1 and 1-3 running.

Initiating Cue: The SFM has requested you to complete a new Outage Safety Checklist for Core Cooling for the new condition.

Task Standard: The Core Cooling Outage Safety Checklist for Mode 6 RCS Level Greater Than or Equal to 111 for current plant conditions is completed and SFM informed of results.

NRCADM02ROREV1. Page 2 of 5 REV.1 DOC

JPM TITLE: PERFORM OUTAGE SAFETY CHECKLIST JPM NUMBER: NRCADM02RO INSTRUCTOR WORKSHEET Start Time:

Step Expected Operator Actions

- 1. Review current Mode 6 Outage 1.1 Compare conditions in Initial Checklist with conditions after the Conditions with the current loss of offsite power. checklist.

- 1.2 Notes the following safety conditions NOT met:

o RHR pump 1-2 NOT operable o SI pump 1-2 NOT operable

- 1.3 Informs SFM Core Cooling function of Outage Safety Checklist NOT met.

Step was: Sat: ______ Unsat _______*

Stop Time:

Total Time: (Enter total time on the cover page)

Denotes an entry required on the JPM cover sheet.

- Denotes a Critical Step.

NRCADM02ROREV1. Page 3 of 5 REV. 1 DOC

JPM NUMBER: NRCADM02RO EXAMINEE CUE SHEET Initial Conditions: Unit 1 was in Mode 6 when a loss of off site power occurred. All three diesels started, but a fault on Bus H occurred, leaving that bus deenergized. Power was restored within 5 minutes and the plant was stabilized, with the exception of Bus H. Plant Conditions are as follows:

o MDAFW Pump 1-3 was cleared.

o S/G 1-1 and 1-4 were drained for SG cleaning related work.

o S/G 1-2 and 1-3 are at 35%.

o CFCUs 1-1 and 1-3 running.

Initiating Cue: The SFM has requested you to complete a new Outage Safety Checklist for Core Cooling for the new condition.

Task Standard: The Core Cooling Outage Safety Checklist for Mode 6 RCS Level Greater Than or Equal to 111 for current plant conditions is completed and SFM informed of results.

NRCADM02ROREV1. Page 4 of 5 REV. 1 DOC

JPM TITLE: PERFORM OUTAGE SAFETY CHECKLIST JPM NUMBER: NRCADM02RO ATTACHMENT 1, SIMULATOR SETUP No simulator associated with this JPM.

NRCADM02ROREV1. Page 5 of 5 REV. 1 DOC

Plant conditions PRIOR to loss of offsite power

NUCLEAR POWER GENERATION DIABLO CANYON POWER PLANT JOB PERFORMANCE MEASURE Number: NRCADM02SRO

Title:

REVIEW OUTAGE SAFETY CHECKLIST Examinee:

Evaluator:

Print Signature Date Results: Sat Unsat Total Time: minutes Comments:

References:

AD8.DC55, Outage Safety Scheduling, Rev. 19 Alternate Path: Yes X No Time Critical: Yes No X Time Allotment: 15 minutes Critical Steps: 1, 2 Job Designation: SRO Task Number: 2.1.32 Rating: 3.8 AUTHOR: JACK BLACKWELL DATE: 01/18/2005 APPROVED BY: N/A DATE:

TRAINING LEADER APPROVED BY: N/A DATE:

LINE MANAGER REV. 1

JPM TITLE: REVIEW OUTAGE SAFETY CHECKLIST JPM NUMBER:

NRCADM02SRO INSTRUCTOR WORKSHEET Directions: No plant controls or equipment are to be operated during the performance of this Job Performance Measure. All actions taken by the examinee should be clearly demonstrated and verbalized to the evaluator. The student will be given the initial conditions, initiating cue, and task standard. The examiner will then ask if any clarifications are needed. After identifying the appropriate procedure for the task, the examinee may be given the procedure and told the step with which to begin.

Required Materials: o Handouts of Mode 6 RCS Level Greater Than or Equal to 111 Initial Conditions: Unit 1 was in Mode 6 when a loss of off site power occurred. All three diesels started, but a fault on Bus H occurred, leaving that bus deenergized. Power was restored within 5 minutes and the plant was stabilized, with the exception of Bus H. Plant Conditions are as follows:

o MDAFW Pump 1-3 was cleared.

o S/G 1-1 and 1-4 were drained for SG cleaning related work.

o S/G 1-2 and 1-3 are at 35%

o CFCUs 1-1 and 1-3 are running The CO has just completed a new Outage Safety Checklist for current plant conditions.

Initiating Cue: The SFM has directed you to review the new Outage Safety Checklist for compliance to the Outage Safety Plan.

Task Standard: The Outage Safety Checklist for current plant conditions is reviewed and SFM informed of your findings.

NRCADM02SROREV Page 2 of 5 REV. 1 1.DOC

JPM TITLE: REVIEW OUTAGE SAFETY CHECKLIST JPM NUMBER:

NRCADM02SRO INSTRUCTOR WORKSHEET Start Time:

Step Expected Operator Actions

- 1. Review current Mode 6 Outage 1.1 Compare conditions in Initial Checklists. Conditions with the current checklist.

- 1.2 Identifies discrepancy with RHR pump 1-2 NOT being operable.

- 1.3 Recognizes Outage Safety Checklist NOT met with RHR 1-2 not operable.

Step was: Sat: ______ Unsat _______*

- 2. Reports discrepancies. ** 2.1 Informs SFM of findings.

Step was: Sat: ______ Unsat _______*

Stop Time:

Total Time: (Enter total time on the cover page)

Denotes an entry required on the JPM cover sheet.

- Denotes a Critical Step.

NRCADM02SROREV Page 3 of 5 REV. 1 1.DOC

JPM NUMBER: NRCADM02SRO EXAMINEE CUE SHEET Initial Conditions: Unit 1 was in Mode 6 when a loss of off site power occurred. All three diesels started, but a fault on Bus H occurred, leaving that bus deenergized. Power was restored within 5 minutes and the plant was stabilized, with the exception of Bus H. Plant Conditions are as follows:

o MDAFW Pump 1-3 was cleared.

o S/G 1-1 and 1-4 were drained for SG cleaning related work.

o S/G 1-2 and 1-3 are at 35%

o CFCUs 1-1 and 1-3 are running The CO has just completed a new Outage Safety Checklist for current plant conditions.

Initiating Cue: The SFM has directed you to review the new Outage Safety Checklist for compliance to the Outage Safety Plan.

Task Standard: The Outage Safety Checklist for current plant conditions is reviewed and SFM informed of your findings.

NRCADM02SROREV Page 4 of 5 REV. 1 1.DOC

JPM TITLE: REVIEW OUTAGE SAFETY CHECKLIST JPM NUMBER:

NRCADM02SRO ATTACHMENT 1, SIMULATOR SETUP No simulator associated with this JPM.

NRCADM02SROREV Page 5 of 5 REV. 1 1.DOC

NUCLEAR POWER GENERATION DIABLO CANYON POWER PLANT JOB PERFORMANCE MEASURE Number: NRCADM03RO

Title:

DETERMINE CLEARANCE POINTS Examinee:

Evaluator:

Print Signature Date Results: Sat Unsat Total Time: minutes Comments:

References:

OP2.ID1, Clearances, Rev. 12 OVID 106713 Alternate Path: Yes No X Time Critical: Yes No X Time Allotment: 15 minutes Critical Steps: 1 Job Designation: RO Task Number: G2.2.13 Rating: 3.6 AUTHOR: JACK BLACKWELL DATE: 01/18/2005 REVIEWED BY: N/A DATE:

JPM COORDINATOR APPROVED BY: N/A DATE:

TRAINING LEADER REV. 0

JPM TITLE: DETERMINE CLEARANCE POINTS JPM NUMBER: NRCADM03RO INSTRUCTOR WORKSHEET Directions: No plant controls or equipment are to be operated during the performance of this Job Performance Measure. All actions taken by the examinee should be clearly demonstrated and verbalized to the evaluator. The student will be given the initial conditions, initiating cue, and task standard. The examiner will then ask if any clarifications are needed. After identifying the appropriate procedure for the task, the examinee may be given the procedure and told the step with which to begin.

Required Materials: Access to Plant Diagrams and Schematics Initial Conditions: A leak on Spent Fuel Pool Cooling pump 1-2 requires maintenance. A Clearance request is part of the work package.

Initiating Cue: The WCSFM has asked you to determine the clearance points for this clearance.

Task Standard: The clearance points are determined and documented on the associated plant drawing.

NRCADM03RO.DOC Page 2 of 7 REV. 0

JPM TITLE: DETERMINE CLEARANCE POINTS JPM NUMBER: NRCADM03RO INSTRUCTOR WORKSHEET Start Time:

Step Expected Operator Actions

- 1. Determine Man On Line 1.1 The following are the MINIMUM clearance points required for SFP clearance points for this action:

1-2 Man On Line tag for SFP 1-2 suction valve 1-61 CLOSED Man On Line tag for SFP 1-2 discharge valve 1-63 CLOSED Man On Line tag for SFP 1-2 Normal Supply Breaker 52-1H-47 OPEN Man On Line tag for SFP 1-2 Backup Supply Breaker 52-1F-33 OPEN Step was: Sat: ______ Unsat _______*

2. Determine Caution clearance 2.1 Determine the following CAUTION points for SFP 1-2 tag points for the clearance.

SFP 1-2 Vent valve 1-66 OPEN SFP 1-2 Drain valve 1-68 OPEN CBI tag on pump controller NOTE: May include Pressure Indicator valves 1-64 and 1-65 as part of clearance, but are NOT required.

Step was: Sat: ______ Unsat _______*

Stop Time:

Total Time: (Enter total time on the cover page)

Denotes an entry required on the JPM cover sheet.

- Denotes a Critical Step.

NRCADM03RO.DOC Page 3 of 7 REV. 0

JPM NUMBER: NRCADM03RO EXAMINEE CUE SHEET Initial Conditions: A leak on Spent Fuel Pool Cooling pump 1-2 requires maintenance. A Clearance request is part of the work package.

Initiating Cue: The WCSFM has asked you to determine the clearance points for this clearance.

Task Standard: The clearance points are determined and documented on the associated plant drawing.

NRCADM03RO.DOC Page 4 of 7 REV. 0

JPM TITLE: DETERMINE CLEARANCE POINTS JPM NUMBER: NRCADM03RO ATTACHMENT 1, SIMULATOR SETUP The simulator is not needed for the performance of this JPM.

NRCADM03RO.DOC Page 5 of 7 REV. 0

NUCLEAR POWER GENERATION DIABLO CANYON POWER PLANT JOB PERFORMANCE MEASURE Number: NRCADM03SRO

Title:

SAFETY FUNCTION DETERMINATION Examinee:

Evaluator:

Print Signature Date Results: Sat Unsat Total Time: minutes Comments:

References:

OP1.DC38, Safety Function Determination Program, Rev. 1 AD4.ID8, Identification and Resolution of Loose, Missing or Damaged Fasterners, Rev. 9 T.S. 3.5.2, ECCS - Operating Alternate Path: Yes X No Time Critical: Yes No X Time Allotment: 15 minutes Critical Steps: 3, 5, 6 Job Designation: SRO Task Number: G2.2.24 Rating: 3.8 AUTHOR: JACK BLACKWELL DATE: 01/18/2005 REVIEWED BY: N/A DATE:

JPM COORDINATOR APPROVED BY: N/A DATE:

TRAINING LEADER REV. 0

JPM TITLE: SAFETY FUNCTION DETERMINATION JPM NUMBER: NRCADM03SRO INSTRUCTOR WORKSHEET Directions: No plant controls or equipment are to be operated during the performance of this Job Performance Measure. All actions taken by the examinee should be clearly demonstrated and verbalized to the evaluator. The student will be given the initial conditions, initiating cue, and task standard. The examiner will then ask if any clarifications are needed. After identifying the appropriate procedure for the task, the examinee may be given the procedure and told the step with which to begin.

Required Materials: OP1.DC38, Safety Function Determination Program AD4.ID8, ID and Resolution of Loose, Missing or Damaged Fasterners Technical Specifications Initial Conditions: Units 1 and 2 are in Mode 1. Unit 1 SSPS Train A Master Relay Testing has been in progress for 3.5 hours5.787037e-5 days 0.00139 hours 8.267196e-6 weeks 1.9025e-6 months . The test was originally scheduled for 3 hours3.472222e-5 days 8.333333e-4 hours 4.960317e-6 weeks 1.1415e-6 months , however a problem developed that will require an additional 4 hours4.62963e-5 days 0.00111 hours 6.613757e-6 weeks 1.522e-6 months to complete repairs. The BOPCO has reported that the upper front door for SIP 1-2 breaker cubicle is missing two bolts, and the bottom front door is missing one bolt. There are no maintenance workers in the area, and no work is in progress. No other equipment is out of service.

Initiating Cue: As the SFM, determine operability and safety function, and any appropriate actions.

Task Standard: The safety function and operability of the affected equipment and any appropriate actions determined.

NRCADM03SRO.DOC Page 2 of 6 REV. 0

JPM TITLE: SAFETY FUNCTION DETERMINATION JPM NUMBER: NRCADM03SRO INSTRUCTOR WORKSHEET Start Time:

Step Expected Operator Actions

1. Obtain the correct procedures. 1.1 References AD4.ID8.

Step was: Sat: ______ Unsat _______*

2. Determine operability of the 2.1 Determines the cubicle is cubicle. INOPERABLE per step 7.1.1.2, 7.1.1.5.a and 7.1.2.2.a.

Cue: IF ASKED, the bolt holes are damaged. Repairs will take approximately 4 hours4.62963e-5 days 0.00111 hours 6.613757e-6 weeks 1.522e-6 months .

Step was: Sat: ______ Unsat _______*

- 3. Determine SIP 1-2 operability. 3.1 Determines SIP 1-2 INOPERABLE per step 7.1.1 NOTE 2. **

Step was: Sat: ______ Unsat _______*

4. Determine operability of the 4.1 References T.S. Table 3.3.2-1 SSPS Train A. Function 1.b and Condition/Required Action C.

4.2 Determines SSPS Train A is INOPERABLE.

Step was: Sat: ______ Unsat _______*

Denotes an entry required on the JPM cover sheet.

- Denotes a Critical Step.

NRCADM03SRO.DOC Page 3 of 6 REV. 0

JPM TITLE: SAFETY FUNCTION DETERMINATION JPM NUMBER: NRCADM03SRO INSTRUCTOR WORKSHEET Step Expected Operator Actions

- 5. Determine SIP 1-1 operability. 5.1 References OP1.DC38 Attachment 8.1.

5.2 Determines SSPS Train A is a support system for SIP 1-1.

5.3 Determines SIP 1-1 is INOPERABLE. **

NOTE: May determine SIP 1-1 inoperable through T.S. 3.5.2 and that the ACTION statement cannot be met, placing the plant in T.S.

3.0.3 Step was: Sat: ______ Unsat _______*

- 6. Determine Safety Function NOT 6.1 Using Attachment 8.2, determines a met. LOSF may exist.

6.2 Determines that both trains of SIP are inoperable, therefore entrance into T.S. 3.0.3 is required.

NOTE: May determine SIP 1-1 inoperable through T.S. 3.5.2 and that the ACTION statement cannot be met, placing the plant in T.S.

3.0.3 6.3 Determines required actions for T.S.

3.0.3 are to implement actions within 1 hour1.157407e-5 days 2.777778e-4 hours 1.653439e-6 weeks 3.805e-7 months to place the unit in Mode 3 within 7 hours8.101852e-5 days 0.00194 hours 1.157407e-5 weeks 2.6635e-6 months . **

Step was: Sat: ______ Unsat _______*

Stop Time:

Total Time: (Enter total time on the cover page)

Denotes an entry required on the JPM cover sheet.

- Denotes a Critical Step.

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JPM NUMBER: NRCADM03SRO EXAMINEE CUE SHEET Initial Conditions: Units 1 and 2 are in Mode 1. Unit 1 SSPS Train A Master Relay Testing has been in progress for 3.5 hours5.787037e-5 days 0.00139 hours 8.267196e-6 weeks 1.9025e-6 months . The test was originally scheduled for 3 hours3.472222e-5 days 8.333333e-4 hours 4.960317e-6 weeks 1.1415e-6 months , however a problem developed that will require an additional 4 hours4.62963e-5 days 0.00111 hours 6.613757e-6 weeks 1.522e-6 months to complete repairs. The BOPCO has reported that the upper front door for SIP 1-2 breaker cubicle is missing two bolts, and the bottom front door is missing one bolt. There are no maintenance workers in the area, and no work is in progress. No other equipment is out of service.

Initiating Cue: As the SFM, determine operability and safety function, and any appropriate actions.

Task Standard: The safety function and operability of the affected equipment and any appropriate actions determined.

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JPM TITLE: SAFETY FUNCTION DETERMINATION JPM NUMBER: NRCADM03SRO ATTACHMENT 1, SIMULATOR SETUP The simulator is not needed for the performance of this JPM.

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Pacific Gas & Electric Company AD4.ID8 Nuclear Power Generation Rev 9 Diablo Canyon Administrative Procedure Page 1 of 10 Identification and Resolution of 7/30/04 Loose, Missing, or Damaged Fasteners Effective Date Sponsoring Organization: Procedure Services Procedure Classification: Quality Related Review Level: "A"

1. SCOPE .................................................................................................................................................. 1

2. DEFINITION .......................................................................................................................................... 2

3. RESPONSIBILITIES ............................................................................................................................. 2

4. INSTRUCTIONS ................................................................................................................................... 2 4.1 IDENTIFYING FASTENER PROBLEMS ............................................................................................... 2 4.2 RESOLVING FASTENER PROBLEMS ................................................................................................ 3 4.3 EVALUATION OF FASTENER PROBLEMS .......................................................................................... 4

5. RECORDS ............................................................................................................................................ 5

6. REFERENCES...................................................................................................................................... 5

7. VITAL 4KV SWITCHGEAR - GUIDANCE FOR EVALUATING OPERABILITY APPENDIX ............... 6 7.1 CUBICLE OPERABILITY .................................................................................................................. 6 7.2 BUS OPERABILITY......................................................................................................................... 8 7.3 SISI CONCERNS ........................................................................................................................ 10 ATTACHMENTS:

1. Index, 07/15/04

1. SCOPE

1) This procedure establishes the requirements for identifying, evaluating and resolving loose, missing or damaged fasteners.T34349/T34350

2) This procedure establishes requirements that allow vital 4kV cubicles to remain operable with cubicle doors open.

3) This procedure applies to fasteners on equipment or systems that:

a) Have OPERABILITY requirements in the technical specifications or equipment control guidelines (ECG).

b) Are covered by the quality assurance program, a graded quality assurance program, or are seismically qualified.

NOTE: Equipment and systems meeting these conditions are classified QA Class Q, R, G, S, T in the Q-List and PIMS component database.

c) Are required by the Seismically Induced System Interaction Program.

4) This procedure may be applied to fasteners on other equipment or systems at the option of the shift foreman.

AD4_ID8.Doc

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Identification and Resolution of AD4.ID8 R9 Loose, Missing, or Damaged Fasteners Page 2 of 10

5) This procedure does not apply to:

a) Loose, missing or damaged fasteners discovered on equipment or systems cleared for maintenance.

NOTE: If these conditions are problems, they should be reported per OM7.ID1.

b) Fasteners lost or damaged during maintenance.

2. DEFINITION Fasteners Screws, bolting material, clips or retaining pins used in or on plant structures, systems and components. Fasteners do not include crimped lug wiring connectors.

Not Properly Installed A term applied to 4kV switchgear bolting when in the following conditions:

The bolt is not fully tightened

The backing washer on the bolt can be rotated by hand

3. RESPONSIBILITIES Individuals discovering loose, missing or damaged fasteners Responsible for initiating an action request.

Operations and Maintenance Responsible for assessing the risk of maintaining 4kV cubicles operable with open doors.

Operations and Engineering Responsible for evaluating the effect of fasteners problems on equipment operability.

4. INSTRUCTIONS 4.1 IDENTIFYING FASTENER PROBLEMS NOTE: Paragraph 4.2.3 may be performed prior to initiating the action request.T34879

1) The individual who discovers a loose, missing or damaged fastener shall:

a) Initiate an action request per OM7.ID1.

AR subtype should be FAST.

b) Bag and tag any loose pieces or parts of a fastener assembly.

(1) On the tag, note the AR number and other pertinent information concerning the fastener, such as its probable location.

(2) Enter the bags storage location on the action request.

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Identification and Resolution of AD4.ID8 R9 Loose, Missing, or Damaged Fasteners Page 3 of 10 4.2 RESOLVING FASTENER PROBLEMS 4.2.1 General

1) Except as noted in paragraph 4.2.3, plan and correct loose or missing fastener problems per AD7.DC8.

2) Procurement and installation of missing or damaged fasteners shall be per AD7.DC8.

3) The following information may be useful in resolving fastener problems or determining operability of some equipment:T34350/T34878

Engineering calculation SQE-42 provides guidance for evaluating panels or covers of cabinets. (SQE-42 is located in RMS at RLOC 04502/4851 through 5792.)

Section 7 provides guidance for evaluating vital 4kV switchgear.

4) If necessary, assistance with or evaluation of any fastener problem may be requested from engineering.

4.2.2 MOV Fasteners

1) Do not tighten any of the following loose fasteners:

Actuator-to-yoke bolting

Yoke-to-bonnet bolting

Body-to-bonnet bolting NOTE: Tightening these fasteners can modify the stiffness of the MOV assembly and invalidate votes sensor calibrations.

2) If any of the above MOV bolting is found loose or missing contact a valve engineer before tightening or reinstalling the fastener.T35171

3) Plan and correct these fastener problems per AD7.DC8.

4.2.3 Non-MOV Fasteners

1) All Fasteners a) Upon discovery, any fastener may be tightened or reinstalled provided:

(1) The function, location, and material type of the fastener are known.

(2) A procedure, drawing, manual, etc. that provides requirements for installing or tightening the fastener is used.

(3) Shift foreman authorization is obtained per OP1.DC18.

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Identification and Resolution of AD4.ID8 R9 Loose, Missing, or Damaged Fasteners Page 4 of 10 b) If any fastener is tightened or reinstalled upon discovery, include the following information on the "FAST" AR.

(1) The "as-found" condition and fastener location with sufficient detail (i.e.,

location, component ID, size of fastener, etc.) to allow someone to find the fastener.

(2) The "as-left" condition and process for installing and/or tightening the fastener including any special tightening requirements, procedure, drawing, manual, etc.

used.

c) If a fastener is tightened or reinstalled upon discovery, maintenance should evaluate the fasteners "as-left" condition to determine if additional corrective action is necessary.

2) 1/2" Diameter Fasteners a) These fasteners may be tightened or reinstalled upon discovery without using a procedure, drawing, manual, etc. if the fastener:

(1) Provides a mechanical function only, and (2) Performs no adjustment function.

b) The following actions are allowed:

(1) A loose fastener may be tightened to snug tight.

(2) A fastener that has fallen out of a panel or cabinet and is recovered may be reinstalled and tightened to snug tight.

4.3 EVALUATION OF FASTENER PROBLEMS

1) When notified, the shift foreman shall review the "FAST" AR and evaluate equipment operability. This evaluation, based on general knowledge of machinery, supports and connections, should address the following operability questions:

Does the loose, missing or damaged fastener have the potential for affecting the operation of systems or equipment as described in the technical specifications or ECGs?

Does the loose, missing or damaged fastener have a direct affect on the operability of systems or equipment? For example, increased vibration, binding, etc.

2) If assistance is needed in determining operability, the shift manager should proceed with the operability assessment per OM7.ID12, "Operability Determination."

a) Details of the problem may be initially communicated by telephone, but the problem shall be documented on an AR/AE.

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Identification and Resolution of AD4.ID8 R9 Loose, Missing, or Damaged Fasteners Page 5 of 10

5. RECORDS None

6. REFERENCES

1) Developmental references are listed in background information document BID AD4.ID8.

This document is in EDMS, NPG Manual, Admin Procedure Info.

2) Licensing Position - Open Doors on Seismically Qualified Cabinets, Revision 2

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Identification and Resolution of AD4.ID8 R9 Loose, Missing, or Damaged Fasteners Page 6 of 10

7. VITAL 4KV SWITCHGEAR - GUIDANCE FOR EVALUATING OPERABILITY APPENDIX When applying the guidance for evaluating vital 4kV switchgear operability, the objective is to ensure vital components are operable based on the bus remaining structurally and electrically operable. The following guidance can be used to navigate through the various sections of this appendix.

To maintain the See the following for requirements Section 7.1 Component or cubicle operable.

Table 1 Section 7.2.1 Bus structurally operable.

Table 2 Section 7.2.2 Bus electrically operable.

Table 3 7.1 CUBICLE OPERABILITY 7.1.1 Open Doors to Support Maintenance NOTE 1: The term "cubicle door" or "cubicle doors" does not include panels.

NOTE 2: The term "cubicle" includes the breaker, the component fed by that breaker, and the components installed in the cubicle.

1) Opening 4kV cubicles doors and maintaining the cubicle operable can be risk significant and should be assessed for risk by maintenance and operations per MA1.DC11 and/or AD7.DC6, as applicable. Troubleshooting should be assessed per MA1.DC10.

2) Provided the conditions of paragraphs 5) and 6) below are met, 4kV bus cubicles may remain OPERABLE when cubicle doors are open to support maintenance.

3) If the conditions of paragraphs 5) and 6) below are not met when cubicle doors are open, the cubicle is INOPERABLE.

4) The following table specifies cubicle configurations and bus combinations allowed when cubicles are operable with open doors.

Table 1: Operable Cubicles with Open Doors Allowed cubicle Allowed cubicles Allowed number Mode configurations per bus of busses 1-4 See Table 2 2 1 5, 6, and See Table 2 2 3 Defueled

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5) In modes 1 - 4, to maintain cubicle operability, the following conditions shall be met:

a) The shift foreman shall grant permission, per OP1.DC18, to open cubicle doors. If the cubicle doors will be open more than one shift, shift foreman permission shall be obtained at the beginning of each shift the cubicle doors are open.

b) The open cubicle shall be attended at all times by a person who is familiar with the maintenance. Anytime the cubicle is unattended, the cubicle doors shall be shut and properly bolted.

c) Cubicle doors should not be open longer than 24 continuous hours.

NOTE: The 24 hour2.777778e-4 days 0.00667 hours 3.968254e-5 weeks 9.132e-6 months period is a nominal period rather than a to-the-minute period.

This guideline is intended to control the duration doors on operable cubicles are open. The expectation is that reasonable efforts will be made to ensure cubicle doors are not open longer than the guideline.

6) In modes 5, 6, and defueled, to maintain cubicle operability, the following conditions shall be met:T36309 a) The shift foreman shall grant permission, per OP1.DC18, to open cubicle doors. If the cubicle doors will be open more that one shift, shift foreman permission shall be obtained at the beginning of each shift the cubicle doors are open.

b) The open cubicle shall be attended at all times by a person who is familiar with the maintenance. Anytime the cubicle is unattended, the cubicle doors shall be shut and properly bolted.

NOTE: The risk assessment for operable cubicles with open doors in modes 5,6, and defueled determined that there is an insignificant increase in risk. Therefore, there is no time limit for having operable cubicle doors open.

7.1.2 Loose or Missing Bolting

1) If a cubicle door or panel is not fully bolted, the cubicle may be INOPERABLE.

2) The criteria for determining operability of an individual cubicle are:

a) Upper front door bolting -- If two or more bolts are missing or not properly installed, the cubicle is INOPERABLE.

b) Lower front door bolting -- If two or more bolts are missing or not properly installed, the cubicle is INOPERABLE.

c) Back door bolting -- If three or more bolts are missing or not properly installed, the cubicle is INOPERABLE.

NOTE 1: For the purposes of determining operability, the vertical panel located just above the back door is be part of the back door.

NOTE 2: The loose or missing bolt criteria stated in a, b, & c above is stand alone and cannot be combined with each other. That is, multiple bolt problems, other than as stated, can make the cubicle inoperable and requires evaluation.

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Identification and Resolution of AD4.ID8 R9 Loose, Missing, or Damaged Fasteners Page 8 of 10 7.2 BUS OPERABILITY NOTE: Bus maintenance that makes bus G inoperable does not make breaker 52-HG-15, startup power, or cross-tie capability inoperable. (See AR A0477404) 7.2.1 Structural Integrity

1) Cubicles a) Two cubicles in any bus can have open doors, open panels, or missing or improperly installed bolts without affecting operability of the entire bus. Refer to paragraph 5) below for allowable configurations.

b) If three or more cubicles have open doors, open panels, or missing or improperly installed bolts, the entire bus is INOPERABLE.

2) Top Horizontal Panel Bolts a) Up to four bolts can be missing or improperly installed on one or two cubicles in a bus without affecting bus operability.

b) If bolts are missing or improperly installed on three or more cubicles, the entire bus is INOPERABLE.

Example:

Three bolts missing or improperly installed on one cubicle and one bolt missing or improperly installed on another cubicle is acceptable.

One missing or improperly installed bolt on each of three cubicles in a bus makes the bus INOPERABLE.

3) Side (End) Panel Bolts a) If two or more bolts are missing or improperly installed on the side (end) panel of a bus, the bus is INOPERABLE.

b) Two bolts, one side (end) panel bolt and one top panel bolt, can be missing or improperly installed without affecting bus operability.

4) PT Drawers a) The auxiliary feeder PT drawers and startup feeder PT drawers are structurally independent from the 4kV bus; therefore, these PT drawers may be opened without affecting bus operability.

b) One PT drawer on a bus may be opened at any time without affecting bus operability.

c) If more than one PT drawer is opened on a bus, the bus is INOPERABLE.

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Identification and Resolution of AD4.ID8 R9 Loose, Missing, or Damaged Fasteners Page 9 of 10

5) Breaker/Ground Buggy Positions a) The door/panel configurations in the following table relate the text of paragraphs 1) through 3) above to breaker/ground buggy positions that have been seismically analyzed to maintain bus structural integrity, thus bus operability.

(1) Maintaining an acceptable breaker/ground buggy position is required for both operable and inoperable cubicles.

(2) Two cubicles per bus can have doors or panels in the indicated configuration.

These door/panel configurations are stand alone and cannot be combined with each other.

(3) Unless otherwise indicated, the breaker/ground buggy positions are analyzed for modes 1 - 6 and defueled.

(4) Having a cubicle with the front and rear door open and the breaker racked in(up) is not analyzed. This configuration makes the bus INOPERABLE.

b) Other configurations may be acceptable, contact engineering for evaluation.

Table 2: Acceptable Breaker/Ground Buggy Positions Door/Panel Configuration Breaker/Ground Buggy Position 1 Each cubicle has the front door open. 2 Breaker rolled out (cubicle empty)

Each cubicle has the rear door open. Breaker down on the floor One cubicle has the front door open and the Ground buggy racked in (up) other cubicle has the rear door open. 2 Breaker racked in (up)

Breaker rolled out (cubicle empty)

Each cubicle has the rear door open and one top Breaker down on the floor panel removed.

Ground buggy racked in (up)

Mode 1-4 Breaker rolled out (cubicle empty)

Breaker down on the floor Each cubicle has the front & rear doors open Mode 5, 6, defueled and both top panels removed. 2 Breaker rolled out (cubicle empty)

Breaker down on the floor Ground buggy racked in (up) 1 Any ONE of the indicated breaker/ground buggy positions is acceptable 2

Front door means both the upper and lower cubicle doors.

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Identification and Resolution of AD4.ID8 R9 Loose, Missing, or Damaged Fasteners Page 10 of 10 7.2.2 Electrical Operability

1) Certain cubicles have door mounted relays that can affect bus operability and the auto transfer scheme. To maintain bus operability or auto transfer capability with these cubicle doors open, the conditions stated in section 7.1.1 shall be met.

2) If the conditions stated in section 7.1.1 are not met, use the following table to determine the affect on equipment operability, bus operability, or auto transfer capability when the indicated 4kV cubicle conditions exist.

Table 3: Bus Operability Impact Matrix Cubicle Condition 1 Operability Impact Tech Spec for the component fed by the Front cubicle door open 2 cubicle breaker Auto transfer to D/G and Front CCW cubicle door open 2 Auto transfer to startup Front D/G cubicle door open 2 Auto transfer to D/G Bus and Front AUX feeder cubicle door open 2 Auto transfer to startup and Auto transfer to D/G Bus and Front STARTUP feeder cubicle door open 2 Auto transfer to startup and Auto transfer to D/G 1

More than one condition may be applicable 2

Front cubicle door means both the upper and lower doors.

7.3 SISI CONCERNS

1) If a breaker is racked out, it may remain in the cell.

2) If the breaker is removed from the cell, the breaker shall be stored in the exciter switchgear room.

a) EXCEPTION 1: During protective relay functional testing, a breaker may be in the TEST position, outside the cubicle and in the 4kV switchgear room.

Comply with the restrictions specified in action request A0400674.

These restrictions are specified in the procedures used for protective relay functional testing.

b) EXCEPTION 2: Following an engineering analysis for floor loading and seismic considerations, storage of breakers in the vital switchgear rooms is permitted.

Floor loading and seismic interaction evaluations are performed by engineering.

See MA1.ID7, "Control of Plant Floor Loading."

Each breaker stored in a vital switchgear room shall have an INFO tag which:

States where the engineering analysis is documented.

Specifies an individual to be contacted in case questions arise.

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(07/15/04) AD4.ID8 Attachment 1 Page 1 of 1 Index discovery, 2 A MOV, 3 non-MOV - 1/2 diameter, 4 action request, initiate, 2 non-MOV - all, 3 B

P bag parts, 2 procedure usage, 3 breaker 52-HG-15, 8 PT drawers, 8 bus G operability, 8 bus operability, 8 S

C shift foreman evaluation, 4 SISI concerns, 10 cross-tie capability, 8 startup power, 8 cubicle operability during maintenance, 6 F

fastener problem correcting, 3 PG&E Diablo Canyon

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PACIFIC GAS AND ELECTRIC COMPANY NUMBER OP1.DC38 NUCLEAR POWER GENERATION REVISION 1 DIABLO CANYON POWER PLANT PAGE 1 OF 10 ADMINISTRATIVE PROCEDURE TITLE: Safety Function Determination Program 07/02/04 EFFECTIVE DATE PROCEDURE CLASSIFICATION: QUALITY RELATED SPONSORING ORGANIZATION: OPERATIONS REVIEW LEVEL: "A"

1. SCOPE 1.1 This procedure implements the Safety Function Determination Program (SFDP) as required by TS 5.5.15.

2. DISCUSSION 2.1 The purpose of the SFDP is to ensure that the proper actions are taken upon failure to meet one or more TS LCOs. It is also the goal of this program to ensure that the allowed out of service time of supported systems is not inappropriately extended as a result of multiple, overlapping support system inoperabilities.

2.2 TS LCO 3.0.2 establishes that upon discovery of a failure to meet an LCO, the associated ACTIONS shall be met. TS LCO 3.0.6 specifies that when a supported system LCO is not met solely due to a support system LCO not being met, the Conditions and Required Actions associated with this supported system are not required to be entered. Only the support system LCO ACTIONS are required to be entered. This is an exception to LCO 3.0.2 for the supported system. If this option is exercised, a safety function determination evaluation shall be made in accordance with TS 5.5.15.

2.3 When a support systems Required Actions directs a supported system to be declared inoperable or directs entry into Conditions and Required Actions for a supported system, LCO 3.0.2 shall be followed.

3. DEFINITIONS 3.1 Support System A support system is a structure, system, or component (SSC) required by Technical Specifications, which provides support for supported system(s) in order for the supported system(s) to perform its safety function. An example of a support system required by Technical Specifications would be Component Cooling Water (CCW). CCW supports the Residual Heat Removal (RHR) system by providing cooling to the pumps and heat exchangers.

An SSC that monitors or maintains a process parameter or operating limit is not a support system for the purpose of implementing TS LCO 3.0.6. For example, if the rod position deviation monitor is inoperable, this does not automatically mean that the control rods are no longer within their required alignment. A process parameter or an operating limit is not a support system. For example, exceeding control rod insertion limits does not automatically mean that hot channel factors are out of limits.

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PACIFIC GAS AND ELECTRIC COMPANY NUMBER OP1.DC38 DIABLO CANYON POWER PLANT REVISION 1 PAGE 2 OF 10 TITLE: Safety Function Determination Program 3.2 Supported System A supported system is a structure, system, or component (SSC) required by Technical Specifications, which requires a support system to ensure its safety function can be performed. Process parameters and operating limits are not supported systems for the purpose of implementing TS LCO 3.0.6.

A support system can also be a supported system. For example, the CCW system supports RHR system operation. As such the CCW system is a support system.

However, the Auxiliary Saltwater System (ASW) supports operation of the CCW system to remove heat. In this case the CCW system is a supported system.

3.3 Safety Function In the SFDP, safety function refers to intended function of the component or system to provide mitigation for those accidents previously analyzed and licensed for DCPP. The safety function for a component or system covered by an TS LCO can be obtained from the applicable TS Bases or in the FSAR. A single component or system may be covered by more than one TS LCO and have more than one safety function.

3.4 Loss of Safety Function 3.4.1 A loss of safety function exists when, assuming no concurrent single failure, and assuming no concurrent loss of offsite power or loss of onsite diesel generators, a safety function assumed in the accident analysis cannot be performed for the mode of applicability. If a loss of safety function is determined to exist by this program, the appropriate Conditions and Required Actions of the LCO in which the loss of safety function exists are required to be entered.

3.4.2 For the purpose of this program, a loss of safety function may exist when a support system is inoperable and:

a. a required system redundant to the system(s) supported by the inoperable support system is also inoperable.

b. a required system redundant to the system(s) in turn supported by the inoperable supported system is also inoperable.

c. a required system redundant to the support system(s) for the supported systems (a) and (b) above is also inoperable.

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PACIFIC GAS AND ELECTRIC COMPANY NUMBER OP1.DC38 DIABLO CANYON POWER PLANT REVISION 1 PAGE 3 OF 10 TITLE: Safety Function Determination Program Generic Diagrammatic Example:

Train A Train B System 1 System 1 Case b System 2 System 2 Case a System 3 (Support System Inoperable) System 3 System 4 System 4 Case c 3.4.3 Due to the 3 vital bus and the cross connected design of the CCW/ASW system, DCPP does not neatly fit into the generic example above. The purpose of the program is to ensure that sufficient cross train checks are performed to ensure that inoperabilities of redundant components (functions) in both trains do not go undetected.

3.5 Safety Function Determination Program This is a program required by TS 5.5.15 to detect a loss of safety function and ensure that appropriate TS actions are implemented.

3.6 Cascading Technical Specifications When a support system is inoperable such that it results in a supported system inoperability the option always exists to enter the Conditions and Required Actions of the LCO for both systems. This is referred to as cascading technical specification Conditions and Required Actions. However, LCO 3.0.6 provides the option to only enter the support system LCO Conditions and Required Actions provided a loss of safety function has not occurred.

4. RESPONSIBILITIES 4.1 The SFM is responsible for:

4.1.1 Determination if implementation of TS LCO 3.0.6 is appropriate for the existing plant conditions and if allowed by the particular support system that is inoperable. Some technical specifications provide actions on declaring supported systems inoperable upon discovery of support system inoperability.

4.1.2 Performing a loss of safety function determination required by TS Administrative Controls 5.5.15, if appropriate.

4.1.3 Ensuring that no inappropriate completion time extensions exist due to multiple support system inoperabilities.

5.

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PACIFIC GAS AND ELECTRIC COMPANY NUMBER OP1.DC38 DIABLO CANYON POWER PLANT REVISION 1 PAGE 4 OF 10 TITLE: Safety Function Determination Program INSTRUCTIONS 5.1 The Safety Function Determination Program (SFDP) as implemented by this procedure does not change the way in which operability of technical specification equipment is determined. The shift foreman (SFM) shall continue to use the guidance provided in OP1.DC17, "Control of Equipment Required by the Plant Technical Specifications," to evaluate individual equipment operability.

5.2 Entry into TS LCO 3.0.6 can be considered whenever the SFM declares a support TS support structure, system, or components (SSC) inoperable. This in no way precludes the shift manager (SM) or SFM from implementing TS LCO 3.0.2 and tracking the Conditions and Required Actions for all supported equipment affected by the support system inoperability. This is referred to as "cascading" technical specifications and is allowed by TS. Review the following criteria prior to implementing TS LCO 3.0.6.

5.2.1 The unit is in Modes 1 - 4. DCPP will only enter TS LCO 3.0.6 in Modes 1- 4.

It has been determined that use of TS LCO 3.0.6 is not advantageous in Modes 5 and 6. See Step 5.4 for more explanation.

5.2.2 Determine if the support systems LCO requires direct entry into the supported systems TS LCO. If so, enter all applicable Required Actions of the support and supported systems TS LCOs. TS LCO 3.0.6 cannot be invoked for that supported system TS LCO.

5.2.3 If the failure of an TS required support system results in the inoperability of an TS supported system, then LCO 3.0.6 may be applied.

5.2.4 If the failure of an TS required support system results in the inoperability of a system outside of TS, and that system is subsequently relied upon by an TS supported system to remain OPERABLE, then LCO 3.0.6 may be applied.

5.2.5 TS LCO 3.0.6 cannot be applied when solely a non TS support item makes an TS LCO item inoperable. There are no Required Actions of the support item to provide the level of protection required for application of TS LCO 3.0.6.

5.2.6 If the failure of an ECG required support system results in the inoperability of an TS supported system, then the ECG and TS LCO Required Actions are required to be entered. TS LCO 3.0.6 cannot be applied.

5.2.7 If there are other support systems which are contributing to the supported systems inoperability, then TS LCO 3.0.6 may NOT be applied without first considering each of the other support systems separately to ensure no loss of safety functions exists.

5.2.8 If the inoperable SSC is not directly addressed by an TS LCO and does not impact the operability of an TS LCO, then no further action with regard to a LOSF evaluation is required and this procedure may be exited.

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PACIFIC GAS AND ELECTRIC COMPANY NUMBER OP1.DC38 DIABLO CANYON POWER PLANT REVISION 1 PAGE 5 OF 10 TITLE: Safety Function Determination Program 5.3 Inoperability of a support system does not necessarily render a supported system inoperable. For example:

5.3.1 Declaring CCW Pump 2 inoperable does not render either RHR pump 1 or 2 inoperable due to the cross connected design of the CCW system.

5.3.2 Supported systems are not declared inoperable when an instrumentation support system TS LCO is not met, unless the failure results in a loss of actuation capability or the support system's Required Action directs the supported system to be declared inoperable.

5.3.3 Supported systems are not declared inoperable solely as a result of inoperability of the normal or emergency electrical power source. The Required Actions for inoperable electrical power sources provide the necessary restrictions.

5.4 TS LCO 3.0.6 does not limit the modes of applicability for implementation of SFDP to only Modes 1-4. However, for simplicity DCPP will not use SFDP in Modes 5 and 6.

This will require the SFM and SM to consider the effect on supported systems when a support system is inoperable and cascade technical specifications as appropriate. It is assumed that implementation of SFDP in Modes 5 and 6 is not advantageous since most systems only require a single train for the safety function to be met. In those cases where 2 trains are required (e.g., RHR when loops are not filled in Mode 5), the SFM needs to address multiple system inoperability and take the TS Required Actions for all support and supported systems.

5.5 An LOSF evaluation is required if TS LCO 3.0.6 is invoked after considering the criteria of Step 5.2. The LOSF evaluation must be performed as soon as practical for each inoperable TS support or TS supported system.

5.6 Documentation of the LOSF evaluation shall be in the PIMS TS tracking module of the inoperable support equipment evaluated.

5.7 If an LOSF is determined to exist, the appropriate Conditions and Required Actions of the LCO in which the LOSF exists shall be entered. If no Condition within the LCO addresses the LOSF, then TS LCO 3.0.3 shall be entered.

5.8 A considerable amount of judgment may be required to perform an LOSF evaluation.

Attachment 8.2, "SFDP Worksheet," is optional for determining if an LOSF evaluation is required. The attachment poses questions to guide the SFM/SM in determining if a more detailed analysis of a loss of safety function is required.

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PACIFIC GAS AND ELECTRIC COMPANY NUMBER OP1.DC38 DIABLO CANYON POWER PLANT REVISION 1 PAGE 6 OF 10 TITLE: Safety Function Determination Program Attachment 8.1, Support System - Supported System Matrix," provides a cross reference of identified support system LCO to supported system LCO relationships. This list is for reference and may be overly conservative depending on the exact cause for declaring the support system inoperable. However, the control room staff can use this list to quickly determine the potential for an LOSF. When a support system TS LCO is not met, Attachment 8.1 can be used to check if any of the listed supported system's TS LCOs are not met. If any supported system Conditions and Required Actions are currently in effect then a more detailed analysis for an LOSF must be performed. This analysis will consist of checking that the supported TS safety function is identified (check TS Bases) and still available assuming no concurrent single failure or loss of offsite power.

5.8.1 Examples

a. Unit 1 is in Mode 1 and the SFM has declared RHR Pp 1-1 inoperable due to a clearance for maintenance. RHR Pp 1-1 is an SSPS Train B actuated component.

The asset team has scheduled reactor trip breaker testing for SSPS Train B during the same shift that the RHR pump is cleared. The SFM reviews Attachment 8.1 and determines that TS LCO 3.3.2 is a support system for TS LCO 3.5.2. Closer inspection reveals that the same train is affected and there is no loss of safety function. This evaluation is documented in the PIMS TS tracking module.

b. Unit 2 is in Mode 1 and the SFM has declared the spray additive tank inoperable. The SFM reviews Attachment 8.1 and determines that TS LCO 3.6.7 is not listed as a support system for any other TS LCO.

Since the spray additive system is NOT a support system for the containment spray system and there are no other TS LCO Conditions in effect, no LOSF exists. This evaluation is documented in the PIMS TS tracking module.

c. Unit 1 is in Mode 1 and the SFM has authorized SSPS Train A testing that makes that train inoperable. A nuclear operator doing rounds in the turbine building discovers a problem with SIP 1-2 4kV breaker cubicle rendering SIP 1-2 inoperable. TS LCO 3.3.2 is listed as support system for TS LCO 3.5.2. In this condition automatic initiation of SIP 1-1 is prevented due to SSPS testing and SIP 1-2 will not start due to a breaker problem. In this case the LOSF evaluation would show a loss of safety function due to both SIPs inoperable and TS LCO 3.0.3 would be entered.

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PACIFIC GAS AND ELECTRIC COMPANY NUMBER OP1.DC38 DIABLO CANYON POWER PLANT REVISION 1 PAGE 7 OF 10 TITLE: Safety Function Determination Program 5.9 Some TS LCOs have Conditions and Required Actions that require technical specification cascading. When a support system's Required Action directs a supported system to be declared inoperable or directs entry into Conditions and Required Actions for a supported system, the applicable conditions and Required Actions are entered in accordance with TS LCO 3.0.2. It should be noted that an LOSF evaluation is still required for the remaining inoperable supported system TS LCOs (Modes 1-4). The directed technical specification cascading may not cover all affected safety functions.

5.10 Common Support Systems 5.10.1 RWST

a. It is recognized that if the RWST is inoperable due to insufficient inventory or inadequate chemical concentration, the acceptance criteria for certain design basis accidents may not be met. Neither the ECCS nor containment spray system can meet their design function with the RWST outside the required TS LCO limits. Since this is clearly stated in the TS Bases for the RWST, the appropriate action is to follow the TS Required Actions for an inoperable RWST and not to enter TS LCO 3.0.3 for the ECCS. The LOSF evaluation will conclude that although there is a degradation in the ECCS, there is not a loss of safety function as long as useable inventory is present. The RWST Required Actions are bounding for this case.

5.10.2 CST/FWST

a. The AFW system will not be able to perform its design function without a supply of water for RCS decay heat removal via the SGs. The Required Actions for inoperability of the CST or FWST is more restrictive than for the case if all three AFW trains are inoperable. The appropriate action is to follow the TS Required Actions for an inoperable CST/FWST and not to enter the Required Actions for an inoperable AFW system. The LOSF evaluation will conclude that although there is a degradation for maintaining an AFW heat sink there is not a loss of safety function as long as there is useable inventory. The CST/FWST Required Actions are bounding for this case.

5.10.3 ULTIMATE HEAT SINK (UHS)

a. The ultimate heat sink provides a heat sink for transferring heat from safety related components during a transient or accident, as well as safety related and nonsafety related heat loads during normal operation. The ASW system is a supported system of the UHS. If the UHS is inoperable the capability to remove heat by the ASW system is impacted. ASW system performance will be degraded with an inoperable UHS but this alone does not make the AWS system inoperable as long as the UHS does not exceed 70degreesF. The appropriate TS Required Actions to ensure the plant is maintained in a safe condition are the Required Actions of TS 3.7.9. Entry into TS 3.7.8 and TS LCO 3.0.3 are not required.

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PACIFIC GAS AND ELECTRIC COMPANY NUMBER OP1.DC38 DIABLO CANYON POWER PLANT REVISION 1 PAGE 8 OF 10 TITLE: Safety Function Determination Program 5.10.4 Diesel Fuel Oil, Starting and Turbo Air

a. The DFO storage volume is based on 7 days of minimum ESF loads during a loss off offsite power. In the event insufficient DFO volume is available the Required Action is to restore inventory within 48 hours5.555556e-4 days 0.0133 hours 7.936508e-5 weeks 1.8264e-5 months .

The diesel generators are supported systems. Although the diesel generators would not be able to support minimum ESF loads for 7 days, they are still meeting their safety function as long as they are running.

The LOSF evaluation for this case concludes that the system is degraded but the safety function of the DGs are met as long as there is useable volume in either DFO storage tank.

b. The TS minimum requirements for DG starting air and turbo air ensure that there is sufficient air capacity for 3 successive DG start attempts. If air pressure is less than 180 psig but greater than 150 psig, there is adequate capacity for one start attempt and the DG can be considered operable until the Completion Time for the Condition expires. The TS LCO Conditions and Required Actions direct declaring any DGs inoperable should they not have at least 150 psig in one starting air receiver or the turbo air receiver. An LOSF evaluation is not required since the TS LCO 3.8.3 Required Actions does not consider the associated DG inoperable within the Completion Time. After expiration of the Completion Time, a directed entry to declare the DG inoperable is required. In this case, entry into TS LCO 3.0.6 is not allowed.

5.11 Cross Connected Systems Since CCW and ASW are cross connected cooling systems, pumps and heat exchangers do not have strict train relationship with respect to cooling ECCS equipment. It should be noted that there is no analysis for one CCW pump during design basis accidents. If one CCW pump is out of service and the DG associated with an operable CCW pump becomes inoperable, TS 3.8.1 Condition B, Required Action B.2 requires declaring that CCW pump inoperable within 4 hours4.62963e-5 days 0.00111 hours 6.613757e-6 weeks 1.522e-6 months . With no vital CCW loop available the SFM will direct entry in TS LCO 3.0.3. Application of TS LCO 3.0.6 is inappropriate in this case.

5.12 If an inoperable support SSC is covered by an ECG and this SSC makes an TS supported system inoperable, it is not allowed to invoke TS LCO 3.0.6. The supported system Conditions and Required Actions must be followed if the supported system is inoperable due to the ECG support system inoperability. TS LCO 3.0.6 is dependent on support system Conditions and Required Actions providing the appropriate level of safety and compensatory actions for supported system inoperability. The ECGs have not been reviewed to provide this level of safety.

