ML110060143

Hope Creek, HC.OP-IO.ZZ-0004(Q), Rev. 82, Shutdown from Rated Power to Cold Shutdown.

ML110060143
Person / Time
Site:	Salem, Hope Creek
Issue date:	01/15/2010
From:	Public Service Enterprise Group
To:	Office of Nuclear Reactor Regulation
References
LR-N10-0355 HC.OP-IO.ZZ-0004(Q), Rev 82
Download: ML110060143 (86)

Text

Page 1 of 1

Packages and Affected Document Number s incorporated into this revision: CP No. CP Rev. AD No. Rev No. None The following OPEX were incorporated into this revision: None The following OTSCs were incorporated into this revision: None Changes the setpoint in Note 5.1.11.D to an allowable band of 4500 to 5100 gpm for the SCP. This was evaluated in DCP 80098725 and is editorial. (80098725-0210) Adds Step 5.1.29.C to ensure that the reactor cooldown is logged. This is the same as Step 5.2.4 and is editorial. (70105925-0010) Corrects a step numbering error on Attachment 12, Step 5.1.13.C. This is an editorial change.

None

1.0 PURPOSE.............................................................................................................2

2.0 PREREQUISITES.................................................................................................2

3.0 PRECAUTIONS

A ND LIMITATIONS....................................................................2

4.0 EQUIPMENT

REQUIRED...................................................................................11

5.0 PROC EDURE.....................................................................................................12

5.1 Load R eduction........................................................................................12 5.2 Reactor Cooldown and De-pressuri zation................................................31

6.0 RECORDS..........................................................................................................

51

7.0 REFERENCES

....................................................................................................51

Attachment 1, Final Checks (Ent ering OPERATIONAL CONDITION 2).................................53 , Final Checks (Ent ering OPERATIONAL CONDITION 3).................................55 , Final Checks (Ent ering OPERATIONAL CONDITION 4).................................57 Attachment 4, Reactor Coolant System Temperature/

Pressure Data....................................59 , Po wer to Flow Maps........................................................................................62 , Placing The Plant In Alternate Decay Heat Removal Mode Of Operation.......64 , Vessel Level Instrument ation Temperature Co mpensation Curves.................66 , Installation of Breaker Overloads for Bkr 52-264042 (BG-HV-F031)...............70 , Main Turbine Shell Cooldown..........................................................................71 0, Nuclear Instrument ation Surveillance Requirements in Operations Condi tions 2, 3, 4 and 5............................................................76 1, Operational Limitations Co mment Page..........................................................80 2, Shut down Flow Chart......................................................................................81

START TIME DATE BY TERMINATION TIME DATE BY COMPLETION TIME DATE BY This procedure provides guidelines for the s hutdown of the plant from rated power to a Cold Shutdown condition. None 3.1 3.1.1. This procedure is to be used as a guideline for the shutdown of the plant from rated power to a Cold Shutdown condition.

IF it is desired to shut down t he plant from other than rated power using this procedure, the proper entry point should be determined by the SM/CRS. It is NOT required that each step

be performed in precise sequence as long as the steps are

performed in a timely manner in keeping with the intent of this

procedure. Changes in sequence must be evaluated for

potential reactivity challenges.

Any deviations and/or limitations of this procedure shall be justified and documented on 1, Operational Limitations Comment Page. ____ 3.1.2. The Flowcharts are to be used as an extension of the procedure.

The procedure user should use a marker to record information

directly on the Flowcharts, and the Flowcharts should be

updated simultaneously with the body of the procedure. ____ 3.1.3. This procedure may be used to perform a controlled shutdown where the Reactor is placed in a Hot Shutdown condition prior to

reaching a low power level provided the sequence of Reactor

operation has been evaluated as part of a "preplanned evolution". ____

3.1.4. IF , while executing this procedure, conditions warrant placing the Reactor in Hot Shutdown without all preparatory actions completed OR this procedure is being used as part of a "preplanned evolution" to place the Reactor in a Hot Shutdown condition prior to reaching a

low power level, THEN final mode change checks should be made in accordance with Step 5.1.29.B, and the Mode Switch may be placed in Shutdown in accordance with Step 5.1.29.D. Once the plant has been stabilized, all remaining steps in this procedure should be reviewed and completed as required. ____ 3.1.5. Control rod insertion and cooldown of the Reactor Coolant System can be performed simultaneously.