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PACIFIC GAS AND ELECTRIC COMPANY NUMBER OP1.DC38 DIABLO CANYON POWER PLANT REVISION 1 PAGE 9 OF 10 TITLE: Safety Function Determination Program 5.13 Supported System Maximum Completion Time NOTE: A supported system Completion Time may only be extended when there is no Loss of Safety Function.

5.13.1 A supported system made inoperable by support system inoperabilities may only remain inoperable for a limited period of time without entering the supported system's Conditions and Required Actions. This time limit is defined as the Maximum Completion Time. The Maximum Completion Time for restoring the supported system to operable status is the Completion Time specified for restoration of the first inoperable support system plus the Completion Time specified for the inoperable supported system.

a. If the supported system is not restored to operable status by restoring the support system(s) to operable status within the Maximum Completion Time, enter the Condition and Required Actions for the inoperable supported system's Completion Time not met.

5.13.2 Example of Completion Time Extension

a. The unit is in Mode 1 when 480 V bus H becomes deenergized due to a feeder breaker problem. TS LCO 3.8.9 specifies that this bus must be restored to operable status in 8 hours9.259259e-5 days 0.00222 hours 1.322751e-5 weeks 3.044e-6 months or a shutdown to Mode 3 is required in the next 6 hours6.944444e-5 days 0.00167 hours 9.920635e-6 weeks 2.283e-6 months . 480 V bus H (TS LCO 3.8.9) is a support system for containment isolation valves (TS LCO 3.6.3). The seal return penetration has CVCS-8112 (bus H) inside containment and CVCS-8100 (bus G) outside containment. TS LCO 3.6.3 Required Actions gives the operator 4 hours4.62963e-5 days 0.00111 hours 6.613757e-6 weeks 1.522e-6 months to complete action to restore or isolate the penetration before a shutdown is required. An LOSF evaluation would conclude that there is no loss of safety function since CVCS-8100 is still powered and able to function on a phase A isolation signal assuming no concurrent single failure or loss of offsite power. The Conditions and Required Actions of TS LCO 3.6.3 are not required to be performed for 8 hours9.259259e-5 days 0.00222 hours 1.322751e-5 weeks 3.044e-6 months since its support system Required Actions are in effect. After 8 hours9.259259e-5 days 0.00222 hours 1.322751e-5 weeks 3.044e-6 months , then the Required Actions of TS LCO 3.6.3 are applied. Isolation of the seal return penetration, assuming power is not restored to CVCS-8112, must be completed within 12 hours1.388889e-4 days 0.00333 hours 1.984127e-5 weeks 4.566e-6 months .

5.13.3 Documentation and Tracking of Maximum Completion Time

a. Initiate a PIMS TS Tracking Sheet listing the supported system Maximum Completion Time for the following instances:

1. The support system restoration Completion Time has expired, or

2. Multiple support system inoperabilities have occurred affecting the same supported system.

6.

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PACIFIC GAS AND ELECTRIC COMPANY NUMBER OP1.DC38 DIABLO CANYON POWER PLANT REVISION 1 PAGE 10 OF 10 TITLE: Safety Function Determination Program RECORDS The LOSF evaluation will be documented in the PIMS TS Tracking Module.

7. REFERENCES 7.1 TS LCO 3.0.6 7.2 TS 5.5.15

8. ATTACHMENTS 8.1 "Support System - Supported System Matrix," 05/21/04 8.2 "SFDP Worksheet," 02/28/2000 OP1_DC38.Doc 01B 0119.1449

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05/21/04 Page 1 of 8 DIABLO CANYON POWER PLANT OP1.DC38 ATTACHMENT 8.1 TITLE: Support System - Supported System Matrix Support Support System Supported Supported System System System TS LCO TS LCO Number 3.3.2 Engineered Safety Feature 3.3.6 Containment Ventilation Isolation Actuation System Instrumentation (ESFAS) instrumentation 3.3.7 Control Room Ventilation System (CRV)

Actuation Instrumentation 3.5.2 ECCS - Operating 3.5.3 ECCS - Shutdown 3.6.3 Containment Isolation Valves 3.6.6 Containment Spray and Cooling Systems 3.6.7 Spray Additive System 3.7.2 Main Steam Isolation Valves 3.7.3 Main Feedwater Isolation, Regulating, and Bypass Valves 3.7.5 Auxiliary Feedwater System 3.7.7 Vital Component Cooling Water (CCW)

System 3.7.8 Auxiliary Saltwater (ASW) System 3.7.12 Auxiliary Building Ventilation System (ABVS) 3.8.1 AC Sources - Operating 3.3.4 Remote Shutdown System 3.4.9 Pressurizer 3.7.5 Auxiliary Feedwater System 3.5.2 ECCS - Operating 3.5.3 ECCS - Shutdown 3.7.7 Vital Component Cooling Water (CCW)

System 3.7.8 Auxiliary Saltwater (ASW) System 3.8.1 AC Sources - Operating 3.8.2 AC Sources - Shutdown 3.3.51 Loss of Power (LOP) Diesel 3.8.1 AC Sources - Operating Generator (DG) Start 3.8.2 AC Sources - Shutdown Instrumentation 1

Required Action A.1 directs entering applicable Condition(s) and Required Action(s) for the associated DG made inoperable by LOP DG start instrumentation (TS LCOs 3.8.1 and 3.8.2), therefore TS LCO 3.0.6 does not apply.

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05/21/04 Page 2 of 8 OP1.DC38 ATTACHMENT 8.1 TITLE: Support System - Supported System Matrix Support Support System Supported Supported System System System TS LCO TS LCO Number 3.3.62 Containment Ventilation Isolation 3.6.3 Containment Isolation Valves.

Instrumentation 3.9.4 Containment Penetrations 3

3.3.7 Control Room Ventilation System 3.7.10 Control Room Ventilation (CRVS)

(CRVS) Actuation Instrumentation 3.3.8 Fuel Building Ventilation System 3.7.13 Fuel Handling Building Ventilation (FBVS) Actuation System (FHVS)

Instrumentation 3.4.144 RCS Pressure Valve (PIV) 3.4.6 RCS Loops - Mode 4 Leakage 3.4.13 RCS Operational Leakage 3.5.2 ECCS - Operating 3.5.3 ECCS - Shutdown 5 5 3.5.4 Refueling Water Storage Tank 3.5.2 ECCS - Operating 5

(RWST) 3.5.3 ECCS - Shutdown 5

3.6.6 Containment Spray and Cooling Systems 6

3.6.2 Containment Air Locks 3.6.1 Containment 2

Required Action B.1 directs entering applicable conditions and Required Actions of LCO 3.6.3 "Containment Isolation Valves," for containment isolation valves made inoperable by isolation instrumentation. Required Action C.2 directs entering Conditions and Required Actions of TS LCO 3.9.4, "Containment Penetrations,"

for containment ventilation isolation valves made inoperable by isolation instrumentation. Therefore TS LCO 3.0.6 is not applicable for these cases.

3 Required Action B.1.2 directs entering applicable Conditions and Required Actions for one CRVS train made inoperable by inoperable CRVS actuation instrumentation (TS LCO 3.7.10). In this case TS LCO 3.0.6 does not apply.

4 Note 2 requires entering applicable Conditions and Required Actions for systems made inoperable by an Inoperable PIV.

5 Although the RWST is a support system of the ECCS and Containment Spray System, TS 3.5.4 contains sufficient Required Actions. See Step 5.10 for explanation.

6 Note 3 directs entering applicable Conditions and Required Actions of TS LCO 3.6.1, "Containment," when air lock leakage results in exceeding the overall containment leakage rate, therefore TS LCO 3.0.6 does not apply.

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05/21/04 Page 3 of 8 OP1.DC38 ATTACHMENT 8.1 TITLE: Support System - Supported System Matrix Support Support System Supported Supported System System System TS LCO TS LCO Number 3.6.37 Containment Isolation Valves 3.5.2 ECCS - Operating 3.5.3 ECCS - Shutdown 3.6.1 Containment 3.6.6 Containment Spray and Cooling Systems 3.6.6 Containment Spray and Cooling 3.6.7 Spray Additive System Systems 3.7.4 10% Atmospheric Dump Valves 3.4.5 RCS Loops-Mode 3 (ADVs) 3.4.6 RCS Loops-Mode 4 3.7.5 Auxiliary Feedwater (AFW) 3.4.5 RCS Loops-Mode 3 System 3.4.6 RCS Loops-Mode 4 8

3.7.6 Condensate Storage Tank (CST) 3.7.5 Auxiliary Feedwater (AFW) System and Fire Water Storage Tank (FWST) 3.7.79 Component Cooling Water 3.4.6 RCS Loops-Mode 4 (CCW) System 3.4.7 RCS Loops-Mode 5, Loops Filled 3.4.8 RCS Loops-Mode 5, Loops Not Filled 3.5.2 ECCS - Operating 3.5.3 ECCS - Shutdown 3.6.6 Containment Spray and Cooling Systems 3.9.5 Residual Heat Removal (RHR) and Coolant Circulation - High Water Level 3.9.6 Residual Heat Removal (RHR) and Coolant Circulation - Low Water Level 7

Note 3 directs entering applicable Conditions and Required Actions for systems made inoperable by containment isolation valves. Note 4 directs entering applicable Conditions and Required Actions of TS LCO 3.6.1, "Containment," when isolation valve leakage results in exceeding the overall containment leakage rate Acceptance Criteria. Therefore TS LCO 3.0.6 does not apply for these cases.

8 Although the CST/FWST is a support system for AFW, TS 3.7.6 contains sufficient Required Actions. See Step 5.10 for explanation.

9 Required Action A.1 Note directs entering applicable Conditions and Required Actions of TS LCO 3.4.6, "RCS Loops - Mode 4," for residual heat removal loops made inoperable by CCW. TS LCO 3.0.6 does not apply to TS LCO 3.4.6.

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05/21/04 Page 4 of 8 OP1.DC38 ATTACHMENT 8.1 TITLE: Support System - Supported System Matrix Support Support System Supported Supported System System System TS LCO TS LCO Number 3.7.810 Auxiliary Saltwater (ASW) 3.7.7 Component Cooling Water (CCW) System System 3.7.911 Ultimate heat Sink (UHS) 3.7.8 Auxiliary Saltwater (ASW) System 3.7.12 Auxiliary Building Ventilation 3.4.6 RCS Loops-Mode 4 System (ABVS) 3.4.7 RCS Loops-Mode 5, Loops Filled 3.4.8 RCS Loops-Mode 5, Loops Not Filled 3.5.2 ECCS - Operating 3.5.3 ECCS - Shutdown 3.7.7 Component Cooling Water (CCW) System 3.9.5 Residual Heat Removal (RHR) and Coolant Circulation - High Water Level 3.9.6 Residual Heat Removal (RHR) and Coolant Circulation - Low Water Level 3.8.112 AC Sources - Operating 3.5.2 ECCS - Operating 3.5.3 ECCS - Shutdown 3.6.6 Containment Spray and Cooling Systems 3.7.5 Auxiliary Feedwater (AFW) System 3.7.7 Vital Component Cooling Water (CCW)

System 3.7.8 Auxiliary Saltwater (ASW) System 3.7.10 Control Room Ventilation System (CRVS) 3.8.7 Inverters - Operating 3.8.913 Distribution Systems - Operating 10 Required Action A.1 Note directs entering applicable Conditions and Required Actions of TS LCO 3.4.6, "RCS Loops - Mode 4," for residual heat removal loops made inoperable by ASW. TS LCO 3.0.6 does not apply to TS LCO 3.4.6.

11 Although the UHS is a support system for ASW, TS 3.7.9 contains sufficient Required Actions. See Step 5.10 for explanation.

12 Required Actions B.2 and C.1 direct declaring required feature(s) inoperable when its required redundant feature(s) is inoperable.

13 A DG inoperable or an offsite circuit inoperable to an ESF bus does not result in TS LCO 3.8.9 not being met.

An LOSF is only required when all AC sources to the ESF bus are inoperable.

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05/21/04 Page 5 of 8 OP1.DC38 ATTACHMENT 8.1 TITLE: Support System - Supported System Matrix Support Support System Supported Supported System System System TS LCO TS LCO Number 3.8.2 AC Sources - Shutdown 3.4.7 RCS Loops-Mode 5, Loops Filled 3.4.8 RCS Loops-Mode 5, Loops Not Filled 3.7.10 Control Room Ventilation System (CRVS) 3.8.8 Inverters - Shutdown 3.8.1013 Distribution Systems - Shutdown 3.9.5 Residual Heat Removal (RHR) and Coolant Circulation - High Water Level 3.9.6 Residual Heat Removal (RHR) and Coolant Circulation - Low Water Level 3.8.314,15 Diesel Fuel Oil, Lube Oil, 3.8.1 AC Sources - Operating Starting Air, and Turbocharger 3.8.2 AC Sources - Shutdown Air Assist 3.8.416 DC Sources - Operating 3.8.1 AC Sources - Operating 3.8.7 Inverters - Operating 3.8.9 Distribution Systems - Operating 16 3.8.5 DC Sources - Shutdown 3.8.2 AC Sources - Shutdown 3.8.8 Inverters - Shutdown 3.8.10 Distribution Systems - Shutdown 17 3.8.6 Battery Cell Parameters 3.8.4 DC Sources - Operating 3.8.5 DC Sources - Shutdown 13 A DG inoperable or an offsite circuit inoperable to an ESF bus does not result in TS LCO 3.8.9 not being met.

An LOSF is only required when all AC sources to the ESF bus are inoperable.

14 Required Actions G.1 and H.1 direct declaring the associated DG inoperable. The associated DG is still considered operable until the Required Action and associated Completion Times of TS LCO 3.8.3 are not met.

TS LCO 3.0.6 does not apply.

15 Although the DFO Storage tank is a support system of the DGs, the safety function of DGs is satisfied as long as DGs are loaded or can be started and loaded. See Step 5.10 for explanation.

16 An LOSF evaluation is only required when all DC sources to the vital bus are inoperable.

17 The affected battery is still considered OPERABLE until the Required Action and associated Completion Time of TS 3.8.6 are not met. TS LCO 3.0.6 does not apply.

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05/21/04 Page 6 of 8 OP1.DC38 ATTACHMENT 8.1 TITLE: Support System - Supported System Matrix Support Support System Supported Supported System System System TS LCO TS LCO Number 3.8.718 Inverters - Operating 3.8.9 Distribution Systems - Operating 3.8.8 Inverters - Shutdown 3.8.10 Distribution Systems - Shutdown 19 3.8.9 Distribution Systems - Operating 3.1.7 Rod Position Indication 3.3.1 Reactor Trip System (RTS)

Instrumentation 3.3.2 Engineered Safety Feature Actuation 3.3.3 Post Accident Monitoring (PAM)

Instrumentation 3.3.4 Remote Shutdown System 3.3.5 Loss of Power (LOP) Diesel Generator (DG) Start Instrumentation 3.3.6 Containment Ventilation Isolation Instrumentation 3.3.7 Control room ventilation System (CRVS)

Actuation Instrumentation 3.3.8 Fuel Building Ventilation System (FBVS)

Actuation Instrumentation 3.4.4 RCS Loops - Modes 1 and 2 3.4.5 RCS Loops - Modes 3 3.4.6 RCS Loops - Modes 4 3.4.9 Pressurizer 3.4.11 Pressurizer Power Operated Relief valves (PORVs) 3.4.12 Low Temperature Overpressure Protection (LTOP) System 3.4.15 RCS Leakage Detection Instrumentation 3.5.2 ECCS - Operating 3.5.3 ECCS - Shutdown 18 Required Action A.1 Note directs entering applicable Conditions and Required Actions of TS LCO 3.8.9, "Distribution Systems - Operating" with any vital 120 V AC bus deenergized. TS LCO 3.0.6 does not apply.

19 There is no redundant system for DRPI, enter and follow the Required Actions for TS LCO 3.1.7.

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05/21/04 Page 7 of 8 OP1.DC38 ATTACHMENT 8.1 TITLE: Support System - Supported System Matrix Support Support System Supported Supported System System System TS LCO TS LCO Number 3.8.9 Distribution Systems - Operating 3.7.3 Main Feedwater Isolation Valves (continued) (MFIVs), Main Feedwater Regulating Valves (MFRVs), and MFRV Bypass Valves 3.7.5 Auxiliary Feedwater (AFW) System 3.7.7 Vital Component Cooling Water (CCW)

System 3.7.8 Auxiliary Saltwater (ASW) System 3.7.10 Control Room Ventilation System (CRVS) 3.7.12 Auxiliary building Ventilation System (ABVS) 3.8.3 Diesel Fuel Oil, Lube Oil, and Starting Air 20 3.8.4 DC Sources - Operating 3.8.7 Inverters - Operating 3.8.10 Distribution Systems - Shutdown 3.3.1 Reactor Trip System (RTS)

Instrumentation 3.3.6 Containment ventilation Isolation Instrumentation 3.3.7 Control Room Ventilation System (CRVS)

Actuation Instrumentation 3.4.7 RCS Loops - Mode 5, Loops Filled 3.4.8 RCS Loops - Mode 5, Loops Not Filled 3.4.12 Low Temperature Overpressure Protection (LTOP) System 3.7.10 Control Room Ventilation System (CRVS) 3.7.13 Fuel Handling Building Ventilation System (FHBVS) 20 Although the Required Actions for TS LCO 3.8.4 could be delayed per TS LCO 3.0.6, it is considered imperative by the TS bases for TS LCO 3.8.4 to place the battery charger on a backup source within 2 hours2.314815e-5 days 5.555556e-4 hours 3.306878e-6 weeks 7.61e-7 months .

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05/21/04 Page 8 of 8 OP1.DC38 ATTACHMENT 8.1 TITLE: Support System - Supported System Matrix Support Support System Supported Supported System System System TS LCO TS LCO Number 3.6.3 Containment Isolation Valves 3.6.6 Containment Spray and Cooling Systems 3.8.10 Distribution Systems - Shutdown 3.8.3 Diesel Fuel Oil, Lube Oil, and Starting Air (continued) 3.8.5 DC Sources - Shutdown 3.9.3 Nuclear Instrumentation 3.9.4 Containment Penetrations 3.9.5 Residual Heat Removal (RHR) and Coolant circulation - High Water Level 3.9.6 Residual Heat Removal (RHR) and Coolant circulation - High Water Level OP1_DC38.Doc 01B 0119.1449

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02/28/2000 Page 1 of 2 DIABLO CANYON POWER PLANT OP1.DC38 ATTACHMENT 8.2 TITLE: SFDP Worksheet

1. Loss of Safety Function (LOSF) Evaluation

a. Has the Unit entered the Required Actions of more than one TS LCO?1

If No, then no LOSF exists. No further evaluation and action is required.

If Yes, Continue with next Step 1.b.

b. Has the Unit entered the Required Actions of other TS LCOs for redundant train equipment?

If No, then no LOSF exists. No further evaluation required.

If Yes, Continue with next Step 1.c.

c. Has the Unit entered the Required Actions of other TS LCOs for redundant train support or supported equipment applicable to this LCO (consult Attachment 8.1)?

If No, then no LOSF exists. No further evaluation required.

If Yes, a LOSF may exist. Perform a LOSF evaluation to ensure that redundant safety equipment is not affected by the support system inoperability.

SUPPORTED SYSTEM REQUIRED ACTION ENTRY TABLE SUPPORT INOPERABLE AFFECTED MAXIMUM INOPERABLE SYSTEM TS LCO TIME/DATE SUPPORTED COMPLETION TIME TIME/DATE OF SYSTEM TS LCO ALLOWED= THE AFFECTED SUPPORTED SUPPORT SYSTEM SYSTEM AOT + SUPPORTED SYSTEM AOT 1

This question provides simplistic screening criteria for an LOSF evaluation. In general, if this is the only TS LCO Condition the unit has entered, then all safety functions should be preserved. Always consider and evaluate common support systems, (i.e., RWST, CST/FWST - See Step 5.11) and single power supply supported systems such as DRPI. The supported system TS LCO list should always be consulted when entering an TS LCO condition to ensure that redundant equipment is available.

OP1_DC38.Doc 01B 0119.1449

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02/28/2000 Page 2 of 2 OP1.DC38 ATTACHMENT 8.2 TITLE: SFDP Worksheet ESF EQUIPMENT POWER SUPPLIES and SSPS TRAIN RELATIONSHIP SAFETY BUS FUNCTION (SSPS Trn)

Vital Bus F Vital Bus G Vital Bus H High head safety injection CCP 1(Trn A) CCP 2 (Trn B)

Medium head safety injection SIP 1 (Trn A) SIP 2 (Trn B)

Low head safety injection RHR Pp 1 (Trn B) RHR Pp 2 (Trn A)

Ultimate heat sink cooling ASW Pp 1 (Trn A) ASW Pp 2 (Trn B)

ESF and decay heat removal CCW Pp 1 (Trn A) CCW Pp 2 (Trn B) CCW Pp 3 (Trn A & B)

Heat sink inventory AFW Pp 3 (Trn A) AFW Pp 2 (Trn B)

Containment cooling CFCU 1, 2 (Trn A) CFCU 3, 5 (Trn B) CFCU 4 (Trn A & B)

Containment cooling CSP 1 (Trn B) CSP 2 (Trn A)

Emergency Vital Power DG 3(Trn A) DG 2 (1) (Trn B) DG 1 (2) (Trn A & B)

OP1_DC38.Doc 01B 0119.1449

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ECCS - Operating 3.5.2 3.5 EMERGENCY CORE COOLING SYSTEMS (ECCS) 3.5.2 ECCS - Operating LCO 3.5.2 Two ECCS trains shall be OPERABLE.

APPLICABILITY: MODES 1, 2, and 3.

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

In MODE 3, both safety injection (SI) pump flow paths may be isolated by closing the isolation valve(s) for up to 2 hours2.314815e-5 days 5.555556e-4 hours 3.306878e-6 weeks 7.61e-7 months to perform pressure isolation valve testing per SR 3.4.14.1.

ACTIONS CONDITION REQUIRED ACTION COMPLETION TIME A. One or more trains A.1 Restore train(s) to ---------NOTE-------

inoperable. OPERABLE status The Completion Time may be extended to AND 7 days for Unit 1 cycle At least 100% of the ECCS 12 for centrifugal flow equivalent to a single charging pump 1-1 OPERABLE ECCS train seal replacement available. -------------------------

72 hours8.333333e-4 days 0.02 hours 1.190476e-4 weeks 2.7396e-5 months B. Required Action and B.1 Be in MODE 3. 6 hours6.944444e-5 days 0.00167 hours 9.920635e-6 weeks 2.283e-6 months associated Completion Time not met. AND B.2 Be in MODE 4. 12 hours1.388889e-4 days 0.00333 hours 1.984127e-5 weeks 4.566e-6 months DIABLO CANYON - UNITS 1 & 2 3.5-3 Unit 1 - Amendment No. 135, 159 TS 3_5_2.Doc - R6 3 Unit 2 - Amendment No. 135, 146, 160

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ECCS - Operating 3.5.2 SURVEILLANCE REQUIREMENTS SURVEILLANCE FREQUENCY SR 3.5.2.1 Verify the following valves are in the listed position 12 hours1.388889e-4 days 0.00333 hours 1.984127e-5 weeks 4.566e-6 months with power to the valve operator removed.

Number Position Function 8703 Closed RHR to RCS Hot Legs 8802A Closed Safety Injection to RCS Hot Legs 8802B Closed Safety Injection to RCS Hot Legs 8809A Open RHR to RCS Cold Legs 8809B Open RHR to RCS Cold Legs 8835 Open Safety Injection to RCS Cold Legs 8974A Open Safety Injection Pump Recirc.

to RWST 8974B Open Safety Injection Pump Recirc.

to RWST 8976 Open RWST to Safety Injection Pumps 8980 Open RWST to RHR Pumps 8982A Closed Containment Sump to RHR Pumps 8982B Closed Containment Sump to RHR Pumps 8992 Open Spray Additive Tank to Eductor 8701 Closed RHR Suction 8702 Closed RHR Suction SR 3.5.2.2 Verify each ECCS manual, power operated, and 31 days automatic valve in the flow path, that is not locked, sealed, or otherwise secured in position, is in the correct position.

SR 3.5.2.3 Verify ECCS piping is full of water. 31 days (continued)

DIABLO CANYON - UNITS 1 & 2 3.5-4 Unit 1 - Amendment No. 135 TS 3_5_2.Doc - R6 4 Unit 2 - Amendment No. 135

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ECCS - Operating 3.5.2 SURVEILLANCE REQUIREMENTS (continued)

SURVEILLANCE FREQUENCY SR 3.5.2.4 Verify each ECCS pump's developed head at the test In accordance with flow point is greater than or equal to the required the Inservice developed head. Testing Program.

SR 3.5.2.5 Verify each ECCS automatic valve in the flow path 24 months that is not locked, sealed, or otherwise secured in position, actuates to the correct position on an actual or simulated actuation signal.

SR 3.5.2.6 Verify each ECCS pump starts automatically on an 24 months actual or simulated actuation signal.

SR 3.5.2.7 Verify, for each ECCS throttle valve listed below, 24 months each mechanical position stop is in the correct position.

Charging Injection Safety Injection Throttle Valves Throttle Valves 8810A 8822A 8810B 8822B 8810C 8822C 8810D 8822D SR 3.5.2.8 Verify, by visual inspection, each ECCS train 24 months containment recirculation sump suction inlet is not restricted by debris and the suction inlet trash racks and screens show no evidence of structural distress or abnormal corrosion.

DIABLO CANYON - UNITS 1 & 2 3.5-5 Unit 1 - Amendment No. 135 TS 3_5_2.Doc - R6 5 Unit 2 - Amendment No. 135

NUCLEAR POWER GENERATION DIABLO CANYON POWER PLANT JOB PERFORMANCE MEASURE Number: NRCADM04RO

Title:

DETERMINE RADIOLOGICAL POSTINGS Examinee:

Evaluator:

Print Signature Date Results: Sat Unsat Total Time: minutes Comments:

References:

Radiation Worker Training Handout RCP D-240, Radiological Posting, Rev. 16 Alternate Path: Yes No X Time Critical: Yes No X Time Allotment: 10 Critical Steps: 1, 2 Job Designation: RO Task Number: 2.3.4 Rating: 2.5 AUTHOR: JACK BLACKWELL DATE: 01/18/2005 APPROVED BY: N/A DATE:

TRAINING LEADER APPROVED BY: N/A DATE:

LINE MANAGER REV. 1

JPM TITLE: DETERMINE Radiological Postings JPM NUMBER: NRCADM04RO INSTRUCTOR WORKSHEET Directions: No plant controls or equipment are to be operated during the performance of this Job Performance Measure. All actions taken by the examinee should be clearly demonstrated and verbalized to the evaluator. The student will be given the initial conditions, initiating cue, and task standard. The examiner will then ask if any clarifications are needed. After identifying the appropriate procedure for the task, the examinee may be given the procedure and told the step with which to begin.

Required Materials: Attached Radiological Maps Copy of RCP D-240, Radiological Posting Initial Conditions: Radiological Surveys have just been completed to update the baseline data in preparation for scheduled work. RP is short on personnel and has requested Operations assistance in preparing for the work by reviewing the surveys and determining the required postings.

Initiating Cue: The SFM has directed you to review the survey forms and make recommendations with regards to posting to the identified areas.

Task Standard: The required postings are documented below and reported to the SFM.

Radiation Area Survey Map 1 Survey Point Rad Posting 1

2 3

4 5

Contamination Survey Map 2 Survey Point SCA Posting 1

2 3

4 NRCADM04ROREV1.DOC Page 2 of 7 REV. 1

JPM TITLE: DETERMINE Radiological Postings JPM NUMBER: NRCADM04RO INSTRUCTOR WORKSHEET Start Time:

Step Expected Operator Actions

- 1. Determines radiological postings 1.1 Identifies the following areas for for survey map one. Posting based on area surveys:

o Point 1 - No posting, less than 5mr/hr, part of the RCA o Point 2 - No posting, less than 5mr/hr, part of the RCA

- o Point 3 - Radiation Area

(>5mr/hr but <100mr/hr) o Point 4 - Radiation Area,

- (>5mr/hr but <100mr/hr) o Point 5 - High Radiation Area

- (>100mr/hr but less than 1000mr/hr)

Step was: Sat: ______ Unsat _______*

Denotes an entry required on the JPM cover sheet.

- Denotes a Critical Step.

NRCADM04ROREV1.DOC Page 3 of 7 REV. 1

JPM TITLE: DETERMINE Radiological Postings JPM NUMBER: NRCADM04RO INSTRUCTOR WORKSHEET Step Expected Operator Actions

- 2. Determines radiological postings 2.1 Identifies the following contamination for contamination survey map area postings based on smears:

two.

o Contamination results 1 - No SCA posting required

(<1000dpm/100cm2) o Contamination results 2 - SCA

- posting required

(>1000dpm/100cm2)

- o Contamination results 3 - SCA posting required

(>1000dpm/100cm2) o Contamination results 4 - No SCA posting required

(<1000dpm/100cm2)

Step was: Sat: ______ Unsat _______*

Stop Time:

Total Time: (Enter total time on the cover page)

Denotes an entry required on the JPM cover sheet.

- Denotes a Critical Step.

NRCADM04ROREV1.DOC Page 4 of 7 REV. 1

JPM NUMBER: NRCADM04RO EXAMINEE CUE SHEET Initial Conditions: Radiological Surveys have just been completed to update the baseline data in preparation for scheduled work. RP is short on personnel and has requested Operations assistance in preparing for the work by reviewing the surveys and determining the required postings.

Initiating Cue: The SFM has directed you to review the survey forms and make recommendations with regards to posting to the identified areas.

Task Standard: The required postings are documented below and reported to the SFM.

Radiation Area Survey Map 1 Survey Point Rad Posting 1

2 3

4 5

Contamination Survey Map 2 Survey Point SCA Posting 1

2 3

4 NRCADM04ROREV1.DOC Page 5 of 7 REV. 1

JPM TITLE: DETERMINE Radiological Postings JPM NUMBER: NRCADM04RO ATTACHMENT 1, SIMULATOR SETUP o No simulator setup is required for this JPM.

NRCADM04ROREV1.DOC Page 6 of 7 REV. 1

JPM TITLE: DETERMINE Radiological Postings JPM NUMBER: NRCADM04RO ANSWER KEY Radiation Area Survey Map 1 Survey Point Rad Posting 1 No Posting, <5mr/hr, RCA only 2 No Posting, <5mr/hr, RCA only 3 Radiation Area (>5mr/hr but <100mr/hr) 4 Radiation Area (>5mr/hr but <100mr/hr) 5 High Radiation Area (>100mr/hr but <1000mr/hr)

Contamination Survey Map 2 Survey Point SCA Posting 1 Contamination results 1 - No SCA posting required (<1000dpm/100cm2) 2 Contamination results 2 - SCA posting required (>1000dpm/100cm2) 3 Contamination results 3 - SCA posting required (>1000dpm/100cm2) 4 Contamination results 4 - No SCA posting required (<1000dpm/100cm2)

NRCADM04ROREV1.DOC Page 7 of 7 REV. 1

Student Handout Student Handout

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PACIFIC GAS AND ELECTRIC COMPANY NUMBER RCP D-240 NUCLEAR POWER GENERATION REVISION 16 DIABLO CANYON POWER PLANT PAGE 1 OF 15 RADIATION CONTROL PROCEDURE UNITS TITLE: Radiological Posting 1 2 08/03/04 AND EFFECTIVE DATE PROCEDURE CLASSIFICATION: QUALITY RELATED LEVEL OF USE: REFERENCE TABLE OF CONTENTS SECTION PAGE SCOPE .............................................................................................................................................................. 1 DISCUSSION ................................................................................................................................................... 1 DEFINITIONS .................................................................................................................................................. 2 RESPONSIBILITIES........................................................................................................................................ 4 PREREQUISITES............................................................................................................................................. 4 PRECAUTIONS ............................................................................................................................................... 4 INSTRUCTIONS .............................................................................................................................................. 4 General .......................................................................................................................................................... 4 Radiation Area Posting.................................................................................................................................. 6 High Radiation Area Posting......................................................................................................................... 6 Locked High Radiation Area Posting............................................................................................................ 6 Very High Radiation Area Posting................................................................................................................ 7 Surface Contamination Area Posting ............................................................................................................ 7 Airborne Radioactivity Area Posting ............................................................................................................ 8 RECORDS ........................................................................................................................................................ 8 APPENDICES................................................................................................................................................... 8 ATTACHMENTS ............................................................................................................................................. 8 REFERENCES.................................................................................................................................................. 9

1. SCOPE 1.1 This procedure describes the proper posting requirements utilized at DCPP for the purpose of radiological control.

2. DISCUSSION 2.1 Routine and special radiological surveys are performed to maintain a knowledge of the radiological conditions of plant areas. Areas with radiological conditions in excess of specified limits are posted to identify the conditions within. Components within these posted areas may require additional markings to identify specific radiological conditions (e.g., contamination under insulation, sample sinks, etc). Consistent and correct radiological posting is essential to maintain compliance with Federal regulations and to inform personnel of the radiological hazards associated with particular areas.

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PACIFIC GAS AND ELECTRIC COMPANY NUMBER RCP D-240 DIABLO CANYON POWER PLANT REVISION 16 PAGE 2 OF 15 TITLE: Radiological Posting UNITS 1 AND 2 2.2 A standard radiation posting sign is used at DCPP. The sign consists of three sections:

a heading, a symbol, and inserts.

2.2.1 The heading contains the words: "CAUTION," "DANGER," or "GRAVE DANGER."

2.2.2 The conventional three bladed radiation symbol is located near the heading.

Normally the bladed area is magenta. It may also be black or purple. The background is yellow.

2.2.3 The CAR system of posting radiological areas requires that each sign has an insert for C-contamination, A-airborne, and R-radiation and that they be in the following order under the three bladed symbol:

a. Contamination

b. Airborne

c. Radiation 2.2.4 Each insert is color coded

a. Green: No radiological concern exists.

b. Yellow: Low to moderate radiological concern exists.

c. Red: A high level of radiological concern exists.

2.3 Additional informational signs, placards, labels or tape may be used in conjunction with the CAR posting to provide more specific detail about the radiological condition.

2.4 Components that require additional radiological information may be identified by barrier tape and/or informational labels instead of the CAR posting.

3. DEFINITIONS 3.1 Accessible - means an area that can be occupied by a major portion of an individual's whole body.

3.2 Accessible Overhead Area - An area greater than eight feet is accessible if a platform or ladder is configured such that the area becomes accessible to an individual. All other overhead areas are inaccessible.

3.3 Airborne Radioactivity Area - is (per 10 CFR 20.1003) a room, enclosure or area in which airborne radioactive materials, composed wholly or partly of licensed material, exist in concentrations:

3.3.1 Exceeding 100% of the derived air concentrations (DAC) specified in Appendix B of 10 CFR 20.1001 - 20.2402.

OR 3.3.2 To such a degree that an individual present in the area without respiratory protective equipment could exceed, during the hours an individual is present in a week, an intake of 0.6 percent of the annual limit on intake (ALI), or 12 DAC-hours.

3.4 Barricade - a door, gate, chain, rope or any such item that obstructs passage.

3.5 Component - equipment, piping, valves and other parts within an area.

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PACIFIC GAS AND ELECTRIC COMPANY NUMBER RCP D-240 DIABLO CANYON POWER PLANT REVISION 16 PAGE 3 OF 15 TITLE: Radiological Posting UNITS 1 AND 2 3.6 Controlled Area - is (per 10 CFR 20.1003) an area, outside of a restricted area but inside the site boundary, access to which can be limited by the licensee for any reason.

3.7 High Radiation Area (HRA) - is ( per 10 CFR 20.1601) an area accessible to personnel with radiation levels that could result in an individual receiving a deep dose equivalent (DDE) of greater than 100 mrem in one hour measured at 30 cm from the radiation source or from any surface that the radiation penetrates.

3.8 Locked High Radiation Area (LHRA) - is an area accessible to personnel with radiation levels that could result in an individual receiving a DDE of greater than 1000 mrem PER HOUR measured at 30 cm from the radiation source or from any surface that the radiation penetrates.

3.9 In One Hour - means a cumulative dose averaged over a period of one hour, as opposed to a constant dose rate measured "per hour."

3.10 Radiation Area - is (per 10 CFR 20.1003) an area accessible to personnel with radiation levels that could result in an individual receiving a DDE of greater than 5 mrem in one hour measured at 30 cm from the radiation source or from any surface that the radiation penetrates.

3.11 Radioactive Material Area - is (per 10 CFR 20.1902) a room or area accessible to personnel in which radioactive material is used or stored that exceeds ten times the amounts specified in Appendix C of 10 CFR 20.

3.12 Radiological Controls Area (RCA) - is ( per DCPP administrative control) an area in which access is controlled for the purpose of radiation protection, in part, through the use of a Radiation Work Permit. The permanent RCA includes the Containment Buildings, the Fuel Handling Buildings, most of the Auxiliary Building, the Radwaste and Laundry Buildings, the area between the Auxiliary Building and the Radwaste Buildings, and the Calibration Facilities located in the Turbine and Buttress Buildings.

3.13 Restricted Area - is (per 10 CFR 20.1003) an area, access to which is limited for the purpose of protecting individuals against undue risks from exposure to radiation and radioactive materials. Normally a restricted area boundary is the same as either the permanent or temporary RCA boundary with which it is associated. Restricted areas not associated with an RCA may be setup with the approval of the RPM or designee provided they are documented in an AR within two working days. The AR should contain the rationale for the setup and steps taken to insure compliance with regulatory and procedural requirements.

3.14 Surface Contamination Area (SCA) - is (per DCPP administrative control) an area accessible to personnel in which smear surveys indicate removable contamination equal to or greater than 20 DPM/100 cm2 alpha.

OR Surface Contamination Area (SCA) - is (per DCPP administrative control) an area accessible to personnel in which smear surveys indicate removable contamination equal to or greater than 1000 DPM/100 cm2 beta-gamma.

RCP D-240, Radiological Postings.Doc 07 1118.0224

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PACIFIC GAS AND ELECTRIC COMPANY NUMBER RCP D-240 DIABLO CANYON POWER PLANT REVISION 16 PAGE 4 OF 15 TITLE: Radiological Posting UNITS 1 AND 2 3.15 Very High Radiation Area (VHRA) - is an area (per 10 CFR 20.1602) accessible to personnel with radiation levels that could result in an individual receiving an absorbed dose of greater than 500 rads in one hour measured at one meter from the radiation source or from any surface that the radiation penetrates.

4. RESPONSIBILITIES 4.1 Radiation protection is responsible for maintaining the radiological postings in all plant areas in accordance with this procedure.

5. PREREQUISITES None

6. PRECAUTIONS 6.1 Posting Placement 6.1.1 Where practical, placement of posting materials should avoid attachment to plant piping or components.

a. Posting and barricade material should be attached using metal hasps and an approved adhesive (stock code 73-0664 or 72-6333 or an approved equal). Temporary adhesive attachment anchors, i.e., wall stickies, should not be used. Wire-ties are not temporary adhesive attachments and are acceptable anchor points.

b. The requirements of CF4.ID8, "Temporary Attachments," shall be followed where attachment to piping or components cannot be reasonably avoided.

c. The requirements of CF5.ID12, "Consumable Material Control," shall be followed where contact with affected corrosion resistant alloys cannot reasonably be avoided.

7. INSTRUCTIONS 7.1 General 7.1.1 Signs and postings that meet the wording requirements of sections 7.2 through 7.9 of this procedure, and the requirement for the magenta radiation symbol on a yellow background, are in compliance with the regulations and as such are considered acceptable.

a. The CAR posting should use a standard 3-pocket sign. Barrier rope should be used to identify the area boundaries.

b. Typically such signs and postings are located in infrequently accessed areas and are "holdovers" from the time period before the CAR system was introduced at this plant. Such signs and postings should be brought up to the CAR standards in a timely manner.

7.1.2 When used as required, a "CAUTION," "DANGER," or "GRAVE DANGER" sign shall be visible from each accessible point of entry into the posted area.

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PACIFIC GAS AND ELECTRIC COMPANY NUMBER RCP D-240 DIABLO CANYON POWER PLANT REVISION 16 PAGE 5 OF 15 TITLE: Radiological Posting UNITS 1 AND 2 7.1.3 When practicable, discrete areas meeting the Radiation Area, High Radiation Area or Locked High Radiation Area criteria should be individually posted.

Posting of a very large area or building is generally inappropriate if most of the area does not meet the applicable criteria.

If most of the area within a building or on a floor meets the Radiation Area or High Radiation Area criteria, all entrances to the area may be posted in lieu of posting each discrete area within.

If rooms or areas have components, equipment, or work evolutions which cause variable dose rates, a larger boundary may be established with the postings based on the higher of the expected conditions.

a. Examples of appropriate use of these larger boundaries include:

1. Posting of the area on 140' elevation between Unit 1 and Unit 2 Containment Buildings or large portions of the 115' elevation backyard area during periods of bulk movement of radioactive material, such as during outages.

2. Posting of large portions of the 55' elevation Auxiliary Building due to the automatic discharge function of the Reactor Coolant Drain Tank.

7.1.4 If rooms or areas have work evolutions which cause short term airborne radioactivity, a larger boundary may be established with the postings based on the higher of the expected conditions.

a. The postings within the larger boundary do not require updating.

b. The posting at the larger boundary contains additional information which describes the reason for the short term posting.

7.1.5 Additional informational signs, placards, labels or tape may be used in conjunction with the CAR posting to provide more specific detail about the radiological condition. See Appendix 9.4 for descriptions of the more commonly used informational signs.

7.1.6 Where uses of the color magenta is specified for purposes of posting, the following substitutions may be made:

a. Purple or red may be used for the radiation symbol or radiological barricade rope/tape. Black may also be used for the radiation symbol.

7.1.7 General Posting Exceptions

a. Overhead areas that are inaccessible do not require posting and/or barricading. It is considered a prudent action to place a posting if the area is considered easy to post.

b. Except for a VHRA, the need for any other required posting may be deleted for periods of less than eight hours IF personnel responsible for positive control over access to the affected area are in attendance sufficient to either prevent access, or to control access in accordance with an applicable Radiation Work Permit.

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PACIFIC GAS AND ELECTRIC COMPANY NUMBER RCP D-240 DIABLO CANYON POWER PLANT REVISION 16 PAGE 6 OF 15 TITLE: Radiological Posting UNITS 1 AND 2

c. Standard signs using the CAR system are not required for permanently installed postings such as the signs at the entrances to inside a containment crane wall. Each sign, either standing alone or in conjunction with other postings, sign shall meet the wording requirements of Sections 7.2 through 7.10 as well as the requirement for the magenta radiation symbol on a yellow background.

NOTE: The first three slots below the radiation symbol should always contain inserts. If conditions do not exist that would require a yellow or red insert, then a green unlabeled insert should be used.

7.2 Radiation Area Posting 7.2.1 Each Radiation Area shall be conspicuously posted as follows:

a. A standard sign shall be used.

b. The heading shall contain the word "CAUTION".

c. The insert in the third slot below the tri-foil should be colored yellow and shall contain the words "Radiation Area".

7.3 High Radiation Area Posting 7.3.1 Each High Radiation Area shall be conspicuously posted as follows:

a. A standard sign shall be used.

b. The heading shall contain the word "CAUTION" or the word "DANGER".

1. The preferred wording is "DANGER".

c. The insert in the third slot below the tri-foil should be colored red and shall contain the words "High Radiation Area".

7.3.2 Magenta and yellow rope, or other similar physical barricade, shall be used in conjunction with the posting requirements of 7.3.1.

7.3.3 Pink Stop signs with contrasting lettering should be posted at unlocked HRAs.

7.3.4 Access controls to High Radiation Areas are discussed in RCP D-220.

7.4 Locked High Radiation Area Posting 7.4.1 Each LHRA shall be conspicuously posted as follows:

a. A standard sign shall be used.

b. The heading shall contain the word "DANGER."

c. The insert in the third slot below the tri-foil should be colored red and shall contain the words "Locked High Radiation Area."

RCP D-240, Radiological Postings.Doc 07 1118.0224

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PACIFIC GAS AND ELECTRIC COMPANY NUMBER RCP D-240 DIABLO CANYON POWER PLANT REVISION 16 PAGE 7 OF 15 TITLE: Radiological Posting UNITS 1 AND 2 7.4.2 Magenta and yellow rope, or other similar physical barricade, shall be used in conjunction with the posting requirements of 7.4.1.

7.4.3 Pink Stop signs with contrasting lettering should be posted at unlocked LHRAs.

7.4.4 Whenever practical, LHRA postings should be at or on the gate or door that is locked to control access to the area.

7.4.5 Access controls and locking requirements for LHRAs are discussed in RCP D-220.

7.5 Very High Radiation Area Posting 7.5.1 Each VHRA shall be conspicuously posted as follows:

a. A standard sign shall be used.

b. The heading shall contain the words "GRAVE DANGER."

c. The insert in the third slot below the radiation symbol should be colored red and shall contain the words "Very High Radiation Area."

7.5.2 Magenta and yellow rope, or other similar physical barricade, shall be used in conjunction with the posting requirements of 7.5.1.

7.5.3 Whenever practical, VHRA postings should be at or on the gate or door that is locked to control access to the area.

7.5.4 Access controls and locking requirements for VHRAs are discussed in RCP D-220.

7.6 Surface Contamination Area Posting 7.6.1 Each Surface Contamination Area should be conspicuously posted as follows:

a. A standard sign should be used.

b. For contamination levels of 1K dpm/100 cm2 but 100K dpm/100 cm2

- (or 20 dpm/100 cm2) the insert in the first slot below the tri-foil should be colored yellow and contain the words "Surface Contamination Area."

c. For contamination levels of >100K dpm/100 cm2 - the insert in the first slot below the tri-foil should be colored red and contain the words "Surface Contamination Area."

7.6.2 The boundaries of Surface Contamination Areas on floors or other surfaces should normally be designated with yellow and magenta rope, or similar physical barrier, to prevent inadvertent entry into the area. Yellow and magenta tape may be used to further delineate the area.

a. When a temporary wall preventing inadvertent access is utilized to delineate a Surface Contamination Area boundary, yellow and magenta rope are not necessary.

RCP D-240, Radiological Postings.Doc 07 1118.0224

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PACIFIC GAS AND ELECTRIC COMPANY NUMBER RCP D-240 DIABLO CANYON POWER PLANT REVISION 16 PAGE 8 OF 15 TITLE: Radiological Posting UNITS 1 AND 2

b. Small contaminated components such as pump bases, filter housings, etc.,

are exempt from the above method of posting signs or erecting rope if the boundaries are identified with yellow and magenta tape. If yellow drip bags are used, no tape or wording is needed.

7.6.3 Areas with significantly higher contamination levels than the surrounding contaminated area that mandate different protective clothing entry requirements should be bounded with yellow and magenta rope or yellow and magenta tape (if practicable) to mark the boundaries and posted with the proper contamination level information.

Large areas that have been painted to fix contamination should be identified.

Examples of such methods include the use of a designated paint color, stencils, lamicoids or labels.

7.7 Airborne Radioactivity Area Posting 7.7.1 Each Airborne Radioactivity Area shall be conspicuously posted as follows:

a. A standard sign shall be used.

b. The heading shall contain the word "CAUTION" or the word "DANGER".

c. The insert in the second slot below the tri-foil should be colored red and shall contain the words "Airborne Radioactivity Area."

7.8 Radiography Posting Requirements-See Appendix 9.2.

7.9 Radioactive Material Area Posting-see Appendix 9.3.

7.10 For establishment of an RMA outside the permanent RCA, see Appendix 9.3.

7.11 Labeling of containers of radioactive materials is discussed in RCP D-610.

7.12 Labeling or radioactive tools and equipment is discussed in AD4.ID5.

8. RECORDS None

9. APPENDICES 9.1 Radiological Controls Area: Boundaries, Postings and Special Requirements 9.2 Radiography Posting Requirements 9.3 Radioactive Material Area Posting Requirements 9.4 Description of the some commonly used informational signs/labels/placards

10. ATTACHMENTS None RCP D-240, Radiological Postings.Doc 07 1118.0224

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PACIFIC GAS AND ELECTRIC COMPANY NUMBER RCP D-240 DIABLO CANYON POWER PLANT REVISION 16 PAGE 9 OF 15 TITLE: Radiological Posting UNITS 1 AND 2

11. REFERENCES 11.1 10 CFR 20, "Standards for Protection Against Radiation."

11.2 CF4.ID8, "Temporary Attachments."

11.3 CF5.ID12, "Consumable Material Control."

11.4 AD4.ID5, "Job Site Tool Control."

11.5 RCP D-220, "Control of Access to High Radiation Areas, High-High Radiation and Very High Radiation Areas."

11.6 RP1.ID7, "Control of Radiography."

11.7 RCP D-500, "Radiation and Contamination Surveys."

11.8 Information Notice No. 84-82, "Guidance for Posting Radiation Areas."

11.9 Nonconformance Report DCO-91-TC-N093, "Radiological Labeling and Posting."

11.10 Quality Evaluation, Q0009704, "Lights Found Not Flashing."

11.11 Information Notice No. 88-79, "Misuse of Flashing Lights for High Radiation Area Controls."

11.12 NCRP Report No. 59, 09/15/76.

11.13 SER 10-97, "Unplanned Exposure During Spent Fuel Pool Diving Operations."

11.14 Information Notice No. 97-68, "Loss of Control of Diver in a Spent Fuel Storage Pool."

11.15 Action Request #A0545467 11.16 RCP EM-4, "Area TLD Monitoring."

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PACIFIC GAS AND ELECTRIC COMPANY NUMBER RCP D-240 DIABLO CANYON POWER PLANT REVISION 16 PAGE 10 OF 15 TITLE: Radiological Posting UNITS 1 AND 2 APPENDIX 9.1 Radiological Controls Area: Boundaries, Postings and Special Requirements

1. SCOPE This appendix describes the posting of RCA boundaries (excluding radiography) and special controls for limiting dose at those boundaries.

2. DISCUSSION RCA boundaries are important for controlling dose to Members of the Public. These RCA boundary dose rates have administrative limits that are controlled by radiation surveys, Area TLD Monitoring, and personnel occupancy times.

3. DEFINITIONS

a. Occupancy

1) High Occupancy - means areas such as offices, laboratories and other similar work stations occupied continuously on an annual basis. (i.e., approximately 100% occupancy; 40 hours4.62963e-4 days 0.0111 hours 6.613757e-5 weeks 1.522e-5 months per normal work week).

2) Intermediate Occupancy - means areas which are populated for shorter periods, such as temporary work stations. (i.e., approximately 25% occupancy; 10 hours1.157407e-4 days 0.00278 hours 1.653439e-5 weeks 3.805e-6 months per normal work week).

3) Low Occupancy - means areas such as walkways or roads used for pedestrian or vehicular traffic. (i.e., approximately 6.25% occupancy; 2.5 hours5.787037e-5 days 0.00139 hours 8.267196e-6 weeks 1.9025e-6 months per normal work week).

4. RESPONSIBILITIES

a. Radiation protection is responsible for maintaining the radiological postings in all plant areas in accordance with this procedure.

b. The REMP engineer (or designee) is responsible for communicating to RP Supervision posting changes as a result of Area TLD Monitoring.

c. The RP technician or supervision is responsible for initiating an AR (AT-REMP) within two working days when the conditions of step 7.b.3) of this appendix apply

5. PREREQUISITES None RCP D-240, Radiological Postings.Doc 07 1118.0224

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PACIFIC GAS AND ELECTRIC COMPANY NUMBER RCP D-240 DIABLO CANYON POWER PLANT REVISION 16 PAGE 11 OF 15 TITLE: Radiological Posting UNITS 1 AND 2 APPENDIX 9.1 (Continued)

6. PRECAUTIONS

a. Posting Placement

1) Where practical, placement of posting materials should avoid attachment to plant piping or components.

a) Posting and barricade material should be attached using metal hasps and an approved adhesive (stock code 73-0664 or 72-6333 or an approved equal). Temporary adhesive attachment anchors, i.e., wall stickies, should not be used. Wire-ties are not temporary adhesive attachments and are acceptable anchor points.

b) The requirements of CF4.ID8, "Temporary Attachments," shall be followed where attachment to piping or components cannot be reasonably avoided.

c) The requirements of CF5.ID12, "Consumable Material Control," shall be followed where contact with affected corrosion resistant alloys cannot reasonably be avoided.

7. INSTRUCTIONS

a. General Requirements:

1) The radiation level at the RCA boundary shall not exceed 2 mrem in one hour.

2) The total effective dose equivalent (TEDE) to individual Members of the Public shall not exceed 100 mrem in a year.

b. Radiological Controls Area

1) Standard signs using the color coded CAR system are not required at entrances to the RCA.

2) All personnel entrances to an RCA shall be conspicuously posted as follows:

a) The sign or signs shall have the magenta radiation symbol on a yellow background.

b) The sign should include the words "CAUTION, RADIOLOGICAL CONTROLS AREA, PERSONNEL MONITORING DEVICES REQUIRED BEYOND THIS POINT" or other similar wording.

c) In addition, all personnel entrances to an RCA shall be posted as a Radioactive Material Area.

(1) The Radioactive Material designation may be contained on an RCA sign or on a separate RMA posting as described in appendix 9.3 of this procedure.

3) The RP technician or supervisor shall initiate an AR (AT-REMP) within two working days when:

a) RCA boundary dose rates exceed 0.5 mr/hr OR b) RCA boundary occupancy is intermediate or high by unmonitored personnel and the boundary is not monitored with Area TLDs (RCP EM-4).

NOTE: The permanent RCA boundary is normally monitored with Area TLDs.

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PACIFIC GAS AND ELECTRIC COMPANY NUMBER RCP D-240 DIABLO CANYON POWER PLANT REVISION 16 PAGE 12 OF 15 TITLE: Radiological Posting UNITS 1 AND 2 APPENDIX 9.1 (Continued)

4) The radiation level at an unattended RCA boundary shall not exceed 2 mrem in one hour.

5) Radioactive material staged for a period of less than 24 hours2.777778e-4 days 0.00667 hours 3.968254e-5 weeks 9.132e-6 months outside the permanent RCA, and labeled and packaged in accordance with DOT regulations is exempt from the requirement to generate an Action Request.

6) All access points to and from a permanent RCA not staffed by RP personnel should remain locked when practicable to prevent unauthorized entry without the knowledge of the RP personnel. Doors which remain unlocked to allow egress from a permanent RCA in the event of an emergency shall have a posting visible at the door requiring notification of RP upon exit at this point. These requirements are not applicable to the normally staffed access control.

7) Other RCA access points may be established by RP if provisions are made for personnel and material access and egress in accordance with appropriate Radiation Control Procedures, Radiation Work Permit or policies.

8) The contiguous boundary of the permanent RCA may be changed by RP supervision provided that the posting of and control of access to the RCA meets the above requirements.

9) Establishing a new RCA/RMA outside of the permanent RCA.

NOTE: Radioactive material staged due to shipment (incoming or outgoing) is exempt from the requirements of this section provided it is for a period of less than 8 hours9.259259e-5 days 0.00222 hours 1.322751e-5 weeks 3.044e-6 months and the radioactive material is constantly attended by an individual who takes the precautions necessary to prevent the exposure of individuals to radiation or radioactive materials in excess of regulatory limits.

a) Select an appropriate area.

(1) To the extent practicable, RCAs outside of the permanent RCA should be maintained within a lockable enclosure.

b) Consider potential airborne or liquid effluent pathway that may exist due to the storage activities under normal conditions.

c) Consider any postulated airborne and/or liquid effluents due to a fire in the proposed storage area and firefighting water used to control such a fire.

d) Ensure the criteria of this appendix steps 7.b.3) and 7.b.4) are met.

e) After completing all applicable steps above, obtain approval to establish the RCA from the radiation protection manager or designee.

f) Establish and post the new area as applicable.

g) Notify chemistry that a new RCA has been established.

h) Notify the RP access foreman that a new RCA has been established.

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PACIFIC GAS AND ELECTRIC COMPANY NUMBER RCP D-240 DIABLO CANYON POWER PLANT REVISION 16 PAGE 13 OF 15 TITLE: Radiological Posting UNITS 1 AND 2 APPENDIX 9.2 Radiography Posting Requirements

1. SCOPE This appendix describes the posting of radiography boundaries and special controls for limiting dose at those boundaries.

2. DISCUSSION None

3. DEFINITIONS None

4. RESPONSIBILITIES

a. Radiation protection is responsible for implementing radiological postings that may be needed in addition to the radiographer requirements.

5. PREREQUISITES None

6. PRECAUTIONS None

7. INSTRUCTIONS

a. Standard signs using the CAR system are not required for radiography postings.

b. Radiography postings are to include the following:

1) The sign or signs shall have the magenta radiation symbol on a yellow background.

2) Radiography postings used solely for radiography exposures should include wording similar to the following: "RADIOGRAPHY IN PROGRESS:" and "NO ENTRY".

3) The wording for postings of radiation areas and high and very high radiation areas shall be in accordance with 7.2 through 7.5 of this procedure.

4) LHRAs caused by radiography exposures do not need to be posted as a LHRA.

c. Areas controlled solely due to radiography exposures do not need to include the wording required for an RCA posting as described in Appendix 9.1.

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PACIFIC GAS AND ELECTRIC COMPANY NUMBER RCP D-240 DIABLO CANYON POWER PLANT REVISION 16 PAGE 14 OF 15 TITLE: Radiological Posting UNITS 1 AND 2 APPENDIX 9.3 Radioactive Material Area Posting Requirements

1. SCOPE This appendix describes the posting of radioactive material area boundaries.

2. DISCUSSION None

3. DEFINITIONS None

4. RESPONSIBILITIES None

5. PREREQUISITES None

6. PRECAUTIONS None

7. INSTRUCTIONS

a. Normally a Radioactive Material Area is contained within a Radiological Controls Area and the RMA posting is located at all personnel entry points to the RCA. (Appendix 9.1 of this procedure)

b. Under some circumstances an RMA may be designated outside of an RCA. The following areas are examples of RMAs outside of RCAs at DCPP. Any other area located outside of an RCA requires the approval of the RPM or designee to be designated as an RMA.

1) Areas with low activity sources used for the purpose of source checking, performance checking, or calibrating instruments.

2) Areas used for storing smoke detectors.

3) The areas surrounding the setup of the steam generator chemical cleaning equipment located outside of the RCA and also the protected area.

c. Standard signs using the color coded CAR system are not required for Radioactive Material Area Posting.

d. The sign or signs shall include the following:

1) The magenta radiation symbol on a yellow background.

2) The words "CAUTION RADIOACTIVE MATERIAL" or the words "DANGER RADIOACTIVE MATERIAL".

3) The preferred wording is "CAUTION RADIOACTIVE MATERIAL".

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PACIFIC GAS AND ELECTRIC COMPANY NUMBER RCP D-240 DIABLO CANYON POWER PLANT REVISION 16 PAGE 15 OF 15 TITLE: Radiological Posting UNITS 1 AND 2 APPENDIX 9.4 Description of the some Commonly used Informational Signs/Labels/Placards

1. SCOPE This appendix lists some of the commonly used informational signs/labels/placards (called signs here for purposes of simplicity) that may be used in conjunction with the CAR posting or to identify sources of radiation and contamination on components. These informational signs are not required to be used.

This information in this Appendix is not intended to be and inclusive list of all the various signs. Signs containing handwritten radiological information that are not described in this Appendix may also be used.

2. DESCRIPTION

a. LOCALIZED RADIATION - used to define and alert personnel to sources of radiation and smaller areas within larger areas in which the exposure rates are be significantly higher than the general area dose rate.

b. HOT SPOT - used to define a specific radiation source where the physical contact reading is

> 100 mrem per hour and is at least 5 times the general area dose rates. See RP1.DC4, "Radiological Hot Spot Identification and Control Program" for specific conditions when posting is required, and when a Hot Spot database entry is required.

c. COLD AREA - used to define areas which have substantially lower dose rates than the surrounding general area so that personnel may use these areas to maintain exposure ALARA.

d. RADIOLOGICAL CONDITIONS HAVE CHANGED - used to emphasize significant changes in radiological conditions

e. STOP - used when a High Radiation Area or Locked High Radiation Area is not locked to control access.

RCP D-240, Radiological Postings.Doc 07 1118.0224

NUCLEAR POWER GENERATION DIABLO CANYON POWER PLANT JOB PERFORMANCE MEASURE Number: NRCADM04SRO

Title:

APPROVE EMERGENCY EXPOSURE Examinee:

Evaluator:

Print Signature Date Results: Sat Unsat Total Time: minutes Comments:

References:

EP RB-2, Emergency Exposure Guides, Rev. 5 EP RB-3, Stable Iodine Thyroid Blocking, Rev. 4 Alternate Path: Yes X No Time Critical: Yes No X Time Allotment: 15 minutes Critical Steps: 2, 3, 4, 5 Job Designation: RO/SRO Task Number: G2.3.4 Rating: 2.5/3.1 AUTHOR: JACK BLACKWELL DATE: 01/18/2005 REVIEWED BY: N/A DATE:

JPM COORDINATOR APPROVED BY: N/A DATE:

TRAINING LEADER REV. 0

JPM TITLE: APPROVE EMERGENCY EXPOSURE JPM NUMBER: NRCADM04SRO INSTRUCTOR WORKSHEET Directions: No plant controls or equipment are to be operated during the performance of this Job Performance Measure. All actions taken by the examinee should be clearly demonstrated and verbalized to the evaluator. The student will be given the initial conditions, initiating cue, and task standard. The examiner will then ask if any clarifications are needed. After identifying the appropriate procedure for the task, the examinee may be given the procedure and told the step with which to begin.

Required Materials: EP RB-2, Emergency Exposure Guides EP RB-3, Stable Iodine Thyroid Blocking Attached Attachments from EP RB-2, Att. 9.1, 9.6 and 9.7 and RB-3 Att. 5.1 Unit 1 was at 100% power when an earthquake resulted in major Initial Conditions:

equipment damage, especially in the GE 100 penetration area. A Large Break LOCA is in progress and a Site Area Emergency has been declared. The TSC has not been manned yet, and you are the ISEC. Two employees were last seen in GE 100 penetration area. Radiation Protection estimates radiation exposure to be 55 Rem/hour whole body with airborne contamination, therefore requiring SCBAs. There have been five volunteers to perform a search and rescue operation. They are:

Frank Fireman, Fireman, male, age 37

Fred Fireman, Fireman, male, age 47

Joe Operator, Nuclear Operator, male, age 50

Rebecca Radman, RP Tech., female, age 32, declared pregnant woman.

Oscar Operator, Licensed Operator, male, age 47, prior emergency exposure at another utility.

All volunteers have been briefed, special hazards identified and protective measures implemented. The expected stay time is from 20 to 50 minutes. All operators are self-monitoring trained. A backup team is being assembled.

Initiating Cue: The RP Supervisor has presented forms for KI distribution and authorization for emergency exposure for your approval.

Task Standard: The search and rescue is approved and the appropriate forms signed by the ISEC.

NRCADM04SRO.DOC Page 2 of 6 REV. 0

JPM TITLE: APPROVE EMERGENCY EXPOSURE JPM NUMBER: NRCADM04SRO INSTRUCTOR WORKSHEET Start Time:

Step Expected Operator Actions

1. Obtain the correct procedure. 1.1 Refers to EP RB-2, attachment 9.1.

Step was: Sat: ______ Unsat _______*

- 2. Determines volunteers meet ** 2.1 Determines Rebercca Radman and requirements and are briefed. Oscar Operator do not qualify.

NOTE: May also determine Frank Fireman not eligible due to age.

2.2 Initials appropriate block.

Step was: Sat: ______ Unsat _______*

- 3. Ensure activity necessary, ** 3.1 Recognizes the necessity of the hazards identified, protective activity, that the team has been measures implemented, and briefed on hazards and protective backup team established. measures, and a backup team is being assembled.

Cue: If asked, all hazards have been discussed and protective measures implemented.

3.2 Initials appropriate blocks.

Step was: Sat: ______ Unsat _______*

Denotes an entry required on the JPM cover sheet.

- Denotes a Critical Step.

NRCADM04SRO.DOC Page 3 of 6 REV. 0

JPM TITLE: APPROVE EMERGENCY EXPOSURE JPM NUMBER: NRCADM04SRO INSTRUCTOR WORKSHEET Step Expected Operator Actions

- 4. Implement EP RB-3, Stable 4.1 Reviews procedure.

Iodine Thyroid Blocking, and directs RA to administer KI.

- 4.2 Removes Rebecca Radman and Oscar Operator from list. (May remove Frank Fireman also)

- 4.3 Approves administering KI to remaining volunteers.

4.4 Initials appropriate block.

Step was: Sat: ______ Unsat _______*

Denotes an entry required on the JPM cover sheet.

- Denotes a Critical Step.

NRCADM04SRO.DOC Page 4 of 6 REV. 0

JPM TITLE: APPROVE EMERGENCY EXPOSURE JPM NUMBER: NRCADM04SRO INSTRUCTOR WORKSHEET Step Expected Operator Actions

- 5. Sign Permit to approve ** 5.1 Removes Rebecca Radman and Oscara Authorized Limit. Operator from list. (May remove Frand Fireman also)

- 5.2 Verifies Permit accurate.

Nuclear Operator and older Fireman assigned for entry. (May also assign other Fireman) **

Max TEDE Rate of 55 Rem/hr Stay time of 50 minutes Anticipated TEDE of 46 Rem NO LIMIT checked

- 5.3 Signs for approval of Permit.

5.4 Initials appropriate block on Attachment 9.1.

Step was: Sat: ______ Unsat _______*

Stop Time:

Total Time: (Enter total time on the cover page)

Denotes an entry required on the JPM cover sheet.

- Denotes a Critical Step.

NRCADM04SRO.DOC Page 5 of 6 REV. 0

JPM NUMBER: NRCADM04SRO EXAMINEE CUE SHEET Unit 1 was at 100% power when an earthquake resulted in major Initial Conditions:

equipment damage, especially in the GE 100 penetration area. A Large Break LOCA is in progress and a Site Area Emergency has been declared. The TSC has not been manned yet, and you are the ISEC. Two employees were last seen in GE 100 penetration area. Radiation Protection estimates radiation exposure to be 55 Rem/hour whole body with airborne contamination, therefore requiring SCBAs. There have been five volunteers to perform a search and rescue operation. They are:

Frank Fireman, Fireman, male, age 37

Fred Fireman, Fireman, male, age 47

Joe Operator, Nuclear Operator, male, age 50

Rebecca Radman, RP Tech., female, age 32, declared pregnant woman.

Oscar Operator, Licensed Operator, male, age 47, prior emergency exposure at another utility.

All volunteers have been briefed, special hazards identified and protective measures implemented. The expected stay time is from 20 to 50 minutes. All operators are self-monitoring trained. A backup team is being assembled.

Initiating Cue: The RP Supervisor has presented forms for KI distribution and authorization for emergency exposure for your approval.

Task Standard: The search and rescue is approved and the appropriate forms signed by the ISEC.

NRCADM04SRO.DOC Page 6 of 6 REV. 0

JPM TITLE: APPROVE EMERGENCY EXPOSURE JPM NUMBER: NRCADM04SRO ATTACHMENT 1, SIMULATOR SETUP The simulator is not needed for the performance of this JPM.

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PACIFIC GAS AND ELECTRIC COMPANY NUMBER EP RB-2 NUCLEAR POWER GENERATION REVISION 5 DIABLO CANYON POWER PLANT PAGE 1 OF 6 EMERGENCY PLAN IMPLEMENTING PROCEDURE UNITS TITLE: EMERGENCY EXPOSURE GUIDES 1 2 03/23/03 AND EFFECTIVE DATE PROCEDURE CLASSIFICATION: QUALITY RELATED TABLE OF CONTENTS SECTION PAGE SCOPE .............................................................................................................................................................. 1 DISCUSSION ................................................................................................................................................... 2 DEFINITIONS .................................................................................................................................................. 3 RESPONSIBILITIES........................................................................................................................................ 4 PREREQUISITES............................................................................................................................................. 4 PRECAUTIONS ............................................................................................................................................... 5 INSTRUCTIONS .............................................................................................................................................. 5 RECORDS ........................................................................................................................................................ 5 ATTACHMENTS ............................................................................................................................................. 6 Recovery Manager (or SEC) Checklist ......................................................................................................... 6 TSC Radiological Advisor Checklist ............................................................................................................ 6 OSC Site Radiation Protection Coordinator Checklist.................................................................................. 6 OSC Emergency Maintenance Coordinator Checklist .................................................................................. 6 EOF Radiological Manager Checklist........................................................................................................... 6 DCPP Emergency Exposure Guidelines........................................................................................................ 6 Emergency Exposure Permit ......................................................................................................................... 6 REFERENCES.................................................................................................................................................. 6 SPONSOR ......................................................................................................................................................... 6

1. SCOPE 1.1 This procedure provides guidance in the process of determining the need for authorizing and controlling emergency radiological exposure to selected individuals that is beyond the 10 CFR 20 annual exposure limits.

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PACIFIC GAS AND ELECTRIC COMPANY NUMBER EP RB-2 DIABLO CANYON POWER PLANT REVISION 5 PAGE 2 OF 6 TITLE: EMERGENCY EXPOSURE GUIDES UNITS 1 AND 2

2. DISCUSSION 2.1 Authorization of emergency exposure is an extraordinary measure, but justifiable under four sets of circumstances.

When the intended action requiring a potential overexposure to an emergency worker, is expected to result in;

Saving or preserving the quality of a human life that would otherwise be lost.

Significant projected dose saving to others.

Protection of valuable property.

Sampling results required to redefine or adjust existing Protective Actions for the public or site personnel.

All of the above situations require that no reasonable method is immediately available (or readily apparent) to avoid exceeding the established annual limits and that every effort will be made to keep the emergency exposure ALARA.

2.2 The emergency exposure guidelines implemented at DCPP are consistent with the Environmental Protection Agency (EPA) guidance for controlling doses to workers under emergency conditions. (Reference 10.3) 2.3 Authorized dose limits for workers during emergencies are based on avoiding acute health effects and limiting the risk of delayed health effects.

2.4 Emergency worker exposures are not controlled by Planned Special Exposures.

NOTE: Planned Special Exposures may be implemented during non-emergency situations including Recovery Operations. Refer to RP1.ID8, "Planned Special Exposures."

2.5 An emergency exposure should be authorized only once in an individual's lifetime and is in addition to any prior occupational exposure from normal or planned special exposures.

2.6 The emergency exposure limits specified in this procedure are applicable to both in-plant team response activities and off-site field monitoring by the company's ERO personnel.

2.7 Emergency exposure above 25 rem TEDE shall require the voluntary consent of the authorized individual.

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PACIFIC GAS AND ELECTRIC COMPANY NUMBER EP RB-2 DIABLO CANYON POWER PLANT REVISION 5 PAGE 3 OF 6 TITLE: EMERGENCY EXPOSURE GUIDES UNITS 1 AND 2

3. DEFINITIONS 3.1 Annual Administrative Exposure Guidelines An administrative dose restriction for individual occupational radiation exposure established by the company to control personnel exposures within non-regulatory recommendations prescribed by NCRP and ICRP.

3.2 Annual Administrative Exposure Limits Dose limits established by the company to ensure that personnel do not exceed regulatory limits.

3.3 Committed Dose Equivalent (CDE) The dose to the organs or tissues that would be received from an intake of radioactive material by an individual during the 50 years following the intake.

3.4 Committed Effective Dose Equivalent (CEDE) The sum of the products of the weighting factors applicable to each of the body organs or tissues that are irradiated and the CDE to these organs or tissues.

3.5 Corrective Actions Those emergency measures taken to mitigate or terminate an emergency situation at or near the source of the problem in order to prevent an uncontrolled release of radioactive material or reduce the magnitude of a release.

3.6 Declared Pregnant Woman (DPW) A woman who has voluntarily informed her supervision, in writing, of her pregnancy and the estimated date of conception.

3.7 Deep Dose Equivalent (DDE) Dose associated with external exposure of the whole body at a depth of 1 cm.

3.8 Lens Dose Equivalent (LDE) External exposure to the lens of the eye at a depth 0.3 cm.

3.9 Lifesaving Action Any of several activities that are necessary to save human life including search and rescue, first aid, transport and emergency medical care.

3.10 Occupational Dose Dose received by an individual in the course of employment in which the individual's assigned duties involve exposure to radiation and to radioactive material.

3.11 Planned Special Exposure Dose received in addition to and accounted for separately from the doses received under the limits of 10 CFR 20.1201 as a planned and specially authorized exposure in accordance with 10 CFR 20.1206.

3.12 Shallow Dose Equivalent (SDE) External exposure of the skin or any extremity (depth 0.007 cm).

3.13 Total Effective Dose Equivalent (TEDE) The sum of the DDE (for external exposure) and CEDE (for internal exposure).

4.

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PACIFIC GAS AND ELECTRIC COMPANY NUMBER EP RB-2 DIABLO CANYON POWER PLANT REVISION 5 PAGE 4 OF 6 TITLE: EMERGENCY EXPOSURE GUIDES UNITS 1 AND 2 RESPONSIBILITIES 4.1 The Recovery Manager (RM) or Site Emergency Coordinator (SEC) prior to turnover, has the unilateral authority and non-delegable responsibility for authorizing an individual emergency worker to exceed normal 10 CFR 20 exposure limits.

The RM/SEC is furthermore responsible for ensuring that the NRC is notified of any overexposure that may result.

4.2 The TSC Radiological Advisor (RA) is responsible for evaluating the conditions requiring an emergency exposure authorization and advising the RM (or SEC) on its justification and when all prerequisite requirements have been met.

4.3 The EOF Radiological Manager (ERM) is responsible for evaluating radiological conditions and exposures to off-site emergency response personnel and advising the RM when an emergency exposure authorization is justified.

4.4 The Site Radiation Protection Coordinator (SRPC) is responsible for identifying the necessity of obtaining an emergency exposure authorization and in assisting with volunteer selection as needed.

4.5 The Emergency Maintenance Coordinator (EMC) is responsible for ensuring that the maximum protection and support is provided to those personnel dispatched from the OSC under the extraordinary conditions of emergency exposure.

4.6 The emergency worker is responsible for knowing the potential health consequences of the emergency exposure and for signing the Emergency Exposure Permit when volunteering for potential emergency exposures of > 25 Rem TEDE.

The emergency worker is responsible for maintaining his/her emergency exposure ALARA consistent with the successful completion of the emergency activity.

5. PREREQUISITES 5.1 Emergency classification of Alert or higher has been declared.

5.2 An essential emergency action is required (refer to Attachment 9.6) and cannot be performed without one or more workers potentially exceeding 10 CFR 20 annual exposure limits.

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PACIFIC GAS AND ELECTRIC COMPANY NUMBER EP RB-2 DIABLO CANYON POWER PLANT REVISION 5 PAGE 5 OF 6 TITLE: EMERGENCY EXPOSURE GUIDES UNITS 1 AND 2

6. PRECAUTIONS 6.1 Selection of volunteers shall be based upon established criteria and on the specific skills and knowledge of the workers needed to successfully complete the activity.

6.2 Individuals shall not be authorized to enter any area where exposure rates are unknown or beyond the highest range of portable monitoring instruments.

6.3 Any individual who receives (or is suspected to have received) an actual overexposure shall be removed from further participation in the emergency response.

6.4 Those personnel receiving a dose of 25 rem TEDE or greater shall be promptly transported off-site for evaluation by appropriate medical personnel.

6.5 An Emergency Exposure Authorization considers only the radiological hazards involved.

Other potential hazards to health (i.e., heat stress, hazardous chemicals, biological hazards, confined space entry, etc.) shall be taken into consideration as well and shall be explained to the emergency workers prior to dispatching the team.

7. INSTRUCTIONS NOTE 1: Emergency exposure authorization is specific to each volunteer (individually) for performing the specific activity authorized. Any changes in the specific conditions that established the basis for the authorization are not valid until approved by both the volunteer(s) (for potential emergency exposures > 25 Rem TEDE) and the RM/SEC.

NOTE 2: Individual voluntary emergency exposure should be limited to once in a lifetime.

7.1 The Recovery Manager (or Site Emergency Coordinator) shall implement Attachment 9.1 of this procedure.

7.2 The TSC Radiological Advisor (RA) shall implement Attachment 9.2 of this procedure.

7.3 The OSC Site Radiation Protection Coordinator (SRPC) shall implement Attachment 9.3 of this procedure.

7.4 The OSC Emergency Maintenance Coordinator (EMC) shall implement Attachment 9.4 of this procedure.

7.5 The EOF Radiological Manager (ERM) shall implement Attachment 9.5 of this procedure.

8. RECORDS 8.1 All records generated by the utilization of this procedure for an exercise or emergency shall be forwarded the next working day to the emergency planning supervisor, for review and retention.

8.1.1 Completed forms and documents generated during drills are non-quality related records and shall be retained a minimum of 3 years in accordance with AD10.ID2.

8.1.2 Completed forms and documents generated during real events are quality related records and shall be retained in accorance with AD10.ID1.

9.

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PACIFIC GAS AND ELECTRIC COMPANY NUMBER EP RB-2 DIABLO CANYON POWER PLANT REVISION 5 PAGE 6 OF 6 TITLE: EMERGENCY EXPOSURE GUIDES UNITS 1 AND 2 ATTACHMENTS 9.1 Form 69-20628, "Recovery Manager (or Sec) Checklist," 01/09/03 9.2 Form 69-20629, "TSC Radiological Advisor Checklist," 01/09/03 9.3 Form 69-20630, "OSC Site Radiation Protection Coordinator Checklist," 01/09/03 9.4 Form 69-20631, "OSC Emergency Maintenance Coordinator Checklist," 01/09/03 9.5 Form 69-20632, "EOF Radiological Manager Checklist," 01/09/03 9.6 "DCPP Emergency Exposure Guidelines," 10/07/93 9.7 Form 69-10554, "Emergency Exposure Permit," 01/09/03

10. REFERENCES 10.1 NUREG-0737, November 1980.

10.2 Title 10, Code of Federal Regulations, Part 20.

10.3 "Manual of Protective Actions for Nuclear Incidents," USEPA, 400-R-92-001, May 1992.

10.4 RP1.ID6, "Personnel Dose Limits and Monitoring Requirements."

10.5 EP RB-1, "Personnel Dosimetry."

10.6 EP RB-3, "Stable Iodine Thyroid Blocking."

10.7 CP M-13, "Personnel Injury (or Illness) with Radioactive Contamination or Personnel Overexposure.

10.8 EP G-2, "Activation and Operation of the Interim Site Emergency Organization."

10.9 EP EF-1, "Activation and Operation of the Technical Support Center."

10.10 EP EF-2, "Activation and Operation of the Operational Support Center."

10.11 EP EF-3, "Activation and Operation of the Emergency Operations Facility."

11. SPONSOR Gray EP RB-2.Doc 03B 1116.0244

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69-20628 01/09/03 Page 1 of 1 DIABLO CANYON POWER PLANT TITLE:

EP RB-2 ATTACHMENT 9.1 Recovery Manager (or Sec) Checklist 1 2 AND Actions Initial NOTE: All RCA Qualified personnel are automatically authorized to receive a dose up to, but not exceeding, the DCPP Administrative Limits for Calendar Year exposure (4.5 rem TEDE) during an Alert or higher emergency classification event, EXCEPT as may be limited by lifetime and current year occupational dose already received or other restrictions such as a declared pregnancy.

1. INITIAL ACTIONS Review the completed Emergency Exposure Permit, Form 69-10554, with the RA (or ERM) and evaluate the justification for authorization.

1.1 Ensure volunteers (if necessary) have been obtained and thoroughly briefed on the potential health consequences of this exposure. (See Criteria in Attachment 9.3) ______

1.2 Ensure emergency activity is necessary (no reasonable alternatives) and can be successful in outcome. ______

1.3 Ensure special hazards have been identified and protective measures implemented. ______

1.4 Direct the EMC to establish a back-up team of volunteers for rotation, relief, or rescue if very high dose rates or other life threatening conditions are applicable. ______

1.5 Implement EP RB-3, "Stable Iodine Thyroid Blocking," and direct the RA to Administer KI distribution, if needed. ______

1.6 Sign the Permit to approve the Authorized Limit. (Refer to Attachment 9.6 for Exposure Limits.) LOG NOTE: Each Permit is specific to the individuals or volunteers identified and specified activity. Any changes or additions require a new authorization.

2. SUBSEQUENT ACTIONS 2.1 Direct the Administrative Advisor to callout anticipated replacement personnel for the potentially overexposed volunteers. ______

2.2 Ensure other emergency measures taken concurrently do not increase the accepted risks to the volunteers or jeopardize a successful outcome. ______

2.3 Ensure overexposed personnel are promptly transported to off-site medical facilities for evaluation and treatment. (Refer to CP M-13.) ______

2.4 Ensure that the NRC is notified immediately in accordance with 10 CFR 20.2202(a) for any individual exposure of 25 rem TEDE, 250 rad SDE, or 75 rem LDE. ______

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69-20629 01/09/03 Page 1 of 2 DIABLO CANYON POWER PLANT TITLE:

EP RB-2 ATTACHMENT 9.2 TSC Radiological Advisor Checklist 1 2 AND Actions Initial NOTE: All RCA Qualified personnel are automatically authorized to receive a dose up to, but not exceeding, the DCPP Administrative Limits for Calendar Year exposure (4.5 rem TEDE) during an Alert or higher emergency classification event, EXCEPT as may be limited by lifetime and current year occupational dose already received or other restrictions such as a declared pregnancy.

1. INITIAL ACTIONS Obtain a FAXed copy of Emergency Exposure Permit, Form 69-10554, from the OSC and confirm it's completeness by contacting the SRPC.

1.1 Volunteers (if needed) have been obtained and thoroughly briefed on the potential health consequences of this exposure. (See Criteria in Attachment 9.3.) ______

1.2 Emergency activity is necessary (no reasonable alternatives) and can be successful in outcome. ______

1.3 Ensure special hazards have been identified and protective measures implemented. ______

1.4 Recommend a back-up team to be assembled and standing by if very high dose rates are anticipated. ______

1.5 Obtain authorization from the RM/SEC for Thyroid Blocking Agent per EP RB-3, if necessary. ______

1.6 Evaluate justification for the Authorized Limit and advise the RM/SEC to authorize the permit. (Refer to Attachment 9.6 for Exposure Limits.) ______

NOTE: TEDE exposure is the controlling limit for continuous monitoring of the team. Other exposures (SDE, LDE, and CDE+DDE) require appropriate protective measures (i.e., KI, respirator use, clothing, etc.) and are important for planning purposes only, unless capability of direct monitoring exists.

1.7 Determine any appropriate Dose Correction Factors to adjust the Authorized TEDE Limit if conditions indicate that other doses are more likely to be limiting and notify the SRPC. ______

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69-20629 01/09/03 Page 2 of 2 EP RB-2 (UNITS 1 AND 2)

ATTACHMENT 9.2 TITLE: TSC Radiological Advisor Checklist Actions Initial

2. SUBSEQUENT ACTIONS 2.1 Notify SRPC that RM/SEC authorization has been obtained and provide any special instructions, conditions or revised limits, if needed. ______

2.2 Direct SRPC to prepare an SWP, if not already done. ______

NOTE: If the situation requires immediate action the SWP may be completed afterward, but verbal authorization is required beforehand.

2.3 Ensure that CP M-13, "Personal Injury (or Illness) with Radioactive Contamination or Personnel Overexposure, is implemented in anticipation of overexposed personnel. ______

2.4 Implement EP RB-3, "Stable Iodine Thyroid Blocking," as directed by the RM/SEC. ______

2.5 Implement EP RB-1, "Personnel Dosimetry," to ensure that; ______

adequate personnel exposure monitoring is provided for the extraordinary conditions

dosimetry devices are collected and dose evaluated promptly after task completion

dosimetry incidents (off-scale or erratic readings) are evaluated promptly EP RB-2.Doc 03B 1116.0244

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69-20630 01/09/03 Page 1 of 2 DIABLO CANYON POWER PLANT TITLE:

EP RB-2 ATTACHMENT 9.3 OSC Site Radiation Protection Coordinator Checklist 1 2 AND Actions Initial NOTE: All RCA Qualified personnel are automatically authorized to receive a dose up to, but not exceeding, the DCPP Administrative Limits for Calendar Year exposure (4.5 rem TEDE) during an Alert or higher emergency classification event, EXCEPT as may be limited by lifetime and current year occupational dose already received or other restrictions such as a declared pregnancy.

1. INITIAL ACTIONS When a pre-departure analysis of radiological conditions, in accordance with EP EF-2, indicates that the planned or anticipated dose to any emergency response team member will exceed 10 CFR 20 annual limits, perform the following; 1.1 Review your dose evaluation with the EMC and EOC to determine if any alternative actions can achieve the desired results without requiring an emergency exposure. ______

1.2 Obtain qualified volunteers (if needed) from those personnel available.

(Criteria in Section 3.0, next page) ______

1.3 Obtain a working copy of Form 69-10554, Emergency Exposure Permit (Attachment 9.7), and fill in the required information. ______

NOTE: Complete a new Permit form for each team activity that is analyzed to require emergency exposure. ______

1.4 Calculate and record the anticipated exposure to the most limiting team member and determine the authorized limit appropriate to the activity in accordance with Attachment 9.6, DCPP Emergency Exposure Guidelines. ______

1.5 Brief the volunteers on the radiological hazards and ensure they are informed about the potential health consequences associated with authorized exposure. ______

1.6 For potential exposures of > 25 Rem TEDE, obtain the signature on the Emergency Exposure Permit of each volunteer, including the C&RP technician assigned to monitor the team. ______

1.7 Obtain authorization from the RM/SEC for Thyroid Blocking Agent per EP RB-3, if necessary. ______

1.8 FAX the completed form to the Recovery Manager (or SEC if EOF is not activated) and contact the RA to review the Permit and advise the RM/SEC. ______

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69-20630 01/09/03 Page 2 of 2 EP RB-2 (UNITS 1 AND 2)

ATTACHMENT 9.3 TITLE: OSC Site Radiation Protection Coordinator Checklist Actions Initial

2. SUBSEQUENT ACTIONS When the emergency exposure authorization is approved by the RM/SEC then ensure that all conditions and limitations are understood by the response team prior to departure in accordance with EP EF-2 and SWP documentation requirements.

With regard to the extraordinary circumstances of this activity ensure that the following additional actions are taken; 2.1 Ensure that appropriate personnel dosimetry is issued in accordance with EP RB-1, "Personnel Dosimetry." ______

2.2 Ensure that a portable radiation monitoring instrument with adequate range capability is supplied to the C&RP Technician. ______

CAUTION: IT IS FORBIDDEN TO ENTER ANY AREA WHERE THE DOSE RATES ARE UNKNOWN OR BEYOND THE RANGE OF INSTRUMENTATION AVAILABLE.

2.3 Ensure that the Team Leader understands that whenever practical (without compromising the mission) ALARA principles should be used to minimize team exposure. ______

NOTE: If the situation requires immediate action, the SWP may be completed afterward, but verbal authorization is required beforehand.

3. CRITERIA FOR VOLUNTEER SELECTION 3.1 Professional rescue personnel for lifesaving activities who volunteer by choice of employment should be chosen for search and rescue.

3.2 RCA Qualified personnel should be selected for missions involving very high dose rates and high contamination levels.

3.3 Volunteers shall be fully aware of the risks involved.

3.4 Volunteers should be above the age of 45 years old.

3.5 Declared Pregnant Women (DPW) shall not be chosen.

3.6 Individuals who have already received an emergency exposure should not be chosen.

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69-20631 01/09/03 Page 1 of 2 DIABLO CANYON POWER PLANT TITLE:

EP RB-2 ATTACHMENT 9.4 OSC Emergency Maintenance Coordinator Checklist 1 2 AND Actions Initial

1. INITIAL ACTIONS When a pre-departure analysis of radiological conditions, in accordance with EP EF-2, indicates that the planned or anticipated dose to the emergency response team members will exceed 10 CFR 20 annual limits, perform the following; 1.1 Contact the Operations Coordinator and the SEC to determine if any alternative actions can achieve the desired results without requiring an emergency exposure. ______

Consideration may be given to any one or combination of the following possible alternatives and should be pursued in parallel, as time permits to avoid unnecessary risk to individuals;

use of robotics or fabrication of special tools

use of temporary shielding

changing plant system lineups to reduce background exposure rates near operating equipment

use of short cut procedures, elimination of double checks and hold points, non-QA parts, tools, etc.

installation of jumpers and bypasses to achieve remote operation of equipment from lower dose areas NOTE: Some of the above options may involve intentionally violating Technical Specifications*, written procedures, or Quality Standards, but may be equally justifiable to the RM/SEC as emergency exposure of personnel, depending on circumstances.

1.2 Determine the optimum team composition in terms of skills and experience to ensure the highest degree of confidence in mission success in the least amount of time for exposure of personnel available. ______

NOTE: Emergency exposure is unwarranted in circumstances where alternative actions can achieve equal or better results.

1.3 Review the Permit form prepared by the SRPC and concur with seeking emergency dose authorization. ______

NRC Notification required per EP G-3 EP RB-2.Doc 03B 1116.0244

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69-20631 01/09/03 Page 2 of 2 EP RB-2 (UNITS 1 AND 2)

ATTACHMENT 9.4 TITLE: OSC Emergency Maintenance Coordinator Checklist Actions Initial

2. SUBSEQUENT ACTIONS 2.1 Ensure that the Team Leader is briefed on potential hazards that are expected and the limits of authority that he/she may exercise in making ad hoc decisions in the field. ______

2.2 Ensure that a back-up team is chosen and prepared for immediate dispatch to rotate in, relieve, or rescue the primary response team, as needed. ______

NOTE: Back-up Team shall be briefed, dressed out, pre-authorized, pre-staged in low dose area, standing by if needed for immediate action.

2.3 Control any other concurrent activities that may hamper, impede, or otherwise increase the risk to the primary emergency response team. ______

2.4 Ensure that the emergency operations coordinator is aware of the team location so that Operations activities from the Control Room do not change radiological conditions adversely without warning. ______

2.5 Maintain an open communication line with the EOC, as needed, to ensure that changes in plant status are immediately recognized and factored into ongoing risk assessment. ______

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69-20632 01/09/03 Page 1 of 2 DIABLO CANYON POWER PLANT TITLE:

EP RB-2 ATTACHMENT 9.5 EOF Radiological Manager Checklist 1 2 AND Actions Initial NOTE: All RCA Qualified personnel are automatically authorized to receive a dose up to, but not exceeding, the DCPP Administrative Limits for Calendar Year exposure (4.5 Rem TEDE) during an Alert or higher emergency classification event, EXCEPT as may be limited by lifetime and current year occupational dose already received or other restrictions such as a declared pregnancy.

1. INITIAL ACTIONS When an analysis of radiological conditions, in accordance with EP EF-3, indicates that the planned or anticipated dose to any emergency response off-site field team member will exceed 10 CFR 20 annual limits, perform the following:

1.1 Review the dose evaluation with the ESE and RMD to determine if any alternative actions can achieve the desired results without requiring an emergency exposure such as rotation or replacement of team members, shorter sampling times, team movement tactics to avoid higher exposures, etc. ______

1.2 Obtain a working copy of Form 69-10554, "Emergency Exposure Permit" (Attachment 9.7), and fill in the required information. ______

NOTE: Complete a new Permit form for each off-site field team that is analyzed to require emergency exposure authorization, when needed.

1.3 Calculate and record the anticipated exposure to the most limiting team member and ensure that the authorized limit of 5 rem TEDE (total emergency exposure, not counting occupational dose prior to the emergency) will not be exceeded. ______

1.4 Provide the completed Form 69-10554 to the RM for emergency exposure authorization. ______

NOTE: Voluntary consent is not necessary for emergency exposures authorized at less than 25 rem TEDE, but written authorization is required.

1.5 Obtain authorization from the RM for Thyroid Blocking Agent per EP RB-3, if necessary. ______

1.6 Notify the RMD to communicate the authorizations to the Field Team Leaders affected when obtained from the RM. ______

NOTE: These authorizations are for company personnel only and separate authorizations for SLO County team members, if needed, must be obtained through the UDAC from the County Health Officer (CHO).

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69-20632 01/09/03 Page 2 of 2 EP RB-2 (UNITS 1 AND 2)

ATTACHMENT 9.5 TITLE: EOF Radiological Manager Checklist Actions Initial

2. SUBSEQUENT ACTIONS When the emergency exposure authorization is approved by the RM then ensure that all conditions and limitations are understood by the field monitoring team prior to directing them to continue with their activities in the plume or plume affected areas.

With regard to the extraordinary circumstances of this activity ensure that the following additional actions are taken:

2.1 Ensure that each Team Member understands that whenever practical (without compromising the mission) ALARA principles should be used to minimize team exposure. ______

2.2 Ensure that Turn-back dose rates are re-analyzed for the present or projected characteristics of the plume and revised as needed. ______

CAUTION: IT IS FORBIDDEN TO ENTER ANY AREA WHERE THE DOSE RATES ARE UNKNOWN OR BEYOND THE RANGE OF INSTRUMENTATION AVAILABLE.

2.3 Ensure that the RMD makes more frequent checks on accumulated dose (SRD readings) and is controlling team deployment to minimize unnecessary exposures. ______

2.4 Begin the process of obtaining reliefs and replacements for the field monitoring team members, if necessary, to ensure continuous monitoring capability. ______

2.5 Consider deployment of additional teams in standby locations in case an active team can no longer function due to any of the following: ______

gross contamination of vehicle, equipment, or personnel requires decontamination efforts.

respirator use in field conditions creating additional heat stress and fatigue.

higher than anticipated dose rates cause authorized exposure limits to be exceeded.

vehicle breakdown or accident necessitates assistance to personnel stuck in plume pathway.

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10/07/93 Page 1 of 2 DIABLO CANYON POWER PLANT TITLE:

EP RB-2 ATTACHMENT 9.6 DCPP Emergency Exposure Guidelines 1 2 AND The following table contains guidelines for use in authorizing emergency exposures when lower doses are not practicable:

RADIOLOGICAL PROPERTY DOSE SAVING LIFESAVING ASSESSMENT SAVING TO TO SAMPLING POPULATION* INDIVIDUAL*

Emergency Sampling Under Mitigating Damage Corrective Actions, Lifesaving Actions, Actions----> Emergency to Valuable stop/reduce a 1st Aid, Search and Part of Body Conditions Property release rescue Irradiated Whole Body 5 rem TEDE 10 rem TEDE 25 rem TEDE 25 rem TEDE Skin & any 50 rem SDE 100 rem SDE 250 rem SDE 250 rem SDE Extremity Lens of the Eye 15 rem LDE 30 rem LDE 75 rem LDE 75 rem LDE Any Organ or 50 rem 100 rem 250 rem 250 rem Tissues (CDE+DDE) (CDE+DDE) (CDE+DDE) (CDE+DDE)

NOTES: 1. Radiological Assessment Sampling, includes collection of atmospheric, liquid, and environmental radiological activity samples as well as chemistry samples involving high activity or high radiation. Emergency exposure limits may be authorized for selected individuals, for emergency assessment functions, in addition to annual occupational dose to date.

2. Property Saving, for example, might be dispatching the Fire Brigade to extinguish a fire in a Very High Radiation Area to protect plant equipment though no immediate threat exists to compromising Plant Safety.

3. Dose Saving to Population, includes activities that justify a potential overexposure to a few workers in order to save even a small average dose in a large population. (May also include Traffic Control for Evacuees or other Security Plan Functions.)

4. Lifesaving to Individual, includes the activity of search and rescue in very high dose rates or high airborne activity.

Extreme situations may occur in which a dose in excess of 25 rem TEDE would be unavoidable for either Dose Saving to (Large) Population or Lifesaving to (An)

Individual.

An authorization of emergency exposure with NO LIMITS may be made under those conditions, but only to volunteers who are fully aware of the risks involved, including the numerical levels of dose at which acute effects of radiation will be incurred and the numerical estimates of the risk of delayed effects.

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10/07/93 Page 2 of 2 EP RB-2 (UNITS 1 AND 2)

ATTACHMENT 9.6 TITLE: DCPP Emergency Exposure Guidelines NOTES: (Continued)

5. If any of the above emergency exposure limits would prevent successful completion of the activity then the RM or SEC should ensure that back-up teams are standing by to rotate in and relieve the primary responders.

6. Volunteers for any authorized exposures above 25 rem TEDE should be made aware that there is some risk of acute health effects involved, however remote.

The dose limit of 75 rem to the whole body previously recommended by the EPA for lifesaving action represents a very high level of risk of both acute and delayed effects.

A dose of 100 rem is expected to result in an approximately 15 percent risk of temporary incapacity from non lethal acute effects and an indeterminate, but less than 5 percent, chance of death within 60 days. This is in addition to a risk of about 1 in 30 of incurring fatal cancer.

Such high risk levels can only be accepted by a recipient who has been made aware of the risks involved.

(Reference, EPA Manual of Protective Action Guides and Protective Actions for Nuclear Incidents, May 1992)

NOTE: Although EPA-400 guidelines say that no limit is applicable under extreme situations it is also true that the RM/SEC must make the authorization and may impose a more restrictive limit if so desired consistent with the availability of personnel resources, alternative actions and the desire to avoid acute health effects of the volunteers.

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69-10554 01/09/03 Page 1 of 1 DIABLO CANYON POWER PLANT TITLE: Emergency Exposure Permit EP RB-2 ATTACHMENT 9.7 1 2 AND Date: Time: Permit #:

Responder(s):

(Print)

RP Support:

Description of Activity:

Special Hazards:

Special Instructions:

Anticipated TEDE Rate: (rem/hr) AUTHORIZED LIMIT: [ ] 5 rem TEDE (Check One) [ ] 10 rem TEDE Anticipated Stay Time: (hr) [ ] 25 rem TEDE

[ ] NO LIMIT Anticipated TEDE: (rem)

Voluntary Consent (For potential exposures of > 25 rem TEDE): I hereby volunteer to perform the activity described above and I acknowledge having received a radiological briefing. I am fully aware of the health risks associated with the anticipated exposure. (Sign Below.)

Authorization of Site Emergency Coordinator or Recovery Manager: Time:

EP RB-2.Doc 03B 1116.0244

PACIFIC GAS AND ELECTRIC COMPANY NUMBER EP RB-3 NUCLEAR POWER GENERATION REVISION 4 DIABLO CANYON POWER PLANT PAGE 1 OF 3 EMERGENCY PLAN IMPLEMENTING PROCEDURE UNITS TITLE: Stable Iodine Thyroid Blocking 1 2 03/30/00 AND EFFECTIVE DATE PROCEDURE CLASSIFICATION: QUALITY RELATED TABLE OF CONTENTS SECTION PAGE SCOPE ...................................................................................................................................................1 RESPONSIBILITIES ...............................................................................................................................1 Radiological Advisor............................................................................................................................1 Site Emergency Coordinator ...............................................................................................................1 TSC Liaison Coordinator .....................................................................................................................1 Radiological Manager..........................................................................................................................2 Recovery Manager ..............................................................................................................................2 INSTRUCTIONS .....................................................................................................................................2 Evaluating When to Administer KI .......................................................................................................2 Approval of KI Administration ..............................................................................................................3 Administration of KI .............................................................................................................................3 RECORDS ..............................................................................................................................................3 Drills ....................................................................................................................................................3 Emergency ..........................................................................................................................................3 ATTACHMENTS .....................................................................................................................................3 Form 69-9395, "Record of Distribution of Potassium Iodide ...................................................................3

1. SCOPE 1.1 This procedure provides instructions for the administration of stable iodine in the form of Potassium Iodide (KI) under emergency conditions for emergency personnel.

1.2 This procedure was rewritten; therefore, revision bars are not included.

2. RESPONSIBILITIES 2.1 Radiological Advisor 2.1.1 Responsible for evaluating when KI should be administered.

2.1.2 Responsible for coordinating the issuance of KI to onsite personnel.

2.2 Site Emergency Coordinator 2.2.1 Responsible for authorizing administration of KI to onsite personnel.

2.2.2 Responsible for authorizing administration of KI to offsite personnel until relieved by the Recovery Manager.

2.3 TSC Liaison Coordinator 2.3.1 Responsible for informing onsite personnel of the decision to administer KI.

EP RB-3.doc 3B 1

PACIFIC GAS AND ELECTRIC COMPANY NUMBER EP RB-3 DIABLO CANYON POWER PLANT REVISION 4 PAGE 2 OF 3 TITLE: Stable Iodine Thyroid Blocking UNITS 1 AND 2 2.4 Radiological Manager 2.4.1 Responsible for evaluating when KI should be administered.

2.4.2 Responsible for advising the Recovery Manager when the County Health Officer has elected to issue KI to emergency workers.

2.4.3 Responsible for coordinating the issuance of KI to offsite personnel.

2.5 Recovery Manager 2.5.1 Responsible for authorizing administration of KI to offsite personnel.

3. INSTRUCTIONS 3.1 Evaluating When to Administer KI 3.1.1 KI is most effective when administered immediately prior to exposure to radioiodine, therefore administration of KI should be considered when:

a. Exposure situations exist where calculated iodine dose equivalent to the thyroid can be 25 rem or greater.

NOTE: Refer to Figure 1 below to determine thyroid dose equivalent as a function of the airborne I-131 concentration.

b. No current air analysis is available and high levels of radio-iodine release are suspected prior to undertaking an emergency response operation.

NOTE: If the County Health Officer approves KI administration to the County emergency workers, then approving KI administration for PG&E workers may be considered below the 25 rem exposure Protective Action Guideline.

FIGURE 1 EP RB-3.doc 3B 2

PACIFIC GAS AND ELECTRIC COMPANY NUMBER EP RB-3 DIABLO CANYON POWER PLANT REVISION 4 PAGE 3 OF 3 TITLE: Stable Iodine Thyroid Blocking UNITS 1 AND 2 3.2 Approval of KI Administration 3.2.1 Radiological Advisor shall obtain Site Emergency Coordinator authorization prior to administering KI to onsite personnel.

3.2.2 Radiological Manager shall obtain Recovery Manager authorization prior to administering KI to offsite personnel.

3.3 Administration of KI CAUTION: Personnel with sensitivity to iodine may develop adverse symptoms from KI tablet ingestion. A history of shellfish allergies may indicate iodine sensitivity.

3.3.1 Prior to issue of KI, warn personnel of the possible effects to personnel with iodine sensitivity.

3.3.2 Instruct personnel to review the "Patient Package Insert for THYRO-BLOCK Tablets, Wallace Laboratories."

3.3.3 Ensure personnel complete Attachment 5.1, "Record of Distribution of Potassium Iodide."

3.3.4 Instruct affected personnel to take one 130 mg KI tablet.

3.3.5 Tablets should be administered for ten days after verified exposure.

Dosage is one tablet, once a day.

3.3.6 Individuals suspected of inhalation of airborne contamination should receive thyroid counts on a regular basis throughout the KI treatment period to verify effectiveness of treatment and to estimate dose commitment.