WHEN this occurs, the cooldown rate AND neutron flux should be closely monitored for any sudden changes. ____

3.1.6. IF control rod insertion is stopped pr ior to all rods being inserted, re-criticality must be anticipated due to cooldown. The Reactor

Operator shall NOT be distracted for any reason until Rx power is stable, or all rods are fully inserted. ____ 3.1.7. The following Abnormal Operating Procedures may be applicable during a plant Shut-down, and should be reviewed as

applicable: HC.OP-AB.RPV-0001(Q), Reactor Power. ____ HC.OP-AB.RPV-0003(Q), Recirculation System. ____ HC.OP-AB.RPV-0004(Q), Reactor Level Control. ____ HC.OP-AB.IC-0001(Q), Control Rod. ____ HC.OP-AB.IC-0004(Q), Neutron Monitoring. ____ 3.1.8. Values of Megawatts Elec tric (MWe), throughout this procedure are approximate. These values can be affected by Seasonal conditions and/or Plant conditions, such as Degraded Vacuum Operations. ____

3.2

3.2.1. WHEN a thermal power change exceeding 15% of rated thermal power occurs within a one-hour period, the Chemistry Department shall be notified to obtain the required samples as specified in Technical Specific ation 3/4.4.5, and the Radiation Protection Department shall be notified to obtain the required samples as specified in ODCM Table 4.11.2.1.2-1. ____ 3.2.2. The oxygen c oncentration limits of Technical Specification 3.6.6.2 shall be complied with. ____ 3.2.3. Vessel metal temperatures above and below the water level and Reactor Coolant System Temperat ure/Pressure Data should be monitored to ensure the TS Cooldown limits are NOT exceeded while raising Reactor Vessel Level. ____ 3.2.4. Technical Specification 4.6.

1.3.c Primary Containment Air Lock operability requirements (and it s associated note) shall be observed. ____ 3.2.5. With NO Reactor Recirculation Pumps in service, AND the Reactor is "Critical"; the Mode Switch shall be LOCKED in "Shutdown".[] ____ 3.3 3.3.1. The single rod scram test switches are intended for test purposes and should NOT be used to bypass the requirements

for banked control rod movement below the RWM low power

setpoints. These test switches are NOT to be used for power

control or rapid power reduction purposes.[

]____ 3.3.2. Directions from Reactor Engineering should be adhered to when any steps which require the mo vement of control rods are performed. ALL power changes should be done with directions

provided by Reactor Engineering or designated representative.[] ____ 3.3.3. IF immediate Reactor power reduction is required AND there is no dedicated reactivity plan THEN the Standard Power Reduction Instructions.

[] ____

3.3.4. IF the Crossflow Correction Factor is "Applied" / "Automatic" (Mode A) and is frozen, operation at Licensed Thermal Power Limit may continue for up to 24 hours2.777778e-4 days <br />0.00667 hours <br />3.968254e-5 weeks <br />9.132e-6 months <br /> with the frozen Correction

Factor. Within 24 hours2.777778e-4 days <br />0.00667 hours <br />3.968254e-5 weeks <br />9.132e-6 months <br />, one of the below actions must be

completed. The problem causing the freeze is resolved and the Correction Factor is unfrozen (i.e. remain in Mode A). ____

A manual Correction Factor is implemented (i.e. transition to Mode B). ____ The Correction Factor is toggled to "Not Applied" (i.e. transition to Mode C). ____

IF plant conditions change significantly during this period, THEN the validity of the frozen Correction Factor should be evaluated. ____ 3.3.5. When the OPRM's are "Oper able", Operation within the OPRM Enable Region of the Power to Flow Map will allow a Reactor Scram due to OPRM input to RPS. ____ 3.3.6. When the OPRM's are "Inoperable" AND operating in or near Region 2 of the Power to Flow Map, nuclear instrumentation

should be closely monitored for Reactor Core instability. ____ 3.3.7. Reactor operation shall be c onsistent with the Power to Flow Maps on Attachment 5. ____

3.3.8. WHEN repositioning IRM RANGE SELECT Switches, only one switch should be operated at a time. ____ 3.3.9. All IRM RANGE SELECT Switc hes should be in RANGE 10 prior to IRM insertion. ____ 3.3.10. WHEN reducing Thermal Power, the RWM Low Power Alarm Point (LPAP) should be reached by 17% power, but may be reached at a higher power level. The Low Power Set Point (LPSP) shall be reached by 8.6% power, but may be reached at a higher power level. ____

3.3.11. A rise in RPV Level could occur as the RPV depressurizes due to "flashing" in the Feedwater lines. This is caused when flow from the Feedwater system is no longer required to make up for

steam loss from the RPV, which a llows the Feedwater to cool at a slower rate than the RPV. As the RPV depressurizes, this

higher temperature water expands as it changes phase, causing

flow from the Feedwater system to the RPV.

If a Steam Bubble has formed in the Feedwater lines (as indicated by a sudden rise in RPV level), Feedwater Flow

should not be initiated until the Bubble has condensed.

The recovery of RPV level, in the absence of water loss from

steaming or letdown, would be an indication that the Bubble has condensed. ____ 3.3.12. The IRM/APRM Recorders have dual scales (0 - 40 and 0 -125), but are only configured for the 0 - 125 scale.

During Startup, the IRM signals are s ent to the recorder at either the 0 - 40 or the 0 - 125 scale, based on the position of the

IRM Range Switch. At power, the APRM signals are sent to the

recorders using the 0 - 125 scale only. The recorders do not

change scale, therefore, the 0 - 40 scales, when using the IRMs, will not be accurate on the recorders. ____

3.4

3.4.1. During

low flow conditions F eedwater flow to the Reactor should be maintained relatively constant to minimize thermal transients on the Reactor Vessel. Opening a bypass valve may be necessary to achieve steady Feedwater flow. ____ 3.4.2. To avoid thermal stress to the Feedwater Nozzles, maximum RWCU flow should be maintained,

WHEN a low Feedwater flow condition exists. ____ 3.4.3. A 150°F/hr Cooldown rate on the Main Turbine first stage shell temperature is NOT to be exceeded. ____ 3.4.4. The Mechanical Vacuum Pu mp(s) are NOT to be started OR operated if Reactor Thermal Power is above 5%.[] ____ 3.4.5. The Main Turbine s hould NOT be operated with exhaust pressure above the variable alarm setpoint. Under low-load conditions, exhaust pressure in excess of 4.0 Inches Hg Abs

should be avoided. ____ 3.4.6. This procedure does NOT r equire that the Reactor Building Sample Station Drains be diverted to CRW; however, if it is

deemed necessary to do so, Condenser Vacuum should be

monitored when repositioning 1-RC-V005. ____

3.4.7. At low loads, backpressure should be maintained 1.5" Hg Abs, (degraded vacuum) to mitigate shell and rotor distortion, which could result in a rub induced vibration condition. Degraded

vacuum should be established gradually over a 2 Hour period, at approximately 25% Reactor Power. Vibration should be closely

monitored when establishing degr aded vacuum operation. ____ 3.4.8. Extended low power operation with 3 or more Station Service Water pumps in service may result in overflowing the Cooling Tower Basin.

WHEN operating in this mode, the Cooling Tower Blowdown Flow should be monitored for extended High Flow conditions (Ex: PNL 10C604, 0SP-RI-4168)

AND the Cooling Tower Basin Level monitored locally. The SSW pump(s) should be secured as necessary. ___

3.5 3.5.1. During Rx depressurization, flashing may occur in the RWCU System piping, causing spurious Hi Delta Flow isolation signals or RWCU Pump trips on low flow to occur. ____ 3.5.2. During plant Cooldown/Depressu rization, similar Rx water level instrumentation should be monito red for significant deviation, indicating possible reference line de-gassing. Also, all

maintenance activities which have the potential for draining the Rx Vessel should be terminated. ____ 3.5.3. Excessive cooldown rates may be experienced with small amounts of decay heat present. Removal of loads from the Main

Steam Header

OR closing the MSIVs and using the Main Steam Line Equalizer Valve AB-HV-F020 will help to control cooldown rate. ____ 3.5.4. During plant Cooldown the following guidance should be adhered to in order to minimize shutdown radiation levels from CRUD release and transport: [] ____ A. Recirculation Pumps should be maintained in operation as long as possible in order to assist in CRUD Burst Cleanup ____ B. RWCU Filter Demin flow should be maximized to remove CRUD released during the cooldown. (90 gpm Demin flow (single pump ops) may be the max flow while

Depressurizing/Cooldown, due to suction venturi flashing causing inadvertent pump trips.) ____ C. Chemistry Department shoul d be notified of changes in plant condition that may reduce CRUD removal.