4. RECORDS 4.1 Drills 4.1.1 When used for drills, Attachment 5.1 is a good business record and shall be retained by Emergency Planning for 3 years.

4.2 Emergency 4.2.1 When used for an actual emergency, Attachment 5.1 shall be retained as a quality record in accordance with AD10.ID1.

5. ATTACHMENTS 5.1 Form 69-9395, "Record of Distribution of Potassium Iodide," 03/23/00 EP RB-3.doc 3B 3

69-9395 03/23/00 Page 1 of 1 DIABLO CANYON POWER PLANT TITLE: Record of Distribution of Potassium Iodide EP RB-3 ATTACHMENT 5.1 1 2 AND

1. Fill out time and date KI is administered.

2. Your initials indicate you have been made aware of possible adverse effects to iodine sensitive personnel.

Date Time Dosage Name Initials SSN Organization Address EP RB-3.doc 3B 4 3

NUCLEAR POWER GENERATION DIABLO CANYON POWER PLANT JOB PERFORMANCE MEASURE Number: NRCADM-05SRO

Title:

GDT RUPTURE - DOSE ASSESSMENT/PAR/EAL Examinee:

Evaluator:

Print Signature Date Results: Sat Unsat Total Time: minutes Comments: The Simulator is not required for the performance of this JPM EP R-2, Attachment 10.1 & 10.2 answer key is included for evaluator use

References:

EP R-2, Release of Airborne Radioactive Materials Initial Assessment, Rev. 22 EP G-1, Emergency Classification and Emergency Plan Activation, Rev. 33B Alternate Path: Yes X No Time Critical: Yes No X Time Allotment: 15 minutes Critical Steps: 2, 3, 4, 5, 7 Job Designation: SRO Task Number: 2.4.41 Rating: 4.1 AUTHOR: JACK BLACKWELL DATE: 01/18/2005 REVIEWED BY: N/A DATE:

TRAINING LEADER APPROVED BY: N/A DATE:

LINE MANAGER REV. 0

JPM TITLE: GDT RUPTURE - DOSE ASSESSMENT/PAR/EAL JPM NUMBER: NRCADM05SRO INSTRUCTOR WORKSHEET Directions: No PLANT controls or equipment are to be operated during the performance of this Job Performance Measure. All actions taken by the examinee should be clearly demonstrated and verbalized to the evaluator. The student will be given the initial conditions, initiating cue, and task standard. The examiner will then ask if any clarifications are needed. The examinee may be given the applicable procedure and step with which to begin.

Required Materials: Calculator, and copies of EP R-2.

Initial Conditions: Both units are at 100% power, MOL, equilibrium conditions. Gas Decay Tank 11 rupture disk failed and the relief valve was isolated after 45 minutes of release. The Shift Manager has activated the Emergency Response Organization and is currently the ISEC. The following conditions exist:

o 1 FHB Exhaust Fan running o 2 Aux Blg Exhaust Fans running o 1 GE/GW Area Fan running o 1 Containment Purge fan running o RM-14/87 is OOS o RM-29 is 25 mR/hr o Wind is from 294° at 1.84 m/s from the backup tower. X/Q is not available.

Initiating Cue: The Shift Manager has directed you to perform the necessary assessments to determine the event classification. The PPC program for R-2 calculations is unavailable.

Task Standard: Assessments made and classification of event ready for the ISEC.

NRCADM05SRO.DOC PAGE 2 OF 12 REV. 0

JPM TITLE: GDT RUPTURE - DOSE ASSESSMENT/PAR/EAL JPM NUMBER: NRCADM05SRO INSTRUCTOR WORKSHEET Start Time:

Step Expected Operator Actions

1. Obtain the correct procedure. 1.1 References EP R-2.

Step was: Sat: ______ Unsat _______*

- 2. Determine the plant vent flow 2.1 References Attachment 10.1, rate. page 1, of EP R-2.

2.2 Fills out section 1.

2.3 Uses alternate method to determine plant vent flow rate.

NOTE:This information is on the turnover. 1 FHB fan, 2 aux building fan, 1 GE/GW area fan and 1 Cont. Purge fan are running.

- 2.4 Calculates plant vent flow rate is 262,750 cfm.

Step was: Sat: ______ Unsat _______*

Denotes an entry required on the JPM cover sheet.

- Denotes Critical Step and Sub Steps.

NRCADM05SRO.DOC PAGE 3 OF 12 REV. 0

JPM TITLE: GDT RUPTURE - DOSE ASSESSMENT/PAR/EAL JPM NUMBER: NRCADM05SRO INSTRUCTOR WORKSHEET Step Expected Operator Actions

- 3. Determine the Noble Gas ** 3.1 Calculates noble gas release rate to be Release Rate 31 +1 Ci/sec.

Step was: Sat: ______ Unsat _______*

- 4. Determine the total effluent 4.1 References Attachment 10.1, release rate. page 3, of EP R-2 and determines GDT Rupture = RCS source term.

4.2 Determines Total Effluent Conversation Factor to be 1.00 (RCS).**

4.3 Calculates total effluent release rate to be 31 +1 Ci/sec.**

Step was: Sat: ______ Unsat _______*

Denotes an entry required on the JPM cover sheet.

- Denotes Critical Step and Sub Steps.

NRCADM05SRO.DOC PAGE 4 OF 12 REV. 0

JPM TITLE: GDT RUPTURE - DOSE ASSESSMENT/PAR/EAL JPM NUMBER: NRCADM05SRO INSTRUCTOR WORKSHEET Step Expected Operator Actions

- 5. Perform dose calculations. 5.1 References Attachment 10.2 of EP R-2.

5.2 Observes met data from PPC not available.

5.3 Determines Site Boundary X/Q at 0.8km using Default Values.

- 5.4 Determines DCF to be 1.1 E+5 (RCS).

- 5.5 Calculates TEDE rate of 1804 + 10 mrem/ hr.

- 5.6 Using .75 hours8.680556e-4 days 0.0208 hours 1.240079e-4 weeks 2.85375e-5 months duration, calculates total dose of 1353+10 mrem.

5.7 Determines thyroid CDE calculation to be N/A.

Step was: Sat: ______ Unsat _______*

Denotes an entry required on the JPM cover sheet.

- Denotes Critical Step and Sub Steps.

NRCADM05SRO.DOC PAGE 5 OF 12 REV. 0

JPM TITLE: GDT RUPTURE - DOSE ASSESSMENT/PAR/EAL JPM NUMBER: NRCADM05SRO INSTRUCTOR WORKSHEET Step Expected Operator Actions

6. Obtain correct procedure. 6.1 References EP G-1, Attachment 7.1 Step was: Sat: ______ Unsat _______*

- 7. Recommend event ** 7.1 Determines event classification as a classification. GENERAL EMERGENCY #4 (due to exceeding 1,000 mRem TEDE)

Step was: Sat: ______ Unsat _______*

Stop Time:

Total Time: (Enter total time on the cover page)

Denotes an entry required on the JPM cover sheet.

- Denotes Critical Step and Sub Steps.

NRCADM05SRO.DOC PAGE 6 OF 12 REV. 0

JPM NUMBER: NRCADM-05SRO EXAMINEE CUE SHEET Initial Conditions: Both units are at 100% power, MOL, equilibrium conditions. Gas Decay Tank 11 rupture disk failed and the relief valve was isolated after 45 minutes of release. The Shift Manager has activated the Emergency Response Organization and is currently the ISEC. The following conditions exist:

o 1 FHB Exhaust Fan running o 2 Aux Blg Exhaust Fans running o 1 GE/GW Area Fan running o 1 Containment Purge fan running o RM-14/87 is OOS o RM-29 is 25 mR/hr o Wind is from 294° at 1.84 m/s from the backup tower. X/Q is not available.

Initiating Cue: The Shift Manager has directed you to perform the necessary assessments to determine the event classification. The PPC program for R-2 calculations is unavailable.

Task Standard: Assessments made and classification of event ready for the ISEC.

NRCADM05SRO.DOC PAGE 7 OF 12 REV. 0

JPM TITLE: GDT RUPTURE - DOSE ASSESSMENT/PAR/EAL JPM NUMBER: NRCADM05SRO ATTACHMENT 1, SIMULATOR SETUP The Simulator is not required for the performance of this JPM.

NRCADM05SRO.DOC PAGE 8 OF 12 REV. 0

10/31/02 Page 1 of 3 DIABLO CANYON POWER PLANT EP R-2 (UNITS 1 AND 2)

ATTACHMENT 10.1 TITLE: Release Rate Calculations PLANT VENT RELEASE

1. GENERAL INFORMATION Date: Today Time: Now Assessment No. 1 Assessment By: Name of Examinee Unit Releasing 1

2. PLANT VENT FLOW RATE DETERMINATION 4

A. DIRECT - Plant Vent Flow Rate FR-12 (0-30x10 CFM (CFM) OOS (CFM)

OR B. ALTERNATE - Operating Ventilation Equipment (Max No. possible) #Fans (CFM/Fan)

FHB Exhaust (1) 1 x 35,750 = 35,750 (CFM)

Aux Bldg Exhaust (2) 2 x 73,500 = 147,000 (CFM)

GE/GW Area (1) 1 x 25,000 = 25,000 (CFM)

Cont. Purge (1) 1 x 55,000 = 55,000 (CFM)

Cont. Hydrogen (1) x 300 = (CFM)

Plant Vent Flow Rate = 262,750 (CFM)

3. RELEASE RATE CALCULATION CAUTION: Do NOT use SPDS to obtain monitor readings.

A. NOBLE GAS RELEASE RATE Circle Reading (Units) Conversion Plant Vent Noble Gas Monitor Factor Flow Rate Release Used (CFM) Rate (Ci/sec)

Primary RE-14/14R/87 µCi/cc x 4.72E-04 x Backup RE-29 25 mR/hr x 4.72E-06 x 262,750 = 31 B. TOTAL EFFLUENT RELEASE RATE NOTE: Refer to Page 3 for criteria in choosing RCS, GAP, or CORE below.

Noble Gas Release Total Effluent Total Effluent Rate (Ci/sec) Conversion Factor Release Rate (Ci/sec) 31 x 1.00 (RCS) = 31 1.11 (GAP) 1.50 (CORE)

NOTE: If it is not possible to calculate a release rate, refer to the DEFAULT RELEASE RATES on Page 3 of this attachment.

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10/31/02 Page 2 of 3 DIABLO CANYON POWER PLANT EP R-2 (UNITS 1 AND 2)

ATTACHMENT 10.1 TITLE: Release Rate Calculations

1. GENERAL INFORMATION ATMOSPHERIC STEAM RELEASE Date: Time: Assessment No.

Assessment By: Unit Releasing CAUTION: WHEN CRITICAL, N-16 ACTIVITY SEEN BY MSL RAD MONITORS CAUSES INVALID READINGS FOR OFFSITE DOSE. POST-TRIP, RE-7X READING IS VALID IF THE RE-7X MONITOR SHOWED AN INITIAL N-16 RESPONSE, OR RESPONDS TO CHECKSOURCE.

NOTE: If it is not possible to calculate a release rate, refer to the DEFAULT RELEASE RATES on Page 3.

2. STEAM RELEASES - Use this form to calculate steam releases to the atmosphere WHEN NOT critical.

A. Required Information (RUPTURED GENERATOR ONLY)

Check MSL Rad Reading S/G Lvl Level S/G Flow Flow Rate Ruptured S/G Monitor (cpm) Narrow (%) Rate (lbs/hr)

Range If <4E5 use 4E5 SG 1 RE-71 LI-517 FI-512 SG 2 RE-72 LI-527 FI-522 SG 3 RE-73 LI-537 FI-532 SG 4 RE-74 LI-547 FI-542 B. Alternate Steam Flow Rate (Only if the RUPTURED S/G Flow Rate is otherwise not available)

Valve Type # Valves Capacity (lbs/hr) Flow Rate (lbs/hr)

Lifted 10% Steam Dump (1 per S/G) x 4.0E+05 =

Safety Reliefs (5 per S/G) x 8.5E+05 =

Total Steam Flow Rate (lbs/hr) (lbs/hr)

3. RADIATION MONITOR FACTORS (Determined based on S/G NR Level indication) (Enter in Section 4 below.)

S/G Level EMPTY NORMAL FLOODED Narrow Range < 4% 4% - 96% > 96%

Monitor Factor 6.08E-10 6.75E-10 3.07E-10 (DEFAULT)

4. RELEASE RATE CALCULATIONS A. TOTAL EFFLUENT RELEASE RATE (RE-7x)

MSL Monitor Reading Flow Rate Monitor Factor Total Effluent Release (cpm) (lbs/hr) Rate (Ci/sec) x x GO TO ATTACHMENT 10.2 NRCADM05SRO.DOC PAGE 10 OF 12 REV. 0

10/31/02 Page 3 of 3 DIABLO CANYON POWER PLANT EPR-2 (UNITS 1 AND 2)

ATTACHMENT 10.1 TITLE: Release Rate Calculations

1. SOURCE TERM SELECTION AND DEFAULT RELEASE RATES NOTE: Use default release rate only if actual data is not available or if the release is not being monitored.

A. Check the accident type which most closely resembles the current event.

Default Release Source Rate (Ci/sec) Term Accident Source Condition LOCA (w/ core melt) 1.74 E+1 RE-30 or 31 >300R/hr CORE LOCA (w/o core melt) 5.74 E+0 RE-30 or RE-31 <300R/hr GAP RE-30 or RE-31 not on scale RCS Main Steam Line Break 8.61 E-3 RCS Feedwater Line Break 8.61 E-3 RCS Blackout 8.62 E-1 RCS Locked Rotor 1.57 E-2 GAP FHB Accident 1.45 E+1 GAP Rod Ejection 1.08 E-2 GAP X GDT Rupture 4.14 E+1 RCS LHUT Rupture 3.10 E+1 RCS VCT Rupture 8.29 E-2 RCS S/G Tube Rupture 1.65 E+0 NR S/G Level < 4% SG - Empty NR S/G Level 4-96% SG- Normal NR S/G Level > 96% SG - Flooded Containment FHA S.B. Dose TEDE = 13.4 mrem/hr Go Accident with Equip. Rates Thy.CDE = 51.4 mrem/hr Directly to Hatch Open S.B. Doses TEDE = 6.7 mrem EP G-1 Thy. CDE = 25.7 mrem B. Record the Default Release Rate in Attachment 10.2, Section 4 and use the DCF choice that is listed for the specific accident source above.

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08/05/94 Page 1 of 1 DIABLO CANYON POWER PLANT EP R-2 (UNITS 1 AND 2)

ATTACHMENT 10.2 TITLE: Off-Site Dose Calculations

1. GENERAL INFORMATION Date: Today Time: Now Assessment No. 1 Assessment By: Name of Examinee Unit Releasing 1

2. METEOROLOGICAL DATA - PPC (Plant Process Computer)

Turn On Codes for Met Data are "METP" (Primary Data) or "METB" (Back-up Data)

Parameter Reading Units DEFAULT Wind Speed (10 Meter Level) 1.86 meters/sec Wind Direction (10 Meter Level) 294 Degrees 3

Site Boundary X/Q (0.8 km) Sec/m 5.29E-04

3. DCF Determination - Select the most appropriate source term for the DCF using the criteria in Attachment 10.1. Circle the corresponding DCF in Section 4 below.

4. DOSE CALCULATIONS - (From data calculated using Attachment 10.1)

A. TOTAL EFFECTIVE DOSE EQUIVALENT (TEDE)

Total Effluent or Site Projected TEDE Default Release Boundary DCF TEDE Release (mrem)

Rate X/Q (0.8 km) (circle one) Rate Duration (hr) 3 (Ci/sec) (Sec/m ) (mrem/hr (DEFAULT 3 hrs)

)

1.1E + 05 (RCS) 3.0E + 06 (Gap )

31 x 5.29E-04 x 1.1E + 07 (Core) = 1804 x .75 = 1353 Attachment 10.1 1.1E + 05 (SG-Empty) 4.3E + 04 (SG-Normal )

9.3E + 05 (SG-Flooded)

B. THYROID COMMITTED DOSE EQUIVALENT (CDE) (DO NOT COMPLETE FOR GDT, LHUT, OR VCT RUPTURE)

Total Effluent or Site Projected Default Release Boundary DCF Thyroid Release Thyroid Rate X/Q (circle one) CDE Duration (hr) CDE (Ci/sec) (0.8 km) Rate (DEFAULT 3 hrs) (mrem)

(Sec/m3) (mrem/hr) 1.5E + 06 (RCS) 6.5E + 07 (Gap )

x 7.7E + 07 (Core) = x =

Attachment 10.1 1.5E + 06 (SG-Empty) 1.5E + 05 (SG-Normal )

1.4E + 07 (SG-Flooded)

5. REPORTING THE RESULTS - (Refer to Section 7.3 of Instructions for details)

A. Refer to EP G-1 for EAL criteria.

B. Implement EP RB-10 for PAR criteria NRCADM05SRO.DOC PAGE 12 OF 12 REV. 0

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PACIFIC GAS AND ELECTRIC COMPANY NUMBER EP R-2 NUCLEAR POWER GENERATION REVISION 22 DIABLO CANYON POWER PLANT PAGE 1 OF 6 EMERGENCY PLAN IMPLEMENTING PROCEDURE UNITS TITLE: Release of Airborne Radioactive Materials Initial Assessment 1 2 AND 03/18/04 EFFECTIVE DATE PROCEDURE CLASSIFICATION: QUALITY RELATED

1. SCOPE 1.1 This procedure describes the steps to be taken by on-shift personnel to initially evaluate the off-site consequences of an accidental airborne release that may result in Emergency Plan Activation.

1.2 It does not describe the operation of the plant equipment necessary to terminate or minimize the release. This latter subject is covered in the appropriate E, ECA, and FR series Emergency Procedures for the particular release mechanism.

CAUTION: Revisions to this procedure require the PPC display be updated (Reference A0595224).

2. DISCUSSION 2.1 An accidental airborne release of radioactive materials that may result in site boundary dose rates in excess of the limits specified in the EP G-1 shall require a prompt initial assessment by the operating staff. This initial release rate and dose assessment is performed using either the Plant Process Computer (PPC) program "EPR2," or manually using Section 7 of this procedure.

2.2 This procedure shall only be used by Control Room personnel to perform initial accident dose assessments. This procedure shall not be used to evaluate compliance with Technical Specification limits during planned effluent releases conducted as part of normal plant operations. The methodology contained in this procedure is intended to provide a rapid and conservative calculation of the projected off-site doses due to an accidental release of airborne radioactive materials. More advanced methodologies are contained in procedures EP RB-9 and EP RB-11 or the appropriate chemistry procedures.

3. DEFINITIONS 3.1 Accidental Release - A release of radioactive material unrelated to any planned effluent release evolutions.

3.2 Committed Dose Equivalent (CDE) - The dose to the organs or tissues that would be received from an intake of radioactive material by an individual during the 50 years following the intake.

3.3 Committed Effective Dose Equivalent (CEDE) - The sum of the products of the weighting factors applicable to each of the body organs or tissues that are irradiated and the CDE to these organs or tissues.

3.4 EP R-2.Doc 03B 0119.0419

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PACIFIC GAS AND ELECTRIC COMPANY NUMBER EP R-2 DIABLO CANYON POWER PLANT REVISION 22 PAGE 2 OF 6 TITLE: Release of Airborne Radioactive Materials Initial UNITS 1 AND 2 Assessment Deep Dose Equivalent (DDE) - Dose associated with exposure of the whole body (depth of 1 cm).

3.5 Total Effective Dose Equivalent (TEDE) - The sum of the DDE (for external exposure) and CEDE (for internal exposure).

3.6 TEDE Rate - The time rate of change of Total Effective Dose Equivalent as a function of immersion and inhalation exposure time.

3.7 Thyroid CDE Rate - The time rate of change of Thyroid Committed Dose Equivalent as a function of immersion and inhalation exposure time.

4. RESPONSIBILITIES 4.1 Emergency Evaluation Coordinator (EEC) is responsible for performing an initial assessment of an airborne radiological release when directed by the ISEC.

4.2 Interim Site Emergency Coordinator (ISEC) is responsible for determining when an assessment is needed and directing the EEC to implement this procedure based on emergency evaluation priorities.

5. PREREQUISITES 5.1 Unified Dose Assessment Center (UDAC) is not activated and performing the function of radiological assessment.

5.2 Interim Site Emergency Coordinator (ISEC) has determined, based on plant accident conditions or symptoms of an accidental radiological release, that an initial assessment of projected off-site doses has priority over other actions being performed by the EEC.

The following listed symptoms indicate that an airborne release may be occurring from within the RCA as guidance to the ISEC:

There is actual or suspected leakage of water, steam, or noncondensible gases from any vessel or piping system containing primary coolant, liquid radwaste, or gaseous radwaste.

Damage occurs to a submerged, irradiated fuel assembly with the resultant release of significant quantities of noncondensible gases.

Alarms occur on CAMs.

A fire occurs involving radioactive materials.

(Refer to EP M-6)

Verified alarm on radiation monitors RE-14/14R, RE-28/28R, RE-29, RE-15/15R, or RE-24/24R.

A major radioactive material spill occurs.

6.

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PACIFIC GAS AND ELECTRIC COMPANY NUMBER EP R-2 DIABLO CANYON POWER PLANT REVISION 22 PAGE 3 OF 6 TITLE: Release of Airborne Radioactive Materials Initial UNITS 1 AND 2 Assessment PRECAUTIONS 6.1 Do not use SPDS to obtain RMS readings. Radiological Monitor readings off SPDS may be based on different units of measurement than required as input to the calculations.

6.2 If the Main Condenser is available during a SGTR event with a stuck open Safety Relief or 10% Steam Dump to atmosphere, there are two release pathways.

6.3 Obtain an independent verification of your calculation whenever time permits to confirm no errors or incorrect assumptions about plant conditions.

6.4 Default release rates are extremely conservative and may result in higher classifications or PARs than would be warranted if actual release indications were available.

6.5 N-16 will be detected on the MSL Radiation Monitors while at power and may cause a false high off-site dose calculation.

6.6 This procedure shall not be used to evaluate compliance to Technical Specifications during planned effluent releases. Such evaluations shall be performed by the Chemistry Department.

6.7 Fuel Handling Accident (FHA) in Containment with Equipment Hatch open is a special case. Use the analyzed default dose rates and doses listed in Attachment 10.1 and go directly to EP G-1 for comparison to the Emergency Action Levels (EALs).

7. INSTRUCTIONS NOTE: This calculation can be performed on the PPC using the turn-on code "EPR2."

7.1 RELEASE RATE CALCULATIONS 7.1.1 Obtain a working copy of Attachment 10.1.

7.1.2 Determine release source location as Plant Vent, Atmospheric Steam Release, or Unmonitored.

CAUTION: Do NOT use SPDS to obtain radiation monitor readings.

7.1.3 Gather and record the required information in accordance with the appropriate section of the form.

NOTE: Plant Vent Extended Range Rad Monitor RE-87 will automatically activate if the Normal Range Gas Monitors RE-14/14R approach their maximum reading.

7.1.4 Perform the required calculation to determine the release rate of Total Effluent and record the results in both this Attachment and Attachment 10.2.

7.1.5 If it is not possible to calculate a release rate, refer to the DEFAULT RELEASE RATES on Page 3 of Attachment 10.1 and choose the most appropriate value for input to Attachment 10.2. For an FHA in containment with equipment hatch open, use default dose rates and EP R-2.Doc 03B 0119.0419

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PACIFIC GAS AND ELECTRIC COMPANY NUMBER EP R-2 DIABLO CANYON POWER PLANT REVISION 22 PAGE 4 OF 6 TITLE: Release of Airborne Radioactive Materials Initial UNITS 1 AND 2 Assessment doses from Attachment 10.1 and go directly to EP G-1 for comparison to the EALs.

EP R-2.Doc 03B 0119.0419

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PACIFIC GAS AND ELECTRIC COMPANY NUMBER EP R-2 DIABLO CANYON POWER PLANT REVISION 22 PAGE 5 OF 6 TITLE: Release of Airborne Radioactive Materials Initial UNITS 1 AND 2 Assessment 7.2 OFF-SITE DOSE CALCULATIONS NOTE: Calculations may be performed using the PPC routine "EPR2," or by hand, as follows:

7.2.1 Obtain a working copy of Attachment 10.2.

7.2.2 Gather and record the required information in accordance with the appropriate section of the form.

NOTE: Plant Process Computer (PPC) Meteorological Data turn on codes are "METP" (Primary Data) and "METB" (Back-up Data).

7.2.3 Determine the appropriate activity source term and circle the associated DCFs to be used in Section 4A and 4B.

7.2.4 Perform the required calculations to determine the TEDE and THYROID CDE RATES.

7.2.5 Project the RELEASE DURATION in hours as input to determining projected doses.

7.2.6 If a duration cannot be projected, use the DEFAULT DURATION of 3 hours3.472222e-5 days 8.333333e-4 hours 4.960317e-6 weeks 1.1415e-6 months .

7.2.7 Perform the required calculations to determine the TEDE and THYROID CDE at the Site Boundary (800 meters).

7.3 REPORTING THE RESULTS 7.3.1 Refer to EP G-1 and compare the results of the above calculations with the Emergency Action Levels.

7.3.2 Refer to EP RB-10 and compare the results of the dose calculations with the PAR determination criteria.

7.4 Advise the ISEC of any EAL thresholds that are exceeded based on site boundary dose rates and doses, or the need to revise PARs due to changing conditions.

7.5 CONTINUOUS ACTIONS 7.5.1 As directed by the ISEC, continue to perform assessment of airborne releases to support evaluation of EAL status and PARs by repeating the above instructions.

7.5.2 Contact Chemistry to request:

a. A sample of the radioactive effluent (if possible) and in-plant airborne activity.

b. A confirmatory assessment of the site boundary dose rate from the release.

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PACIFIC GAS AND ELECTRIC COMPANY NUMBER EP R-2 DIABLO CANYON POWER PLANT REVISION 22 PAGE 6 OF 6 TITLE: Release of Airborne Radioactive Materials Initial UNITS 1 AND 2 Assessment

8. RECORDS 8.1 All checklists generated during activation of the EOF for drills and exercises are non-quality Good Business Records and shall be retained by Emergency Planning Group for three years.

8.2 All checklists generated during activation of the EOF for a real event are non-quality records and shall be retained in RMS in accordance with AD10.ID2.

9. APPENDICES None

10. ATTACHMENTS 10.1 ""Release Rate Calculations," 10/31/02 10.2 ""Off-Site Dose Calculations," 08/05/94

11. REFERENCES 11.1 EP G-1, "Accident Classification and Emergency Plan Activation."

11.2 EP G-2, "Activation and Operation of the Interim Site Emergency Organization (Control Room)."

11.3 EP RB-9, "Calculation of Release Rate."

11.4 EP RB-10, "Protective Action Recommendations."

11.5 EP RB-11, "Emergency Off-site Dose Calculations."

11.6 EP RB-12, "Mid and High Range Plant Vent Radiation Monitors."

11.7 EP M-6, "Fire."

11.8 NRS-RES Calculation No. RA 93-12, New Dose Conversion Factors for EP R-2 and RB-11, Validation and Verification, Rev. 1, 12/15/93.

11.9 NOS-RECE Calculation No. RA 93-04, EP RB-9, Calculation of Release Rate, Rev. 7 and R-2, Release of Airborne Radioactive Materials, Rev. 12, Validation and Verification, Rev. 0, 4/12/93.

11.10 SH&ES Calculation No. EP-94-01, Rev 0, EP R-2, Release of Airborne Radioactive Materials, Rev 17, Validation and Verification.

11.11 PG&E Calculation PAM-0-04-517, Rev. 4, 4/6/97 "Steam Generator Narrow Range Level Uncertainty."

11.12 PG&E Calculation STA-160, Freq., "Estimate of Expected Exposures Associated with a Fuel Handling Accident with Containment Open."

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10/31/02 Page 1 of 3 DIABLO CANYON POWER PLANT TITLE: Release Rate Calculations EP R-2 ATTACHMENT 10.1 1 2 AND PLANT VENT RELEASE

1. GENERAL INFORMATION Date: Time: Assessment No.

Assessment By: Unit Releasing

2. PLANT VENT FLOW RATE DETERMINATION A. DIRECT - Plant Vent Flow Rate FR-12 (0-30x104 CFM (CFM) = (CFM)

OR B. ALTERNATE - Operating Ventilation Equipment (Max No. possible) #Fans (CFM/Fan)

FHB Exhaust (1) x 35,750 = (CFM)

Aux Bldg Exhaust (2) x 73,500 = (CFM)

GE/GW Area (1) x 25,000 = (CFM)

Cont. Purge (1) x 55,000 = (CFM)

Cont. Hydrogen (1) x 300 = (CFM)

Plant Vent Flow Rate = (CFM)

3. RELEASE RATE CALCULATION CAUTION: Do NOT use SPDS to obtain monitor readings.

A. NOBLE GAS RELEASE RATE Circle Reading (Units) Conversion Plant Vent Noble Gas Monitor Factor Flow Rate Release Used (CFM) Rate (Ci/sec)

Primary RE-14/14R/87 µCi/cc x 4.72E-04 x Backup RE-29 mR/hr x 4.72E-06 x =

B. TOTAL EFFLUENT RELEASE RATE NOTE: Refer to Page 3 for criteria in choosing RCS, GAP, or CORE below.

Noble Gas Release Total Effluent Total Effluent Rate (Ci/sec) Conversion Factor Release Rate (Ci/sec) x 1.00 (RCS) =

1.11 (GAP) 1.50 (CORE)

NOTE: If it is not possible to calculate a release rate, refer to the DEFAULT RELEASE RATES on Page 3 of this attachment.

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10/31/02 Page 2 of 3 EP R-2 (UNITS 1 AND 2)

ATTACHMENT 10.1 TITLE: Release Rate Calculations

1. GENERAL INFORMATION ATMOSPHERIC STEAM RELEASE Date: Time: Assessment No.

Assessment By: Unit Releasing CAUTION: WHEN CRITICAL, N-16 ACTIVITY SEEN BY MSL RAD MONITORS CAUSES INVALID READINGS FOR OFFSITE DOSE. POST-TRIP, RE-7X READING IS VALID IF THE RE-7X MONITOR SHOWED AN INITIAL N-16 RESPONSE, OR RESPONDS TO CHECKSOURCE.

NOTE: If it is not possible to calculate a release rate, refer to the DEFAULT RELEASE RATES on Page 3.

2. STEAM RELEASES - Use this form to calculate steam releases to the atmosphere WHEN NOT critical.

A. Required Information (RUPTURED GENERATOR ONLY)

Check MSL Rad Reading S/G Lvl Level S/G Flow Flow Rate Ruptured S/G Monitor (cpm) Narrow (%) Rate (lbs/hr)

Range If <4E5 use 4E5 SG 1 RE-71 LI-517 FI-512 SG 2 RE-72 LI-527 FI-522 SG 3 RE-73 LI-537 FI-532 SG 4 RE-74 LI-547 FI-542 B. Alternate Steam Flow Rate (Only if the RUPTURED S/G Flow Rate is otherwise not available)

Valve Type # Valves Capacity (lbs/hr) Flow Rate (lbs/hr)

Lifted 10% Steam Dump (1 per S/G) x 4.0E+05 =

Safety Reliefs (5 per S/G) x 8.5E+05 =

Total Steam Flow Rate (lbs/hr) = (lbs/hr)

3. RADIATION MONITOR FACTORS (Determined based on S/G NR Level indication) (Enter in Section 4 below.)

S/G Level EMPTY NORMAL FLOODED Narrow Range < 4% 4% - 96% > 96%

Monitor Factor 6.08E-10 6.75E-10 3.07E-10 (DEFAULT)

4. RELEASE RATE CALCULATIONS A. TOTAL EFFLUENT RELEASE RATE (RE-7x)

MSL Monitor Reading Flow Rate Monitor Factor Total Effluent Release (cpm) (lbs/hr) Rate (Ci/sec) x x =

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10/31/02 Page 3 of 3 EP R-2 (UNITS 1 AND 2)

ATTACHMENT 10.1 TITLE: Release Rate Calculations

1. SOURCE TERM SELECTION AND DEFAULT RELEASE RATES NOTE: Use default release rate only if actual data is not available or if the release is not being monitored.

A. Check the accident type which most closely resembles the current event.

Default Release Source Rate (Ci/sec) Term Accident Source Condition LOCA (w/ core melt) 1.74 E+1 RE-30 or 31 >300R/hr CORE LOCA (w/o core melt) 5.74 E+0 RE-30 or RE-31 <300R/hr GAP RE-30 or RE-31 not on scale RCS Main Steam Line Break 8.61 E-3 RCS Feedwater Line Break 8.61 E-3 RCS Blackout 8.62 E-1 RCS Locked Rotor 1.57 E-2 GAP FHB Accident 1.45 E+1 GAP Rod Ejection 1.08 E-2 GAP GDT Rupture 4.14 E+1 RCS LHUT Rupture 3.10 E+1 RCS VCT Rupture 8.29 E-2 RCS S/G Tube Rupture 1.65 E+0 NR S/G Level < 4% SG - Empty NR S/G Level 4-96% SG - Normal NR S/G Level > 96% SG -

Flooded Containment FHA S.B. Dose TEDE = 13.4 mrem/hr Go Accident with Equip. Rates Thy.CDE = 51.4 mrem/hr Directly to Hatch Open S.B. Doses TEDE = 6.7 mrem EP G-1 Thy. CDE = 25.7 mrem B. Record the Default Release Rate in Attachment 10.2, Section 4 and use the DCF choice that is listed for the specific accident source above.

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08/05/94 Page 1 of 1 DIABLO CANYON POWER PLANT TITLE: Off-Site Dose Calculations EP R-2 ATTACHMENT 10.2 1 2 AND

1. GENERAL INFORMATION Date: Time: Assessment No.

Assessment By: Unit Releasing

2. METEOROLOGICAL DATA - PPC (Plant Process Computer)

Turn On Codes for Met Data are "METP" (Primary Data) or "METB" (Back-up Data)

Parameter Reading Units DEFAULT Wind Speed (10 Meter Level) meters/sec Wind Direction (10 Meter Level) Degrees Site Boundary X/Q (0.8 km) Sec/m3 5.29E-04

3. DCF Determination - Select the most appropriate source term for the DCF using the criteria in Attachment 10.1. Circle the corresponding DCF in Section 4 below.

4. DOSE CALCULATIONS - (From data calculated using Attachment 10.1)

A. TOTAL EFFECTIVE DOSE EQUIVALENT (TEDE)

Total Effluent or Site Projected TEDE Default Release Boundary DCF TEDE Release (mrem)

Rate X/Q (0.8 km) (circle one) Rate Duration (hr) 3 (Ci/sec) (Sec/m ) (mrem/hr (DEFAULT 3 hrs)

)

1.1E + 05 (RCS) 3.0E + 06 (Gap )

x x 1.1E + 07 (Core) = x =

Attachment 10.1 1.1E + 05 (SG-Empty) 4.3E + 04 (SG-Normal )

9.3E + 05 (SG-Flooded)

B. THYROID COMMITTED DOSE EQUIVALENT (CDE) (DO NOT COMPLETE FOR GDT, LHUT, OR VCT RUPTURE)

Total Effluent or Site Projected Default Release Boundary DCF Thyroid Release Thyroid Rate X/Q (circle one) CDE Duration (hr) CDE (Ci/sec) (0.8 km) Rate (DEFAULT 3 hrs) (mrem)

(Sec/m3) (mrem/h r) 1.5E + 06 (RCS) 6.5E + 07 (Gap )

x x 7.7E + 07 (Core) = x =

Attachment 10.1 1.5E + 06 (SG-Empty) 1.5E + 05 (SG-Normal )

1.4E + 07 (SG-Flooded)

5. REPORTING THE RESULTS - (Refer to Section 7.3 of Instructions for details)

A. Refer to EP G-1 for EAL criteria.

B. Implement EP RB-10 for PAR criteria EP R-2.Doc 03B 0119.0419

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PACIFIC GAS AND ELECTRIC COMPANY NUMBER EP G-1 NUCLEAR POWER GENERATION REVISION 33B DIABLO CANYON POWER PLANT PAGE 1 OF 3 EMERGENCY PLAN IMPLEMENTING PROCEDURE UNITS TITLE: Emergency Classification and Emergency Plan Activation 1 2 AND 07/30/04 EFFECTIVE DATE PROCEDURE CLASSIFICATION: QUALITY RELATED TABLE OF CONTENTS SECTION PAGE SCOPE ............................................................................................................................................. 1 DISCUSSION.................................................................................................................................... 1 DEFINITIONS ................................................................................................................................... 1 RESPONSIBILITIES ......................................................................................................................... 2 INSTRUCTIONS ............................................................................................................................... 3 RECORDS ........................................................................................................................................ 3 ATTACHMENTS ............................................................................................................................... 3 REFERENCES ................................................................................................................................. 3

1. SCOPE 1.1 This procedure describes accident classification guidelines and Emergency Plan activation responsibilities.

2. DISCUSSION 2.1 The steps required by this procedure are in addition to the steps required to maintain the plant in, or restore the plant to, a safe condition.

2.2 Events not meeting the minimum classification criteria contained in this procedure should be reviewed for reportability in XI1.ID2, "Regulatory Reporting Requirements and Reporting Process."

2.3 Copies of the Emergency Action Level Classification Chart (Attachment 7.1) are provided as job aids in the following locations: JMC EPIM Office (2), EOF, Recovery Manager Office, TSC Site Emergency Coordinator Office, Unit 1 crash cart, Unit 2 crash cart, and the Simulator crash cart (Ref. OP1.DC23, "Control of Posted Plant Signs and Information").

3. DEFINITIONS 3.1 Emergency Classification Levels (ECLs) 3.1.1 Notification of Unusual Event (NUE) - characterized by off-normal conditions that:

a. May not in themselves be particularly significant from an emergency preparedness standpoint, but could reasonably indicate a potential degradation of the level of safety of the plant if proper action is not taken or if circumstances beyond the control of the operating staff render the situation more serious from a safety stand point. No releases of radioactive material requiring off-site response or monitoring are expected.

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PACIFIC GAS AND ELECTRIC COMPANY NUMBER EP G-1 DIABLO CANYON POWER PLANT REVISION 33B PAGE 2 OF 3 TITLE: Emergency Classification and Emergency Plan Activation UNITS 1 AND 2 3.1.2 Alert - events in progress or having occurred, involving an actual or potentially substantial degradation of the plant safety level.

a. Small releases of radioactivity may occur (greater than Technical Specification limits for normal operation, but only a small fraction of the EPA Protective Action Guideline (PAG) exposure levels at the site boundary). It is the lowest level where emergency offsite response may be anticipated.

b. The lowest classification level where off-site emergency response is anticipated.

3.1.3 Site Area Emergency (SAE) - events which are in progress or have occurred involving actual or likely major failures of plant functions needed for protection of the public, but a core meltdown situation is not indicated based on current information.

a. Any releases are not expected to exceed EPA Protective Action Guides except near the site boundary. However, because the possible release is significant, care must be taken in alerting offsite authorities to distinguish whether the release is merely potential, likely, or actually occurring. Response of offsite authorities will be guided initially by this determination.

3.1.4 General Emergency (GE) - event(s) in progress or having occurred which indicate:

a. Imminent substantial core degradation or melting.

b. Potential for containment loss.

c. Radioactive releases can be reasonably expected to exceed EPA PAGs off-site for more than the immediate area.

4. RESPONSIBILITIES 4.1 Interim Site Emergency Coordinator (Interim SEC or ISEC) - Control room shift manager is responsible for initial event classification and emergency plan activation. The ISEC may upgrade the event classification until relieved by either the SEC or RM. In addition, the ISEC may downgrade a NUE to no ECL.

4.2 Site Emergency Coordinator (SEC) - The SEC may upgrade the classification of an event until relieved by the recovery manager.

4.3 Recovery Manager (RM) - The RM, once staffed, is responsible for upgrading or downgrading ECLs, and may direct the SEC to change ECLs.

5.

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PACIFIC GAS AND ELECTRIC COMPANY NUMBER EP G-1 DIABLO CANYON POWER PLANT REVISION 33B PAGE 3 OF 3 TITLE: Emergency Classification and Emergency Plan Activation UNITS 1 AND 2 INSTRUCTIONS 5.1 The Interim Site Emergency Coordinator shall:

5.1.1 Initially classify and declare the event using ONLY the guidance in Attachment 7.1 of this procedure.

NOTE: Simultaneous EALs that increase the probability of release require escalation of the ECL to one level above the higher EAL.

5.1.2 Formally announce all emergency classification declarations to the control room, TSC, or EOF, respectively.

5.2 The ISEC or SEC may:

5.2.1 Upgrade the event to a higher ECL until the recovery manager arrives at and assumes responsibility in the EOF. However, the ISEC and SEC shall not downgrade an event classified at the Alert or higher level at any time. The ISEC may downgrade a NUE to no ECL.

5.2.2 Only the recovery manager may downgrade an ECL at the Alert or higher level according to the most current controlling EAL.

6. RECORDS 6.1 There are no quality or nonquality records generated by this procedure.

7. ATTACHMENTS 7.1 "Emergency Action Level Classification Chart," 07/28/04

8. REFERENCES 8.1 EP EF-1, "Activation and Operation of the Technical Support Center."

8.2 EP EF-2, "Activation and Operation of the Operational Support Center."

8.3 EP EF-3, "Activation and Operation of the Emergency Operations Facility."

8.4 EP OR-3, "Emergency Recovery."

8.5 EP G-3, "Emergency Notification of Off-Site Agencies."

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07/28/04 Page 1 of 17 DIABLO CANYON POWER PLANT TITLE: Emergency Action Level Classification Chart EP G-1 ATTACHMENT 7.1 1 2 AND UNUSUAL EVENT ALERT SITE AREA EMERGENCY GENERAL EMERGENCY I. FIRE 1. Fire not under control within 15 1. Fire not under control within 1. Fire causing the complete 1. Site Emergency (All Modes) minutes of initiating fire fighting 15 minutes of initiating fire loss of function of any one of Coordinator judges that a efforts AND affecting plant fighting efforts AND the following safety related fire could cause common equipment or power supplies in threatening the loss of systems required for safe damage to plant systems or near the Protected Area(s). function of any of the shutdown: which is determined to have following Safety Related - Vital Power Supplies: the potential to release systems required for safe D/Gs, DFOT, Vital 4kV, radioactive material in shutdown: 480V, 120VAC, or quantities sufficient to

- Vital Power Supplies: 125VDC cause exposures D/Gs, DFOT, Vital 4kV, - Primary Systems and comparable to General 480V, 120VAC, or Auxiliaries: RCS, CCW, Emergency #4.

125VDC RHR, or Charging and

- Primary Systems and Boration Auxiliaries: RCS, CCW, - Heat Sinks: AFW, ASW, RHR, or Charging and 10% Dumps, S/G Boration Safeties, or MSIVs

- Heat Sinks: AFW, ASW, - Control Room, Cable 10% Dumps, S/G Spreading Rooms, or Safeties, or MSIVs HSDP.

- Control Room, Cable Spreading Rooms, or HSDP.

II. FUEL 2. Indication of Fuel Damage as 2. Indication of Fuel Damage as See SAE #14 for Steam Line 2. Degraded core with DAMAGE shown by: shown by: Break possible loss of coolable OR geometry as indicated by:

VESSEL Confirmed RCS sample shows Confirmed RCS sample DAMAGE >300 µCi/cc of equivalent 5 or more thermocouple

>100/ E µCi/gm specific (Modes 1-4) I-131 specific activity OR readings activity (Tech Spec 3.4.16) equivalent fuel failure is > 1200 deg. F.

OR measured by exposure rate OR Confirmed RCS sample shows from systems carrying LOCA with no indication of dose equivalent I-131 activity > reactor coolant per ECCS flow AND indication Tech Spec limit for Iodine EP RB-14A of fuel damage (See Alert Spike (Tech Spec Fig. 3.4-1). #2)

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ATTACHMENT 7.1 TITLE: Emergency Action Level Classification Chart OR LOCA with containment rad levels > values for 100%

gap release in EP RB-14.

Category II Continued on next Category II Continued on Category II Continued on next Category II Continued on page. next page. page. next page.

NOTE: SIMULTANEOUS EALS THAT INCREASE THE PROBABILITY OF RELEASE REQUIRE ESCALATION OF THE CLASSIFICATION TO ONE LEVEL ABOVE THE HIGHER EAL.

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ATTACHMENT 7.1 TITLE: Emergency Action Level Classification Chart UNUSUAL EVENT ALERT SITE AREA EMERGENCY GENERAL EMERGENCY II. FUEL 3. Pressurized Thermal Shock is 3. Loss of 2 of 3 Fission DAMAGE verified by entry into EOP Product Barriers:

OR FR-P.1 VESSEL AND A) Indication of fuel damage DAMAGE Left of Limit A curve (EOP F-0). (See Alert #2)

(Modes 1-4) AND (Continued) Determination of a Steam Generator Tube Rupture (SGTR) which requires entry into EOP E-3 AND Steam release from ruptured S/G, either used for plant cooldown purposes or due to a steamline break.

B) Indication of Fuel Damage (See Alert #2)

AND Determination of a SGTR requiring entry into EOP E-3 AND Indication of a steam line break inside containment AND High potential for loss of containment integrity (e.g., loss of function of both Containment Spray trains OR loss of function of one Containment Spray train and four CFCUs).

NOTE: SIMULTANEOUS EALS THAT INCREASE THE PROBABILITY OF RELEASE REQUIRE ESCALATION OF THE CLASSIFICATION TO ONE LEVEL ABOVE THE HIGHER EAL.

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ATTACHMENT 7.1 TITLE: Emergency Action Level Classification Chart UNUSUAL EVENT ALERT SITE AREA EMERGENCY GENERAL EMERGENCY II. FUEL 3. Pressurized Thermal Shock is C) Indication of Fuel DAMAGE verified by entry into EOP Damage (See Alert #2)

OR FR-P.1 AND VESSEL AND Determination of a SGTR DAMAGE Left of Limit A curve (EOP F-0). which requires entry into (Modes 1-4) (Continued) EOP E-3 (Continued) AND Indication of a steam line break outside containment with inability to isolate the break.

D) Potential fuel damage indicated by 5 or more thermocouple readings

>700 deg. F or RVLIS

<32%

AND LOCA as indicated by RCS leakage and SI AND Loss of containment integrity.

III. FUEL 3. Fuel Handling Accident 2. Fuel Handling Accident HANDLIN causing a release in causing a release in G Containment or the Fuel Containment or the Fuel ACCIDEN Handling Building Handling Building T WITH WITH (All Modes) The potential to exceed the The potential to exceed the criteria listed in Alert #4 or criteria listed in SAE #3.

5.

IV. LOSS OF 4. Projected dose rate at the Site 4. Projected dose rate at the 3. Projected dose at the Site 4. Projected dose at the Site CONTROL Boundary (800 meters) is Site Boundary (800 meters) Boundary (800 meters) is Boundary (800 meters) is OR 0.057 mRem/hr TEDE is 100 mRem TEDE 1,000 mRem TEDE RELEASE OR 0.57 mRem/hr TEDE OR OR OF 0.170 mRem/hr Thyroid CDE OR 500 mRem Thyroid CDE for 5,000 mRem Thyroid CDE RADIOAC for actual or expected release. 1.7 mRem/hr Thyroid CDE actual or expected release. for actual or expected TIVE for actual or expected release.

MATERIAL release.

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ATTACHMENT 7.1 TITLE: Emergency Action Level Classification Chart (All Modes)

Category IV Continued on next Category IV Continued on next Category IV Continued on next Category IV Continued on next page. page. page. page.

NOTE: SIMULTANEOUS EALS THAT INCREASE THE PROBABILITY OF RELEASE REQUIRE ESCALATION OF THE CLASSIFICATION TO ONE LEVEL ABOVE THE HIGHER EAL.

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ATTACHMENT 7.1 TITLE: Emergency Action Level Classification Chart UNUSUAL EVENT ALERT SITE AREA EMERGENCY GENERAL EMERGENCY IV. LOSS OF 5. A valid reading in excess of the 5. Valid alarm on plant vent CONTROL isolation setpoint, which fails to high range noble gas monitor OR isolate the release on any of RE-29.

RELEASE the Radiological Process NOTE: ALARMS AT STATE OF Effluent Monitors:

OES SACRAMENTO.

RADIOAC RE-18 OR RE-23 TIVE During discharge only.

MATERIAL (All Modes)

(Continued

)

6. An actual liquid release which 6. An actual liquid release exceeds the limits of which exceeds 10x the limits 10 CFR 20, Appendix B, of 10 CFR 20, Appendix B, Table 2, Col. 2 per CY2.ID1. Table 2, Col. 2 per CY2.ID1.

7. Radiological Effluent Process 7. Unplanned or unanticipated Monitor High Radiation Alarm increase of 1 R/hr or greater with valid reading in excess of in any of the following areas:

alarm setpoint on any of the Passageways, OR following monitors: Normally occupied areas, OR RE-14/14R Accessible areas normally RE-24/24R < 100 mR/hr, OR RE-28/28R. Outside boundaries of Radiologically Controlled Areas AND, for any area above, a potential exists for EITHER an uncontrolled release to the environment OR a loss of ability to maintain plant safety functions.

8. Unplanned or uncontrolled 8. Unexplained increase of release to the environment 50 X DAC in airborne exceeding alarm setpoints on radioactivity outside the RE-3. boundary of the Radiologically Controlled Areas, but within the Plant EP G-1.Doc 03B 0119.1622

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ATTACHMENT 7.1 TITLE: Emergency Action Level Classification Chart Protected Area.

NOTE: SIMULTANEOUS EALS THAT INCREASE THE PROBABILITY OF RELEASE REQUIRE ESCALATION OF THE CLASSIFICATION TO ONE LEVEL ABOVE THE HIGHER EAL.

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ATTACHMENT 7.1 TITLE: Emergency Action Level Classification Chart UNUSUAL EVENT ALERT SITE AREA EMERGENCY GENERAL EMERGENCY V. LOSS OF 9. Entry into OP AP-8A, 4. Entry into OP AP-8A, "Control CONTROL "Control Room Accessibility," Room Accessibility," AND ROOM AND controls established controls not established within (All Modes) within 15 minutes. 15 minutes.

VI. LOSS 9. Plant is not brought to required 5. Complete loss for greater 5. Loss of Heat Sink indicated OF operating Mode within any than 15 minutes of any of the by:

ENGINEER applicable Tech Spec Action following functions needed to Entry into EOP FR H.1 ED Statement time limit (Modes reach or maintain Hot AND SAFETY 1-4). Shutdown (while in Modes Loss of water inventory in 3 FEATURE 1-4): S/Gs (<23% [34%] Wide Range).

10. Loss of function of both 10. Loss of function of both RHR AFW capability RHR trains for greater than 15 trains for greater than Steam Dump System and minutes while in Mode 5-or 6. 15 minutes in Modes 1-4.

S/G Safety Valves

11. A loss of function of all 11. An unplanned shutdown of Loss of the capability to charging pumps for greater the RHR System (while in maintain RCS inventory as than 15 minutes when normally Mode 5 or 6) for > 1 hour1.157407e-5 days 2.777778e-4 hours 1.653439e-6 weeks 3.805e-7 months evidenced by a loss of all used for RCS inventory control with no other normal means charging pumps coincident (Modes 1-4). of decay heat removal with the inability to available (e.g., flooded depressurize and inject with reactor cavity or steam the Safety Injection pumps generators with loops filled).