(i.e., RWCU flow changes) ____

3.5.5. Reactor

Pressure and/or Level control may be significantly challenged following a Reactor Shutdown in which the Plant Heat Loads exceed the Decay Heat generation. Isolation of the MSIVs and cycling of the SRVs may be required to control

Reactor Pressure. Consideration should be given to the

implementation of a Post Scra m Cooldown Strategy with low Decay Heat Load. ____ 3.6

3.6.1. During

plant start up, run t he Reactor Recirc Pumps at vessel head pressure for the minimal possible time. In addition, maintain pump speed as low as practical, avoiding speeds >30%

and oscillations. If plant conditions will result in extended pump

operation, greater than 24 hours2.777778e-4 days <br />0.00667 hours <br />3.968254e-5 weeks <br />9.132e-6 months <br />, then consideration should be given to removing the pumps from se rvice if a plant startup is not in progress. Plant Engineering should be consulted prior to exceeding this limit. ____ 3.6.2. Operation of the Reactor Reci rc Pump above 200 psig results in better seal operations. At approximately 200 psig the Reactor Recirc Pumps experience thrust changeover from lower thrust

shoes to upper thrust shoes. As reactor pressure is decreased

during shutdown, the plant should not be allowed to "hover" in this range. ____ Effective Core Flow shall be the core flow that would result if both recirculation loop flows

were assumed to be at the smaller value of the two loop flows.

3.6.3. Recirculation

loop flow mi smatch shall be maintained within:[

] A. 5% of rated core flow with effective core flow 70% of rated core flow. ____ B. 10% of rated core flow with effective core flow 70% of rated core flow. ____ 3.6.4. The time when neither the RHR System (operating in the Shutdown Cooling Mode) nor the Reactor Recirculation System is in operation should be minimized. ____

3.6.5. During

the transition from no rmal Reactor Recirculation System operations to establishment of Shutdown Cooling, only the AP201 Reactor Recirc Pump may be left in operation until the BP202 (only) RHR Pump is operating satisfactorily, and then

only until the required B RHR Loop flow of approximately 10,000 gpm is achieved. This limitation does NOT apply when Noble Metals Chemical Application is to be performed during plant shutdown. ____ 3.6.6. The discharge valve of any Reactor Recirculation Pump, which is NOT in operation, should remain closed throughout Shutdown

Cooling operations. IF it is required to stroke the discharge valve of an out-of-service Reactor Recirculation Pump, the pump's

suction valve should be verified to be closed

AND the suction valve's power supply breaker cleared and tagged open. ____ 3.6.7. While the RHR System is operating in the Shutdown Cooling Mode of operation, any valve mani pulations that would prevent ANY of the rated Shutdown Cooling Flow (approximately 10,000 gpm), from returning to the Reactor

Vessel via the respective Recirculation System discharge piping

and jet pumps are NOT to be performed. For example, recirculation suction and discharge valves being open

simultaneously on the loop seeing shutdown cooling return flow

would result in a portion of the return flow being diverted back through the Recirculation Loop in t he reverse direction, rather than into the respective Jet Pumps where forced circulation

through the core would occur. This limitation does NOT preclude intentionally reducing Shutdown Cooling flow to support Noble Metals Chemical Application. ____ 3.6.8. While the RHR System is operating in the Shutdown Cooling Mode of operation, maintaining t he rated shutdown cooling flow

to the Reactor Vessel via the respective Recirculation System

discharge piping and Jet Pumps is essential to assure that the

RHR Heat Exchanger inlet temper ature is representative of actual bulk coolant temperature. ____

3.6.9. WHEN the average Reactor coolant temperature is below 200°F, periods with the Reactor Vessel level 80 inches should be minimized, to ensure that natur al circulation will be immediately available IF forced circulation is lost or terminated for any reason. ____

3.7 3.7.1. Cold Shutdown IST Valve Testing should commence within 16 hours1.851852e-4 days <br />0.00444 hours <br />2.645503e-5 weeks <br />6.088e-6 months <br /> but must commence within 48 hours5.555556e-4 days <br />0.0133 hours <br />7.936508e-5 weeks <br />1.8264e-5 months <br /> of achieving Cold Shutdown, and continue until testing is complete or the plant is ready to return to power. There is NO requirement to keep the plant in Cold Shutdown solely to complete Cold Shutdown

Testing. For extended outages, testing need NOT begin in 48 hours5.555556e-4 days <br />0.0133 hours <br />7.936508e-5 weeks <br />1.8264e-5 months <br />, provided all valves r equired to be tested during Cold Shutdown will be re-tested before plant startup. ____ IF an outage lasts beyond 92 days, then all Cold Shutdown Testing shall be completed. Additionally, Cold Shutdown

Testing shall continue such that all applicable components

have been tested within the last 92 days of the shutdown. ____ WHEN an extended Cold Shutdown occurs which necessitates de-inerting the containment, then testing of

valves that require this condition is discretionary. The length

of the shutdown and the extent of other outage activities could be factored into a decision. ____ 3.7.2. As part of Station Blackout considerations, valves AB-HV-F020 AND AB-HV-F021 will be tagged in thei r required position during power operations due to their inaccessibility (Main Steam Line valves). [] ____ 3.7.3. All initiating actions in EHC include a confirmation message.

Initiating actions include valve testing, setpoint changes, resetting trips, etc. Specific direction to confirm the action is not included with each procedural step to perform an action. All terminating

actions in EHC do not require confirmation. Terminating actions

include stopping testing, terminating cooldown, adjusting with RAISE or LOWER pushbuttons, etc. ____ 3.7.4. The blowdown rate from the Reactor Water Cleanup (RWCU)

System should be limited to prev ent RWCU Filter/Demineralizer inlet temperature from exceeding 130°F. ____

3.7.5. IF this procedure is being performed in preparation for Refueling activities, THEN consideration should be given to performing controlled flushes of systems which have the potential to affect

Vessel Cavity clarity during refueling operations.

(i.e., Shutdown Cooling Loops). ____

3.7.6. IF this procedure is being performed in preparation for Refueling activities AND the plant is in Mode 4, THEN consideration should be given to defeating the sec ondary containment air lock doors (Rx 102' & 145') in order to prev ent damage to the doors, reduce the chance of injury and minimize transit times.

[] ____ 3.7.7. IF CRIDS is lost during the plant shutdown THEN Field Operators shall take continuous rounds AND log keeping. Normal rounds and log keeping can resume WHEN the plant is in Hot Shutdown.[] ____ None

5.1 When lowering load IAW this procedure DO NOT exceed a rate of change of 1% per minute unless the change is due to a single control rod movement and positioning the control rod at an intermediate position is not recommended by Reactor Engineering.

All power changes shall be done with directions provided by Reactor Engineering or a designated representative. Detailed directi ons from Reactor Engineering will be provided when performing any steps which require the movement of control rods. [

] The Main Turbine should NOT be operated with exhaust pressure above the variable alarm setpoint. Operation at low or minimum load should be performed at the best attainable exhaust pressure. Under low-load conditi ons, exhaust pressure in excess of

4.0 Inches

Hg Abs should be avoided.

5.1.1. the following steps have been completed before commencing a shutdown: A. System Operator notified of the shutdown. ____ B. Reactor Engineer notified of the shutdown. ____ C. Steam Lead Drain #4 (AC-HV-1018A) is in AUTO. ____

Recirculation loop flow mismatch shall be maintained within: [

] 1. 5% of rated core flow with effective core flow ** greater than or equal to 70% of rated core flow. 2. 10% of rated core flow with effective core flow** less than 70% of rated core flow. ** Effective core flow shall be the core flow that would result if both recirculation loop flows were assumed to be at the sma ller value of the two loop flows. If immediate Reactor power reduction is requir ed, and there is no dedicated reactivity plan, then implement the Standard Powe r Reduction Instructions. [

]

5.1.2. Reactor

power by reducing Reactor Recirculation Pump A and B speed IAW HC.OP-SO.BB-0002(Q), REACTOR RECIRCULATION SYSTEM OPERATION. ____

5.1.3.