12. An unplanned loss of function Loss of capability to increase of the RHR System (Mode 5 the Boric Acid concentration or 6) for greater than sufficient to maintain Keff less 15 minutes than .99 in Mode 4 with a loss AND of capability to trip control RCS thermocouple rods temperature is projected to ASW or CCW Systems exceed 200 deg.F within 1 hour1.157407e-5 days 2.777778e-4 hours 1.653439e-6 weeks 3.805e-7 months of RHR loss (see Loss of electrical power or Appendix B of OP AP SD I&C for any of the above series) listed systems, causing a OR complete loss of function.

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ATTACHMENT 7.1 TITLE: Emergency Action Level Classification Chart deg.F.

NOTE: SIMULTANEOUS EALS THAT INCREASE THE PROBABILITY OF RELEASE REQUIRE ESCALATION OF THE CLASSIFICATION TO ONE LEVEL ABOVE THE HIGHER EAL.

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ATTACHMENT 7.1 TITLE: Emergency Action Level Classification Chart UNUSUAL EVENT ALERT SITE AREA EMERGENCY GENERAL EMERGENCY VII. 12. Loss of all off-site power for 13. Loss of all off-site power for 6. Loss of all on-site AND off-site See General Emergency LOSS OF greater than 15 minutes AND greater than 15 minutes AND AC power for > 15 minutes Condition #5 under LOSS POWER OR at least 2 D/Gs are supplying only 1 D/G is supplying its (Modes 1-4). OF ENGINEERED SAFETY ALARMS OR their vital busses (Modes 1-4). vital bus (Modes 1-4). FEATURE.

ASSESSMENT OR COMMUNICAT 13. Loss of all off-site power for 14. Loss of all off-site and on-site IONS greater than 15 minutes AND AC power for greater than at least 1 D/G is supplying its 15 minutes in Modes 5 or 6.

vital bus (Modes 5 and 6).

14. Loss of all vital DC power as 15. Loss of all vital DC power as 7. Loss of all vital DC power as indicated by DC Bus 11(21), indicated by DC Bus 11(21), indicated by DC Bus 11 (21), 12 12(22), and 13(23) 12 (22) and 13 (23) (22) and 13 (23) undervoltage undervoltage for > 15 minutes undervoltage for < for > 15 minutes (Modes 1-4).

(Modes 5-and 6) 15 minutes (Modes 1-4).

15. Loss of assessment capabilities as indicated by a total loss of SPDS in the Control Room AND simultaneous loss of all displays for any "Accident Monitoring" variable in Tech Spec Table 3.3.3-1 for > 1 hour1.157407e-5 days 2.777778e-4 hours 1.653439e-6 weeks 3.805e-7 months while in Modes 1, 2 or 3.

16. Main Control Room 16. Main Control Room 8. Main Control Room Annunciators PKs 1 through 5 Annunciators PKs 1 through Annunciators PKs 1 through 5 AND display capabilities AND 5 AND display capabilities AND display capabilities AND the seismically qualified AND the seismically qualified the seismically qualified annunciator display all do not annunciator display all do not annunciator display all do not respond to an alarm condition respond to an alarm condition respond to an alarm condition in Modes 1-4 for over in MODES 1-4 for over in MODES 1-4 for over 15 minutes. 15 minutes 15 minutes AND AND the plant is in a significant the plant is in a significant transient (plant trip, SI, or transient AND backup, generator runback nonannunciating systems are EP G-1.Doc 03B 0119.1622

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ATTACHMENT 7.1 TITLE: Emergency Action Level Classification Chart

>25 Mw/min), not available (PPC, SPDS).

nonannunciating systems available.

NOTE: SIMULTANEOUS EALS THAT INCREASE THE PROBABILITY OF RELEASE REQUIRE ESCALATION OF THE CLASSIFICATION TO ONE LEVEL ABOVE THE HIGHER EAL.

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ATTACHMENT 7.1 TITLE: Emergency Action Level Classification Chart UNUSUAL EVENT ALERT SITE AREA EMERGENCY GENERAL EMERGENCY VII. 17. Total loss of communication LOSS OF capability with off-site agencies POWER OR (all Modes) as indicated by the ALARMS OR inability to communicate with ASSESSMENT SLO County (by telephone and OR radio)

COMMUNICAT OR IONS the NRC Operations Center.

(Continued)

VIII. 18. Ground motion felt and 17. Earthquake > 0.2 g verified 9. Earthquake > 0.4 g verified by 6. Site Emergency NATURAL recognized as an earthquake by Seismic Monitors. Seismic Monitors. Coordinator's judgment PHENOMENA by a consensus of Control that major internal or (All Modes) Room operators on duty AND external events (e.g.,

measuring greater than 0.01g earthquakes, wind on the Earthquake Force damage, explosions, etc.)

Monitor. which could cause massive common damage to plant systems which is determined to have the potential to release radioactive material in quantities sufficient to cause exposures comparable to General Emergency #4.

19. Flooding of any plant structure 18. High water exceeding Intake 10. High water causing that causes initiation of entry to Structure main deck elevation flooding of ASW pump Mode 3 due to a Tech Spec or low water causing compartments or low water action statement. cavitation and shutdown of causing the shutdown of both both ASW pumps for < 15 ASW pumps for > 15 minutes.

minutes.

20. Tsunami or Hurricane Warning 19. Sustained wind of 85 mph 11. Sustained wind speed >

from the State, NOAA, NWS, (38 m/sec) at any elevation 100 mph (45 m/sec). at any Coast Guard or System on the Met. Tower. elevation on the Met. Tower.

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ATTACHMENT 7.1 TITLE: Emergency Action Level Classification Chart water levels at the Intake Structure indicative of a Tsunami or Hurricane.

21. A tornado sighted within Site 20. Tornado strikes the plant Boundary. protected area.

NOTE: SIMULTANEOUS EALS THAT INCREASE THE PROBABILITY OF RELEASE REQUIRE ESCALATION OF THE CLASSIFICATION TO ONE LEVEL ABOVE THE HIGHER EAL.

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ATTACHMENT 7.1 TITLE: Emergency Action Level Classification Chart UNUSUAL EVENT ALERT SITE AREA EMERGENCY GENERAL EMERGENCY IX. 22. Report of airplane crash within 21. Confirmed missile, airplane 12. Missile, airplane crash or See General Emergency #6 OTHER the Site Boundary or unusual crash or explosion involving a explosion causing complete above.

HAZARDS airplane activity threatening plant structure in the loss of a safety system (All Modes) the plant. protected area. function that causes entry into a Tech Spec Action Statement.

23. Confirmed explosion on-site.

24. Turbine failure causing casing 22. Turbine failure generating penetration OR damage to missiles that cause visual turbine or generator seals damage to other safety related structures, equipment, controls OR power supplies.

25. Significant release of 23. Release of flammable OR flammable OR toxic gas OR toxic gas OR liquid that liquid that prevents, even with jeopardizes operation of SCBAs, operations inside the safety related systems by power block OR intake either preventing required structure (ref. CP M-9a). access OR by threatening imminent damage.

X. 26. RCS unidentified OR pressure 24. Primary leak rate >50 gpm. 13. Known primary system LOCA See General Emergency #3 PRIMARY OR boundary leakage that during which RCS subcooling under Fuel or Vessel PRI/SEC OR exceeds 10 gpm OR identified cannot be maintained >20F Damage.

SECONDARY leakage that exceeds 25 gpm. OR PZR level cannot be LEAK) maintained >4% (28% with adverse containment).

(Modes 1-4 27. SI Actuation with ECCS 25. Determination of a SGTR 14. Determination of a SGTR injection into the RCS which results in entry into coincident with steam release resulting from a valid signal EOP E-3. from ruptured S/G, either used based on actual plant for plant cooldown purposes or conditions. due to a steamline break.

NOTE: SI ACTUATION ALSO ALARMS AT OES IN SACRAMENTO.

28. Steam line break which 26. Determination of a steam line results in SI actuation. break with >10 gpm Primary EP G-1.Doc 03B 0119.1622

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ATTACHMENT 7.1 TITLE: Emergency Action Level Classification Chart to Secondary leakage.

29. Failure of a PZR PORV AND Block Valve OR Safety Valve fails to reseat, excluding allowable leakage, following a pressure reduction below the reset point.

NOTE: SIMULTANEOUS EALS THAT INCREASE THE PROBABILITY OF RELEASE REQUIRE ESCALATION OF THE CLASSIFICATION TO ONE LEVEL ABOVE THE HIGHER EAL.

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ATTACHMENT 7.1 TITLE: Emergency Action Level Classification Chart UNUSUAL EVENT ALERT SITE AREA EMERGENCY GENERAL EMERGENCY XI. 27. Anticipated Transient Without 15. An ATWS condition with no 7. ATWS with Fuel Damage REACTOR Scram (ATWS) as indicated fuel damage evident indications (see Alert PROTECTION by: AND Condition #2 under FUEL SYSTEM Failure of an automatic An additional failure of a DAMAGE)

FAILURE reactor trip to trip the reactor. system required for Hot OR (Modes 1-4) Shutdown (See SAE #5) to ATWS with potential Core actuate. Melt indicated by 5 or more thermocouple readings > 700 deg. F AND RVLIS < 32%.

XII. 30. Security reports the notification 28. Security reports ongoing 16. Security reports ongoing 8. Security reports ongoing SECURITY of a credible site-specific security threat involving physical attack on the facility or security threat which THREAT (All security threat or attempted physical attack on the facility a sabotage device causing a causes loss of control of Modes) entry or attempted sabotage. or a sabotage device has confirmed loss of function of the operations of the been detected that threatens any one of the following safety plant to hostile forces.

the operability of safety related systems required for related equipment (see safe shutdown:

Alert #1). - Vital Power Supplies:

D/Gs, DFOT, Vital 4kV, 480V, 120VAC, or 125VDC

- Primary Systems and Auxiliaries: RCS, CCW, RHR, or Charging and Boration

- Heat Sinks: AFW, ASW, 10% Dumps, S/G Safeties, or MSIVs

- Control Room, Cable Spreading Rooms, or HSDP.

XIII. 31. Site Emergency Coordinator 29. Site Emergency Coordinator 17. Site Emergency Coordinator 9. Site Emergency SITE determines conditions warrant judges plant conditions exist judges that conditions exist Coordinator judges EMERGENCY increased awareness on the that warrant precautionary that warrant activation of the conditions exist which COORDINATO part of off-site authorities of activation of the TSC and emergency centers and have a potential to R'S initiation of a plant shutdown placing the EOF and other monitoring teams or a release radioactive JUDGMENT per Tech Spec LCOs or involve key emergency personnel on precautionary notification to material in quantities (All Modes) other than normal controlled stand-by. the public near the site. sufficient to cause EP G-1.Doc 03B 0119.1622

- - UNCONTROLLED PROCEDURE - DO NOT USE TO PERFORM WORK or ISSUE FOR USE ***

07/28/04 Page 17 of 16 EP G-1 (UNITS 1 AND 2)

ATTACHMENT 7.1 TITLE: Emergency Action Level Classification Chart shutdown. exposures comparable to General Emergency #4.

NOTE: SIMULTANEOUS EALS THAT INCREASE THE PROBABILITY OF RELEASE REQUIRE ESCALATION OF THE CLASSIFICATION TO ONE LEVEL ABOVE THE HIGHER EAL.

EP G-1.Doc 03B 0119.1622

ES-301 Control Room/In-Plant Systems Outline Form ES-301-2 Facility: ______DCPP______________________ Date of Examination: __02/07/2005__

Exam Level (circle one): RO / SRO-I / SRO-U Operating Test No.: ______01_________

Control Room Systems@ (8 for RO; 7 for SRO-I; 2 or 3 for SRO-U)

System / JPM Title Type Code* Safety Function

a. 004 - Dilution w/o Makeup Control Operable (NRCLJC-301) RO/SROI A/N/S 01

b. W/E05 - Initiate Bleed and Feed for Loss of Heat Sink (NRCLJC-116) D/E/L/P/S 04p RO/SROI

c. 009 - Respond to Loss of RHR Inventory (Mode 5) (NRCLJC-093) A/E/L/M/S 03 RO/SROI/SROU

d. 008 - Respond to a Loss of CCW (NRCLJC-103) RO/SROI A/E/L/M/S 08

e. 026 - Secure Containment Spray (NRCLJC-081) RO/SROI E/D/S 05

f. 059- Establish MFW (NRCLJC-052) RO/SROI D/S 04s

g. 062 - Crosstie Vital Bus G to H (NRCLJC-032) RO D/E/L/S 06 A/E/M/S 02

h. 006 - Align SIS for Hot Leg Recirc (NRCLJC-123) RO/SROI/SROU In-Plant Systems@ (3 for RO; 3 for SRO-I; 3 or 2 for SRO-U)

i. 024 - Align Emergency Boration (NRCLJP-088) RO/SROI/SROU D/E/R 01

j. 062 - Transfer PZR Htr Group 13 to Backup (NRCLJP-079) D/L 06 RO/SROI/SROU D/E/L/P/R 04S

k. 061 - Reset TDAFW Pump (NRCLJP-012) RO/SROI/SROU

@ All control room (and in-plant) systems must be different and serve different safety functions; in-plant systems and functions may overlap those tested in the control room.

Type Codes Criteria for RO / SRO-I / SRO-U (A)lternate path 4-6 / 4-6 / 2-3 (C)ontrol room (D)irect from bank 9/8/4 (E)mergency or abnormal in-plant 1/1/1 (L)ow-Power 1/1/1 (N)ew or (M)odified from bank including 1(A) 2/2/1 (P)revious 2 exams 3 / 3 / 2 (randomly selected)

(R)CA 1/1/1 (S)imulator

NUCLEAR POWER GENERATION DIABLO CANYON POWER PLANT JOB PERFORMANCE MEASURE Number: NRCLJC-032

Title:

CROSSTIE OF VITAL BUS G TO H Examinee:

Evaluator:

Print Signature Date Results: Sat Unsat Total Time: minutes Comments:

References:

EOP ECA-0.3, Restore 4kV Buses, Appendix X, Rev. 12 Alternate Path: Yes No X Time Critical: Yes No X Time Allotment: 15 minutes Critical Steps: 2, 3, 4, 5, 8, 9, 10, 11, 12 Job Designation: RO/SRO Task Number: 062/06/A2.05 Rating: 2.9/3.3 AUTHOR: JACK BLACKWELL DATE: 01/18/2005 REVIEWED BY: N/A DATE:

TRAINING LEADER APPROVED BY: N/A DATE:

LINE MANAGER REV.01

JPM TITLE: CROSSTIE OF VITAL BUS G TO H JPM NUMBER: NRCLJC-032 INSTRUCTOR WORKSHEET Directions: No PLANT controls or equipment are to be operated during the performance of this Job Performance Measure. All actions taken by the examinee should be clearly demonstrated and verbalized to the evaluator. The student will be given the initial conditions, initiating cue, and task standard. The examiner will then ask if any clarifications are needed. The examinee may be given the applicable procedure and step with which to begin.

Required Materials: None Initial Conditions: A reactor trip and safety injection has occurred concurrent with a loss of all off-site power. Diesel generator 11 and diesel generator 13 have failed due to lube oil pressure problems. Diesel generator 12 is supplying 4kV bus G. CCW Pp 12 has failed resulting in a complete loss of CCW flow.

Initiating Cue: The Shift Foreman directs you to crosstie 4kV bus G to 4kV bus H per EOP ECA-0.3, Appendix X, commencing at step 3. Steps 1 and 2 have been completed. The Site Emergency Coordinator has concurred with this implementation.

Task Standard: 4kV and 480V bus H are energized after being crosstied to 4kV bus G in accordance with ECA-0.3.

NRCLJC032.DOC PAGE 2 OF 7 REV.01

JPM TITLE: CROSSTIE OF VITAL BUS G TO H JPM NUMBER: NRCLJC-032 INSTRUCTOR WORKSHEET Start Time:

Step Expected Operator Actions

1. Obtain the correct procedure. 1.1 References ECA-0.3, Appendix X.

Step was: Sat: ______ Unsat _______*

- 2. Cut in the DIR PWR, LOSS OF 2.1 Places D/G DIR PWR, LOSS OF FIELD, & BKR OC PROT RLYS FLD & BKR OC PROT RLYS C/O for diesel generator 12. SW to CUT-IN. **

Step was: Sat: ______ Unsat _______*

- 3. Reset SI. 3.1 Checks PK08-21 Safety Injection Actuation status.

3.2 Manually depresses both SI Reset pushbuttons, if required. **

3.3 Checks at least one of the following:

Monitor Light Box B Safety Injection red light OFF, OR

PK08-21, Safety Injection Actuation not ON.

Step was: Sat: ______ Unsat _______*

- 4. Cutout the auto transfer FCOs for 4.1 Places all Xfer to S/U PWR C/O 4kV and 12kV buses. toggle switch to CUT-OUT. **

Step was: Sat: ______ Unsat _______*

- 5. Depress all auto transfer reset 5.1 Reads NOTE.

pushbuttons.

5.2 Depresses all AUTO XFER RESET pushbuttons, if required. **

5.3 Verifies that all Auto Xfer indicating blue lights are off.

Step was: Sat: ______ Unsat _______*

Denotes an entry required on the JPM cover sheet.

- Denotes Critical Step and Sub Steps.

NRCLJC032.DOC PAGE 3 OF 7 REV.01

JPM TITLE: CROSSTIE OF VITAL BUS G TO H JPM NUMBER: NRCLJC-032 INSTRUCTOR WORKSHEET Step Expected Operator Actions

6. Verify OPEN all vital 4kV bus 6.1 Verifies all vital 4kV bus aux feeder auxiliary feeder breakers. breakers are OPEN:

52-HH-13 OPEN

52-HG-13 OPEN

52-HF-13 OPEN Step was: Sat: ______ Unsat _______*

7. Verify OPEN all vital 4kV bus 7.1 Verifies all vital 4kV bus startup startup feeder breakers. feeder breakers are OPEN:

52-HH-14 OPEN

52-HG-14 OPEN

52-HF-14 OPEN Step was: Sat: ______ Unsat _______*

- 8. Verify OPEN the 4kV startup 8.1 Opens 52-HG-15. **

feeder breaker 52-HG-15.

8.2 Verifies that 52-HG-15 has opened.

Step was: Sat: ______ Unsat _______*

- 9. Verify OPEN the 4kV to 480 9.1 Opens 52-HH-10. **

VAC bus feeder breaker for the deenergized bus to be reenergized. 9.2 Verifies that 52-HH-10 has opened.

Step was: Sat: ______ Unsat _______*

- 10. Close 4kV startup feeder breaker 10.1 Reads CAUTION and NOTE.

for the deenergized bus being reenergized. 10.2 Inserts sync key for 4kV bus H startup feeder breaker 52-HH-14.

10.3 Turns sync switch to ON. **

10.4 Closes 52-HH-14. **

10.5 Verifies that 52-HH-14 has closed.

Step was: Sat: ______ Unsat _______*

Denotes an entry required on the JPM cover sheet.

- Denotes Critical Step and Sub Steps.

NRCLJC032.DOC PAGE 4 OF 7 REV.01

JPM TITLE: CROSSTIE OF VITAL BUS G TO H JPM NUMBER: NRCLJC-032 INSTRUCTOR WORKSHEET Step Expected Operator Actions

- 11. Close the 4kV startup feeder 11.1 Inserts sync key for 4kV bus G breaker for the bus that will be startup feeder breaker 52-HG-14.

supplying power to the deenergized bus. 11.2 Turns sync switch to ON. **

11.3 Closes 52-HG-14. **

11.4 Verifies that 52-HG-14 has closed.

11.5 Verifies running diesel generator remains stable.

Step was: Sat: ______ Unsat _______*

- 12. Close the 4kV to 480V bus feeder 12.1 Reads CAUTION.

breaker for the reenergized bus.

Cue: An Operator has been stationed at VB4 to monitor the diesel generator.

12.2 Closes 52-HH-10. **

12.3 Verifies that 52-HH-10 has closed.

Step was: Sat: ______ Unsat _______*

Stop Time:

Total Time: (Enter total time on the cover page)

Denotes an entry required on the JPM cover sheet.

- Denotes Critical Step and Sub Steps.

NRCLJC032.DOC PAGE 5 OF 7 REV.01

JPM NUMBER: NRCLJC-032 EXAMINEE CUE SHEET Initial Conditions: A reactor trip and safety injection has occurred concurrent with a loss of all off-site power. Diesel generator 11 and diesel generator 13 have failed due to lube oil pressure problems. Diesel generator 12 is supplying 4kV bus G. CCW Pp 12 has failed resulting in a complete loss of CCW flow.

Initiating Cue: The Shift Foreman directs you to crosstie 4kV bus G to 4kV bus H per EOP ECA-0.3, Appendix X, commencing at step 3. Steps 1 and 2 have been completed. The Site Emergency Coordinator has concurred with this implementation.

Task Standard: 4kV and 480V bus H are energized after being crosstied to 4kV bus G in accordance with ECA-0.3.

NRCLJC032.DOC PAGE 6 OF 7 REV.01

JPM TITLE: CROSSTIE OF VITAL BUS G TO H JPM NUMBER: NRCLJC-032 ATTACHMENT 1, SIMULATOR SETUP Initialize the simulator to the IC-510 (100%, MOL).

Enter drill file 1032 or manually insert the following:

Command Description

1. mal deg1a act,2,0,0,d,0 Fails DG 11

2. mal deg1c act,2,0,0,d,0 Fails DG 13

3. mal syd1 act,1,1,0,d,0 Loss of offset power

4. mal ppl2a act,0,0,0,d,2 Inadvertent SI, Train A

5. mal ppl2b act,0,0,0,d,2 Inadvertent SI, Train B

6. pmp ccw2 4,0,0,4,d,0 CCW pp 1-2 OC trip

7. loa afw14 act,f,0, 60,d,0 Opens knife switch for AFW pp 1-2

8. loa css8 act,f,0,60,d,0 Opens knife switch for cont. spray pp 1-2

9. loa rhr10 act,f,0,60,d,0 Opens knife switch for RHR pp 1-2

10. loa ccw31 act,f,0,60,d,0 Opens knife switch for CCW pp 1-3

11. loa sis2 act,f,0,60,d,0 Opens knife switch for SI pp 1-2

12. dsc ven14 act,f,0,60,d,0 Opens breaker for CFCU 1-4

13. run 90 freezes simulator after 90 seconds Inform the examiner that the simulator setup is complete.

Go to RUN when the examinee is given the cue sheet.

NRCLJC032.DOC PAGE 7 OF 7 REV.01

- - UNCONTROLLED PROCEDURE - DO NOT USE TO PERFORM WORK or ISSUE FOR USE ***

PACIFIC GAS AND ELECTRIC COMPANY NUMBER EOP ECA-0.3 DIABLO CANYON POWER PLANT REVISION 12 PAGE 27 OF 29 TITLE: Restore 4KV Buses UNIT 1 APPENDIX X CROSSTIE OF VITAL BUS SCOPE Implementation of this Appendix requires approval of the Site Emergency Coordinator. This Appendix should be performed to energize two vital buses from one diesel. If two vital buses are energized, enough ESF equipment should be energized to establish one ESF train.

DISCUSSION As a general guideline, in this situation several options may be available depending on plant conditions and RCS status. If SI is not required, the Shift Foreman may elect to stay in a hot shutdown status awaiting restoration of off-site power or he may decide to cooldown. If SI is required, a minimum of one ECCS flow path must be established. The type of ECCS flowpath desired would depend on RCS conditions.

PRECAUTIONS AND LIMITATIONS

1. Some equipment may have to be operated on a continuous basis and some on an "as needed" basis. Existing conditions will determine which equipment is needed (plant cooldown, SI, etc.).

2. The maximum capacity of the diesel generator should not be exceeded. Appendix Q identifies the diesel generator load limits. Appendix Q, Table 1 provides loads for various vital 4KV and 480 vital equipment. STP M-9M also contains specific loads on all 480 volt vital equipment.

3. The DG that is to be used is assumed to be supplying its own vital bus and running in Auto. The DG should remain in Auto during the performance of this procedure to allow the Isoc feature to maintain proper frequency. If manual is used the operator will need to make frequency adjustments as the DG is loaded.

PROCEDURE

1. Obtain permission from the Site Emergency Coordinator.

2. On the deenergized bus being reenergized, verify ALL the breakers AND DC control power switches are OPEN for the following loads AND ALL the 480V Breakers are open. This prevents automatic loading and overloading the diesel. Continue with steps 3 through 10 while performing this step.

F VITAL BUS G VITAL BUS H VITAL BUS ASW Pp 1 (52-HF-08) ASW Pp 2 (52-HG-06) AFW Pp 2 (52-HH-08)

AFW Pp 3 (52-HF-09) CS Pp 1 (52-HG-07) CS Pp 2 (52-HH-09)

CCP 1 (52-HF-11) RHR Pp 1 (52-HG-08) RHR Pp 2 (52-HH-11)

CCW Pp 1 (52-HF-12) CCW Pp 2 (52-HG-12) CCW Pp 3 (52-HH-12)

SI Pp 1 (52-HF-15) CCP 2 (52-HG-09) SI Pp 2 (52-HH-15)

PDP 3 (52-HG-11)

- - UNCONTROLLED PROCEDURE - DO NOT USE TO PERFORM WORK or ISSUE FOR USE ***

PACIFIC GAS AND ELECTRIC COMPANY NUMBER EOP ECA-0.3 DIABLO CANYON POWER PLANT REVISION 12 PAGE 28 OF 29 TITLE: Restore 4KV Buses UNIT 1 APPENDIX X (Continued)

3. Cutin the D/G DIR PWR, LOSS OF FIELD & BKR OC PROT RLYS C/O SW for the diesel generator selected to supply power to the deenergized bus.

4. Reset SI (if applicable) so that the affected bus will not try to auto load when the bus becomes energized.

5. Cutout the AUTO Transfer FCO's for 4KV buses and 12KV buses.

NOTE: If the D/G associated with the deenergized bus is running but will not load on the bus, it must be shutdown to permit the Auto Transfer Relay to be reset.

6. Depress all AUTO Transfer Reset Push Buttons, verify the BLUE lights go OUT.

7. Verify OPEN all vital bus 4KV auxiliary feeder breakers, 52-HH-13, 52-HG-13 and 52-HF-13.

8. Verify OPEN all vital bus 4KV startup feeder breakers, 52-HH-14, 52-HG-14 and 52-HF-14.

9. Verify OPEN startup feeder breaker 52-HG-15, to the vital buses F, G and H.

10. Verify OPEN the 4KV to 480V bus feeder breaker for the deenergized bus to be reenergized:

DEENERGIZED BUS FEEDER BREAKER F 52-HF-10 G 52-HG-10 H 52-HH-10 CAUTION: o The breaker alignment in step 2 must be completed prior to performing Step 11.

o If the diesel generator appears unstable at any time beyond this point in the procedure, immediately open startup feeder breaker for the operable bus to separate the diesel from the inoperable bus.

NOTE: Although the DG will not be synchronized with other buses during the performance of these steps, most of the breakers to be closed will require the sync key to be on.

11. CLOSE the 4KV startup feeder breaker for the deenergized bus being reenergized.

DEENERGIZED BUS FEEDER BREAKER F 52-HF-14 G 52-HG-14 H 52-HH-14

- - UNCONTROLLED PROCEDURE - DO NOT USE TO PERFORM WORK or ISSUE FOR USE ***

PACIFIC GAS AND ELECTRIC COMPANY NUMBER EOP ECA-0.3 DIABLO CANYON POWER PLANT REVISION 12 PAGE 29 OF 29 TITLE: Restore 4KV Buses UNIT 1 APPENDIX X (Continued)

12. Close the 4KV startup feeder breaker for the bus that will be supplying power to the deenergized bus.

OPERATING D/G FEEDER BREAKER No. 1 52-HH-14 No. 2 52-HG-14 No. 3 52-HF-14 CAUTION: Station an operator at the VB4 to monitor the diesel generator. If an SI should occur, immediately open the 4KV startup feeder breaker for the bus with the operable diesel generator to prevent overloading the operable diesel generator when the SI loads sequence on the bus.

13 CLOSE the 4KV to 480V bus feeder breaker for the reenergized bus:

BUS BREAKER F 52-HF-10 G 52-HG-10 H 52-HH-10 14 Return to procedure and step in effect and IMPLEMENT Appendix Q for equipment starting instructions and diesel generator load limits.

NUCLEAR POWER GENERATION DIABLO CANYON POWER PLANT JOB PERFORMANCE MEASURE Number: NRCLJC-052

Title:

ESTABLISH MAIN FEEDWATER FLOW Examinee:

Evaluator:

Print Signature Date Results: Sat Unsat Total Time: minutes Comments:

References:

EOP FR-H.1, Loss of Secondary Heat Sink, Rev. 19 Alternate Path: Yes No X Time Critical: Yes No X Time Allotment: 30 minutes Critical Steps: 4, 5, 6, 8, 9, 10, 11, 15, 16 Job Designation: RO/SRO Task Number: 059/04S /A4.11 Rating: 3.1/3.3 AUTHOR: JACK BLACKWELL DATE: 01/18/2005 REVIEWED BY: N/A DATE:

TRAINING LEADER APPROVED BY: N/A DATE:

MANAGER - OPERATIONS REV. 01

JPM TITLE: ESTABLISH MAIN FEEDWATER FLOW JPM NUMBER: NRCLJC-052 INSTRUCTOR WORKSHEET Directions: No PLANT controls or equipment are to be operated during the performance of this Job Performance Measure. All actions taken by the examinee should be clearly demonstrated and verbalized to the evaluator. The student will be given the initial conditions, initiating cue, and task standard. The examiner will then ask if any clarifications are needed. The examinee may be given the applicable procedure and step with which to begin.

Required Materials: None Initial Conditions: A plant trip has occurred. The crew has diagnosed a total loss of AFW flow with no immediate prospects for regaining AFW flow. Actions of EOP FR-H.1, up to and including Step 6, have been completed.

Initiating Cue: The Shift Foreman directs you to start the # 1 MFW pump and establish flow to the steam generators, in accordance with EOP FR-H.1, Step 7.

Task Standard: The #1 MFW pump is started from the Control Room and main feedwater flow is established to at least one steam generator in accordance with EOP FR-H.1.

NRCLJC052.DOC PAGE 2 OF 9 REV. 01

JPM TITLE: ESTABLISH MAIN FEEDWATER FLOW JPM Number: NRCLJC-052 INSTRUCTOR WORKSHEET Start Time:

Step Expected Operator Actions

1. Obtains the correct procedure. 1.1 References EOP FR-H.1.

Step was: Sat: ______ Unsat _______*

2. Check at least one 2.1 Reads CAUTION prior to step.

condensate/booster pump set running in recirc. 2.2 Observes that at least one condensate/booster pump set is already running.

Step was: Sat: ______ Unsat _______*

3. Check main feedwater isolation 3.1 Observes that all feedwater isolation valves - OPEN. valves are CLOSED.

Step was: Sat: ______ Unsat _______*

- 4. Reduce RCS pressure to less than 4.1 Observes that letdown is NOT in 1915 psig. service.

4.2 Positions any PZR PORVs control switch to OPEN. **

4.3 Lowers PZR pressure to less than 1915 psig. **

4.4 Returns selected PZR PORVs control switch to AUTO.

Cue: Another operator will maintain RCS pressure less than 1865 psig.

Note: An instructor should open a PZR PORV as necessary to ensure that PZR pressure remains less than 1865 psig. Be careful NOT to reset P-11, since this could lead to an SI actuation.

Step was: Sat: ______ Unsat _______*

Denotes an entry required on the JPM cover sheet.

- Denotes Critical Step and Sub Steps.

NRCLJC052.DOC PAGE 3 OF 9 REV. 01

JPM TITLE: ESTABLISH MAIN FEEDWATER FLOW JPM Number: NRCLJC-052 INSTRUCTOR WORKSHEET Step Expected Operator Actions

- 5. Block the Low PZR Pressure SI. 5.1 Observes PK08-06, PZR S.I.

PERMISSIVE P-11, ON.

5.2 Positions the PZR SI RESET/BLOCK TRAIN A and TRAIN B switches to BLOCK. **

5.3 Verifies that PZR Low Pressure SI has blocked by observing PK08-16, PZR S.I. BLOCKED - ON.

Step was: Sat: ______ Unsat _______*

- 6. Block the Low Steam Line 6.1 Positions the LO STM LINE PRESS Pressure Pressure SI SI RESET/BLOCK TRAIN A and TRAIN B switches to BLOCK. **

6.2 Verifies that Low Steamline Presure SI has blocked by observing PK08-17, LO STM LINE PRESSURE S.I. BLOCKED - ON.

Step was: Sat: ______ Unsat _______*

7. Maintain RCS pressure 1500 - **********************************

1865 psig. Cue: Another operator will maintain RCS pressure between 1500 &

1865 psig.

- 8. Reset SI. 8.1 Depresses the SAFETY INJECTION RESET TRAIN A and TRAIN B pushbuttons. **

8.2 Verifies that SI is reset by observing PK08-22 ON and/or SI Monitor Box red status light OFF.

Step was: Sat: ______ Unsat _______*

- 9. Cycle Reactor Trip Breakers. 9.1 Momentarily positions the REACTOR TRIP RESET/TRIP switch to RESET (CC1). **

Step was: Sat: ______ Unsat _______*

Denotes an entry required on the JPM cover sheet.

- Denotes Critical Step and Sub Steps.

NRCLJC052.DOC PAGE 4 OF 9 REV. 01

JPM TITLE: ESTABLISH MAIN FEEDWATER FLOW JPM Number: NRCLJC-052 INSTRUCTOR WORKSHEET Step Expected Operator Actions

- 10. Reset Feedwater Isolation. 10.1 Resets feedwater isolation by depressing the FDWTR ISOL RESET pushbuttons.**

10.2 Verifies feedwater isolation has reset by observing F.W. ISOL red light OFF and/or PK09-11 OFF.

Step was: Sat: ______ Unsat _______*

- 11. Open Main Feedwater Isolation 11.1 Opens MFW isolation valves Valves. FCV-438, 439, 440, and 441.**

Note: Opening only one isolation valve satisfies critical task.

11.2 Verifies Main Feedwater isolation valves have opened.

Step was: Sat: ______ Unsat _______*

12. Verify condenser - AVAILABLE. 12.1 Observes PK08-14, CONDENSER AVAILABLE C ON.

or Observes adequate condenser vacuum on PI-44 and one circulating water pump running.

Step was: Sat: ______ Unsat _______*

13. Verifies MSIVs - OPEN. 13.1 Observes that all MSIVs are open.

Step was: Sat: ______ Unsat _______*

14. Verify manual isolation for HP 14.1 Requests that another operator verify Steam to MFW Pumps - OPEN. MS-1-95 and MS-1-92 OPEN.

MS-1-95 (MFW Pp1-1)

MS-1-92 (MFW Pp1-2) **********************************

Cue: MS-1-95 and MS-1-92 are OPEN.

Step was: Sat: ______ Unsat _______*

Denotes an entry required on the JPM cover sheet.

- Denotes Critical Step and Sub Steps.

NRCLJC052.DOC PAGE 5 OF 9 REV. 01

JPM TITLE: ESTABLISH MAIN FEEDWATER FLOW JPM Number: NRCLJC-052 INSTRUCTOR WORKSHEET Step Expected Operator Actions

- 15. Restart MFW pumps. Note: Starting MFP 1-2 also satisfies the critical task.

15.1 Verifies MFWP latched.

15.2 Verifies FCV-53 and FCV-54 switches in RECIRC.

15.3 Presses ALARM/TRIP RESET on the MFWP 1-1 S/U STATION. **

15.4 Latches the MFW pump turbine by holding the PUMP 1-1 TRIP/LATCH SELECT switch in RESET until the LATCHED light is ON. **

Note: Latch time is 2 minute.

Operator may elect to have pump latched locally. If so, latch the pump from the Sim booth.

15.5 Presses RAMP UP TO IDLE. **

15.6 Observes speed rising to IDLE setpoint verifies to ~ 600 RPM.

15.7 Presses IDLE TO STBY. **

15.8 Observes speed rising to STBY setpoint.

15.9 When speed reaches 3000, raises MFP speed until discharge pressure is approximately 100 psig greater than S/G pressure (PI-509A or PI-509). **

Note: Operator may raise speed by pressing the RAISE pushbutton at the S/U station, or by selecting DFW CONTROL at the S/U station and then raising the output of the CC3 controller.

Step was: Sat: ______ Unsat _______*

Denotes an entry required on the JPM cover sheet.

- Denotes Critical Step and Sub Steps.

NRCLJC052.DOC PAGE 6 OF 9 REV. 01

JPM TITLE: ESTABLISH MAIN FEEDWATER FLOW JPM Number: NRCLJC-052 INSTRUCTOR WORKSHEET Step Expected Operator Actions

- 16. Throttle open MFW control 16.1 Throttles open at least one MFW bypass valves. bypass valve and establishes flow.**

Note: Opening a MFW control valve satisfies the critical task.

16.2 Verifies feedwater flow to at least one S/G.

Step was: Sat: ______ Unsat _______*

Stop Time:

Total Time: (Enter total time on the cover page)

Denotes an entry required on the JPM cover sheet.

- Denotes Critical Step and Sub Steps.

NRCLJC052.DOC PAGE 7 OF 9 REV. 01

JPM NUMBER: NRCLJC-052 EXAMINEE CUE SHEET Initial Conditions: A plant trip has occurred. The crew has diagnosed a total loss of AFW flow with no immediate prospects for regaining AFW flow. Actions of EOP FR-H.1, up to and including Step 6, have been completed.

Initiating Cue: The Shift Foreman directs you to start the # 1 MFW pump and establish flow to the steam generators, in accordance with EOP FR-H.1, Step 7.

Task Standard: The #1 MFW pump is started from the Control Room and main feedwater flow is established to at least one steam generator in accordance with EOP FR-H.1.

NRCLJC052.DOC PAGE 8 OF 9 REV. 01

JPM TITLE: ESTABLISH MAIN FEEDWATER FLOW JPM NUMBER: NRCLJC-052 ATTACHMENT 1, SIMULATOR SETUP Initialize the simulator to the RELAP INIT 510 (100%, MOL).

If possible, a second instructor should be available during this JPM to control PZR pressure when required.

Enter drill file 1052 or manually insert the following:

Command Description

1. mal afw1 act,0,0,d,0 Trips AFW pp 1-1

2. pmp afw1 4,0,0,0,d,0 Trips AFW pp 1-2 from starting

3. pmp afw2 4,0,0,0,d,0 Trips AFW pp 1-3 from starting

4. ovr xrei022h act,1,0,0,c,fnispr.1t.10,5 Reset MSRS

5. delm bsgnwrr1 Removes bsgnwrr1 from monitor

6. monv bsgnwrr1 Monitors steam generator wide range level

7. run 120

8. mal pp12a act,0,0,0,d,2 Inadvertent SI, Train A

9. mal pp12b act,0,0,0,d,2 Inadvertent SI, Train B

10. ovr xv2i260o act,1,0,0,c,fnispr.1t.10,0 Trips RCP 11

11. ovr xv2i261o act,1,0,0,c,fnispr.1t.10,0 Trips RCP 12

12. ovr xv2i262o act,1,0,0,c,fnispr.1t.10,0 Trips RCP 13

13. ovr xv2i263o act,1,0,0,c,fnispr.1t.10,0 Trips RCP 14 Perform the following:

1. Place FCV-53/54 in RECIRC.

2. Place Steam Dump Control in Steam Pressure Mode.

3. Place LCV-12 in CONT ONLY.

4. Stop all but one Condensate/Booster Pump set.

Inform the examiner that the simulator setup is complete.

Go to RUN when the examinee is given the cue sheet.

NRCLJC052.DOC PAGE 9 OF 9 REV. 01

PACIFIC GAS AND ELECTRIC COMPANY NUMBER EOP FR-H.1 DIABLO CANYON POWER PLANT REVISION 19 PAGE 5 OF 28 TITLE: Response to Loss of Secondary Heat Sink UNIT 1 ACTION/EXPECTED RESPONSE RESPONSE NOT OBTAINED

6. CHECK If Condenser Steam Dump IF Condenser Steam Dump is Should Be In Pressure Control Mode: NOT available, THEN Adjust 10% Steam Dump controllers as needed to

a. Check MSIVs - OPEN maintain S/G pressure LESS THAN OR EQUAL TO

b. Check Condenser - AVAILABLE 1005 PSIG (8.38 turns)

c. Increase setting on HC-507 to achieve 0% demand OR Transfer HC-507 to manual and decrease demand to 0%. -----------------------------

d. Place Steam Dump in Steam Pressure Mode

e. Adjust Steam Dump controller as needed to maintain S/G pressure LESS THAN OR EQUAL TO 1005 PSIG (8.38 turns)

7. TRY To Establish Mn Fdwtr Flow To At Least One S/G:

CAUTION: Hotwell level should be monitored when supplying S/Gs with Condensate/Booster Pps and Mn Fdwtr Pps.

a. Check Condensate System - IN a. Try to place condensate system in SERVICE service. REFER TO OP C-7A series.

1) At least one Condensate/Booster Pp Set IF NOT, running in recirc THEN GO TO Step 11 (Page 12).

THIS STEP CONTINUED ON NEXT PAGE

PACIFIC GAS AND ELECTRIC COMPANY NUMBER EOP FR-H.1 DIABLO CANYON POWER PLANT REVISION 19 PAGE 6 OF 28 TITLE: Response to Loss of Secondary Heat Sink UNIT 1 ACTION/EXPECTED RESPONSE RESPONSE NOT OBTAINED

7. TRY To Establish Mn Fdwtr Flow To At Least One S/G:

(Continued)

b. Check Mn Fdwtr Isol Vlvs - OPEN b. Open Mn Fdwtr Isol Vlvs as follows:

1) Reduce RCS Pressure to LESS THAN 1915 PSIG as follows:

o Use PZR PORV OR o If Letdown is in service, use Aux Spray

2) Block the Low PZR Pressure SI

3) Block the Low Stmline Pressure SI

4) Maintain RCS pressure 1500 -

1865 PSIG.

5) Reset SI

6) Cycle Reactor Trip Bkrs (CC1)

7) Reset Fdwtr Isolation

8) Open Mn Fdwtr Isol Vlvs OR Locally Open Mn Fdwtr Isol Vlvs IF NO Mn Fdwtr path can be opened, THEN GO TO Step 11 (Page 12)

THIS STEP CONTINUED ON NEXT PAGE

PACIFIC GAS AND ELECTRIC COMPANY NUMBER EOP FR-H.1 DIABLO CANYON POWER PLANT REVISION 19 PAGE 7 OF 28 TITLE: Response to Loss of Secondary Heat Sink UNIT 1 ACTION/EXPECTED RESPONSE RESPONSE NOT OBTAINED

7. TRY To Establish Mn Fdwtr Flow To At Least One S/G:

(Continued)

c. Establish Mn Fdwtr flow capability: c. GO TO Step 9 (Page 9).

1) Verify Condenser - AVAILABLE

2) Verify MSIVs - OPEN

3) Verify manual isolation for HP steam to MFW Pumps - OPEN o MS-1-95 (MFW Pp 1-1) o MS-1-92 (MFW Pp 1-2)

4) Check ANY MFW Pp - LATCHED 4) Restart a MFW Pp as follows:

(a) Verify FCV-53 AND FCV-54 switches in RECIRC.

(b) Press ALARM/TRIP RESET on MFW Pp S/U station (VB3).

(c) Take Trip/Latch switch to RESET to latch the MFW Pp Turbine (Hold until latched, ~ 2 min).

(d) Press RAMP UP TO IDLE, verify ramp to ~ 600 RPM.

(e) Press IDLE TO STANDBY, verify ramp to ~ 3000 RPM.

IF MFW Pps will not start, THEN REFER TO APPENDIX K to restart locally.

5) Increase MFW Pp speed until discharge pressure is 100 PSIG GREATER THAN S/G Pressure THIS STEP CONTINUED ON NEXT PAGE

PACIFIC GAS AND ELECTRIC COMPANY NUMBER EOP FR-H.1 DIABLO CANYON POWER PLANT REVISION 19 PAGE 8 OF 28 TITLE: Response to Loss of Secondary Heat Sink UNIT 1 ACTION/EXPECTED RESPONSE RESPONSE NOT OBTAINED

7. TRY To Establish Mn Fdwtr Flow To At Least One S/G:

(Continued)

6) Check PK09-11, FEEDWATER 6) Reset Fdwtr Isolation.

ISOLATION - OFF

7) Throttle open: 7) Locally Throttle open Mn Fdwtr Cont Vlvs o Mn Fdwtr Cont Bypass Vlvs IF NO Mn Fdwtr path can be OR opened o Mn Fdwtr Cont Vlvs THEN GO TO Step 11 (Page 12)

d. IF This Step was implemented from Step 22, THEN GO TO Step 23 (Page 18)

8. CHECK S/G NR Levels:

a. S/G NR Level in at least one S/G - a. IF Feedflow to at least one GREATER THAN 6% [16%] S/G - verified, THEN Maintain flow to restore S/G NR Level to GREATER THAN 6% [16%].

IF Feedflow NOT verified, THEN GO TO Step 9 (Next Page).

b. RETURN TO procedure and step in effect

NUCLEAR POWER GENERATION DIABLO CANYON POWER PLANT JOB PERFORMANCE MEASURE Number: NRCLJC-081

Title:

SECURE CONTAINMENT SPRAY Examinee:

Evaluator:

Print Signature Date Results: Sat Unsat Total Time: minutes Comments:

References:

EOP E-1, Loss of Reactor or Secondary Coolant, Rev. 19 Alternate Path: Yes No X Time Critical: Yes No X Time Allotment: 10 minutes Critical Steps: 3, 4, 5, 6, 7, 9 Job Designation: RO/SRO Task Number: 026/05/A4.01 Rating: 4.5/4.3 AUTHOR: JACK BLACKWELL DATE: 01/18/2005 REVIEWED BY: N/A DATE:

TRAINING LEADER APPROVED BY: N/A DATE:

LINE MANAGER REV. 01

JPM TITLE: SECURE CONTAINMENT SPRAY JPM NUMBER: NRCLJC-081 INSTRUCTOR WORKSHEET Directions: No PLANT controls or equipment are to be operated during the performance of this Job Performance Measure. All actions taken by the examinee should be clearly demonstrated and verbalized to the evaluator. The student will be given the initial conditions, initiating cue, and task standard. The examiner will then ask if any clarifications are needed. The examinee may be given the applicable procedure and step with which to begin.

Required Materials: None Initial Conditions: Unit 1 has experienced a steam line break inside Containment. The faulted steam generator was isolated and the crew has transitioned to EOP E-1.

Initiating Cue: The Shift Foreman directs you to evaluate and secure, as appropriate, Containment Spray per EOP E-1, Step 5.

Task Standard: The criterion for stopping Containment Spray has been evaluated, and the system is aligned as required by EOP E-1.

NRCLJC081.DOC PAGE 2 OF 7 REV. 01

JPM TITLE: SECURE CONTAINMENT SPRAY JPM NUMBER: NRCLJC-081 INSTRUCTOR WORKSHEET Start Time:

Step Expected Operator Actions

1. Obtain the correct procedure. 1.1 References EOP E-1.

Step was: Sat: ______ Unsat _______*

2. Check PK01-18, CONTMT 2.1 Observes that PK01-18 is ON.

SPRAY ACTUATION, ON.

Step was: Sat: ______ Unsat _______*

- 3. Check containment radiation 3.1 Observes PK11-21, HIGH levels. RADIATION is OFF. **

3.2 Observes normal indication on RE-2/RE-7. (SPDS, PPC, RNRM-A) **

3.3 Observes PK11-19 CONTMT RADIATION is OFF. **

3.4 Observes normal indication on R-30/R-31. (PAM2) **

Step was: Sat: ______ Unsat _______*

- 4. Check containment pressure less 4.1 Observes containment pressure than 20 psig. (PI-934-937) less than 20 psig. **

Step was: Sat: ______ Unsat _______*

Denotes an entry required on the JPM cover sheet.

- Denotes Critical Step and Sub Steps.

NRCLJC081.DOC PAGE 3 OF 7 REV. 01

JPM TITLE: SECURE CONTAINMENT SPRAY JPM NUMBER: NRCLJC-081 INSTRUCTOR WORKSHEET Step Expected Operator Actions

- 5. Reset Containment Spray Trains A 5.1 Depresses both Containment Spray and B. reset push buttons. **

5.2 Checks that PK01-18, CONTMT SPRAY ACTUATION is OFF.

Step was: Sat: ______ Unsat _______*

- 6. Stop containment spray pumps. 6.1 Stops both Containment Spray pumps. **

6.2 Verifies both Containment Spray pumps have stopped.

Step was: Sat: ______ Unsat _______*

- 7. Close 9001A and B. 7.1 Closes 9001A and B. **

7.2 Verifies 9001A and B have closed.

Step was: Sat: ______ Unsat _______*

8. Verify 9003A and B closed. 8.1 Observes that 9003A and B are closed.

Step was: Sat: ______ Unsat _______*

- 9. Close 8994A and B. 9.1 Closes 8994A and B. **

9.2 Verifies that valves have closed.

Step was: Sat: ______ Unsat _______*

Stop Time:

Total Time: (Enter total time on the cover page)

Denotes an entry required on the JPM cover sheet.

- Denotes Critical Step and Sub Steps.

NRCLJC081.DOC PAGE 4 OF 7 REV. 01

JPM NUMBER: NRCLJC-081 EXAMINEE CUE SHEET Initial Conditions: Unit 1 has experienced a steam line break inside Containment. The faulted steam generator was isolated and the crew has transitioned to EOP E-1.

Initiating Cue: The Shift Foreman directs you to evaluate and secure, as appropriate, Containment Spray per EOP E-1, Step 5.

Task Standard: The criterion for stopping Containment Spray has been evaluated, and the system is aligned as required by EOP E-1.

NRCLJC081.DOC PAGE 5 OF 7 REV. 01

JPM TITLE: SECURE CONTAINMENT SPRAY JPM NUMBER: NRCLJC-081 ATTACHMENT 1, SIMULATOR SETUP Initialize the simulator to IC-510 (100%, MOL).

Enter drill file 1081 or manually insert the following:

Command Description mal mssla act,4e+07,0,0,d,0 Break SG 1 INSIDE CNMT vlv afw3 2,0,0,0,D,0 #rafl106 Isolate afw to s/g 11 cnv afw1 2,0,0,0,D,0 #rafl110 ovr xv3i149m act,1,0,0,d,5 #vb3024I lcv-110 cntlr to man ovr XV1I113C act,1,0,0,d,5 #vb1106a ACTUATE PHASE B (CNMT ovr XV1I114C act,1,0,0,d,5 #vb1107a SPRAY) ovr XV2I2600 act,1,0,0,d,5 #vb2163e Selects stop for each rcp ovr XV2I2610 act,1,0,0,d,5 #vb2164e ovr XV2I2620 act,1,0,0,d,5 #vb2165e ovr XV2I2630 act,1,0,0,d,5 #vb2166e cnh mss2 1,0,0,0,d,0 #xcnh516e Stop 10% dumps from opening.

cnh mss3 1,0,0,0,d,0 #xcnh526e cnh mss4 1,0,0,0,d,0 #xcnh536c cnh mss5 1,0,0,0,d,0 #xcnh546c Ovr xreo006h act,1,0,0,c,fnispr (1) .lt.10,5 RESET MSRS

vb3164r ovr xc3I136M act,1,0,0,c,fnispr (1) .lt.10,5 TAKE FWRVS TO MANUAL

cc3050c AND CLOSE.

ovr xc3I136L act,1,0,0,c,fnispr (1) .lt.10,60

cc3050f ovr xc3I137M act,1,0,0,c,fnispr (1) .lt.10,5

cc3051c ovr xc3I137L act,1,0,0,c,fnispr (1) .lt.10,60

cc3051f ovr xc3I138M act,1,0,0,c,fnispr (1) .lt.10,5

cc3052c ovr xc3I138L act,1,0,0,c,fnispr (1) .lt.10,60

cc3052f ovr xc3I139M act,1,0,0,c,fnispr (1) .lt.10,5

cc3053c ovr xc3I139L act,1,0,0,c,fnispr (1) .lt.10,60

cc3053f NRCLJC081.DOC PAGE 6 OF 7 REV. 01

JPM TITLE: SECURE CONTAINMENT SPRAY JPM NUMBER: NRCLJC-081 ATTACHMENT 1, SIMULATOR SETUP Command Description ovr xv3i224o act,1,0,0,c,fnispr.lt.10,5 STOP CND/BSTR PP 1-2 & 1-3

vb3062e ovr xv3i180c act,0,0,0,d,0 #vb3060b ovr xv3i194c act,1,0,0,c,fnispr.lt.10,0 Recirc on fw pp recirc valves.

vb3131b ovr xv3i197c act,1,0,0,c,fnispr.lt.10,0

vb3132b Ovr xv4i388o act,0,0,0,d,0 #vb4303a Turn on charcoal filter preheater.

ovr xv4i388c act,1,0,1,d,0 #vb4303b Run 60 Runs 60 seconds When simulator freezes, place:

FCV-53 & 54 in RECIRC

Cnd/Bstr set 13 to MAN

Char Fltr Prehtr to ON Inform the examiner that the simulator setup is complete.