the Load Setpoint at 100 % OR as directed by CRS. ____

5.1.4. WHEN the #4 CONTROL VALVE is Full Closed, the #4 STEAM LEAD DRAIN (HV-1018A) is OPEN. ____

5.1.5. WITH Feedwater flow < 15.27 Mlbm/hr ("Less Than 95% Flag" set)

OR consistent with Reactor Engineering guidance, the Crossflow Correction Factor to "Not Applied" IAW HC.RE-RA.ZZ-0011(Q). ____

Typically the first RFP removed from service for a planned maintenance outage has been designated for maintenance activities and may be removed from service IAW

HC.OP-SO.AE-0001(Q), Feedwater System Operat ion. However, with this pump out of service (Tripped - yielding a 2 of 3 low control oil signal), any subsequent transient causing RPV level to reach Level 4 (30") will enforce an Intermediate Recirc Runback.

5.1.6. At approximately 70% power the third RFP in Recirc. Operation

IAW HC.OP-SO.AE-0001(Q), Feedwater System Operation. ____

5.1.7. WHEN the Reactor Recirculation Pump speeds are between 45 and 50%, the Reactor Recirculation Pumps are in Individual Manual Control IAW HC.OP-SO.BB-0002(Q),

Reactor Recirculation System Operation. ____ 5.1.8. At approximately 50% power, H1CA -CA-HV-1991 is open to provide steam to the Steam Seal evaporator from Main Steam. ____

5.1.9. reducing

Reactor power as follows:

A. IF needed, PRIOR to reducing power below 40% of rated (507 MWe) the Throttle Pressure Set, Pressure Setpoint has been returned to normal (905 psig ) as follows: ____

1. Control , Pressure Control ____

2. IF needed, Ramp Rate AND desired rate. ____ 3. IF needed, Setpoint AND 905 psig. ____ 4.

Throttle Pressure Set, "Pressure Reference" is equal to "Pressure Setpoint". ____ B.

Reactor Recirculation Pump A AND B speed UNTIL minimum speed is reached ____ AND/OR C. control rods IAW Reactor Engineering Guidance. ____

At approximately 5000 gpm RFP flow to the ve ssel, flow oscillations could occur due to opening of RFP Minimum Flow Control Valves. This could cause RPV Level and Power perturbations.

Placing a second RFP in recirc before removing the first RFP placed in recirc from service (i.e., tripped - yielding a 2 of 3 low control oil signal) will ensure the Intermediate Recirc Runback circuit is not activated if the associat ed level transient causes RPV level to reach Level 4 (30").

Designating the RFP with the lowest di scharge flow to the vessel as the 2 nd RFP to be placed in Recirc operation reduces the amount of flow t he last in-service RFP will have to assume to maintain a steady feed rate - this will minimize the level transient.

SCP Minimum Flow Valves will begin to open when total feed and condensate flow lowers to 13,500 gpm (the sum of RFP flow to the vessel and RFP minimum flow); this will occur at approximately 30% load (380 MWe). This step contains actions at an approximate power value and may be performed earlier if SCP Minimum Flow Valve performance is challenging level control. Based on RFP capacity, this step should be performed at less than or equal to

38% power (482 MWe).

5.1.10. At approximately 30% power, (386 MWe) WHEN the RFP with the lowest discharge flow (flow to the vessel) approaches 5,500 gpm, THEN ,

the following:

A. the RFP operating with the lowest discharge flow to the vessel (the 2nd RFP) in Recirc operation IAW

HC.OP-SO.AE-0001 (Q), Feedwater System Operation. ____

B. one Secondary Condensate Pump A(B,C)P137 IAW HC.OP-SO.AD-0001(Q), Condensate System Operation (leaving two Secondary Condensate Pumps in service). ____

C. one Primary Condensate Pump A (B, C) P102 IAW HC.OP-SO.AD-0001(Q), Condensate System Operation (leaving two Primary Condensate Pumps in service). ____

D. one RFP previously placed in Recirc operation has been removed from service IAW

HC.OP-SO.AE-0001 (Q), Feedwater System Operation. ____

E.

1 RFP in service AND 1 RFP in Recirc operation IAW HC.OP-SO.AE-0001(Q), Feedwater System Operation. ____ (Continued on next page)

5.1.10 (Continued)

F.

that the "SINGLE ELEMENT CONTROL", block on the DFCS Main Screen #1 is illuminated yellow. ____ G.