Go to RUN when the examinee is given the cue sheet.

NRCLJC081.DOC PAGE 7 OF 7 REV. 01

- - UNCONTROLLED PROCEDURE-DO NOT USETO PERFORM WORK orISSUEFOR USE ***

PACIFIC GAS AND ELECTRIC COMPANY NUMBER EOP E-1 DIABLO CANYON POWER PLANT REVISION 19 PAGE 7 OF 30 TITLE: Loss of Reactor or Secondary Coolant UNIT 1 ACTION/EXPECTED RESPONSE RESPONSE NOT OBTAINED

5. CHECK If Containment Spray Should Be Stopped:

a. Check PK01-18 CONTAINMENT a. GO TO Step 6 (Next Page).

SPRAY ACTUATION - ON

b. Check Containment Radiation b. IF Containment Radiation Levels Levels are above normal, THEN Verify spray system is still o PK11-21, HIGH RADIATION - in operation OFF AND o RE-2/RE-7-NORMAL GO TO Step 6 (Next Page).

o PK11-19, CONTMT RADIATION - OFF -----------------------------

o R-30/R-31-NORMAL (PAM 2)

c. Check Containment Pressure - c. Perform the following:

LESS THAN 20 PSIG

1) Verify Containment Spray system is still in operation.

2) WHEN Containment Pressure is LESS THAN 20 PSIG, THEN Perform Steps 5d through 5.h

d. Reset Containment Spray Trains A and B

e. Stop Containment Spray Pps

f. Close 9001A & B

g. Verify 9003A & B - Closed

h. Close 8994A & B

NUCLEAR POWER GENERATION DIABLO CANYON POWER PLANT JOB PERFORMANCE MEASURE Number: NRCLJC-093

Title:

RESPOND TO LOSS OF RHR INVENTORY IN MODE 5 Examinee:

Evaluator:

Print Signature Date Results: Sat Unsat Total Time: minutes Comments:

References:

OP AP SD-2, Loss of RCS Inventory, Rev. 15 Alternate Path: Yes X No Time Critical: Yes No X Time Allotment: 15 minutes Critical Step(s): 4, 5 Job Designation: RO/SRO Task Number: 009/03/EA1.04 Rating: 3.7/3.5 AUTHOR: JACK BLACKWELL DATE: 01/18/2005 REVIEWED BY: N/A DATE:

TRAINING LEADER APPROVED BY: N/A DATE:

LINE MANAGER REV. 01

JPM TITLE: RESPOND TO LOSS OF RHR INVENTORY IN MODE 5 JPM NUMBER:NRCLJC-093 INSTRUCTOR WORKSHEET Directions: No PLANT controls or equipment are to be operated during the performance of this Job Performance Measure. All actions taken by the examinee should be clearly demonstrated and verbalized to the evaluator. The student will be given the initial conditions, initiating cue, and task standard. The examiner will then ask if any clarifications are needed. The examinee may be given the procedure and told the step with which to begin.

Required Materials: None Initial Conditions: Unit 1 is in MODE 5, eight days after a plant shutdown for refueling.

RCS level has been at 109. No work is currently in progress on the RCS.

Initiating Cue: Reactor Vessel level has just started decreasing, as noted on wide range and narrow range RVRLIS on the PPC. The Shift Foreman has directed you to respond to the loss of inventory in accordance with OP AP SD-2, Task Standard: Required actions have been taken to stabilize reactor vessel level in accordance with OP AP SD-2.

NRCLJC093.DOC PAGE 2 OF 7 REV. 01

JPM TITLE: RESPOND TO LOSS OF RHR INVENTORY IN MODE 5 JPM NUMBER: NRCLJC-093 INSTRUCTOR WORKSHEET Start Time:

Step Expected Operator Actions

1. Obtain the correct procedure. 1.1 Refers to OP AP SD-2.

1.2 Reads CAUTIONs prior to Step 1.

Step was: Sat: ______ Unsat _______*

2. Check RVRLIS level <108 feet or 2.1 Checks any or all of the following inventory loss is rapid. RVRLIS indications:

o WR RVRLIS (PPC Pt. U2012).

o NR RVRLIS (PPC Pt. U2014).

o RVRLIS Ultrasonic (PPC Pt.

L0470A).

o Standpipe level Cue: If Containment contacted, report level at approx. 107.5.

2.2 Determines RVRLIS level <108 feet or Decreasing Rapidly.

Step was: Sat: ______ Unsat _______*

3. Check if RHR pumps should be 3.1 Observes that RHR pump 1-1 is stopped. running 3.2 Observes RVRLIS level greater than 1073.

3.3 Observes RHR flow at 2000 gpm.

3.4 Reduces flow to around 1550 -

1675 gpm using HCV-637 or 638.

3.4 Checks RHR pump not cavitating by observing RHR flow and amps.

3.5 Observes that RCS level is STILL decreasing and continues with step 2 RNO.

Step was: Sat: ______ Unsat _______*

Denotes an entry required on the JPM cover sheet.

- Denotes Critical Step and Sub Steps.

NRCLJC093.DOC PAGE 3 OF 7 REV. 01

JPM TITLE: RESPOND TO LOSS OF RHR INVENTORY IN MODE 5 JPM NUMBER: NRCLJC-093 INSTRUCTOR WORKSHEET Step Expected Operator Actions

- 4. Isolate letdown and increase RCS ** 4.1 Closes HCV-133 or PCV-135.

makeup.

4.2 Verifies that letdown is isolated.

Note: A 500 gpm RHR leak is used to simulate a letdown leak. The malfunction will NOT clear when letdown is isolated.

- 4.3 Increases RCS makeup by using any of the following methods:

o Open 8805A or 8805B o Open 8980 o Open FCV-128 and HCV-142 o Increases charging o Start an SI pump 4.4 Checks all known drain paths closed.

Cue: No known drain paths exist.

4.5 Observes that RCS level is still decreasing.

4.6 Sounds Containment Evacuation.

4.7 Verifies personnel clear of SG manways.

Cue: Manways and immediate area are clear.

Step was: Sat: ______ Unsat _______*

Denotes an entry required on the JPM cover sheet.

- Denotes Critical Step and Sub Steps.

NRCLJC093.DOC PAGE 4 OF 7 REV. 01

JPM NUMBER: NRCLJC-093 EXAMINEE CUE SHEET Step Expected Operator Actions

- 5. Check RHR system intact. 5.1 Checks PRT level normal 5.2 Verifies RHR pump room sump alarm, PK02-16 ON.

- 5.3 Stops RHR pump.

- 5.4 Closes:

o 8701 or 8702 o HCV-133, Letdown to CVCS 5.5 Checks RCPs secured.

Step was: Sat: ______ Unsat _______*

6. Depressurize RCS to Atmospheric 6.1 Verifies PORV block valves open.

pressure.

6.2 Verifies PORV open.

Step was: Sat: ______ Unsat _______*

7. Restore RCS inventory. 7.1 Add makeup as needed by either:

o Increase charging o Open 8805 A or B o Open 8980 o Open 8741 o Any ECCS pump/path 7.2 Verify RCS level stable or increasing.

Step was: Sat: ______ Unsat _______*

Stop Time:

Total Time: (Enter total time on the cover page)

NRCLJC093.DOC PAGE 5 OF 7 REV. 01

JPM NUMBER: NRCLJC-093 EXAMINEE CUE SHEET Initial Conditions: Unit 1 is in MODE 5, eight days after a plant shutdown for refueling.

RCS level has been at 109. No work is currently in progress on the RCS.

Initiating Cue: Reactor Vessel level has just started decreasing, as noted on wide range and narrow range RVRLIS on the PPC. The Shift Foreman has directed you to respond to the loss of inventory in accordance with OP AP SD-2, Task Standard: Required actions have been taken to stabilize reactor vessel level in accordance with OP AP SD-2.

NRCLJC093.DOC PAGE 6 OF 7 REV. 01

JPM TITLE: RESPOND TO LOSS OF RHR INVENTORY IN MODE 5 JPM NUMBER: NRCLJC-093 ATTACHMENT 1, SIMULATOR SETUP Initialize the simulator to IC_704 (109, one RHR pump operating).

Set Group Display RVRLIS on a PPC and QP RVRLIS on another PPC screen. Set BIG to U2014 on the PPC screen by the crash cart.

Put RHR lamicoids on 8726A and B and 8734A and B (red OPEN valve lamicoids).

OR Perform the following:

Initialize the simulator to IC_537 Enter drill file 6501 (a modified 1093 from LJC-093)

OR manually enter the following:

Command Description delm bsiscore monitors RVRLIS level monv bsiscore monitors RVRLIS level mal rhr2 RHR System Break, clears when act,500,120,0,d,wldsldhx.lt.0.5 letdown isolated run sim in RUN Allow the simulation to run until RCS level is at 107.9 then go to freeze.

Inform the instructor the simulation is ready.

NRCLJC093.DOC PAGE 7 OF 7 REV. 01

- - UNCONTROLLED PROCEDURE - DO NOT USE TO PERFORM WORK or ISSUE FOR USE ***

1&2 D IABLO C ANYON P OWER P LANT OP AP SD-1 ABNORMAL OPERATING PROCEDURE REV. 14 UNITS PAGE 1 OF 22 Loss of AC Power 08/31/04 EFFECTIVE DATE PROCEDURE CLASSIFICATION: QUALITY RELATED

1. SCOPE 1.1 This procedure is used in Mode 5 or 6 when it is apparent that the electrical buses needed for effective Decay Heat Removal are not energized and measures more complex than closing or reclosing the bus feeder breakers are required.

1.2 This procedure may be entered from Shutdown Emergency Procedure OP AP SD-0, or directly when the loss of AC power is recognized.

1.3 This procedure provides guidance for regaining AC power to the Vital 4kV buses and the Nonvital 4kV buses.

1.3.1 The possible Recovery power sources are:

a. Steps 2 and 3 - Affected Unit's Aux Power system

b. Steps 5 and 6 - Affected Unit's SU Power system

c. Steps 8 and 9 - Other Unit's SU Power system

d. Steps 11 thru 16 - Other Unit's Aux Power system

e. Appendix X - Cross-tie of Vital buses using an operating Diesel Generator

f. Appendix N - Energizing Non-Vital buses using an operating Diesel Generator 1.3.2 The selection of the best Recovery power source will depend on factors that are impossible to predict during an outage, therefore it is not necessary to select the Recovery power source in the order given in this procedure. The shift foreman may choose not to use a Recovery power source, the procedure reader may then assume that power source is not available and follow the instructions in the Response not Obtained column to get to the implementation instructions for the Recovery power source of choice.

2. SYMPTOMS 2.1 Loss of AC power to any required electrical bus.

02 01069314.DOC 0831.0934 Page 1 of 23

- - UNCONTROLLED PROCEDURE - DO NOT USE TO PERFORM WORK or ISSUE FOR USE ***

Loss of AC Power U1&2 OP AP SD-1 REV. 14 PAGE 2 OF 22 ACTION / EXPECTED RESPONSE RESPONSE NOT OBTAINED CAUTION: RHR pumps and the SFP cooling pump must be manually STARTED following power restoration to the bus.

NOTE: If there are three or less empty fuel assembly locations in the core, then the SRO should consider placing the hanging fuel assembly in the upender and lowering the upender.

1. CHECK Fuel Handling Equipment:

a. Manipulator Crane - NO FUEL a. IF The assembly is NOT a new fuel ASSEMBLY LATCHED assembly, THEN Manually crank the Bridge until over the core AND lower the assembly until approximately one foot above the lower core support plate. Turn power on the bridge - OFF CAUTION: DO NOT lower the fuel assembly without power to the load cell.

b. Spent Fuel Pool Bridge Crane - NO b. Open the power supply breaker to the FUEL ASSEMBLY SUSPENDED Bridge Crane.

2. CHECK Status of 500kV System: GO TO step 5.

VERIFY PCB 532(542) or 632(642) -

CLOSED

Aux Transformer 1-2 (2-2) Power Available White Lights - ON (VB5)

3. IMPLEMENT OP J-2:V To Backfeed From GO TO step 5.

500kV System

4. RETURN To Procedure And Step In Effect

5. CHECK Status Of Own Unit's Startup GO TO step 8.

Power:

OCB-212 - CLOSED

S/U Transformer 1-1 (2-1) Power Available White Status Light - ON (VB5) 02 01069314.DOC 0831.0934 Page 2 of 23

- - UNCONTROLLED PROCEDURE - DO NOT USE TO PERFORM WORK or ISSUE FOR USE ***

Loss of AC Power U1&2 OP AP SD-1 REV. 14 PAGE 3 OF 22 ACTION / EXPECTED RESPONSE RESPONSE NOT OBTAINED

6. Make S/U Transformer 1-2 (2-2) Available GO TO step 8.

AND Energize Desired 4kV Buses:

a. Place AUTO TRANSFER TO START-UP CUTOUT Switch in the CUTOUT position for ALL 4kV AND 12kV buses (affected Unit only)

b. Depress the AUTO BUS TRANSFER Reset Pushbuttons on ALL 4kV and 12kV buses (affected Unit only)

Blue lights - OFF

c. Implement OP J-2:II to make S/U Transformer 1-2 (2-2) available and energize desired 4kV buses

7. RETURN To Procedure And Step In Effect

8. CHECK Status Of Other Unit's S/U Power:

GO TO step 11.

OCB-212 - CLOSED

S/U Transformer 2-1 (1-1) Power Available White Status Light - ON (OTHER UNIT'S VB5) 02 01069314.DOC 0831.0934 Page 3 of 23

- - UNCONTROLLED PROCEDURE - DO NOT USE TO PERFORM WORK or ISSUE FOR USE ***

Loss of AC Power U1&2 OP AP SD-1 REV. 14 PAGE 4 OF 22 ACTION / EXPECTED RESPONSE RESPONSE NOT OBTAINED

9. ENERGIZE 4kV Buses From Other Unit's S/U Power:

a. Place AUTO TRANSFER TO START-UP CUTOUT Switch in the CUTOUT position for ALL 4kV AND 12kV buses (affected Unit only)

b. Depress the AUTO BUS TRANSFER Reset Pushbuttons on ALL 4kV and 12kV buses (affected Unit only)

Blue lights - OFF

c. Verify 52-VU-12 (52-VU-24) - OPEN

d. Open S/U Feeder Bkrs on the affected Unit for ALL 4kV AND 12kV buses:

UNIT 1 UNIT 2 52-HF-14 (52-HF-14) 52-HG-14 (52-HG-14) 52-HH-14 (52-HH-14) 52-HD-14 (52-HD-05) 52-HE-03 (52-HE-13) 52-VD-04 (52-VD-06) 52-VE-06 (52-VE-04)

e. CLOSE 52-VU-11

f. VERIFY 52-VU-14 (52-VU-23) -

CLOSED

g. VERIFY 52-VU CLOSED THIS STEP CONTINUED ON NEXT PAGE 02 01069314.DOC 0831.0934 Page 4 of 23

- - UNCONTROLLED PROCEDURE - DO NOT USE TO PERFORM WORK or ISSUE FOR USE ***

Loss of AC Power U1&2 OP AP SD-1 REV. 14 PAGE 5 OF 22 ACTION / EXPECTED RESPONSE RESPONSE NOT OBTAINED

9. ENERGIZE 4kV Buses From Other Unit's S/U Power: (Continued)

CAUTION: Closely monitor the unaffected units' Startup transformer loading as additional loads are placed on the Startup bus. The transformer is capable of 75 MVA if forced air and oil cooling are available.

h. Limit the added load on Startup transformer 2-1(1-1) to 30 MVA Recall that MVA =

(MW 2+ MVAR2)

NOTE: 1 RCP equals approx. 6.6 MVA 1 CWP equals approx. 11.4 MVA

i. CLOSE S/U Feeder Bkrs to Desired 4kV Vital AND Non-Vital buses

10. RETURN To Procedure And Step In Effect

11. CHECK Other Unit's Aux Power - GO TO step 18.

AVAILABLE:

Aux Transformer 2-1 (1-1) Power Available White Status Lights - ON (Other Unit's VB5) 02 01069314.DOC 0831.0934 Page 5 of 23

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Loss of AC Power U1&2 OP AP SD-1 REV. 14 PAGE 6 OF 22 ACTION / EXPECTED RESPONSE RESPONSE NOT OBTAINED

12. ENERGIZE Both Unit's 12kV S/U Buses From Other Unit's Aux Power:

a. VERIFY the following Bkrs - OPEN

1) 52-VU-12

2) 52-VU-14

3) This Unit's 12kV Bus E S/U Feeder Bkr

Unit 1- 52-VE-06

(Unit 2- 52-VE-04)

4) This Unit's 12kV Bus D S/U Feeder Bkr

Unit 1- 52-VD-04

(Unit 2- 52-VD-06)

5) 52-VU-24

6) 52-VU-23

b. Place AUTO TRANSFER TO STARTUP CUTOUT Switch in the CUTOUT position for ALL 4kV AND 12kV buses on the affected Unit only

c. Depress AUTO TRANSFER reset pushbuttons for all 4kV and 12kV buses (for the affected unit only)

Blue lights - OFF THIS STEP CONTINUED ON NEXT PAGE 02 01069314.DOC 0831.0934 Page 6 of 23

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Loss of AC Power U1&2 OP AP SD-1 REV. 14 PAGE 7 OF 22 ACTION / EXPECTED RESPONSE RESPONSE NOT OBTAINED

12. ENERGIZE Both Unit's 12kV S/U Buses From Other Unit's Aux Power: (Continued)

CAUTION: Both Startup feeders 52-VU-15 and 52-VU-20 to the 12kV Underground system must be opened since the aux transformer's ground system is inadequate in the event of a fault on the 12kV underground loop.

NOTE: Permission to open 52-VU-15 and 52-VU-20 from MBSC need not be obtained.

d. Open the feeders to the 12kV Underground system:

1) 2-VU-15

2) 52-VU-20

e. CLOSE the following Bkrs:

1) Other Unit's 12kV Bus D S/U 1) Other Unit's 12kV Bus E S/U Feeder Bkr Feeder Bkr

Unit 1 VD-04

Unit 1-52-VE-06

(Unit 2 VD-06) * (Unit 2-52-VE-04)

2) 52-VU-11

13. CHECK Status of Own Unit's 12kV Bus:

a. 12kV S/U Bus Power Available White a. VERIFY the following Bkrs - CLOSED Status Light - ON

52-VU-21

52-VU-22 CAUTION: Closely monitor other unit's Aux transformer loading as additional loads are placed on the Startup bus. The transformer is capable of 56.25 MVA.

b. Total load must be limited to the load capacity of the transformer (56.25 MVA - forced cooling)

Recall that MVA =

(MW 2+ MVAR2) 02 01069314.DOC 0831.0934 Page 7 of 23

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Loss of AC Power U1&2 OP AP SD-1 REV. 14 PAGE 8 OF 22 ACTION / EXPECTED RESPONSE RESPONSE NOT OBTAINED

14. ENERGIZE Own Unit's 12/4kV S/U Transformer:

a. Depress the AUTO BUS TRANSFER Reset Pushbuttons on ALL 4kV and 12kV buses

Blue lights - OFF

b. Open S/U Feeder Bkrs for ALL 4kV buses:

UNIT 1 UNIT 2 52-HF-14 (52-HF-14) 52-HG-14 (52-HG-14) 52-HH-14 (52-HH-14) 52-HD-14 (52-HD-05) 52-HE-03 (52-HE-13)

c. CLOSE 52-VU-14 (52-VU-23)

15. ENERGIZE The Vital S/U Feeder Bkrs:

VERIFY 52-HG CLOSED

16. CLOSE S/U Feeder Bkrs To Desired 4kV Vital AND Nonvital Buses

Refer to OP AP-26, section B for desired buses and loads

17. RETURN To Procedure And Step In Effect

18. VERIFY AT LEAST ONE 4kV Vital Bus Refer to AR PK16, 17, OR 18 to restart a Diesel Energized From Associated Diesel Generator.

Generator 02 01069314.DOC 0831.0934 Page 8 of 23

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Loss of AC Power U1&2 OP AP SD-1 REV. 14 PAGE 9 OF 22 ACTION / EXPECTED RESPONSE RESPONSE NOT OBTAINED CAUTION: Monitor Diesel Generator loading while cross-tying buses. Refer to Appendix Q for load limits.

19. CROSS-TIE 4kV Buses As Required:

a. IMPLEMENT Appendix X to cross-tie Vital Buses

b. IMPLEMENT Appendix N to energize 480V Nonvital Buses as required

20. VERIFY Required 4kV AND 480V buses - IF Power is NOT sufficient to ensure ENERGIZED sufficient Decay Heat Removal, THEN IMPLEMENT OP AP SD-0, LOSS OF OR INADEQUATE DECAY HEAT REMOVAL, Step 7 AND RETURN To Step 2 in this procedure.

21. RETURN To Procedure And Step In Effect END 3.

02 01069314.DOC 0831.0934 Page 9 of 23

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Loss of AC Power U1&2 OP AP SD-1 REV. 14 PAGE 10 OF 22 APPENDICES 3.1 Appendix B, Estimation of Decay Heat and Heatup Rate 3.2 Appendix N, Energizing Nonvital 480V Buses With Diesel Generator 3.3 Appendix Q, Diesel Generator Load Limits 3.4 Appendix X, Crosstie of Vital Bus

4. ATTACHMENTS 4.1 "FoldOut Page," 12/30/03

5. REFERENCES 5.1 PG&E NOS/ISAG Calculational File No. 920815-0, "Heatup Rates During an Outage", August 21, 1992.

5.2 PG&E NOS/ISAG Calculational File No. 920831-0, "Revised Inventory Factors for Reduced Inventory Operations", September 1, 1992.

5.3 NESNE Calculational file No. N-147, "Inventory Factors for RCS Heatup",

August 19, 1994.

02 01069314.DOC 0831.0934 Page 10 of 23

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Loss of AC Power U1&2 OP AP SD-1 REV. 14 PAGE 11 OF 22 APPENDIX B ESTIMATION OF DECAY HEAT AND HEATUP RATE

1. PREDICTED HEAT LOAD

2. REDUCTION FACTOR FOR REFUELED CORES MW X = MW Predicted Fraction of Previously Estimated Heat Load Used assemblies Decay Heat Installed in Core

Load

Use 1.0 if unknown 02 01069314.DOC 0831.0934 Page 11 of 23

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Loss of AC Power U1&2 OP AP SD-1 REV. 14 PAGE 12 OF 22 APPENDIX B (Continued)

3. HEAT UP RATE PREDICTION MW X = Degrees per Estimated Inventory Predicted Minute Decay Heat Factor Heat Up Load Rate

a. INVENTORY FACTOR - Degrees/MW Min 107' 0.52 108' 0.45 Nozzle Dams Installed OR NO Nozzle Dams Installed AND SG Tubes Voided SG Tubes Not Voided 110' 0.40 112' 0.36 0.29 114' 0.33 0.27 116' 0.31 0.26 118' 0.31 0.054 Upper Internals Removed (Use 118' if Upper Internals Installed) 120' 0.06 130' 0.03 138' 0.02

4. ESTIMATED TIME TO REACH 200 DEGREES 200 _____________ Delta Temp Existing

-______ Temperature ÷ Actual or = _________

Predicted Minutes to

_______ Delta Temp _____________ Heat Up Rate reach 200 02 01069314.DOC 0831.0934 Page 12 of 23

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Loss of AC Power U1&2 OP AP SD-1 REV. 14 PAGE 13 OF 22 APPENDIX N ENERGIZING NONVITAL 480V BUSES WITH DIESEL GENERATOR

1. SCOPE 1.1 This Appendix provides general instructions for restoring power to plant auxiliaries which will facilitate plant recovery until off-site power is restored.

1.2 Use of this Appendix requires the approval of the Shift Manager.

2. INSTRUCTIONS 2.1 Verify Auto Transfer Cutouts for all 4kV and 12kV Buses - CUTOUT.

2.2 Reset Auto Bus Transfer for all 4kV and 12kV Buses - BLUE LIGHT OFF.

2.3 Verify All Vital 4kV Bus Auxiliary Feeder Breakers - OPEN.

52-HH-13, Bus H

52-HG-13, Bus G

52-HF-13, Bus F 2.4 Verify All Vital 4kV Bus Startup Feeder Breakers - OPEN.

52-HH-14, Bus H

52-HG-14, Bus G

52-HF-14, Bus F 2.5 Verify Vital 4kV Bus Common Startup Feeder Breaker - OPEN.

52-HG-15 2.6 Verify Nonvital 4kV Bus D Auxiliary Feeder Breaker - OPEN.

52-HD-15 (52-HD-4) 2.7 Verify Nonvital 4kV Bus E Auxiliary Feeder Breaker - OPEN.

52-HE-2 (52-HE-14) 2.8 Verify Nonvital 4kV Bus D Startup Feeder Breaker - OPEN.

52-HD-14 (52-HD-5) 2.9 Verify Nonvital 4kV Bus E Startup Feeder Breaker - OPEN.

52-HE-3 (52-HE-13) 2.10 Verify Startup Transformer Feeder Breaker - OPEN.

52-VU-14 (52-VU-23) 02 01069314.DOC 0831.0934 Page 13 of 23

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Loss of AC Power U1&2 OP AP SD-1 REV. 14 PAGE 14 OF 22 APPENDIX N (Continued) 2.11 Verify 4kV Bus D Feeder Breakers to 480V Buses - OPEN.

52-HD-6, Bus 15D (52-HD-6, Bus 23D)

52-HD-8, Bus 14D (52-HD-7, Bus 22D)

52-HD-10, Bus 11D (52-HD-9, Bus 21D)

52-HD-11, 230kV SWYD (52-HD-11, Bus 24D)

52-HD-12, Bus 12D (52-HD-13, Bus 25D)

52-HD-13, Bus 13D 2.12 Verify 4kV Bus E Feeder Breakers To 480V Buses - OPEN.

52-HE-4, Bus 13E (52-HE-4, Bus 25E)

52-HE-5, Bus 12E (52-HE-6, Bus 24E)

52-HE-7, 500kV SWYD (52-HE-8, Bus 21E)

52-HE-8, Bus 11E (52-HE-11, Bus 22E)

52-HE-10, Bus 14E (52-HE-12, Bus 23E)

52-HE-12, Bus 15E 2.13 Dispatch operators to ALL Nonvital 480V load centers on the affected Unit. Open all individual load supply breakers.

11D (21D) 11E (21E)

12D (22D) 12E (22E)

13D (23D) 13E (23E)

14D (24D) 14E (24E)

15D (25D) 15E (25E) 2.14 Diesel Generator Protection At SFM discretion, cutin the FCOs for the diesel generator selected to supply the Nonvital Buses.

CAUTION: The following steps will reenergize nonvital buses. If the diesel generator appears unstable, immediately reopen nonvital bus supply breakers.

2.15 CLOSE the startup feeder breaker for the D/G feeding the nonvital loads.

52-HF-14 for D/G 1-3 (2-3)

52-HG-14 for D/G 1-2 (2-1)

52-HH-14 for D/G 1-1 (2-2) 02 01069314.DOC 0831.0934 Page 14 of 23

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Loss of AC Power U1&2 OP AP SD-1 REV. 14 PAGE 15 OF 22 APPENDIX N (Continued)

NOTE: There will be a load surge on the diesel generator as the startup transformer is reenergized when 52-HG-15 is closed.

2.16 CLOSE breaker 52-HG-15, startup power common supply to vital Buses F, G, and H.

2.17 CLOSE startup feeder to 4kV Bus D.

52-HD-14 (52-HD-05) 2.18 CLOSE startup feeder to 4kV Bus E.

52-HE-03 (52-HE-13).

2.19 Determine desired loads. Refer to Table 1 to determine power supply and power requirements of key plant auxiliaries. When determining load power requirements, consider the starting current surge.

2.20 Evaluate diesel generator reserve capacity (REFER TO APPENDIX Q).

2.21 Determine power requirements of desired load.

2.22 Determine power supply of desired load.

2.23 Verify load control switch position - OFF.

2.24 Verify the desired load center is energized.

CAUTION: Evaluate diesel generator stability as each additional load is energized and immediately shed nonvital load if the diesel generator appears overloaded or unstable.

2.25 Close the load supply breaker.

2.26 Refer to OP AP-26 for other non-vital loads which may be desirable if it is determined that normal non-vital power supplies will not be available for an extended length of time.

02 01069314.DOC 0831.0934 Page 15 of 23

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Loss of AC Power U1&2 OP AP SD-1 REV. 14 PAGE 16 OF 22 APPENDIX N (Continued)

TABLE 1 LOAD BUS PWR REQ BREAKER

1. Screen Wash Pp 1-1 14D 290 KW 52-14D-03 1-2 14E 290 KW 52-14E-03 2-1 24D 290 KW 52-24D-03

2. Service Cooling Wtr Pp 1-1 11D 83 KW 52-11D-05 1-2 11E 83 KW 52-11E-05 2-1 21D 83 KW 52-21D-05 2-2 21E 83 KW 52-21E-05

3. Air Compressors (Control Power for 0-1 to 0-4 is selectable from 52-1F-27 OR 52-11E-27) 0-1 15D 62 KW 52-15D-05 0-2 15E 62 KW 52-15E-05 0-3 25D 62 KW 52-25D-05 0-4 25E 62 KW 52-25E-05 0-5* 25D 124 KW 52-25D-11 0-6* 11E 124 KW 52-11E-15 0-7 15E 124 KW 52-15E-37

Requires either SCW Booster Pp to be running SCW Booster Pumps 0-1 11E 2 KW 52-11E-03 0-2 25D 2 KW 52-25D-31

4. Nonvital 280VDC Battery Chargers ED15 (25) 15D (25D) 23 (34) KW 52-15D-36 (52-25D-36)

ED16 (26) 15E (25E) 23 (34) KW 52-15E-13 (52-25E-13)

5. Digital FW Cont Sys Rect/Chgr 12J (22J) 10 KVA 52-12J-26 (52-22J-05)

6. Plt Process Computer Inverter - IC111 12I (22I) 30 KVA 52-12I-17 (52-22I-36)

7. SPDS UPS Battery Charger EJBC (25D) 10 KVA (52-25D-39)

Inverter Alternate AC (25E/I) 10 KVA (52-25I-28A) 02 01069314.DOC 0831.0934 Page 16 of 23

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Loss of AC Power U1&2 OP AP SD-1 REV. 14 PAGE 17 OF 22 APPENDIX N (Continued)

LOAD BUS PWR REQ BREAKER NOTE: Power requirements indicated for the 500kV and 230kV Station Services Transformers represent transformer ratings for THPW-1 and THPF-1 (powered from Unit 1 only).

Normally, the transformers carry almost no load. Coordinate with the switchyard operator to summarize initial load requirements by placing switchyard loads in service sequentially.

8. 230kV Swyd Sta 4kV 150 KVA 52-HD-11 Serv Trans BUS D

9. 500kV Swyd Sta 4kV 750 KVA 52-HE-7 Serv Trans BUS E 02 01069314.DOC 0831.0934 Page 17 of 23

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Loss of AC Power U1&2 OP AP SD-1 REV. 14 PAGE 18 OF 22 APPENDIX Q DIESEL GENERATOR LOAD LIMITS 02 01069314.DOC 0831.0934 Page 18 of 23

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Loss of AC Power U1&2 OP AP SD-1 REV. 14 PAGE 19 OF 22 APPENDIX X CROSSTIE OF VITAL BUS

1. SCOPE 1.1 Implementation of the Appendix requires approval of the Site Emergency Coordinator or his designate. This Appendix should be performed to energize two vital buses from one diesel. If two vital buses are energized, enough ESF equipment should be energized to establish effective Decay Heat Removal.

2. DISCUSSION 2.1 As a general guideline, in this situation several options may be available depending on plant conditions and RCS status.

3. PREREQUISITES 3.1 Verify all the breakers AND DC control power switches are OPEN for the 4kV loads AND ALL 480V Breakers are racked out on the deenergized bus being reenergized to prevent automatic loading and overloading the diesel. Refer to list below:

F VITAL BUS G VITAL BUS H VITAL BUS ASW Pp 1 52-HF-08 ASW Pp 2 52-HG-06 AFW Pp 2 52-HH-08 AFW Pp 3 52-HF-09 CS Pp 1 52-HG-07 CS Pp 2 52-HH-09 CCP 1 52-HF-11 RHR Pp 1 52-HG-08 RHR Pp 2 52-HH-09 CCW Pp 1 52-HF-12 CCP 2 52-HG-09 CCW Pp 3 52-HH-12 SI Pp 1 52-HF-15 PDP 52-HG-11 SI Pp 2 52-HH-15 CCW Pp 2 52-HG-12

4. PRECAUTIONS AND LIMITATIONS 4.1 Some equipment may have to be operated on a continuous basis and some on an "as needed" basis. Existing conditions will determine which equipment is needed.

4.2 The maximum capacity of the diesel generator should not be exceeded.

Appendix Q identifies the diesel generator load limits. Table 1 provides loads for various vital 4kV and 480V vital equipment. STP M-9M also contains specific loads on all 480V vital equipment.

4.3 Start only one piece of equipment at a time, allowing at least 4 seconds between each start, since starting current may cause bus failure.

02 01069314.DOC 0831.0934 Page 19 of 23

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Loss of AC Power U1&2 OP AP SD-1 REV. 14 PAGE 20 OF 22 APPENDIX X (Continued)

5. INSTRUCTIONS 5.1 Obtain permission from the Site Emergency Coordinator or his designate.

5.2 Reset SI (if applicable) so that the affected bus will not try to auto load when the bus becomes energized.

5.3 Cutout the AUTO Transfer FCOs for 4kV Buses and 12kV Buses.

NOTE: If the D/G associated with the deenergized bus is running but will not load on the bus, it must be shutdown to permit the Auto Transfer Relay to be reset.

5.4 Depress all AUTO Transfer Reset Pushbuttons, verify the BLUE lights go OUT.

5.5 Verify OPEN all vital Bus 4kV auxiliary feeder breakers, 52-HH-13, 52-HG-13 and 52-HF-13.

5.6 Verify OPEN all vital Bus 4kV startup feeder breakers, 52-HH-14, 52-HG-14 and 52-HF-14.

5.7 Verify OPEN startup feeder breaker 52-HG-15, to the vital Buses F, G and H.

5.8 Verify OPEN the 4kV to 480V Bus feeder breaker for the deenergized Bus to be reenergized:

DEENERGIZED BUS FEEDER BUS F 52-HF-10 G 52-HG-10 H 52-HH-10 5.9 Determine the D/G to supply the deenergized bus. Station an operator at VB4 to monitor the diesel generator to supply the deenergized bus.

CAUTION: The prerequisites of this Appendix must be completed prior to performing step 5.10.

5.10 CLOSE the 4kV startup feeder breaker for the Bus that will be supplying power to the deenergized Bus.

OPERATING D/G CLOSE No. 1 52-HH-14 (52-HG-14)

No. 2 52-HG-14 (52-HH-14)

No. 3 52-HF-14 (52-HF-14) 02 01069314.DOC 0831.0934 Page 20 of 23

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Loss of AC Power U1&2 OP AP SD-1 REV. 14 PAGE 21 OF 22 APPENDIX X (Continued)

CAUTION: If the diesel generator appears unstable during the performance of step 5.11 or at any time beyond this point in the procedure, immediately open startup feeder breaker for the inoperable bus to separate the diesel from the inoperable bus.

5.11 CLOSE the 4kV startup feeder breaker for the deenergized bus being reenergized.

DEENERGIZED BUS FEEDER BREAKER F 52-HF-14 G 52-HG-14 H 52-HH-14 5.12 CLOSE the 4kV to 480V bus feeder breaker for the reenergized bus:

BUS BREAKER F 52-HF-10 G 52-HG-10 H 52-HH-10 CAUTION: CCP and CCW Pps require their 480V aux lube oil Pp and the ASW Pp requires its 480V exhaust fan breaker to be shut prior to starting the Pp.

5.13 Operate only the equipment needed for the existing conditions. Evaluate the load on the diesel generator prior to energizing each additional load to ensure its capacity limit is not exceeded REFER TO APPENDIX Q. Note that this curve is based on the number of hours the diesel was operated in an overloaded condition, not total run hours. Refer to Table 1 to identify the expected maximum loads for each piece of equipment. If the diesel overloads, it may only be necessary to trip the last load added rather than the entire bus.

NOTE: Battery chargers are necessary within two hours to ensure continued instrument AC power.

5.14 Energize DC buses from their battery chargers within the limits of the diesel capacity (max 64 KW each). Refer to OP J-9:II, "Operating the Battery Chargers."

11 (21) 52-1F-42 (2F-42) 121 (221) 52-1H-60 (2H-60) 12 (22) 52-1G-42 (2G-42) 131 (231) 52-1F-52 (2F-52) 132 (232) 52-1H-34 (2H-34) 02 01069314.DOC 0831.0934 Page 21 of 23

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Loss of AC Power U1&2 OP AP SD-1 REV. 14 PAGE 22 OF 22 APPENDIX X (Continued)

TABLE 1: EQUIPMENT LOADS Rating Max Demand in KW Load Qty (Each) Bus F Bus G Bus H

1. Centrifugal Charging Pumps 2 600 Hp 515 515 -----

2. Safety Injection Pumps 2 400 Hp 330 ----- 330

3. Containment Spray Pumps 2 400 Hp ----- 350 350

4. Residual Heat Removal Pumps 2 400 Hp ----- 333 333

5. Containment Fan Cooler 5

a. Slow Speed 100 Hp 82 ea 82 ea 82

b. Fast Speed 300 Hp 240 ea 240 ea 240

6. Component Cooling Water 3 400 Hp 342 342 342 Pumps

7. Auxiliary Saltwater Pumps 2 440 Hp 361 361 -----

8. Auxiliary Feedwater Pumps 2 600 Hp 395 ----- 395

9. Fire Pumps 2 200 Hp 147 ----- 147

10. Pressurizer Heaters

2 483/207 483/207

11. Remaining 480V loads are ----- ----- ----- ----- -----

extensive (Refer to STP M-9M for a specific listing)

483 KW for 7 Heaters; 207 for 3 Heaters 02 01069314.DOC 0831.0934 Page 22 of 23

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DCPP (12/30/03) FOLDOUT PAGE FOR PAGE 1 OF 1 U1&2 OP AP SD-1 1.0 EVALUATION OF HEATUP RATE - STA If Decay heat removal is lost for > 2 minutes:

CAUTION: In core T/C's will not reflect actual core exit temperatures if ECCS injection is into RCS hot legs.

Evaluate rate of RCS heatup using Appendix B and change in actual In-core T/C temperatures.

Determine time until RCS will exceed 200°, inform SS and SFM.

2.0 CONTAINMENT CLOSURE INITIATION CRITERIA Initiate Containment closure if:

RCS pressurization takes place due to loss of decay heat removal.

RCS temperature is projected to increase to > 200° in < one hour.

RCS refilling efforts may cause a spill of the RCS into containment.

Rx Vessel level decreases to < 107' 3" with fuel in vessel.

RHR not restored within 10 minutes with fuel in vessel.

3.0 CONTAINMENT CLOSURE ACTIONS If containment closure is required:

Sound the Containment Evacuation alarm

Evacuate non-essential personnel from containment

Periodically monitor Containment Radiation monitors RM 2, 7, 30, 31

Verify Equipment hatch closed

Verify at least one personnel hatch door closed

Verify at least one emergency personnel hatch door closed

Verify Steam generator secondary sides isolated

Verify SFS-50 closed or transfer tube flange installed

Verify Containment Ventilation Isolation Operable

Run all available CFCUs in fast speed 4.0 ALTERNATIVE HEAT REMOVAL METHODS IF -

RCS begins to pressurize due to loss of RHR

Reactor Vessel level falls below 106' 1"

RHR cooling unavailable

RCS temperature is projected to increase above 200° THEN -

Refer to OP AP SD-0 step 7 to select and implement the alternative method(s) of decay heat removal.

02 01069314.DOC 0831.0934 Page 23 of 23

NUCLEAR POWER GENERATION DIABLO CANYON POWER PLANT JOB PERFORMANCE MEASURE Number: NRCLJC-103

Title:

RESPOND TO A LOSS OF CCW FLOW TO ONE RCP Examinee:

Evaluator:

Print Signature Date Results: Sat Unsat Total Time: minutes Comments:

References:

AR PK01-08, CCW HEADER C, Rev 16 OP AP-11, Malfunction of Component Cooling Water System, Rev 21 Alternate Path: Yes X No Time Critical: Yes No X Time Allotment: 15 minutes Critical Steps: 4,5,6 Job Designation: RO/SRO Task Number: 008/08/A2.01 Rating: 3.3/3.6 AUTHOR: JACK BLACKWELL DATE: 01/18/2005 REVIEWED BY: N/A DATE:

TRAINING LEADER APPROVED BY: N/A DATE:

LINE MANAGER REV. 01

JPM TITLE: RESPOND TO A LOSS OF CCW FLOW TO ONE RCP JPM NUMBER: NRCLJC-103 INSTRUCTOR WORKSHEET Directions: No PLANT controls or equipment are to be operated during the performance of this Job Performance Measure. All actions taken by the examinee should be clearly demonstrated and verbalized to the evaluator. The student will be given the initial conditions, initiating cue, and task standard. The examiner will then ask if any clarifications are needed. The examinee may be given the applicable procedure and step with which to begin.

Required Materials: None Initial Conditions: Unit 1 is operating at 100% power.

Initiating Cue: PK01-08, CCW HEADER C, has just alarmed. Input 428, RCP Thermal Barrier CCW Flow Lo is causing the alarm.

Task Standard: The alarms have been responded to and appropriate actions have been taken in accordance with applicable plant procedures.

NRCLJC103.DOC PAGE 2 OF 6 REV. 01

JPM TITLE: RESPOND TO A LOSS OF CCW FLOW TO ONE JPM NUMBER: NRCLJC-103 RCP INSTRUCTOR WORKSHEET Start Time:

Step Expected Operator Actions

1. Obtain the correct procedure. 1.1 References AR PK01-08.

Note: Operator may go directly to OP AP-11 Step was: Sat: ______ Unsat: _______*

2. Perform actions for RCP lube oil 2.1 Observes that two CCW pumps are cooler low flow. running.

2.2 Observes that FCV-355 and FCV-356 are open.

Note: Operator may use PPC PICTURE RCP or Group Display PK05-02 to monitor RCP 1-2.

2.3 Observes RCP lower bearing temps normal and proper seal injection flow on RCPs.

2.4 Refers to OP AP-11, Section E.

Step was: Sat: ______ Unsat _______*

3. Verify CCW Flow To All RCP 3.1 Reads CAUTION.

Lube Oil Coolers:

3.2 Observes that the following valves

a. Verify CCW Vlvs - OPEN are open:

b. RCP L.O. Clr CCW Flow LO

FCV-355 Alarm (PK01-08) - NOT IN

FCV-356

c. RCP Temp PPC Alarm

FCV-749 (PK05-01), 02, 03, 04) - NOT

FCV-363 IN 3.3 Observes that PK01-08 is in alarm.

3.4 Determines RCP Lube Oil coolers have CCW flow.

Step was: Sat: ______ Unsat _______*

Denotes an entry required on the JPM cover sheet.

- Denotes Critical Step and Sub Steps NRCLJC103.DOC PAGE 3 OF 6 REV. 01

JPM TITLE: RESPOND TO A LOSS OF CCW FLOW TO ONE JPM NUMBER: NRCLJC-103 RCP INSTRUCTOR WORKSHEET Step Expected Operator Actions

- 4. VERIFY RCP Seal Injection In 4.1 Observes Seal Injection between 8 Service. and 13 gpm.

4.2 Observes RCP Seal #1 Outlet Temps and Radial Brg Outlet Temps NORMAL.

Step was: Sat: ______ Unsat _______*

- 5. VERIFY CCW Flow to All RCP 5.1 Reads Caution.

Thermal Barriers Normal.

5.2 Verifies FCV-357 Closed and PK01-08 IN.

- 5.3 Goes to Step 5.b of Section B.

Step was: Sat: ______ Unsat _______*

- 6. Isolate Leak. 6.1 Closes FCV-750.

6.2 Locally closes CCW valves for RCPs 1, 2, 3, 4.

Cue: An Operator in the field will close the valves.

6.3 Monitors containment sump for expected level increase.

6.4 Implements OP AP-1 for excessive RCS leakage.

Cue: The SFM will take care of sump monitoring and AP-1.

Step was: Sat: ______ Unsat _______*

Stop Time:

Total Time: (Enter total time on the cover page)

Denotes an entry required on the JPM cover sheet.

- Denotes Critical Step and Sub Steps NRCLJC103.DOC PAGE 4 OF 6 REV. 01

JPM NUMBER: NRCLJC-103 EXAMINEE CUE SHEET Initial Conditions: Unit 1 is operating at 100% power.

Initiating Cue: PK01-08, CCW HEADER C, has just alarmed. Input 428, RCP Thermal Barrier CCW Flow Lo is causing the alarm.

Task Standard: The alarms have been responded to and appropriate actions have been taken in accordance with applicable plant procedures.

NRCLJC103.DOC PAGE 5 OF 6 REV. 01

JPM TITLE: RESPOND TO A LOSS OF CCW FLOW TO ONE RCP JPM NUMBER: NRCLJC-103 ATTACHMENT 1, SIMULATOR SETUP Initialize the simulator to IC-510 (100%, MOL).

Manually insert the following:

Command Description

1. vlv ccw8 2,0,0,0,d,0 CCW RCP Thermal Barrier Return Isolation FCV-357 Ensure the annunciator CRT and alarm viewer contain the alarm inputs required by the JPM.

Ensure PPC alarms acknowledged.

Inform the examiner that the simulator setup is complete.

Go to RUN when the examinee completes reading the cue sheet.

NRCLJC103.DOC PAGE 6 OF 6 REV. 01

- - UNCONTROLLED PROCEDURE - DO NOT USE TO PERFORM WORK orISSUE FOR USE ***

PACIFIC GAS AND ELECTRIC COMPANY NUMBER AR PK01-08 NUCLEAR POWER GENERATION REVISION 16 DIABLO CANYON POWER PLANT PAGE 1 OF 3 ANNUNCIATOR RESPONSE UNIT TITLE: CCW HEADER C 102/14/03 EFFECTIVE DATE PROCEDURE CLASSIFICATION: QUALITY RELATED

1. LOGIC DIAGRAM FS88 FS137 FS95 FS92A ALARM FS92B FS96 FS89

2. ALARM INPUT DESCRIPTION DEVICE ALARM ANNUNCIATOR TYPEWRITER NUMBER INPUT PRINTOUT SETPOINT FS 92A 264 RCP Thermal Barrier CCW Flo Hi GT 220 GPM (Hi Flow Isolation at 250 +/- GPM)

FS 92B 428 RCP Thermal Barrier CCW Flo Lo LT 140 GPM FS 88 265 RCP 1-1 or 1-3 L.O. Clr CCW Flo Lo LT 106 GPM FS 137 265 RCP 1-1 or 1-3 L.O. Clr CCW Flo Lo LT 106 GPM FS 96 1372 RCP 1-2 or 1-4 L.O. Clr CCW Flo Lo LT 106 GPM FS 89 1372 RCP 1-2 or 1-4 L.O. Clr CCW Flo Lo LT 106 GPM FS 95 429 CCW Hdr-C Flo Lo LT 2500 GPM 08261516.DOC 16 0214.1227

- - UNCONTROLLED PROCEDURE - DO NOT USE TO PERFORM WORK orISSUE FOR USE ***

PACIFIC GAS AND ELECTRIC COMPANY NUMBER AR PK01-08 DIABLO CANYON POWER PLANT REVISION 16 PAGE 2 OF 3 TITLE: CCW HEADER C UNIT 1

3. PROBABLE CAUSE 3.1 Thermal Barrier High Flow 3.1.1 CCW flow manual control valve set too high.

3.1.2 Failure of RCP thermal barrier pressure integrity.

NOTE: The following alarms will actuate on a Phase B Containment Isolation due to the isolation of CCW header C.

3.2 Thermal Barrier Low Flow 3.2.1 CCW flow manual control valves out of adjustment.

3.2.2 Not sufficient CCW pumps running.

3.2.3 CCW supply valve FCV-355 or FCV-356 closed.

3.3 RCP Lube Oil Cooler Low Flow 3.3.1 CCW manual flow control valves out of adjustment.

3.3.2 Not sufficient CCW pumps running.

3.3.3 CCW supply header valves closed FCV-355 or 356.

3.4 Header C Low Flow 3.4.1 CCW pumps trip without standby start.

3.4.2 Closing of supply or return valve on a large load such as FCV-356 to containment.

3.4.3 Misalignment of FCVs at CCW Hx.

3.4.4 Low frequency on 4KV vital bus F, G, or H.

4. AUTOMATIC ACTIONS 4.1 Thermal Barrier High Flow 4.1.1 Possible isolation of all RCP thermal barrier CCW return.

4.1.2 Possible isolation of CCW surge tank vent valve.

5.

08261516.DOC 16 0214.1227

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PACIFIC GAS AND ELECTRIC COMPANY NUMBER AR PK01-08 DIABLO CANYON POWER PLANT REVISION 16 PAGE 3 OF 3 TITLE: CCW HEADER C UNIT 1 OPERATOR ACTIONS 5.1 Thermal Barrier High Flow, ALARM INPUT 264 5.1.1 Check annunciator printout 5.1.2 Check FCV-357 OC and FCV-750 IC position.

5.1.3 Check RCP lower bearing temps - verify proper RCP Seal Flow 5.1.4 Refer to OP AP-11, Section B, "CCW System Inleakage," or OP AP SD-4, "Loss of CCW."

NOTE: If header C has isolated due to a Containment Isolation Phase B, stop the reactor coolant pumps within 5 minutes in accordance with EOP E-0 foldout page.