I&C and Radiation Protection to restore the MSL Rad Monitor Trip and Alarm Setpoints to normal. ____ When the TCV fast closure and MSV Trip By pass Annunciator alarms, a scram may still be possible from the TCVs or MSVs. This alarm annunciates whenever any of the four channels

monitoring first-stage turbine pressure drops bel ow the setpoint. The CRIDS Digital Points (D3467 through D3470) which monitor continuity of the individual logic trains should be checked to determine when the scram function of the TCVs and MSVs is actually bypassed.

5.1.11. At approximatel y 30% Rated Power:

A.

the ROD BLOCK MONITOR, RBM A and B BYPASS light is ON. ____ B. To minimize the potential for RWCU pump trips on plant

shutdown, the following:

1.

Chemistry to reduce RWCU System flow to < 90 gpm with one Demineralizer in service while monitoring suction flow to prevent a trip on low flow. ____

2. either A or B RWCU Pump from service. ____

C. IF not performed in the previous 92 days, Main Turbine Lift Pump test IAW HC.OP-FT.AC-0003(Q). ____ Secondary Condensate Pump (SCP) Min Flow Valves will cycle open and closed when any

SCP flow lowers to an allowable band of 4500 to 5100 gpm. This occurs at approximately 22% power. In order to avoid Min Flow Valve cycling and corresponding Reactor Level

swinging, RFP Min Flow Valves can be taken to MANUAL and opened to achieve a SCP flow

value above 5500 gpm.

D. a RFP Min Flow Valve controller in MANUAL AND Min Flow to achieve 3500 gpm (or as directed by the CRS).

5.1.12. At approximately 25%

power (approx. 317 MWe),

Condenser backpressure 1.5" Hg Abs, over a 2 Hour period using one or both of the following procedures: HC.OP-SO.CG-0001(R), Condenser Air Removal System Operation. ____ HC.OP-SO.DA-0001(Z), Circulating Water System Operation ____

Continued Next Page

Plant shutdown from >20% power will NOT support Turbine Testing per HC.OP-FT.AC-0004(Q), if it is required.

5.1.13. IF directed by the Operations Dir ector to Lock the Mode Switch in Shutdown from between 30% and 20% power with the Main Turbine still on line, THEN

the following:

A.

an operator for local observation, AND operation of the feedwater Startup Level Control Valves by performing the following steps to stroke the Startup Level Control Valves:

1.

the "INS" pushbutton as necessary to select POSN DEMAND on STARTUP LEVEL

CONTROLLER ____ 2. START UP LEVEL CONTROLLER is in "M" (manual). ____ 3.

LV1785 ON pushbutton. ____ 4. Intermittently INCREASE pushbutton on STARTUP LEVEL CONTROLLER UNTIL POSN DEMAND indicator is at 100%.

5.

LV-1785 CLOSE PB to close the Startup Level Control Valves in preparation for Shutdown. ____

B. Main Turbine Oil Pumps in service, the lube oil temperature controller, AND the power system stabilizer IAW HC.OP-SO.AC-0001(Q) Section for shutting down the Main Turbine. ____

Continued Next Page

5.1.13 (continued)

C.

the following Steps at the current power level (all other intermediate steps should be N/A): 1. 5.1.22 for the EOC RPT system. ____ 2. 5.1.27 for opening the FWH 1&2 vents. ____ 3. 5.1.28 for Pressure Setpoint Adjustment. ____ 4. 5.1.29 for performing a Manual Scram. ____ If a power reduction event occurs so that reac tor power is < 20 percent, Control rod motion (except for scram or other emergency condition) SHALL BE PROHIBITED UNTIL the MSL Rad Monitor Trip and Alarm Setpoints have been returned to normal.