5.2 Thermal Barrier Low Flow, ALARM INPUT 428 5.2.1 Check annunciator printout for item.

5.2.2 Check 2 CCW pumps running.

5.2.3 Check indicating lights for FCV-355 and 356 open or red.

5.2.4 Check RCP lower bearing temp - verify proper RCP Seal Injection Flow.

5.2.5 Refer to OP AP-11, Section E, "Loss of CCW Flow to RCPs," or OP AP SD-4, "Loss of CCW."

5.3 RCP Lube Oil Cooler Low Flow, ALARM INPUT 265, 1372 5.3.1 Check annunciator printout for item.

5.3.2 Check 2 CCW pumps running.

5.3.3 Check FCV-355 and 356 open or red light on.

5.3.4 Check RCP bearing temps on PPC.

5.3.5 Refer to OP AP-11, Section E, "Loss of CCW Flow to RCPs," or OP AP SD-4, "Loss of CCW."

5.4 Header C Low Flow, ALARM INPUT 429 5.4.1 Check annunciator printout.

5.4.2 Check FI-46 on VB1 for flow.

5.4.3 If zero flow is indicated and one or more pumps are running check open or open FCV-355.

a. Check vital 4KV busses for low frequency 5.4.4 Monitor RCP Brg temps.

5.4.5 Refer to OP AP-11, Section E, "Loss of CCW Flow to RCPs," or OP AP SD-4, "Loss of CCW."

5.4.6 Refer to OP F-2 as necessary.

08261516.DOC 16 0214.1227

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PACIFIC GAS AND ELECTRIC COMPANY NUMBER OP AP-11 NUCLEAR POWER GENERATION REVISION 21 DIABLO CANYON POWER PLANT PAGE 1 OF 37 ABNORMAL OPERATING PROCEDURE UNITS TITLE: Malfunction of Component Cooling Water System 1 2 AND 03/25/03 EFFECTIVE DATE PROCEDURE CLASSIFICATION: QUALITY RELATED

1. SCOPE 1.1 This procedure covers Component Cooling Water (CCW) System leakage or loss of cooling to various vital components while in MODES 1-4. If in MODE 5 or 6, OP AP SD-4, Loss of Component Cooling Water, should be used if Decay Heat Removal is threatened.

1.2 Prompt corrective action is vital to prevent complete deterioration of the system. The primary action is to isolate the defective component or section and terminate the leakage.

SECTION A: LOSS OF A CCW PUMP/HIGH CCW SYSTEM TEMP - pg. 2 SECTION B: CCW SYSTEM INLEAKAGE - pg. 4 SECTION C: CCW SYSTEM OUTLEAKAGE - pg. 12 SECTION D: LOSS OF CCW FLOW TO THE LETDOWN HX - pg. 15 SECTION E: LOSS OF CCW FLOW TO THE RCPs - pg. 16 SECTION F: LOSS OF SURGE TANK - pg. 18 APPENDIX A: CLEARING A CCW HEADER DUE TO HEADER FAILURE -

pg. 21 APPENDIX B: CCW HEAT LOAD ISOLATION - pg. 27 APPENDIX C: BACKUP COOLING TO A CENTRIFUGAL CHARGING PUMP -

pg. 31 APPENDIX D INSTRUCTIONS FOR LOSS OF ULTIMATE HEAT SINK APPENDIX E ESTIMATION OF DECAY HEAT/HEAT REMOVAL CAPABILITY GRAPHS - pg. 37

2. SYMPTOMS See Appropriate Section 00079821.DOC 02 0325.1036

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PACIFIC GAS AND ELECTRIC COMPANY NUMBER OP AP-11 DIABLO CANYON POWER PLANT REVISION 21 PAGE 4 OF 37 TITLE: Malfunction of Component Cooling Water System UNITS 1 AND 2 SECTION B: CCW SYSTEM INLEAKAGE SYMPTOMS

1. Surge tank level indicators reading high

2. Possible Main Annunciator Alarms

a. CCW SURGE TANK (PK01-07)

CCW Surge Tk Lvl Hi

b. CCW Header C (PK01-08)

RCP Thermal Barrier CCW Flo Hi

c. RCP _____ (PK05-01, 02, 03, 04)

1) RCP _____ Radial Brg Temp Hi

2) RCP _____ No. 1 Seal Outlet Temp Hi

d. HIGH RADIATION (PK11-21)

Process Monitor Hi-Rad (RE-17A and B)

ACTION/EXPECTED RESPONSE RESPONSE NOT OBTAINED

1. CHECK RE-17A AND B NOT In Alarm VERIFY CCW Surge Tk Vent RCV-16 CLOSED.

PK11-21 NOT in alarm CAUTION 1: If RCP No. 1 Seal Outlet temperature exceeds 235°F OR RCP Radial Bearing temperature exceeds 225°F, DO NOT restore RCP seal cooling.

CAUTION 2: If FCV-357 closed on high flow, do not attempt to open FCV-357 until condition causing high flow is cleared.

2. VERIFY RCP Operability:

a. Verify thermal barrier CCW outlet a. VERIFY RCP seal injection flow.

valve FCV-357 OPEN

b. Verify RCP Radial Bearing b. Shutdown the RCPs Temperature LESS THAN 225°F 1) TRIP the reactor AND 2) TRIP affected RCPs

3) GO TO EP E-O, REACTOR TRIP RCP No. 1 Seal Outlet temperature OR SAFETY INJECTION LESS THAN 235° 00079821.DOC 02 0325.1036

- - UNCONTROLLED PROCEDURE - DO NOT USE TO PERFORM WORK orISSUE FOR USE ***

PACIFIC GAS AND ELECTRIC COMPANY NUMBER OP AP-11 DIABLO CANYON POWER PLANT REVISION 21 PAGE 5 OF 37 TITLE: Malfunction of Component Cooling Water System UNITS 1 AND 2 SECTION B: CCW SYSTEM INLEAKAGE (Continued)

ACTION/EXPECTED RESPONSE RESPONSE NOT OBTAINED

3. VERIFY CCW Surge Tank Makeup NOT Locally ISOLATE makeup supply valve.

The Source Of Inleakage:

Check CCW Surge Tank makeup supply valves:

LCV-69 CLOSED

LCV-70 CLOSED - - - - - - - - - - - - - - - - - - - -- - - - -

4. REQUEST Sample Analysis:

Request CARP To Sample CCW To Assist In Leak Location NOTE: Various methods may be used to identify leakage from the following components, including:

Observance of related flows and temperatures.

Radiation surveys of associated lines.

Selective isolation of primary water side of components.

Selective isolation of CCW to components.

5. DETERMINE Leak Location:

Verify the following components are not the source of RCS inleakage:

a. Letdown Heat Exchanger a. ISOLATE heat exchanger per Appendix B Step 3.3.

AND Refer to OP AP-18, LETDOWN LINE FAILURE.

b. RCP Thermal barriers 1) Verify FCV-750 CLOSED

2) Locally ISOLATE Thermal Barrier CCW return (inside containment) by closing as applicable:

RCP 1: CCW-234 RCP 2: CCW-242 RCP 3: CCW-251 RCP 4: CCW-262 THIS STEP CONTINUED ON NEXT PAGE 00079821.DOC 02 0325.1036

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PACIFIC GAS AND ELECTRIC COMPANY NUMBER OP AP-11 DIABLO CANYON POWER PLANT REVISION 21 PAGE 6 OF 37 TITLE: Malfunction of Component Cooling Water System UNITS 1 AND 2 SECTION B: CCW SYSTEM INLEAKAGE (Continued)

ACTION/EXPECTED RESPONSE RESPONSE NOT OBTAINED

5. DETERMINE Leak Location:

(Continued)

b. RCP Thermal barriers (Continued) 3) Monitor containment sump for expected level increase

4) IMPLEMENT OP AP-1, EXCESSIVE REACTOR COOLANT SYSTEM LEAKAGE

c. Excess Letdown Heat Exchanger 1) ISOLATE RCS flow to Heat Exchanger (VB2)

Close CVCS-8166

- OR -

Close CVCS-8167

Close HCV-123

2) ISOLATE CCW flow to Heat Exchanger:

Locally Close CCW-426

Locally Close CCW-431

- OR -

Close FCV-361

3) Adjust charging flow to minimum or restore normal letdown.

THIS STEP CONTINUED ON NEXT PAGE 00079821.DOC 02 0325.1036

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PACIFIC GAS AND ELECTRIC COMPANY NUMBER OP AP-11 DIABLO CANYON POWER PLANT REVISION 21 PAGE 7 OF 37 TITLE: Malfunction of Component Cooling Water System UNITS 1 AND 2 SECTION B: CCW SYSTEM INLEAKAGE (Continued)

ACTION/EXPECTED RESPONSE RESPONSE NOT OBTAINED

5. DETERMINE Leak Location:

(Continued)

d. RHR Heat Exchanger No. 1 1) ISOLATE RCS flow to HX:

Locally CLOSE RHR-8724A, RHR HX No. 1 inlet

CLOSE HCV-638, RHR Hx No.

1 Outlet to RC loops

Locally CLOSE RHR-8734A, RHR No. 1 Train bypass to LTDN HX inlet

CLOSE FCV-641A, RHR PP No. 1, Recirc

VERIFY CLOSED CS-9003A, RHR HX No. 1 Outlet Hdr to CNTMT Spray Hdr A

VERIFY CLOSED SI-8804A, RHR PP disch to Charging PP suction

2) ISOLATE CCW flow to RHR HX No. 1:

Locally CLOSE CCW-457

Locally CLOSE CCW-459

- OR -

Close FCV-365 NOTE: CCW-457/459 are sealed-open valves and require Valve Seal Change Form to break seal.

THIS STEP CONTINUED ON NEXT PAGE 00079821.DOC 02 0325.1036

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PACIFIC GAS AND ELECTRIC COMPANY NUMBER OP AP-11 DIABLO CANYON POWER PLANT REVISION 21 PAGE 8 OF 37 TITLE: Malfunction of Component Cooling Water System UNITS 1 AND 2 SECTION B: CCW SYSTEM INLEAKAGE (Continued)

ACTION/EXPECTED RESPONSE RESPONSE NOT OBTAINED

5. DETERMINE Leak Location:

(Continued)

e. RHR Heat Exchanger No. 2 1) ISOLATE RCS flow to HX:

Locally CLOSE RHR-8724B, RHR HX No. 2 inlet

CLOSE HCV-637, RHR HX No.

2 outlet to RC loops

Locally CLOSE RHR-8734B, RHR No. 2 Train Bypass to LTDN HX inlet

CLOSE FCV-641B, RHR PP No. 2 Recirc

VERIFY CLOSED CS-9003B, RHR HX No. 2 Outlet Hdr to CNTMT Spray Hdr B

VERIFY CLOSED SI-8804B, RHR HX No. 2 Outlet to SI Pp No. 2 suction

2) ISOLATE CCW flow to RHR HX No. 2

Locally CLOSE CCW-150 (50)

Locally CLOSE CCW-151 (50)

- OR -

CLOSE FCV-364 NOTE: CCW-150/151 are sealed-open valves and require a Sealed Component Change Form to break their seals.

THIS STEP CONTINUED ON NEXT PAGE 00079821.DOC 02 0325.1036

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PACIFIC GAS AND ELECTRIC COMPANY NUMBER OP AP-11 DIABLO CANYON POWER PLANT REVISION 21 PAGE 9 OF 37 TITLE: Malfunction of Component Cooling Water System UNITS 1 AND 2 SECTION B: CCW SYSTEM INLEAKAGE (Continued)

ACTION/EXPECTED RESPONSE RESPONSE NOT OBTAINED

5. DETERMINE Leak Location:

(Continued)

f. RHR Pump Seal Coolers f. Locally ISOLATE CCW flow to cooler by closing the following valves, as applicable:

Pump 1: CCW-460 AND CCW 462

Pump 2: CCW-153 AND CCW 154 NOTE: The above valves are sealed-open (S.O.)

valves and require Sealed Component Change Form to break their seal.

g. PZR Steam Space Sample Cooler 1) Locally CLOSE NSS-9371A reactor coolant supply to cooler.

2) Locally ISOLATE CCW flow to cooler:

CLOSE CCW-379

CLOSE CCW-380

h. PZR Liquid Space Sample Cooler 1) Locally CLOSE NSS-9371B, RC supply to cooler

2) Locally ISOLATE CCW flow to cooler:

CLOSE CCW-377

CLOSE CCW-378 THIS STEP CONTINUED ON NEXT PAGE 00079821.DOC 02 0325.1036

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PACIFIC GAS AND ELECTRIC COMPANY NUMBER OP AP-11 DIABLO CANYON POWER PLANT REVISION 21 PAGE 10 OF 37 TITLE: Malfunction of Component Cooling Water System UNITS 1 AND 2 SECTION B: CCW SYSTEM INLEAKAGE (Continued)

ACTION/EXPECTED RESPONSE RESPONSE NOT OBTAINED

5. DETERMINE Leak Location:

(Continued)

i. RCS Hot Legs 1 & 4 Sample Cooler 1) ISOLATE RC supply to cooler by locally closing NSS-9371C.

2) ISOLATE CCW flow to Cooler:

Locally close CCW-375

Locally close CCW-376

6. VERIFY CCW Inleakage Is Isolated:

a. CCW surge tank level - NOT a. Return to Step 1, Page 4.

INCREASING

b. Perform an RCS Water Inventory Balance per STP R-10C

c. Notify CARP before reopening RCV-16 00079821.DOC 02 0325.1036

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PACIFIC GAS AND ELECTRIC COMPANY NUMBER OP AP-11 DIABLO CANYON POWER PLANT REVISION 21 PAGE 11 OF 37 TITLE: Malfunction of Component Cooling Water System UNITS 1 AND 2 SECTION B: CCW SYSTEM INLEAKAGE (Continued)

ACTION/EXPECTED RESPONSE RESPONSE NOT OBTAINED

7. NOTIFY Maintenance Services to institute repair/tube plugging of leaky components

8. RETURN to Procedure and Step in Effect

- END -

00079821.DOC 02 0325.1036

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PACIFIC GAS AND ELECTRIC COMPANY NUMBER OP AP-11 DIABLO CANYON POWER PLANT REVISION 21 PAGE 16 OF 37 TITLE: Malfunction of Component Cooling Water System UNITS 1 AND 2 SECTION E: LOSS OF CCW FLOW TO RCPs SYMPTOMS

1. Thermal barrier and lube oil cooler cooling water return high temperature indication.

2. Possible Main Annunciator Alarms

a. CCW HEADER C (PK01-08)

1) RCP L.O. Clr CCW Flo Lo

2) RCP Thermal Barrier CCW Flo Lo

3) CCW Hdr C Flo Lo

b. RCP No. _____ (PK05-01, 02, 03, 04)

RCP _____ Temp PPC ACTION/EXPECTED RESPONSE RESPONSE NOT OBTAINED CAUTION: IF RCP No. 1 Seal Outlet Temperature exceeds 235°F OR RCP Radial Bearing Temperature exceeds 225°F, DO NOT restore RCP seal cooling.

1. VERIFY CCW Flow To All RCP Lube Oil IF CCW Flow to RCP(s) CANNOT be Coolers: restored to Lube Oil Coolers within 5 minutes,

a. Verify CCW Vlvs - OPEN THEN 1) TRIP reactor.

FCV-355 2) TRIP affected RCP.

FCV-356 3) GO TO EOP E-O, REACTOR TRIP OR SAFETY INJECTION.

FCV-749

FCV-363

b. RCP L.O. Clr CCW Flow LO Alarm (PK01-08) - NOT IN

c. RCP Temp PPC Alarm (PK05-01), 02, 03, 04) - NOT IN 00079821.DOC 02 0325.1036

- - UNCONTROLLED PROCEDURE - DO NOT USE TO PERFORM WORK orISSUE FOR USE ***

PACIFIC GAS AND ELECTRIC COMPANY NUMBER OP AP-11 DIABLO CANYON POWER PLANT REVISION 21 PAGE 17 OF 37 TITLE: Malfunction of Component Cooling Water System UNITS 1 AND 2 SECTION E: LOSS OF CCW FLOW TO RCPs (Continued)

ACTION/EXPECTED RESPONSE RESPONSE NOT OBTAINED

2. VERIFY RCP Seal Injection - IF Both thermal barrier and seal injection flow IN SERVICE: are lost AND CANNOT be immediately restored,

RCP Seal Injection Flow between 8 and THEN 1) Manually TRIP reactor.

13 GPM

2) TRIP affected RCP(s).

RCP Seal No. 1 Outlet Temp - NORMAL 3) GO TO EOP E-O, REACTOR TRIP OR SAFETY INJECTION while

RCP Radial Brg Outlet Temp - NORMAL implementing the next two steps, as applicable.

4) Isolate seal injection to the affected RCP(s) before restarting a charging pump:

Locally close CVCS-8369A,B,C, OR D as appropriate, RCP SEAL INJ WTR (100' Pen Area, GE).

5) If all RCPs are affected, close FCV-357, RCP Thermal Barrier CCW Return Isolation.

3. VERIFY CCW Flow To All RCP Thermal Barriers - NORMAL:

CAUTION: If FCV-357 closed on high flow, do not attempt to open FCV-357 until condition causing high flow is cleared.

a. Verify FCV-357 did not close on high a. GO TO Section B Step 5.b, page 5.

flow

b. Verify Thermal Barrier Return Vlvs FCV-750 and FCV-357 - OPEN

c. Verify RCP Thermal Barrier CCW Flow Lo Alarm (PK01-08) - NOT IN

4. INCREASE Surveillance on RCPs:

Monitor RCP temperatures closely until CCW system can be returned to normal status

- END -

00079821.DOC 02 0325.1036

NUCLEAR POWER GENERATION DIABLO CANYON POWER PLANT JOB PERFORMANCE MEASURE Number: NRCLJC-116

Title:

INITIATE BLEED AND FEED FOR A LOSS OF HEAT SINK Examinee:

Evaluator:

Print Signature Date Results: Sat Unsat Total Time: minutes Comments:

References:

EOP FR-H.1, Response to Loss of Secondary Heat Sink, Rev.19 Alternate Path: Yes No X Time Critical: Yes No X Time Allotment: 15 minutes Critical Steps: 2, 4, 5, 6, 7 Job Designation: RO/SRO Task Number: W/E05/04P/EA1.1 Rating: 4.1/4.0 AUTHOR: JACK BLACKWELL DATE: 01/18/2005 REVIEWED BY: N/A DATE:

TRAINING LEADER APPROVED BY: N/A DATE:

MANAGER - OPERATIONS REV. 01

JPM TITLE: INITIATE BLEED AND FEED FOR A LOSS OF HEAT SINK JPM NUMBER: NRCLJC-116 INSTRUCTOR WORKSHEET Directions: No PLANT controls or equipment are to be operated during the performance of this Job Performance Measure. All actions taken by the examinee should be clearly demonstrated and verbalized to the evaluator. The student will be given the initial conditions, initiating cue, and task standard. The examiner will then ask if any clarifications are needed. The examinee may be given the applicable procedure and step with which to begin.

Required Materials: None Initial Conditions: Unit 1 has experienced a loss of secondary heat sink. EOP FR-H.1 has been implemented and all efforts to establish AFW, MFW, and condensate flow have failed.

Initiating Cue: All steam generator wide range levels are less than 23% and the Shift Foreman directs you to establish and verify RCS bleed and feed using Steps 12 through 18 of EOP FR-H.1.

Task Standard: RCS bleed and feed has been established and verified as required by EOP FR-H.1.

NRCLJC116.DOC PAGE 2 OF 6 REV. 01

JPM TITLE: INITIATE BLEED AND FEED FOR A LOSS OF HEAT SINK JPM NUMBER: NRCLJC-116 INSTRUCTOR WORKSHEET Start Time:

Step Expected Operator Actions

1. Obtain the correct procedure. 1.1 References EOP FR-H.1.

1.2 Reads CAUTION prior to Step 12.

Step was: Sat: ______ Unsat _______*

- 2. Actuate Safety Injection. 2.1 Positions the SAFETY INJECTION ACTUATE switch on CC-2 or VB-1 to ACTUATE. **

Step was: Sat: ______ Unsat _______*

3. Verify RCS feed paths. 3.1 Observes that at least one CCP or one SI pump is running.

3.2 Observes that ECCS valves are in their proper emergency alignment on the VB1 and VB2 mimic.

Step was: Sat: ______ Unsat _______*

- 4. Reset Safety Injection. Note: The 60 second SI timer will have to time out before SI can be reset.

4.1 Depresses the SAFETY INJECTION RESET TRAIN A and TRAIN B pushbuttons. **

4.2 Verifies that SI is reset by observing PK08-22 ON and/or SI Monitor Box red status light OFF.

Step was: Sat: ______ Unsat _______*

- 5. Reset Containment Isolation 5.1 Depresses the CONTMT ISOL Phase A and Phase B. PHASE A RESET pushbuttons. **

5.2 Verifies Phase A red lights are OFF or PK02-01 is OFF.

5.3 Observes that Phase B is NOT actuated or depresses the Phase B RESET pushbuttons.

Step was: Sat: ______ Unsat _______*

Denotes an entry required on the JPM cover sheet.

- Denotes Critical Step and Sub Steps.

NRCLJC116.DOC PAGE 3 OF 6 REV. 01

JPM TITLE: INITIATE BLEED AND FEED FOR A LOSS OF HEAT SINK JPM NUMBER: NRCLJC-116 INSTRUCTOR WORKSHEET Step Expected Operator Actions

- 6. Establish instrument air to 6.1 Opens FCV-584. **

containment.

6.2 Verifies that FCV-584 has opened.

6.3 Observes that instrument air header pressure is > 90 psig on PI-380.

Step was: Sat: ______ Unsat _______*

- 7. Establish RCS bleed path. 7.1 Observes that power is available to the PORV block valves:

8000A 8000B 8000C 7.2 Observes that PORV block valves are already open.

8000A 8000B 8000C 7.3 Opens all PORVs by taking switches to the OPEN position. **

PCV-474 PCV-455C PCV-456 7.4 Verifies all PORVs have opened.

Step was: Sat: ______ Unsat _______*

8. Verify adequate RCS bleed path. 8.1 Observes that at least two PZR PORVs and associated block valves have opened.

Step was: Sat: ______ Unsat _______*

Stop Time:

Total Time: (Enter total time on the cover page)

Denotes an entry required on the JPM cover sheet.

- Denotes Critical Step and Sub Steps.

NRCLJC116.DOC PAGE 4 OF 6 REV. 01

JPM NUMBER: NRCLJC-116 EXAMINEE CUE SHEET Initial Conditions: Unit 1 has experienced a loss of secondary heat sink. EOP FR-H.1 has been implemented and all efforts to establish AFW, MFW, and condensate flow have failed.

Initiating Cue: All steam generator wide range levels are less than 23% and the Shift Foreman directs you to establish and verify RCS bleed and feed using Steps 12 through 18 of EOP FR-H.1.

Task Standard: RCS bleed and feed has been established and verified as required by EOP FR-H.1.

NRCLJC116.DOC PAGE 5 OF 6 REV. 01

JPM TITLE: INITIATE BLEED AND FEED FOR A LOSS OF HEAT SINK JPM NUMBER: NRCLJC-116 ATTACHMENT 1, SIMULATOR SETUP Initialize to JPM IC 716.

This SNAP allows entry into EOP FR-H.1 at Step 12. Steam generator wide range levels are 22% and steam generator pressures are at 1005 psig with the 10% steam dumps in AUTO at 8.38 turns.

Perform the following:

1. Display the E-0 screen on SPDS panel A.

2. Display the CSF-3 screen on SPDS panel B.

Inform the examiner that the simulator setup is complete.

Go to RUN when the examinee is given the cue sheet.

NRCLJC116.DOC PAGE 6 OF 6 REV. 01

PACIFIC GAS AND ELECTRIC COMPANY NUMBER EOP FR-H.1 DIABLO CANYON POWER PLANT REVISION 19 PAGE 13 OF 28 TITLE: Response to Loss of Secondary Heat Sink UNIT 1 ACTION/EXPECTED RESPONSE RESPONSE NOT OBTAINED CAUTION: Steps 12 through 18 must be performed without delay in order to establish RCS heat removal by RCS bleed and feed.

12. ACTUATE SI

13. VERIFY RCS Feed Paths: Manually start ECCS Pps and align ECCS Injection Valves to establish

a. Check ECCS Pp status: RCS feed path.

o CCP - AT LEAST ONE IF An RCS feed path CANNOT RUNNING be established, THEN Activate the monitor lights for OR monitor light Box C by turning the Monitor Test Light Switch o SI Pps - AT LEAST ONE to ON.

RUNNING Use White Status lights to verify ECCS

b. Verify ECCS valve alignment - valve alignment.

PROPER EMERGENCY ALIGNMENT IF An RCS feed path CANNOT be established, THEN Continue attempts to establish RCS feed flow AND RETURN TO Step 4 (Page 3).

PACIFIC GAS AND ELECTRIC COMPANY NUMBER EOP FR-H.1 DIABLO CANYON POWER PLANT REVISION 19 PAGE 14 OF 28 TITLE: Response to Loss of Secondary Heat Sink UNIT 1 ACTION/EXPECTED RESPONSE RESPONSE NOT OBTAINED

14. RESET SI

15. RESET Containment Isolation Phase A And Phase B

16. ESTABLISH Instrument Air To Containment:

a. Open FCV-584

b. Check Instrument Air Header b. IMPLEMENT OP AP-9, LOSS OF Pressure GREATER THAN INSTRUMENT AIR.

90 PSIG, PI-380 (VB4 UNIT 1)

17. ESTABLISH RCS Bleed Path:

a. Verify PZR PORV Block Vlvs - a. Restore power to block valves AND OPEN OPEN:

o 8000A for PCV-474 8000A: 52-1F-40 AND 52-1F-40R o 8000B for PCV-455C 8000B: 52-1G-46 AND 52-1G-46R o 8000C for PCV-456 8000C: 52-1H-33 AND 52-1H-33R

b. Open all PZR PORVs

PACIFIC GAS AND ELECTRIC COMPANY NUMBER EOP FR-H.1 DIABLO CANYON POWER PLANT REVISION 19 PAGE 15 OF 28 TITLE: Response to Loss of Secondary Heat Sink UNIT 1 ACTION/EXPECTED RESPONSE RESPONSE NOT OBTAINED

18. VERIFY Adequate RCS Bleed Path:

CAUTION: The second off head vent valves may open spuriously if the first off valve is opened first.

a. Verify PZR PORVs and associated a. Perform the following:

Block Vlvs - AT LEAST TWO OPEN 1) Open Reactor Vessel Head Vents:

(a) 8078A & D (PAM 1)

(b) 8078B & C (PAM 1)

2) Align any available water source to the S/Gs.

o Main Feed or Condensate. Refer to Step 7 (Page 5) or Step 9 (Page 9).

OR o Any low pressure water source.

Refer to EOP FR-C.1, RESPONSE TO INADEQUATE CORE COOLING, Appendix F, Step 6, for guidance IF No water source can be aligned, THEN GO TO Step 19 (Next Page)

3) IF A low pressure water source is aligned, THEN Depressurize at least one intact S/G to atmospheric pressure using 10% Steam Dump to inject water source. S/G with highest indicated level is preferred.

NUCLEAR POWER GENERATION DIABLO CANYON POWER PLANT JOB PERFORMANCE MEASURE Number: NRCLJC-123

Title:

ALIGN SAFETY INJECTION PUMP 11 FOR HOT LEG RECIRCULATION Examinee:

Evaluator:

Print Signature Date Results: Sat Unsat Total Time: minutes Comments:

References:

EOP E-1.4, Transfer to Hot Leg Recirculation, Rev. 15 Alternate Path: Yes X No Time Critical: Yes No X Time Allotment: 15 minutes Critical Steps: 2,3,4 Job Designation: RO/SRO Task Number: 006/02/A4.05 Rating: 3.9/3.8 AUTHOR: JACK BLACKWELL DATE: 01/18/2005 REVIEWED BY: N/A DATE:

TRAINING LEADER APPROVED BY: N/A DATE:

MANAGER - OPERATIONS REV. 01

JPM TITLE: ALIGN SAFETY INJECTION PUMP 11 FOR HOT LEG JPM NUMBER: NRCLJC-123 RECIRCULATION INSTRUCTOR WORKSHEET Directions: No PLANT controls or equipment are to be operated during the performance of this Job Performance Measure. All actions taken by the examinee should be clearly demonstrated and verbalized to the evaluator. The student will be given the initial conditions, initiating cue, and task standard. The examiner will then ask if any clarifications are needed. The examinee may be given the applicable procedure and step with which to begin.

Required Materials: None Initial Conditions: A Unit 1 Reactor Trip and Safety Injection has occurred due to a LOCA.

Cold leg recirculation was initiated 10.5 hours5.787037e-5 days 0.00139 hours 8.267196e-6 weeks 1.9025e-6 months ago. Preparation for hot leg recirculation per EOP E-1.4, Step 1, is complete.

Initiating Cue: The Shift Foreman directs you to align Safety Injection pump 1-1 for hot leg recirculation per EOP E-1.4, Step 2.

Task Standard: Safety Injection is aligned for hot leg recirculation in accordance with EOP E-1.4.

NRCLJC123.DOC PAGE 2 OF 6 REV. 01

JPM TITLE: ALIGN SAFETY INJECTION PUMP 11 FOR HOT LEG JPM NUMBER: NRCLJC-123 RECIRCULATION INSTRUCTOR WORKSHEET Start Time:

Step Expected Operator Actions

1. Obtain the correct procedure. 1.1 References EOP E-1.4 Step was: Sat: ______ Unsat _______*

- 2. Align SI Pump 1-1 for HL Recirc. 2.1 Observes that RHR pump 11 is NOT running.

2.2 Closes 8804A.

2.3 Verifies SIP 1-1 is stopped.

2.4 Closes 8821A.

2.5 Opens 8802A.

2.6 Start SIP 1-1 but trips after start.

Step was: Sat: ______ Unsat _______*

- 3. Align SI Pump 1-2 for HL Recirc. 3.1 Checks RHRP 1-2 running.

3.2 Verify 8804B Open.

3.3 CUTIN 8809B Series Contractor.

3.4 Close 8809B.

3.5 Verify 9003B Closed.

3.6 Verify SIP 1-2 stopped.

3.7 Close 8821B.

3.8 Close 8835.

3.9 Open 8802B.

Step was: Sat: ______ Unsat _______*

Denotes an entry required on the JPM cover sheet.

- Denotes Critical Step and Sub Steps.

NRCLJC123.DOC PAGE 3 OF 6 REV. 01

JPM TITLE: ALIGN SAFETY INJECTION PUMP 11 FOR HOT LEG JPM NUMBER: NRCLJC-123 RECIRCULATION INSTRUCTOR WORKSHEET Step Expected Operator Actions

- 4. Start SI Pump 1-2. 4.1 Start SIP 1-2.

4.2 Verify RHR 2 < 57 amps.

4.3 Verify SIP 2 discharge flow on FI-922.

Step was: Sat: ______ Unsat _____*

Stop Time:

Total Time: (Enter total time on the cover page)

Denotes an entry required on the JPM cover sheet.

- Denotes Critical Step and Sub Steps.

NRCLJC123.DOC PAGE 4 OF 6 REV. 01

JPM NUMBER: NRCLJC-123 EXAMINEE CUE SHEET Initial Conditions: A Unit 1 Reactor Trip and Safety Injection has occurred due to a LOCA.

Cold leg recirculation was initiated 10.5 hours5.787037e-5 days 0.00139 hours 8.267196e-6 weeks 1.9025e-6 months ago. Preparation for hot leg recirculation per EOP E-1.4, Step 1, is complete.

Initiating Cue: The Shift Foreman directs you to align Safety Injection pump 1-1 for hot leg recirculation per EOP E-1.4, Step 2.

Task Standard: Safety Injection is aligned for hot leg recirculation in accordance with EOP E-1.4.

NRCLJC123.DOC PAGE 5 OF 6 REV. 01

JPM TITLE: ALIGN SAFETY INJECTION PUMP 11 FOR HOT LEG JPM NUMBER: NRCLJC-123 RECIRCULATION ATTACHMENT 1, SIMULATOR SETUP Initialize to JPM IC 780.

Load Drill File 6302, or manually input the following:

pmp sis1 6,8,0,0,c,xv1o240r, (SIP 1-1 trip on overcurrent when pump red light is on)

This SNAP allows entry into EOP E-1.4 at Step 2.

Hang control board CAUTION tags on 8105 and 8106.

Inform the examiner that the simulator setup is complete.

Go to RUN when the examinee is given the cue sheet.

NRCLJC123.DOC PAGE 6 OF 6 REV. 01

- - ISSUED FOR USE BY:_______________________DATE:_____________EXPIRES:_______________***

- - UNCONTROLLED PROCEDURE-DO NOT USETO PERFORM WORK orISSUEFOR USE ***

PACIFIC GAS AND ELECTRIC COMPANY NUMBER EOP E-1.4 NUCLEAR POWER GENERATION REVISION 15 DIABLO CANYON POWER PLANT PAGE 1 OF 11 EMERGENCY OPERATING PROCEDURE UNIT TITLE: TRANSFER TO HOT LEG RECIRCULATION 1 12/01/98 EFFECTIVE DATE PROCEDURE CLASSIFICATION: QUALITY RELATED 1.0 SCOPE 1.1 This procedure provides the necessary instructions for transferring the safety injection system to hot leg recirculation.

1.2 The major actions in EOP E-1.4 are:

o Align the RHR flow path for hot leg recirculation, o Align the SI Pp flow path for hot leg recirculation, o Separate the CCW trains if directed by the TSC.

2.0 VERIFY ENTRY CONDITION FOR EOP E-1.4 2.1 EOP E-1, Step 18

- - UNCONTROLLED PROCEDURE-DO NOT USETO PERFORM WORK orISSUEFOR USE ***

PACIFIC GAS AND ELECTRIC COMPANY NUMBER EOP E-1.4 DIABLO CANYON POWER PLANT REVISION 15 PAGE 2 OF 11 TITLE: TRANSFER TO HOT LEG RECIRCULATION UNIT 1 ACTION/EXPECTED RESPONSE RESPONSE NOT OBTAINED NOTE: It is important during this phase that two separate and redundant trains of recirculation outside containment are established unless an inoperable 4 KV vital bus prevents total separation.

1. PREPARE For Hot Leg Recirculation 10 hours1.157407e-4 days 0.00278 hours 1.653439e-5 weeks 3.805e-6 months After Event Initiation:

a. Check the following control a. Place the Valve Control Switches in switches in their required position: the required position.

o 8802A - CLOSED, SI to Hot ------------------------------

Legs 1 & 2 o 8835 - OPEN, SI Pp to Cold Leg o 8703 - CLOSED, RHR to Hot Legs 1 & 2 o 8802B - CLOSED, SI to Hot Legs 3 & 4

b. Close the following 480V breakers:

o 52-1F-48, 8802A o 52-1G-24, 8835 o 52-1G-56, 8703 o 52-1G-56R, 8703 o 52-1H-26, 8802B

- - UNCONTROLLED PROCEDURE-DO NOT USETO PERFORM WORK orISSUEFOR USE ***

PACIFIC GAS AND ELECTRIC COMPANY NUMBER EOP E-1.4 DIABLO CANYON POWER PLANT REVISION 15 PAGE 3 OF 11 TITLE: TRANSFER TO HOT LEG RECIRCULATION UNIT 1 ACTION/EXPECTED RESPONSE RESPONSE NOT OBTAINED

2. At 10.5 hours5.787037e-5 days 0.00139 hours 8.267196e-6 weeks 1.9025e-6 months ALIGN SI Pp 1 For Hot Leg Recirculation:

a. Verify both RHR Pps are running a. Manually Start any RHR Pp NOT running.

IF RHR Pp 1 is NOT Operable, THEN Close 8804A AND GO TO Step 2f.

IF RHR Pp 2 is NOT Operable, THEN Continue with Step 2b.

b. Verify 8804A, RHR Hx No. 1 to Chg and SI Pps Suction - OPEN

c. Cutin 8809A series contactor toggle switch

d. Close 8809A, RHR to Cold Legs 1 and 2

e. Verify Closed 9003A, RHR Pp 1 to Spray Hdr A - CLOSED

f. Verify SI Pp 1 - STOPPED

g. Close 8821A, SI Pp No. 1 Disch Crosstie Vlv

h. Open 8802A, SI to Hot Legs 1 and 2

i. Perform the following i. IF SI Pp 1 is NOT Operable, THEN GO TO Step 2k (Next

1) Start SI Pp 1 Page).

2) Verify operating RHR Pp ------------------------------

motor current LESS THAN 57 AMPS THIS STEP CONTINUED ON NEXT PAGE

- - UNCONTROLLED PROCEDURE-DO NOT USETO PERFORM WORK orISSUEFOR USE ***

PACIFIC GAS AND ELECTRIC COMPANY NUMBER EOP E-1.4 DIABLO CANYON POWER PLANT REVISION 15 PAGE 4 OF 11 TITLE: TRANSFER TO HOT LEG RECIRCULATION UNIT 1 ACTION/EXPECTED RESPONSE RESPONSE NOT OBTAINED

2. At 10.5 hours5.787037e-5 days 0.00139 hours 8.267196e-6 weeks 1.9025e-6 months ALIGN SI Pp 1 For Hot Leg Recirculation (Continued):

j. Check for SI Pp 1 Disch Flow on j. RETURN TO Step 1 (Page 2),

FI-918 AND reverify system lineup downstream of RHR Pp 1.

k. Verify Both RHR Pps - k. GO TO Step 3 (Next Page).

RUNNING

l. Close 8923A, SI Pp 1 RWST Suction

- - UNCONTROLLED PROCEDURE-DO NOT USETO PERFORM WORK orISSUEFOR USE ***

PACIFIC GAS AND ELECTRIC COMPANY NUMBER EOP E-1.4 DIABLO CANYON POWER PLANT REVISION 15 PAGE 5 OF 11 TITLE: TRANSFER TO HOT LEG RECIRCULATION UNIT 1 ACTION/EXPECTED RESPONSE RESPONSE NOT OBTAINED

3. ALIGN SI Pp 2 for Hot Leg Recirculation:

a. Check RHR Pp 2 - RUNNING a. IF RHR Pp No. 2 is NOT running, THEN Close 8804B RHR to SI Pp 2 AND GO TO Step 3f.

b. Verify 8804B, RHR to SI Pp No. 2 Suction Vlv - OPEN

c. Cutin series contactor toggle switch for 8809B

d. Close 8809B, RHR to Cold Legs 3 and 4

e. Verify 9003B, RHR Pp 2 to Spray Hdr B - CLOSED

f. Verify SI Pp 2 - STOPPED

g. Close 8821B, SI Pp No. 2 Disch Crosstie Vlv

h. Close 8835, SI to Cold Legs Vlv

i. Open 8802B, SI to Hot Legs 3 and 4

j. Perform the following: j. IF SI Pp 2 is NOT operable, THEN GO TO Step 3l.

1) Start SI Pp 2

2) Verify operating RHR Pp motor current LESS THAN 57 AMPS.

k. Verify SI Pp 2 Disch Flow on k. RETURN TO Step 1 (Page 2),

FI-922 AND Reverify system lineup downstream of RHR Pp 2.

l. Verify Both RHR Pps - RUNNING l. GO TO Step 4 (Next Page).

m. Close 8923B, SI Pp 2 RWST Suction

- - UNCONTROLLED PROCEDURE-DO NOT USETO PERFORM WORK orISSUEFOR USE ***

PACIFIC GAS AND ELECTRIC COMPANY NUMBER EOP E-1.4 DIABLO CANYON POWER PLANT REVISION 15 PAGE 6 OF 11 TITLE: TRANSFER TO HOT LEG RECIRCULATION UNIT 1 ACTION/EXPECTED RESPONSE RESPONSE NOT OBTAINED

4. CHECK RHR Pp 2 Is Running IF RHR Pp 2 is NOT operable, THEN GO TO Step 6 (Next Page).

5. ALIGN RHR Pp 2 For Hot Leg Recirculation:

a. Open 8716B, RHR Pp 2 Disch a. Perform the following:

Crosstie Vlv

1) Open 8809B

2) Close 8716B, RHR Pp 2 Disch Crosstie Vlv AND GO TO Step 6 (Next Page). Maintain RHR Pp current between 50 AMPS and 57 AMPS.

b. Open 8703, RHR to Hot Legs 1 b. Perform the following:

and 2

1) Open 8809A AND B

2) Close 8716B, RHR Pp 2 Disch Crosstie Vlv AND GO TO Step 7 (Next Page). Maintain RHR Pp current between 50 AMPS and 57 AMPS.

c. Adjust HCV-637, RHR Hx 2 Outlet Flow Control Vlv to maintain suction to SI Pps AND RHR Pp 2 motor current between 50 AMPS and 57 AMPS

d. GO TO Step 7 (Next Page)

- - UNCONTROLLED PROCEDURE-DO NOT USETO PERFORM WORK orISSUEFOR USE ***

PACIFIC GAS AND ELECTRIC COMPANY NUMBER EOP E-1.4 DIABLO CANYON POWER PLANT REVISION 15 PAGE 7 OF 11 TITLE: TRANSFER TO HOT LEG RECIRCULATION UNIT 1 ACTION/EXPECTED RESPONSE RESPONSE NOT OBTAINED

6. ALIGN RHR Pp 1 For Hot Leg Perform the following:

Recirculation:

1) Open 8809A

a. Open 8716A, RHR Pp 1 Disch Crosstie Vlv 2) Close 8716A, RHR Pp 1 Disch Crosstie Vlv AND GO TO Step 7.

b. Open 8703, RHR to Hot Legs 1 Maintain RHR Pp current between and 2 50 AMPS and 57 AMPS.

c. Adjust HCV-638 to maintain ------------------------------

suction to SI Pps AND RHR Pp 1 motor current between 50 AMPS and 57 AMPS NOTE: The Technical Support Center shall determine train separation requirements within 24 hours2.777778e-4 days 0.00667 hours 3.968254e-5 weeks 9.132e-6 months of event initiation.

7. CONTACT Plant Engineering In Technical Support Center To Evaluate CCW System Train Separation

- - UNCONTROLLED PROCEDURE-DO NOT USETO PERFORM WORK orISSUEFOR USE ***

PACIFIC GAS AND ELECTRIC COMPANY NUMBER EOP E-1.4 DIABLO CANYON POWER PLANT REVISION 15 PAGE 8 OF 11 TITLE: TRANSFER TO HOT LEG RECIRCULATION UNIT 1 ACTION/EXPECTED RESPONSE RESPONSE NOT OBTAINED

8. Technical Support Center Directs GO TO Step 9 (Next Page)

CCW Train Separation:

a. Verify at least two CCW Pps - a. GO TO Step 9 (Next Page).

RUNNING

b. Open FCV-430 or 431, Idle CCW Hx Outlet Stop Vlv

c. Check CCW Pp 1 - RUNNING c. CCW Pp 2 AND CCW Pp 3 are running

1) Open FCV-355, CCW Header C Isol Vlv.

2) Locally Close CCW-19, CCW Pp 2 Discharge Vlv to CCW Header A.

3) Locally Close CCW-17, CCW Pp 3 Discharge Vlv to CCW Header B.

4) Locally Close CCW-23 CCW Header A to C Isol Vlv.

5) Locally Close CCW-5, Suction Header Crosstie Vlv between CCW Headers A and C.

6) GO TO Step 9 (Next Page).

d. Open FCV-355, CCW Header C Isol Vlv

e. Locally close CCW-18, CCW Pp 1 Discharge Vlv to CCW Header A

f. Locally Close CCW-16, CCW Pp 2 Discharge Vlv to CCW Header B

g. Locally Close CCW-17, CCW Pp 3 Discharge Vlv to CCW Header B

h. Locally Close CCW-24, CCW Header B to C Isol Vlv

i. Locally Close CCW-4, Suction Header Crosstie Vlv between CCW Headers B and C

- - UNCONTROLLED PROCEDURE-DO NOT USETO PERFORM WORK orISSUEFOR USE ***

PACIFIC GAS AND ELECTRIC COMPANY NUMBER EOP E-1.4 DIABLO CANYON POWER PLANT REVISION 15 PAGE 9 OF 11 TITLE: TRANSFER TO HOT LEG RECIRCULATION UNIT 1 ACTION/EXPECTED RESPONSE RESPONSE NOT OBTAINED

9. EVALUATE Long Term Plant Status:

a. Maintain Cold Shutdown conditions:

o RCS Temperature LESS THAN 200°F o Keff LESS THAN .99

b. Contact the Chemistry Dept to obtain the following samples:

1) Reactor Coolant System to assess o RCS activity o Fuel damage o Hydrogen concentration

2) Recirculation Sump to determine o Boron Concentration o PH

3) Sample Containment atmosphere:

(a) Hydrogen concentration (a) Consult Plant Engineering LESS THAN 3.5% Staff (TSC) for additional recovery action with potential explosive HYDROGEN/AIR mixture in containment AND GO TO Step 9c.

(b) Hydrogen concentration (b) IMPLEMENT OP H-9, LESS THAN 0.5% INSIDE CONTAINMENT H2 RECOMBINATION SYSTEM, to reduce Hydrogen Concentration.

c. Consult Plant Engineering Staff in Technical Support Center for additional guidance on long term action

- END -

- - UNCONTROLLED PROCEDURE-DO NOT USETO PERFORM WORK orISSUEFOR USE ***

PACIFIC GAS AND ELECTRIC COMPANY NUMBER EOP E-1.4 DIABLO CANYON POWER PLANT REVISION 15 PAGE 10 OF 11 TITLE: TRANSFER TO HOT LEG RECIRCULATION UNIT 1 3.0 APPENDICES 3.1 Appendix A, Blackout Emergency Loading of Vital Buses 4.0 ATTACHMENTS 4.1 "Foldout Page for EOP E-1.4," 2/98 5.0 SPONSOR Steve Derks

NUCLEAR POWER GENERATION DIABLO CANYON POWER PLANT JOB PERFORMANCE MEASURE Number: NRCLJC-301

Title:

Dilution without Makeup Control Operable Examinee:

Evaluator:

Print Signature Date Results: Sat Unsat Total Time: minutes Comments:

References:

OP B-1A:VII, CVCS - Makeup Control System Operation, Rev. 33 Alternate Path: Yes X No Time Critical: Yes No X Time Allotment: 20 minutes Critical Steps: 4 Job Designation: RO K/A

Reference:

004/01/A2.25 RO/SRO Rating: 3.8 AUTHOR: JACK BLACKWELL DATE: 01/18/2005 APPROVED BY: N/A DATE:

TRAINING LEADER APPROVED BY: N/A DATE:

LINE MANAGER REV. 1

JPM TITLE: DILUTION WITHOUT MAKEUP CONTROL OPERABLE JPM NUMBER: NRCLJC-301 INSTRUCTOR WORKSHEET Directions: No plant controls or equipment are to be operated during the performance of this Job Performance Measure. All actions taken by the examinee should be clearly demonstrated and verbalized to the evaluator. The student will be given the initial conditions, initiating cue, and task standard. The examiner will then ask if any clarifications are needed. After identifying the appropriate procedure for the task, the examinee may be given the procedure and told the step with which to begin.

Required Materials: (Required materials here)

Initial Conditions: Unit 1 was at 100% power when a runback occurred to 75%. The condition causing the runback has cleared, approvals have been obtained, and the crew is preparing to ramp back to 100%.

Initiating Cue: The shift foreman has directed you to dilute 200 gallons to compensate for Xenon and in preparation for the ramp up in power.

Task Standard: A dilution is completed per procedure.

NRCLJC-301.DOC Page 2 of 6 REV. 1

JPM TITLE: Dilution without Makeup Control Operable JPM NUMBER: NRCLJC-301 INSTRUCTOR WORKSHEET Start Time:

Step Expected Operator Actions

1. Obtains procedure. 1.1 Obtains a copy of OP B-1A:VII.

1.2 Determines Section 6.2 is appropriate.

Step was: Sat: ______ Unsat _______*

2. Performs Dilution With Makeup 2.1 Places 1/MU to STOP.

System In Automatic per Section 6.2 2.2 Places 43/MU in DILUTE.

2.3 Verifies 200 gal. set in primary water integrator using BATCH function.

2.4 Enables integrator.

2.5 Selects SUM on YIC-111.

2.6 Takes 1/MU to START

- 2.7 Determines 1/MU will NOT START Step was: Sat: ______ Unsat _______*

NOTE: May enter OP AP-19 and complete steps there. When notifiying TM to troubleshoot, then perform CUE below.

3. Notify SFM that Makeup is 3.1 Notifies SFM makeup will not start inoperable. and is inoperable.

Cue: SFM directs you to use section 6.9 of the procedure to dilute.

Step was: Sat: ______ Unsat _______*

Denotes an entry required on the JPM cover sheet.

- Denotes a Critical Step.

NRCLJC-301.DOC Page 3 of 6 REV. 1

JPM TITLE: Dilution without Makeup Control Operable JPM NUMBER: NRCLJC-301 INSTRUCTOR WORKSHEET Step Expected Operator Actions

- 4. Peform Dilution per Section 6.9 4.1 Select and Verify CLOSED:

FCV-111A/B

FCV-110A/B 4.2 Reads CAUTION 4.3 Verify intergrators still set for dilution.

4.4 Places HC-111 in MANUAL and adjusts as necessary.

4.5 OPEN FCV-111A.

4.5 OPEN FCV-111B and confirm flow.

4.6 CLOSE FCV-111B when integrator count complete. (200 gal + 20) 4.7 CLOSE FCV-111A Step was: Sat: ______ Unsat _______*

Stop Time:

Total Time: (Enter total time on the cover page)

Denotes an entry required on the JPM cover sheet.

- Denotes a Critical Step.

NRCLJC-301.DOC Page 4 of 6 REV. 1

JPM NUMBER: NRCLJC-301 EXAMINEE CUE SHEET Initial Conditions: Unit 1 was at 100% power when a runback occurred to 75%. The condition causing the runback has cleared, approvals have been obtained, and the crew is preparing to ramp back to 100%.

Initiating Cue: The shift foreman has directed you to dilute 200 gallons to compensate for Xenon and in preparation for the ramp up in power.

Task Standard: A dilution is completed per procedure.

NRCLJC-301.DOC Page 5 of 6 REV. 1

JPM TITLE: DILUTION WITHOUT MAKEUP CONTROL OPERABLE JPM NUMBER: NRCLJC-301 ATTACHMENT 1, SIMULATOR SETUP Initialize the simulator to IC-511 (75%, MOL).

1. Load drill file 6301 into the the following path:

o T:\simtrn\cmd_file

2. Enter the following on the Expert Screen:

o tc xc2i031b,file drl_6301.txt OR Enter drill file 6301 when 43/MU taken to Dilute, or manually insert the following when 43/MU taken to Dilute:

Command Description ovr xc2i031a act,0,0,0,d,0 #cc2010a Fails 43/MU to OFF ovr xc2i031e act,0,0,0,d,0 #cc2010b ovr xc2i031c act,0,0,0,d,0 #cc2010c ovr xc2i031c act,0,0,0,d,0 #cc2010c ovr xc2i031b act,0,0,0,d,0 #cc2010d ovr xc2i031d act,0,0,0,d,0 #cc2010e ovr xc2i031f act,1,0,0,d,0 #cc2010f Perform the following:

1. None Inform the examiner that the simulator setup is complete.

Go to RUN when the examinee is given the cue sheet.

ON COMPLETION OF JPM: Reset Integrators to 10 gal boration and 20 gal dilution NRCLJC-301.DOC Page 6 of 6 REV. 1

- - UNCONTROLLED PROCEDURE - DO NOT USE TO PERFORM WORK or ISSUE FOR USE ***

PACIFIC GAS AND ELECTRIC COMPANY NUMBER OP B-1A:VII NUCLEAR POWER GENERATION REVISION 33A DIABLO CANYON POWER PLANT PAGE 1 OF 39 OPERATING PROCEDURE UNIT TITLE: CVCS - Makeup Control System Operation 112/03/04 EFFECTIVE DATE PROCEDURE CLASSIFICATION: QUALITY RELATED

1. SCOPE 1.1 This procedure provides instructions for the various modes of operating the Makeup Control System. It also includes instructions for Deborator 1-2 operation at End-of-Life.

1.2 This procedure has been re-written, therefore no revision bars have been included.

2. DISCUSSION 2.1 Sections 6.2 and 6.3 contain the instructions for dilution and boration with the makeup system aligned for automatic operation, respectively. Checklist style forms are also provided in Attachments 9.1 and 9.2 which contain the same essential actions as these sections and can be used as stand alone instructions.

2.2 Attachment 9.3 is a summary of the functions of the boric acid and primary water integrators replaced during 1R12.

2.3 The specific instructions included in this procedure are as follows:

2.3.1 Section 6.1 - Place in Automatic 2.3.2 Section 6.2 - Dilution With Makeup System in Automatic 2.3.3 Section 6.3 - Boration With Makeup System in Automatic 2.3.4 Section 6.4 - Continuous Dilution at Adjustable Flowrates 2.3.5 Section 6.5 - Continuous Boration at Adjustable Flowrates 2.3.6 Section 6.6 - Dilute/Alternate Dilute 2.3.7 Section 6.7 - Borate 2.3.8 Section 6.8 - Manual Operation 2.3.9 Section 6.9 - Manual Operation With Makeup Control System Inoperable 2.3.10 Section 6.10 - Makeup to the RWST 2.3.11 Section 6.11 - Deborator Operation 2.3.12 Section 6.12 - Flush to an LHUT 2.3.13 Section 6.13 - Emergency Boration using CVCS-1-8104.

2.3.14 Section 6.14 - Manual Operation While on Excess Letdown 3.

00122533.DOC 02 1203.0704

- - UNCONTROLLED PROCEDURE - DO NOT USE TO PERFORM WORK or ISSUE FOR USE ***

PACIFIC GAS AND ELECTRIC COMPANY NUMBER OP B-1A:VII DIABLO CANYON POWER PLANT REVISION 33A PAGE 2 OF 39 TITLE: CVCS - Makeup Control System Operation UNIT 1 RESPONSIBILITIES 3.1 SFM is responsible for operation of equipment as described in this procedure.

3.2 Chemistry is responsible for sampling, as required.

4. PREREQUISITES 4.1 Applicable portions of OP B-1A:IX , "CVCS - Alignment Verification For Plant Startup" have been completed.

4.2 The 4% Boric Acid System is in normal operation.

4.3 Primary Water System is in normal operation.

4.4 A charging pump (either reciprocating or centrifugal) is in service.

4.5 If performing section 6.10, a sealed component change form has been prepared for CVCS-1-8428.

5. PRECAUTIONS AND LIMITATIONS 5.1 When boric acid is supplied to the charging system, the flow shall be routed through FCV-110B directly to the charging pump suction. This is necessary since the flowpath to the VCT (FCV-111B) is only partially heat traced. Routing boric acid through this flowpath can result in crystallization of the boric acid and plugging of the volume control tank spray nozzle. If on Excess Letdown, boric acid in concentrations RCS boron may be directed through FCV-111B per section 6.14 of this procedure.

5.2 When volume control tank pressure is increased above the normal range due to level increases, volume control tank pressure and reactor coolant pump No. 1 seal leakoff flows should be monitored. The VCT should be vented to the vent header continuously during significant level increases by opening CVCS-1-8101, "VCT to Vent Hdr Isol Vlv."

5.3 Prior to operations requiring large amounts of makeup, the Gaseous Radwaste System should be verified in service with a Waste Gas Compressor operating.

5.4 Equalization of boron concentration between the RCS and the Pressurizer shall be initiated for major boration/dilution evolutions (i.e. greater than 50 PPM change). The difference in boron concentration between the RCS and Pressurizer should not exceed 50 PPM.

5.5 Towards EOL, when the automatic makeup mode is selected, consideration should be given for maintaining the control switch for FCV-110A in the closed position.

5.6 Review the following Technical Specifications/ECG items:

5.6.1 ECG 8.8 Borated Water Sources - Shutdown 5.6.2 ECG 8.9 Borated Water Sources - Operating 5.6.3 T.S. 3.5.4 Refueling Water Storage Tank 00122533.DOC 02 1203.0704

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PACIFIC GAS AND ELECTRIC COMPANY NUMBER OP B-1A:VII DIABLO CANYON POWER PLANT REVISION 33A PAGE 3 OF 39 TITLE: CVCS - Makeup Control System Operation UNIT 1 5.7 The chemistry technician should be notified whenever RCS makeup addition

> 2000 gal/day is anticipated.

5.8 Following major (greater than 50 PPM change) dilution/boration evolutions, the actual RCS boron concentration should be determined by sampling and compared to the expected value prior to any additional boration or dilution evolutions.

5.9 When changing the setting of the batch integrators, the proper setting should be verified prior to initiating boration or dilution.

5.10 When operating with Hagan controllers (HC-110 and HC-111) in manual, CVCS make up deviation alarm and subsequent termination of the selected operation may occur if the potentiometer on the selected Hagan controller is not adjusted to the actual flow achieved, as set by the manual push buttons (30 second time delay). The tolerance is +/- 0.8 gpm for boric acid flow and +/- 5.0 gpm for primary flow.

5.11 The boric acid integrator will stop counting if the flowrate falls below 0.4 gpm and the primary water integrator will stop counting if the flowrate falls below 2.0 gpm. This prevents spurious counts if the flow transmitter output does not fall to exactly zero when the system is shutdown.

5.12 One or two extra gallons of boric acid may be added at the end of borations due to the time it takes for FCV-110A to stroke closed.

5.13 Two or three extra gallons of water may be added at the end of dilutions due to the time it takes for FCV-110B to stroke closed.

5.14 Sections of this procedure will prevent automatic makeup to the VCT during the evolution. VCT level should be monitored closely to ensure expected response is obtained.

5.15 Some sections of this procedure require manual operation of controllers or control switches. Failure to restore the system to normal following these evolutions may result in unplanned reactivity additions.

5.16 Consider potential changes in reactivity that could occur due to actions taken in this procedure, and perform a reactivity brief if required by the Reactivity Management Program.

6.

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PACIFIC GAS AND ELECTRIC COMPANY NUMBER OP B-1A:VII DIABLO CANYON POWER PLANT REVISION 33A PAGE 4 OF 39 TITLE: CVCS - Makeup Control System Operation UNIT 1 INSTRUCTIONS 6.1 PLACE MAKEUP CONTROL SYSTEM IN AUTOMATIC 6.1.1 Verify the makeup mode selector switch (43/MU) in the "OFF" position.

6.1.2 Verify the following control switches in the "AUTO" position:

FCV-110A

FCV-110B

FCV-111A

FCV-111B 6.1.3 Check open FCV-110A.

6.1.4 Check closed the following valves:

FCV-110B

FCV-111A

FCV-111B NOTE: With HC-111 in "AUTO" and 43/MU in "AUTO", a reference setting between 70 and 120 gpm is preset into the controller. The reference setting is variable and is posted on a lamicoid on HC-111. The 10 turn pot is inoperable in this mode.

6.1.5 Verify primary water blend controller (HC-111) in "AUTO".

6.1.6 Verify boric acid blend controller (HC-110) in "AUTO".

6.1.7 Determine the required boric acid flowrate (0-35 gpm) from the boration spreadsheet program to give a blend equal to the existing reactor coolant system boron concentration. Ensure that the current reference flowrate for primary water is used in the program.

NOTE: If the RCS concentration is higher than the makeup system can supply in automatic, then HC-110 should be set to maximum.

6.1.8 Set the 10 turn pot on HC-110 to the required position.

6.1.9 Place the makeup mode selector switch (43/MU) in the "AUTO" position.

6.1.10 Reset the primary water batch integrator as follows (if desired):

a. Depress the "RESET" key.

b. Depress the "START" key.

6.1.11 Reset the boric acid batch integrator as follows (if desired):

a. Depress the "RESET" key.

b. Depress the "START" key.

NOTE: Once the makeup control switch is placed in "START", automatic makeup should start at 14% VCT level (LI-112) and terminate at 24%.

6.1.12 Turn the makeup control switch (1/MU) to the "START" position.

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PACIFIC GAS AND ELECTRIC COMPANY NUMBER OP B-1A:VII DIABLO CANYON POWER PLANT REVISION 33A PAGE 5 OF 39 TITLE: CVCS - Makeup Control System Operation UNIT 1 6.2 DILUTION WITH MAKEUP SYSTEM IN AUTOMATIC NOTE 1: The following instructions apply to dilution with the makeup system aligned for automatic operation per section 6.1.

NOTE 2: A checklist style form is provided in Attachment 9.1 which contains the same essential actions of this entire section and can be used as a stand alone instruction.

6.2.1 Place the makeup control switch (1/MU) in the "STOP" position.

6.2.2 Place the makeup mode selector switch (43/MU) in the "DILUTE" position.

6.2.3 If not already set for the proper quantity, enter the required number of gallons in the primary water integrator using the BATCH function and data entry keys.

(Refer to Attachment 9.3) 6.2.4 Enable the integrator as follows, if required:

a. Press "RESET" key.

b. Press the "START" key.

6.2.5 Select "SUM" on YIC-111 to display delivered quantity of water, if required.

6.2.6 Turn the makeup control switch (1/MU) to "START".

6.2.7 Confirm expected primary water flow.

6.2.8 When the desired number of gallons is reached on the primary water integrator, verify primary water flow stops.

6.2.9 Place the mode selector switch (43/MU) in the "AUTO" position.

6.2.10 Turn the makeup control switch (1/MU) to "START".

6.2.11 Reset the primary water batch integrator as follows:

a. Depress the "RESET" key.

b. Depress the "START" key.

6.3 00122533.DOC 02 1203.0704

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PACIFIC GAS AND ELECTRIC COMPANY NUMBER OP B-1A:VII DIABLO CANYON POWER PLANT REVISION 33A PAGE 6 OF 39 TITLE: CVCS - Makeup Control System Operation UNIT 1 BORATION WITH MAKEUP SYSTEM IN AUTOMATIC NOTE 1: The following instructions apply to boration with the makeup system aligned for automatic operation per section 6.1.

NOTE 2: A checklist style form is provided in Attachment 9.2 which contains the same essential actions of this entire section and can be used as a stand alone instruction.

6.3.1 Place the makeup control switch (1/MU) in the "STOP" position.

6.3.2 Place the makeup mode selector switch (43/MU) in the "BORATE" position.

6.3.3 If not already set for the proper quantity, enter the required number of gallons in the boric acid integrator using the BATCH function and data entry keys.

(Refer to Attachment 9.3) 6.3.4 Enable the integrator as follows, if required:

a. Press the "RESET" key.

b. Press the "START" key.

6.3.5 Select "SUM" on YIC-110 to display delivered quantity of boric acid, if required.

6.3.6 Turn the makeup control switch (1/MU) to "START".

6.3.7 Confirm expected boric acid flow.

6.3.8 When the desired number of gallons is reached on the boric acid integrator, verify boric acid flow stops.

6.3.9 Place the mode selector switch (43/MU) in the "AUTO" position.

6.3.10 Turn the makeup control switch (1/MU) to "START".

6.3.11 Reset the boric acid batch integrator as follows:

a. Depress the "RESET" key.

b. Depress the "START" key.

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PACIFIC GAS AND ELECTRIC COMPANY NUMBER OP B-1A:VII DIABLO CANYON POWER PLANT REVISION 33A PAGE 23 OF 39 TITLE: CVCS - Makeup Control System Operation UNIT 1 6.9 MANUAL OPERATION WITH MAKEUP CONTROL SYSTEM INOPERABLE NOTE 1: With 43/MU in OFF or removed, the Boric Acid Blend Controller (HC-110) does not work. The Primary Water Blend Controller (HC-111) works in manual when FCV-111A is selected to OPEN.

NOTE 2: This section is also applicable when 120 VAC control power has been removed from the reactor coolant makeup control system control circuit (PY1422, Ref:

Dwg 437596).

6.9.1 Verify the makeup mode selector switch (43/MU) is in OFF or removed.

6.9.2 Initial Lineup:

a. Verify the makeup control switch (1/MU) in the STOP position.

b. Select the following valves to CLOSE:

FCV-111B

FCV-111A

FCV-110B

FCV-110A

c. Verify the following valves are closed:

FCV-111B

FCV-111A

FCV-110B

FCV-110A NOTE: There is no practical way to supply blended makeup in this configuration.

6.9.3 If blended makeup is desired:

a. Calculate the total amount of boric acid and primary water desired.

b. Borate the calculated amount of boric acid using step 6.9.4.

c. When boration is complete, then dilute the calculated amount of primary water using step 6.9.5.

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PACIFIC GAS AND ELECTRIC COMPANY NUMBER OP B-1A:VII DIABLO CANYON POWER PLANT REVISION 33A PAGE 24 OF 39 TITLE: CVCS - Makeup Control System Operation UNIT 1 6.9.4 To Borate:

CAUTION: The system will NOT automatically stop when the integrator count is complete.

Boration must be stopped manually.

a. Determine the number of gallons of boric acid necessary to make the required boron concentration change. As necessary, use the boration spreadsheet program.

b. Set up the boric acid integrator as follows:

1. Set the required number of gallons of boric acid into the integrator using the BATCH function and the data setting keys.

2. Enable the integrator as follows:

a) Press "RESET" key.

b) Press the "START" key.

c. Select FCV-110A to OPEN.

d. Select FCV-110B to OPEN.

e. Confirm boric acid flow.

f. If desired, select Hi Speed on the in-service boric acid transfer pump to raise flow rate.

g. When the boric acid flow integrator count is complete, then select FCV-110B to CLOSE.

h. Verify that boric acid flow stops.

i. Select FCV-110A to CLOSE.

j. If the in-service boric acid transfer pump is in Hi Speed, then return the pump to Lo Speed.

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PACIFIC GAS AND ELECTRIC COMPANY NUMBER OP B-1A:VII DIABLO CANYON POWER PLANT REVISION 33A PAGE 25 OF 39 TITLE: CVCS - Makeup Control System Operation UNIT 1 6.9.5 To Dilute:

CAUTION: The system will NOT automatically stop when the integrator count is complete.

Inadvertent dilution is possible if the flow is not stopped manually.

a. Determine the number of gallons of primary water necessary to make the required boron concentration change. As necessary, use the boration spreadsheet program.

b. Set up the primary water integrator as follows:

1. Set the required number of gallons of primary water into the integrator using the BATCH function and the data setting keys.

2. Enable the integrator as follows:

a) Press "RESET" key.

b) Press the "START" key.

c. Place the Primary Water Blend Controller (HC-111) in MANUAL, and adjust the demand to the desired flow rate.

d. Select FCV-111A to OPEN.

e. Perform ONE of the following:

Select FCV-111B to OPEN, and confirm primary water flow, OR

Select FCV-110B to OPEN (alternate dilute), and confirm primary water flow.

f. When the primary water flow integrator count is complete:

1. Verify FCV-111B selected to CLOSE.

2. Verify FCV-110B selected to CLOSE.

3. Confirm primary water flow stops.

g. Select FCV-111A to CLOSE.

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09/21/04 Page 1 of 1 DIABLO CANYON POWER PLANT TITLE: Dilution Checklist OP B-1A:VII ATTACHMENT 9.1 1

DATE: ____________ TIME: ____________

DILUTE

[ ] 1. Verify at least once/shift that this checklist is the proper version by comparing the date in the upper left corner to the date on a Priority 1 copy of the procedure or Procedure Navigator.

[ ] 2. 1/MU to STOP

[ ] 3. 43/MU to DILUTE

[ ] 4. Verify gallons - BATCH key

[ ] 5. Verify integrator enabled - RESET / START keys

[ ] 6. Verify SUM on YIC-111

[ ] 7. 1/MU to START

[ ] 8. Verify expected primary water flow

[ ] 9. Verify flow stops

[ ] 10. 43/MU to AUTO

[ ] 11. 1/MU to START

[ ] 12. Reset integrator - RESET / START keys Total Amount Added This Dilution: _______________ Gallons Completed form should be given to the SFM for tracking of reactivity changes. This form is for information purposes only and there are no retention requirements.

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09/21/04 Page 1 of 1 DIABLO CANYON POWER PLANT TITLE: Boration Checklist OP B-1A:VII ATTACHMENT 9.2 1

DATE: ____________ TIME: ____________

BORATE

[ ] 1. Verify at least once/shift that this checklist is the proper version by comparing the date in the upper left corner to the date on a Priority 1 copy of the procedure or Procedure Navigator.

[ ] 2. 1/MU to STOP

[ ] 3. 43/MU to BORATE

[ ] 4. Verify gallons - BATCH key

[ ] 5. Verify integrator enabled - RESET / START keys

[ ] 6. Verify SUM on YIC-110

[ ] 7. 1/MU to START

[ ] 8. Verify expected boric acid flow

[ ] 9. Verify flow stops

[ ] 10. 43/MU to AUTO

[ ] 11. 1/MU to START

[ ] 12. Reset integrator - RESET / START keys Total Amount Added This Boration: _______________ Gallons Completed form should be given to the SFM for tracking of reactivity changes. This form is for information purposes only and there are no retention requirements.

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07/01/04 Page 1 of 1 DIABLO CANYON POWER PLANT TITLE:

OP B-1A:VII ATTACHMENT 9.3 Boration/Dilution Batch Integrator Function 1

STATUS LIGHTS LOAD Lit during batching. Flashes when batch is interrupted.

PRE Lit when batching is secured and reset.

Flashes when batch is interrupted.

END Lit when batch ends, off when batch is reset. Flashes when batch is interrupted.

ALM Flashes when memory backup battery voltage is low.

FAIL Lit if the controller fails, i.e. loss of power (PY-2119) or major malfunction.

SEQUENCE CONTROL SWITCHES START Initiates batch, i.e. closes the integrator output relay allowing batching to start.

When starting a new batch, reset must be pressed first. When resuming an interrupted batch, press only this switch to restart.

RESET Enables the batch sequence and resets the batch totalizer.

STOP Stops a batch in the middle of a sequence, i.e. output relay opens.

FLOW LEDS Instantaneous bar display of input from flow transmitter Integrator Setup Sequence:

BATCH DISPLAY Displays current batch set value in gallons. [ ] Select BATCH with Data display Selector [].

DATA DISPLAY [ ] Set desired gallons with SHIFT [] & INCR []

keys.

Displays the data from the list selected below.

[ ] Press SET twice to 'load' setting, verify value DATA SETTING SWITCHES displayed in batch window.

SET Pressed after setting data quantity to enter the value into the integrator. [ ] Press RESET and START to enable integrator.

SHIFT Selects the digit in the data display that is [ ] Select SUM with Data Display Selector [],

to be changed by the INCR switch. verify reading at zero, to view count up to batch INCR Increments the digit selected in the data setting.

display. [ ] Select FLOW with Data Display Selector [] to MAIN DATA LIST AND DATA DISPLAY view flowrate.

SELECTOR SWITCHES NOTE 1: Once a batch is complete, if the same batch SUM Displays the total number of gallons is desired again, just press RESET and START to pumped since last reset. enable integrator.

BATCH Displays batch quantity in gallons set by NOTE 2: A batch may be interrupted by pressing the operator. STOP button and restarted with the START button.

FLOW LIMIT This function disabled.

FLOW Displays instantaneous flow in gallons per minute.

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NUCLEAR POWER GENERATION DIABLO CANYON POWER PLANT JOB PERFORMANCE MEASURE Number: NRCLJP-012

Title:

RESET THE TURBINE DRIVEN AUX FEEDWATER PUMP Examinee:

Evaluator:

Print Signature Date Results: Sat Unsat Total Time: minutes Comments:

References:

OP D-1:IV, Steam-Driven Auxiliary Feed Pump - Restart or Make Available After Overspeed Trip, Rev. 15 Alternate Path: Yes No X Time Critical: Yes No X Time Allotment: 10 minutes Critical Steps: 3, 4, 5 Job Designation: RO/SRO Task Number: 061/04S/A2.04 Rating: 3.4/3.8 AUTHOR: JACK BLACKWELL DATE: 01/18/2005 REVIEWED BY: N/A DATE:

TRAINING LEADER APPROVED BY: N/A DATE:

LINE MANAGER REV. 0

JPM TITLE: RESET THE TURBINE DRIVEN AUX FEEDWATER PUMP JPM NUMBER: NRCLJP-012 INSTRUCTOR WORKSHEET Directions: No PLANT controls or equipment are to be operated during the performance of this Job Performance Measure. All actions taken by the examinee should be clearly demonstrated and verbalized to the evaluator. The student will be given the initial conditions, initiating cue, and task standard. The examiner will then ask if any clarifications are needed. The examinee may be given the applicable procedure and step with which to begin.

Required Materials: Copy of OP D-1:IV.

Initial Conditions: Unit 1 has tripped from 100% power. All four Steam Generator narrow range levels are below 4%. AFW pump 1-1 has tripped on overspeed and is needed for a plant heat sink.

Initiating Cue: The Shift Foreman directs you to restart AFW pump 1-1 in accordance with OP D-1:IV.

Task Standard: AFW pump 1-1 has been restarted in accordance with OP D-1:IV.

NRCLJP012.DOC PAGE 2 OF 5 REV. 0

JPM TITLE: RESET THE TURBINE DRIVEN AUX FEEDWATER PUMP JPM NUMBER: NRCLJP-012 INSTRUCTOR WORKSHEET Start Time:

Step Expected Operator Actions

1. References procedure. 1.1 References step 6.1 of OP D-1:IV.

2. Verify that the speed setting 2.1 Verifies the speed setting knob on knob on the turbine govenor is FCV-15 is positioned to the maximum positioned to the maximum speed setting (fully clockwise.)

setting. ************************************

Cue: Knob is fully clockwise.

Step was: Sat: ______ Unsat _______*

- 3. Turn the turbine throttle trip 3.1 Turns FCV-152 in the clockwise valve MS-1-FCV-152 handwheel direction until the spring is fully in the clockwise direction. compressed.**

Cue: Spring is fully compressed.

Step was: Sat: ______ Unsat _______*

- 4. Latch up the latching lever by 4.1 Reads Note and refers to Att. 9.1.

means of the trip hook.

4.2 Checks latch plate fully depressed into the latching mechanism.

Cue: Latch plate is NOT depressed into the latching mechanism.

4.3 Presses down on the treaded stud until the latch plate is fully seated on the latching mechanism.**

Cue: Latch plate is fully seated into the latching mechanism.

Step was: Sat: ______ Unsat _______*

Denotes an entry required on the JPM cover sheet.

- Denotes Critical Step and Sub Steps.

NRCLJP012.DOC PAGE 3 OF 5 REV. 0

JPM TITLE: RESET THE TURBINE DRIVEN AUX FEEDWATER PUMP JPM NUMBER: NRCLJP-012 INSTRUCTOR WORKSHEET Step Expected Operator Actions

- 5. Open MS-1-FCV-152 fully by 5.1 Turns the handwheel in the counter turning the handwheel in the clockwise direction until FCV-152 is counter clockwise direction. fully open.**

Cue: Turbine speed is increasing as FCV-152 is manually opened.

Cue: FCV-152 is fully open and the turbine has not tripped.

Step was: Sat: ______ Unsat _______*

6. Check that the Governor is 6.1 Locates the local RPM indication to controlling speed properly. verify the turbine is at full speed.

OR 6.2 Contacts the Control Room to verify turbine speed.

Cue: Turbine speed indicates approximately 4200 RPM.

Step was: Sat: ______ Unsat _______*

Stop Time:

Total Time: (Enter total time on the cover page)

Denotes an entry required on the JPM cover sheet.

- Denotes Critical Step and Sub Steps.

NRCLJP012.DOC PAGE 4 OF 5 REV. 0

JPM NUMBER: NRCLJP-012 EXAMINEE CUE SHEET Initial Conditions: Unit 1 has tripped from 100% power. All four Steam Generator narrow range levels are below 4%. AFW pump 1-1 has tripped on overspeed and is needed for a plant heat sink.

Initiating Cue: The Shift Foreman directs you to restart AFW pump 1-1 in accordance with OP D-1:IV.

Task Standard: AFW pump 1-1 has been restarted in accordance with OP D-1:IV.

NRCLJP012.DOC PAGE 5 OF 5 REV. 0

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PACIFIC GAS AND ELECTRIC COMPANY NUMBER OP D-1:IV NUCLEAR POWER GENERATION REVISION 15 DIABLO CANYON POWER PLANT PAGE 1 OF 3 OPERATING PROCEDURE UNIT TITLE: Steam-Driven Auxiliary Feed Pump - Restart or Make Available After Overspeed Trip 107/06/04 EFFECTIVE DATE PROCEDURE CLASSIFICATION: QUALITY RELATED

1. SCOPE 1.1 This procedure provides instructions for relatching the Steam-Driven Auxiliary Feed Pump Turbine, and returning it to service.

2. DISCUSSION 2.1 The following sections should be used as applicable after an overspeed trip of the auxiliary feed pump turbine:

Section 6.1 - Restart of the Steam-Driven Auxiliary Feed Pump After an Overspeed Trip Section 6.2 - Make the Steam-Driven Auxiliary Feed Pump Available After an Overspeed Trip

3. RESPONSIBILITIES 3.1 The shift foreman (SFM) is responsible for proper alignment and operation of equipment discussed in this procedure.

4. PREREQUISITES 4.1 The following systems should be in service:

4.1.1 Main steam supply to AFW Pump 1-1 via FCV-37 and FCV-38.

AFW Pump 1-1 is operable only if it is capable of being powered from two operable and redundant steam supply sources.

4.1.2 Condensate Storage Tank or alternate auxiliary feedwater supply. (Refer to OP D-1:V.)

5. PRECAUTIONS AND LIMITATIONS 5.1 Review Technical Specifications 3.7.5 and 3.7.6.

5.2 If the turbine room fills with an excessive amount of steam, the oil in the turbine bearings and pump bearings should be changed and sampled for water. An excessive amount of steam would be enough to condense and drip onto the floor such that the better part of the pump room floor is slick or has puddles.

5.3 If the MSIVs are isolated, ensure that the steam traps upstream of the MSIVs are in service with the MSIV bypass valves open for loops 2 and 3. If the air operated MSIV bypass valves can not be opened, locally open the manual bypass valves. If the steam traps for loops 2 and 3 cannot be placed in service, declare AFW Pump 1-1 INOPERABLE, and close FCV-37 and FCV-38.

5.4 If AFW PP 1-1 turbine relief valve RV-57 is found to be lifting, investigate for possible obstruction in the turbine exhaust path.

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PACIFIC GAS AND ELECTRIC COMPANY NUMBER OP D-1:IV DIABLO CANYON POWER PLANT REVISION 15 PAGE 2 OF 3 TITLE: Steam-Driven Auxiliary Feed Pump - Restart or Make UNIT 1 Available After Overspeed Trip

6. INSTRUCTIONS 6.1 Restart of the Steam-Driven Auxiliary Feedwater Pump After an Overspeed Trip 6.1.1 Verify that the speed setting knob on the turbine governor MS-1-FCV-15 is positioned to the maximum speed setting (fully clockwise).

6.1.2 Turn the turbine throttle trip valve MS-1-FCV-152 handwheel in the clockwise direction until the spring is fully compressed.

6.1.3 Latch up the latching lever by means of the trip hook.

6.1.4 Ensure the trip mechanism on top of the bearing housing has been properly reset per Attachment 9.1.

NOTE: This is accomplished by pulling the connecting rod slightly towards the trip valve (FCV-152) while pushing down on the trip tappet and nut. Once the tappet nut is properly seated, gently release tension applied to connecting rod.

6.1.5 Open MS-1-FCV-152 fully by turning the handwheel in the counterclockwise direction.

NOTE: The turbine will roll up to speed as the handwheel is turned in the counterclockwise direction. The governor should come into action to maintain full speed (approximately 4150-4240 RPM). If the turbine again trips on overspeed, reperform steps 6.1.1 through 6.1.4, then open the throttle trip valve handwheel only until full speed is attained. RPM indication is both local and in the Control Room. If manual throttling of MS-1-FCV-152 is required, this indicates a failure of the Woodward Governor.

6.1.6 Leave the throttle trip valve in the latched position.

6.2 Make the Steam-Driven Auxiliary Feed Pump Available After an Overspeed Trip 6.2.1 Verify that plant conditions do not require restart of the Steam-Driven Auxiliary Feedwater Pump.

6.2.2 Close MS-1-FCV-95, Turbine Steam Inlet Valve.

6.2.3 Slowly open MS-1-950, Trap 118 Bypass Valve, to verify steam bypass line is depressurized (refer to PRECAUTIONS section).

6.2.4 Turn the turbine throttle trip valve MS-1-FCV-152 handwheel in the clockwise direction until the spring is fully compressed.

6.2.5 Latch up the latching lever by means of the trip hook.

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PACIFIC GAS AND ELECTRIC COMPANY NUMBER OP D-1:IV DIABLO CANYON POWER PLANT REVISION 15 PAGE 3 OF 3 TITLE: Steam-Driven Auxiliary Feed Pump - Restart or Make UNIT 1 Available After Overspeed Trip 6.2.6 Ensure the trip mechanism on top of the bearing housing has been properly reset per Attachment 9.1.

NOTE: This is accomplished by pulling the connecting rod slightly towards the trip valve (FCV-152) while pushing down on the trip tappet and nut. Once the tappet nut is properly seated, gently release tension applied to connecting rod.

6.2.7 Open MS-1-FCV-152 fully by turning the handwheel in the counterclockwise direction.

6.2.8 Leave the throttle trip valve in the latched position.

6.2.9 Open MS-1-925 and MS-1-926, turbine casing drain valves, to drain any condensate from turbine casing.

6.2.10 Verify that the speed setting knob on the turbine governor MS-1-FCV-15 is positioned to the maximum speed setting (fully clockwise).

6.2.11 Close trap 118 bypass valve (MS-1-950) and turbine casing drain valves (MS-1-925 and MS-1-926).

6.2.12 Verify turbine at rest.

6.2.13 Complete Attachment 9.1 and forward to the SFM.

6.2.14 SFM, review Attachment 9.1 and document with a formal log entry that the Steam-Driven Auxiliary Feed Pump Turbine has been properly relatched.

NOTE: Depending upon the nature of the overspeed trip, initiate an A/R as necessary.

7. REFERENCES 7.1 OVID 106703 SHT 3, and 106704 SHT 4.

8. RECORDS 8.1 There are no formal requirements for record retention.

9. ATTACHMENTS 9.1 "Latching Mechanism for Steam-Driven Auxiliary Feedwater Pump Turbine," 06/25/04 9.2 "Restarting Steam-Driven Auxiliary Feedwater Pump After an Overspeed Trip," 10/01/92 00597115.DOC 02 0706.0142

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06/25/04 Page 1 of 1 DIABLO CANYON POWER PLANT TITLE:

OP D-1:IV ATTACHMENT 9.1 1

Latching Mechanism for Steam-Driven Auxiliary Feedwater Pump Turbine

1. Initial and perform an Independent Verification within 4 hours4.62963e-5 days 0.00111 hours 6.613757e-6 weeks 1.522e-6 months section 6.2 only);

Performed By Indep. Verif.

THROTTLE TRIP VALVE LATCHED IAW PICTURE.

GOVERNOR SPEED SETTING AT MAXIMUM SETTING.

2. Forward this Attachment to SFM for review.

00597115.DOC 02 0706.0142

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10/01/92 Page 1 of 1 DIABLO CANYON POWER PLANT TITLE:

OP D-1:IV ATTACHMENT 9.2 1

Restarting Steam-Driven Auxiliary Feedwater Pump After an Overspeed Trip The following instructions are for emergency use only in restarting the Steam-Driven Auxiliary Feedwater Pump after an overspeed trip. They are posted on lamicoid in the Auxiliary Building at the Steam-Driven Auxiliary Feedwater Pump.

Restarting the Steam-Driven Auxiliary Feedwater Pump after an overspeed trip.

1. VERIFY THE SPEED SETTING ON TURBINE GOVERNOR MS-1-FCV-15 IS SET TO MAXIMUM (fully clockwise).

2. TURN THE TURBINE TRIP THROTTLE VALVE MS-1-FCV-152 HANDWHEEL IN THE CLOCKWISE DIRECTION UNTIL THE SPRING IS FULLY COMPRESSED.

3. LATCH UP THE LATCHING LEVER USING THE TRIP HOOK.

4. ON TOP OF THE OUTBOARD TURBINE BEARING, CHECK THE LATCH PLATE FULLY DEPRESSED INTO THE LATCHING MECHANISM TO VERIFY THE TRIP MECHANISM PROPERLY RESET.

a. If necessary, press down on the threaded stud until the latch plate is fully seated on the latch mechanism.

5. FULLY OPEN TURBINE TRIP THROTTLE VALVE MS-1-FCV-152 BY TURNING THE HANDWHEEL IN THE COUNTERCLOCKWISE DIRECTION. (The turbine will roll up to speed as MS-1-FCV-152 is opened. The governor will maintain full speed of approximately 4150-4240 RPM.)

6. IF THE TURBINE TRIPS ON OVERSPEED, THEN DO THE FOLLOWING:

a. Perform steps 1 through 4.

b. Slowly open Turbine Trip Throttle Valve MS-1-FCV-152 by turning the handwheel in the counterclockwise direction until full speed is attained.

7. LEAVE TURBINE TRIP THROTTLE VALVE MS-1-FCV-152 IN THE LATCHED POSITION.

00597115.DOC 02 0706.0142

NUCLEAR POWER GENERATION DIABLO CANYON POWER PLANT JOB PERFORMANCE MEASURE Number: NRCLJP-079

Title:

TRANSFER PRESSURIZER HEATER GROUP 13 TO BACKUP POWER Examinee:

Evaluator:

Print Signature Date Results: Sat Unsat Total Time: minutes Comments:

References:

OP A-4A:I, Pressurizer - Make Available, Rev 17 Alternate Path: Yes No X Time Critical: Yes No X Time Allotment: 15 minutes Critical Steps: 3, 4, 5, 6, 7, 9, 10, 11, 12 Job Designation: RO/SRO Task Number: 062/06/A2.10 Rating: 3.0/3.3 AUTHOR: JACK BLACKWELL DATE: 01/18/2005 REVIEWED BY: N/A DATE:

TRAINING LEADER APPROVED BY: N/A DATE:

LINE MANAGER REV. 0

JPM TITLE: TRANSFER PRESSURIZER HEATER GROUP 13 TO JPM NUMBER: NRCLJP-079 BACKUP POWER INSTRUCTOR WORKSHEET Directions: No PLANT controls or equipment are to be operated during the performance of this Job Performance Measure. All actions taken by the examinee should be clearly demonstrated and verbalized to the evaluator. The student will be given the initial conditions, initiating cue, and task standard. The examiner will then ask if any clarifications are needed. The examinee may be given the applicable procedure and step with which to begin.

Required Materials: Copy of OP A-4A:I, Section 6.3 Initial Conditions: The Unit 1 Reactor Coolant System is being filled and vented. All house loads are being supplied by startup power.

Initiating Cue: The Shift Foreman directs you to make available pressurizer heater group 13 from its backup power supply in accordance with OP A-4A:I, step 6.3.1.c.

Task Standard: Pressurizer heater group 13 has been made available from its backup power supply in accordance with OP A-4A:I.

NRCLJP079.DOC PAGE 2 OF 7 REV. 0

JPM TITLE: TRANSFER PRESSURIZER HEATER GROUP 13 TO JPM NUMBER: NRCLJP-079 BACKUP POWER INSTRUCTOR WORKSHEET Start Time:

Step Expected Operator Actions

1. Reads Caution. 1.1 Reads Caution.

Cue: Another Operator has been assigned to monitor the loading of Bus H 480V transformer.

2. Place control switch for heater 2.1 Goes to or calls the Control Room to group 13 in the OFF position. check the position of the control switch for heater group 13.

Cue: The control switch for heater group 13 is in the OFF position and the green light is ON.

Step was: Sat: ______ Unsat _______*

- 3. Verify that heater group 13 3.1 Locates the normal breaker for heater normal breaker, 52-13E-2 is group 13 on load center 13E.

open.

3.2 Verifies that the breaker is open. **

Step was: Sat: ______ Unsat _______*

- 4. Place the DC control power 4.1 Locates the DC control power switch switch for pressurizer heater for heater group 13 normal breaker on group 13 normal breaker in the load center 13E.

OFF position.

4.2 Places the control power toggle switch in the OFF position. **

Step was: Sat: ______ Unsat _______*

Denotes an entry required on the JPM cover sheet.

- Denotes Critical Step and Sub Steps.

NRCLJP079.DOC PAGE 3 OF 7 REV. 0

JPM TITLE: TRANSFER PRESSURIZER HEATER GROUP 13 TO JPM NUMBER: NRCLJP-079 BACKUP POWER INSTRUCTOR WORKSHEET Step Expected Operator Actions

- 5. Places 480v transformer THH10 5.1 Locates fan switch 43T in 480V Bus fan switch 43T to MAN. H room.

5.2 Places 43 T switch in MAN. **

5.3 Checks amber FANS ON light is on.

Cue: Amber light is on Step was: Sat: ______ Unsat _______*

- 6. Check heater group 13 backup 6.1 Locates heater group 13 backup breaker, 52-1H-74 open. breaker 52-1H-74 at 480V Bus H.

6.2 Checks that the breaker is open. **

Step was: Sat: ______ Unsat _______*

- 7. Check open the DC control 7.1 Locates the DC knife switch cabinet power knife switch for heater located above the vital breaker.

group 13 backup breaker.

Cue: You may open the cabinet.

7.2 Checks that the knife switch is open. **

Cue: The knife switch is open.

Step was: Sat: ______ Unsat _______*

Denotes an entry required on the JPM cover sheet.

- Denotes Critical Step and Sub Steps.

NRCLJP079.DOC PAGE 4 OF 7 REV. 0

JPM TITLE: TRANSFER PRESSURIZER HEATER GROUP 13 TO JPM NUMBER: NRCLJP-079 BACKUP POWER INSTRUCTOR WORKSHEET Step Expected Operator Actions

8. Verify that both white potential 8.1 Locates the manual transfer switch on lights on the manual transfer the wall next to the 52-1H-74 breaker.

switch are not lit.

Note: Since the normal breaker is available, a white light may be ON.

8.2 Checks that neither light is ON.

Cue: Both lights are OFF.

Step was: Sat: ______ Unsat _______*

- 9. Move the transfer switch down **********************************

to the backup (vital) bus Cue: The Shift Foreman has assigned position. another operator to complete all required seal valve change forms.

9.1 Positions the transfer switch to the backup supply. **

Step was: Sat: ______ Unsat _______*

- 10. Check the heater group 13 10.1 Verifies that the heater group 13 backup breaker, 52-1H-74 backup breaker is racked in. **

racked in.

Step was: Sat: ______ Unsat _______*

- 11. Close the DC control power 11.1 Locates the DC knife switch cabinet knife switch for heater group 13 located above the vital breaker.

backup breaker.

11.2 Places the knife switch in the CLOSE position. **

Cue: The knife switch is closed.

Step was: Sat: ______ Unsat _______*

Denotes an entry required on the JPM cover sheet.

- Denotes Critical Step and Sub Steps.

NRCLJP079.DOC PAGE 5 OF 7 REV. 0

JPM TITLE: TRANSFER PRESSURIZER HEATER GROUP 13 TO JPM NUMBER: NRCLJP-079 BACKUP POWER INSTRUCTOR WORKSHEET Step Expected Operator Actions

- 12. Verify the D.C. Charging Power 12.1 Locates the D.C. charging power Switch for heater group 13 switch on the lower front of backup breaker (52-1H-74) is on 52-1H-74.

and springs charged.

12.2 Verifies the following:

CHARGING POWER switch in the ON position **

SPRINGS CHARGED flag displayed Step was: Sat: ______ Unsat _______*

13. Notify the control room of the 13.1 Notifies the control room that heater status of heater group 13. group 13 is available from the backup power supply.

Cue: The Control Operator will complete the procedure and energize heater group 13.

Step was: Sat: ______ Unsat _______*

Stop Time:

Total (Enter total time on the cover page)

Time:

Denotes an entry required on the JPM cover sheet.

- Denotes Critical Step and Sub Steps.

NRCLJP079.DOC PAGE 6 OF 7 REV. 0

JPM NUMBER: NRCLJP-079 EXAMINEE CUE SHEET Initial Conditions: The Unit 1 Reactor Coolant System is being filled and vented. All house loads are being supplied by startup power.

Initiating Cue: The Shift Foreman directs you to make available pressurizer heater group 13 from its backup power supply in accordance with OP A-4A:I, step 6.3.1.c.

Task Standard: Pressurizer heater group 13 has been made available from its backup power supply in accordance with OP A-4A:I.

NRCLJP079.DOC PAGE 7 OF 7 REV. 0

- - UNCONTROLLED PROCEDURE - DO NOT USE TO PERFORM WORK or ISSUE FOR USE ***

PACIFIC GAS AND ELECTRIC COMPANY NUMBER OP A-4A:I DIABLO CANYON POWER PLANT REVISION 17 PAGE 6 OF 10 TITLE: Pressurizer - Make Available UNIT 1 CAUTION: Monitor the load on 480V Vital Bus/Transformer when transferring pressurizer heaters to the backup power supply. Pressurizer heaters and CFCUs in fast speed could result in exceeding normal maximum load. Load in excess of normal maximum load rating is allowed for short periods of time, see AR A0509579 for description of load and time limits.

6.3 Pressurizer Heaters - Make Available from Backup Power Supply 6.3.1 To energize the pressurizer heaters from the back up power supply, IF off-site power is available THEN go to step 6.3.1c below; IF off-site power is not available (diesels supplying vital busses) THEN perform the following:

a. Select the backup power supply to be used (vital Bus G for heater group 12 or vital Bus H for heater group 13), based on the bus with the lowest load indicated on the diesel.

b. Determine if loads must be stripped from the selected vital bus.

1. IF the bus load is <2.6 MW, THEN go to step c below, it will not be necessary to strip any loads.

CAUTION: Any safety injection signal must be reset before loads can be stripped and before the heaters can be energized. Reset only if applicable reset criteria is met in the specific Emergency Operating Procedures.

2. IF the bus load is >2.6 MW, THEN strip some load using the criteria below:

a) IF all containment fan coolers are running and average containment air temperature is below 120°F, THEN shut down fan cooler 1-3, 1-4 or 1-5, as applicable.

b) IF all three component cooling water pumps are running, THEN one may be shut down. (Either 1-2 or 1-3, as applicable.)

c) IF the ECCS pump shutdown criteria in the applicable Emergency Operating Procedure is met, THEN the following may be shut down as applicable.

1) SI Pump 1-2

2) RHR Pump 1-1 OR RHR Pump 1-2 00280017.DOC 02 1111.0736

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PACIFIC GAS AND ELECTRIC COMPANY NUMBER OP A-4A:I DIABLO CANYON POWER PLANT REVISION 17 PAGE 7 OF 10 TITLE: Pressurizer - Make Available UNIT 1

c. Energize the heaters as follows:

1. Place the control switch for the selected heater group (12 or 13) to the OFF position and check the green light on.

2. Verify that the selected heater group normal breaker (52-13D-6 for heater group 12 or 52-13E-2 for heater group 13) is open, (at the appropriate Bus D or E).

3. Place the D.C. Control Power Switch for the selected heater group normal breaker in the OFF position (located at Load Center 13 D or E).

4. For heater group 13 only, place 480V Transformer THH10 fan switch 43T to MAN. (Located in 480V Bus H room).

a) Check amber FANS ON light is on.

5. Check the heater group backup breaker (52-1G-72 for heater group 12 or 52-1H-74 for heater group 13) open (at the appropriate vital bus room G or H).

6. Check open the D.C. Control Power Knife Switch for the selected heater group backup breaker (located above the vital breaker).

7. Verify that both white potential lights on the manual transfer switch are not lit.

8. Move the transfer switch down to the backup (vital) bus position.

Fill out the Sealed Component Change Form in accordance with OP1.DC20.

9. Rack in, or check racked in, the selected heater group backup breaker (52-1G-72 or 52-1H-74).

10. Close the D.C. Control Power Knife Switch for the selected heater group backup breaker (located above the vital breaker).

11. Verify the D.C. Charging Power Switch for the selected heater group backup breaker is in the ON position (located on the lower front of the vital breaker) and springs are charged.

CAUTION: The pressurizer heater group breaker auto trip on low pressurizer level is defeated when heaters are on backup power supply. Manually turn heaters OFF if pressurizer level drops below 17%.

12. Place the control switch for the selected heater group in the ON position in the Control Room.

NOTE: The indicating lights for this group will not illuminate since they are associated with the normal power supply breaker position.

00280017.DOC 02 1111.0736

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PACIFIC GAS AND ELECTRIC COMPANY NUMBER OP A-4A:I DIABLO CANYON POWER PLANT REVISION 17 PAGE 8 OF 10 TITLE: Pressurizer - Make Available UNIT 1

13. Verify that the heaters are energized by observing the individual wattmeter for the selected heater group.

14. If the selected heater group does not energize as indicated by the associated watt meter, manually close breaker as follows (ref. A0481018).

a) Verify the control switch on CC1 is selected to Auto.

b) Verify closing springs are charged (charging the closing springs electrically requires the local DC knife switch above the breaker to be closed and the toggle switch on the breaker to be in the "on" position).

c) Pull up on the local close lever.

15. Verify that the diesel generator is not overloaded by referencing the capability curve in OP J-6B.

6.4 Pressurizer Heaters - Return to Normal Power Supply from Backup 6.4.1 To energize pressurizer heaters from the normal power supply after being energized from the backup power supply proceed as follows:

a. Place the control switch for the selected heater group (12 or 13), to the OFF position.

b. Verify that the selected heater group backup breaker (52-1G-72 for heater group 12 or 52-1H-74 for heater group 13) is open, (at the appropriate vital bus room G or H).

c. Open the D.C. Control Power Knife Switch for the selected heater group backup breaker (located above the vital breaker).

d. Check the heater group normal breaker (52-13D-6 for heater group 12 or 52-13E-2 for heater group 13) open (at the appropriate Bus D or E).

e. For heater group 13 only, place 480V Transformer THH10 fan switch 43T to AUTO. (Located in 480V Bus H room).

f. Verify the D.C. Control Power Switch for the selected heater group normal breaker is in the OFF position.

g. Verify that both white potential lights on the manual transfer switch are not lit.

h. Move the transfer switch up to the normal bus position. Complete the Sealed Component Change Form in accordance with OP1.DC20.

i. Rack in, or check racked in, the selected heater group normal breaker (52-13D-6 or 52-13E-2).

00280017.DOC 02 1111.0736

NUCLEAR POWER GENERATION DIABLO CANYON POWER PLANT JOB PERFORMANCE MEASURE Number: NRCLJP-088

Title:

ALIGN EMERGENCY BORATION Examinee:

Evaluator:

Print Signature Date Results: Sat Unsat Total Time: minutes Comments: This is a Unit 2 JPM.

References:

OP AP-6, Emergency Boration, Rev 15 Alternate Path: Yes No X Time Critical: Yes No X Time Allotment: 5 minutes Critical Steps: 1 Job Designation: RO/SRO Task Number: 024/01/AA1.20 Rating: 3.2/3.3 AUTHOR: JACK BLACKWELL DATE: 01/18/2005 REVIEWED BY: N/A DATE:

TRAINING LEADER APPROVED BY: N/A DATE:

LINE MANAGER REV. 0

JPM TITLE: ALIGN EMERGENCY BORATION JPM NUMBER: NRCLJP-088 INSTRUCTOR WORKSHEET Directions: No PLANT controls or equipment are to be operated during the performance of this Job Performance Measure. All actions taken by the examinee should be clearly demonstrated and verbalized to the evaluator. The student will be given the initial conditions, initiating cue, and task standard. The examiner will then ask if any clarifications are needed. The examinee may be given the applicable procedure and step with which to begin.

Required Materials: None.

Initial Conditions: Due to a CWP trip and a rapid ramp to 50% power on Unit 2, control rods are below the rod insertion limit. The Unit 2 operating crew has been unable to borate from the Control Room.

Initiating Cue: The Unit 2 Shift Foreman directs you to locally open CVCS-2-8471, manual emergency boration valve, and inform the control room when it is opened.

Task Standard: The manual emergency boration valve is opened.

NRCLJP088.DOC PAGE 2 OF 4 REV. 0

JPM TITLE: ALIGN EMERGENCY BORATION JPM NUMBER: NRCLJP-088 INSTRUCTOR WORKSHEET Start Time:

Step Expected Operator Actions

- 1. Open Unit 2 emergency borate 1.1 Opens local manual emergency valve CVCS-8471. borate valve CVCS-8471 (Unit 2 VCT blender room).**

Note: Actual conditions within the blender room will determine how close the operator can approach the valve. Entry into a surface contamination area should not be allowed. Describing the location and operation of the valve satisfies the critical task.

Step was: Sat: ______ Unsat _______*

2. Notify the Unit 2 Shift Foreman. 2.1 Notifies the control room that CVCS-8471 is open.

Step was: Sat: ______ Unsat _______*

Stop Time:

Total Time: (Enter total time on the cover page)

Denotes an entry required on the JPM cover sheet.

- Denotes Critical Step and Sub Steps.

NRCLJP088.DOC PAGE 3 OF 4 REV. 0

JPM NUMBER: NRCLJP-088 EXAMINEE CUE SHEET Initial Conditions: Due to a CWP trip and a rapid ramp to 50% power on Unit 2, control rods are below the rod insertion limit. The Unit 2 operating crew has been unable to borate from the Control Room.

Initiating Cue: The Unit 2 Shift Foreman directs you to locally open CVCS-2-8471, manual emergency boration valve, and inform the control room when it is opened.

Task Standard: The manual emergency boration valve is opened.

NRCLJP088.DOC PAGE 4 OF 4 REV. 0

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PACIFIC GAS AND ELECTRIC COMPANY NUMBER OP AP-6 DIABLO CANYON POWER PLANT REVISION 15 PAGE 3 OF 6 TITLE: Emergency Boration UNITS 1 AND 2 ACTION/EXPECTED RESPONSE RESPONSE NOT OBTAINED NOTE: Emergency Boration Flowmeter FI-113 may peg high at 50 GPM. XFIT-113 in the Cable Spreading room may be used for higher flowrates or to determine total gallons of boric acid added via the Emergency Boration flowpath.

2. INITIATE Alternate Boration Method

a. OPEN CVCS-8104 and verify a. Perform one of the following in order of approximately 30 GPM or greater preference:

Emergency Boration Flow

1) Swap Charging Pp suction to the RWST.

a. OPEN 8805A AND 8805B.

b. CLOSE LCV-112B AND LCV-112C.

c. VERIFY GREATER THAN 90 GPM charging flow.

OR

2) Locally OPEN CVCS-8471 (100' Blender Room).

3. CHECK Sufficient Boric Acid Available:

In Service Boric Acid Tank level GREATER a. Stop the Boric Acid Transfer Pp not aligned to THAN required gallons of Boric Acid per the blender.

Appendix A

b. Locally OPEN CVCS-8476, Boric Acid Transfer Pp crosstie. (100' Behind Suction to BA Transfer Pp 1-1/2-2).

WHEN Sufficient BA inventory restored, THEN Realign the system per OP B-1C:II, 4% BORIC ACID SYSTEM - PLACE IN SERVICE.

00307115.DOC 02 0826.0836

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