5.1.14.

the MSL Rad Monitor Trip and Alarm Setpoints have been returned to normal PRIOR to decreasing core thermal power 20%.[] ____ 5.1.15. At approximatel y 17% Rated Power, (215 MWe) the following:

A.

that the Low Power Alarm Point (LPAP) on the RWM is reached as follows: ____

1. the MAIN_1 display on the DFCS Console. ____ 2. As indicated Steam Flow decreases to < 2.23 Mlb/hr on the DFCS Console THEN RWM Power Indication changes from "POWER:ABOVE LPAP" to "POWER:TRANSITION" at the RWM Display screen. ____

Continued Next Page

5.1.15 (continued)

The following step is required to be performed within 8 hours9.259259e-5 days <br />0.00222 hours <br />1.322751e-5 weeks <br />3.044e-6 months <br /> PRIOR to RWM automatic initiation when reducing thermal power below the LPSP. [

] If all the control rods in the currently latched step are at the initial or final positions for that

step, then the RWM is at a boundary between adjacent steps. At a step boundary, a selection error is not generated when a control rod in either of the adjacent steps is selected.

In such cases, a control rod must be sele cted from a step other than those adjacent ones.

RWM insert and withdraw blocks are indicated but are NOT enforced in the"transition" zone.

B.

any control rod that is NOT in the currently latched step of the RWM (or an adjacent step if at a boundary) AND

the following: ____ 1.

the below selected indications at the RWM Operators Display: "SR XX - YY : ZZ" where XX - YY is the selected rod and ZZ is its current position ____ "SE" which indicates a selection error ____ "IB" which indicates an insert block (not shown if at 00) ____ "WB" which indicates a withdraw block (not shown if at 48) ____

2. date and time.

Date / Time 5.1.16. At 20% Rated Power (or less) and, IF it is desired to continue Reactor Cooldown AND Depressurization in preparation for Refueling activities, THEN Maintenance Department to commence Reactor Cavity Shield Plug Removal IAW HC.MD-FR.KE-0035(Q), Reactor Pressure Vessel Disassembly. ____

The control rod pattern should be re-established PRIOR to reaching the low power setpoint on the RWM. Failure to do this may result in insert and/or withdraw blocks.

When reducing thermal power, RWM low power se tpoint is nominally reached at 15% power (190 MWe).

PRIOR to load reduction below 8.6% power (109 MWe), automatic initiation of RWM shall be verified by performance of Step 5.1.17. These steps must be completed prior to any control rod movement after the RWM "POW ER" indicates POWER: BELOW LPSP.

5.1.17. WHEN the low power setpoint (LPSP) on the RWM is reached the following:

A. the RWM "POWER" indicates "POWER:BELOW LPSP" ____

OR RWM "POWER" indication changes from

"POWER:TRANSITION" to "POWER:BELOW LPSP". ____

Continued Next Page

5.1.17 (continued)

The following step shall be performed within 1 hour1.157407e-5 days <br />2.777778e-4 hours <br />1.653439e-6 weeks <br />3.805e-7 months <br /> AFTER RWM automatic initiation below the LPSP IAW T/S 4.1.4.1.c. If all the control rods in the currently latched step are at the initial or final positions for that step, then the RWM is at a boundary between adj acent steps. At a step boundary, selection errors are not generated when control rods in eit her of the adjacent steps are selected. In such cases, a control rod must be select ed from a step other than those adjacent ones.

B.

any partially or fully inserted control rod that is NOT in the currently latched step of the RWM (or the adjacent step if on a boundary)

AND the following steps:

1.

the below selected indications at the RWM Operator's Display. "SR XX - YY : ZZ" where XX - YY is the selected rod and ZZ is its current position ____ "SE" which indicates a selection error ____ "IB" which indicates an insert block (not shown if at 00) ____ "WB" which indicates a withdraw block ____

2.

to withdraw the control rod AND that there is no control rod movement. ____ 3. date and time.

Date / Time

5.1.18. At approximately 13% power (165 MWe),

the following: RFP Minimum Flow Valves begin to open at 5000 gpm total RFP flow for pump protection.

To prevent Level / Power fluctuations caused by RFP Minimum Flow Valve operation, the in-service RFP Minimum Flow Valve is placed in MANUAL at 3500 gpm BEFORE the in-service RFP discharge flow to the vessel goes below 5000 gpm.

The RFP Woodward governor's calibrated lower control band is 650 rpm (1500 gpm equivalent). To prevent level fluctuations caused by RFP Woodward Governor performance, power should not be lowered in subsequent steps below that which would cause the in-service Feed Pump's discharge flow to the vessel to go below 2000 gpm WITH its associated minimum flow valve in manual at 3500 gpm.

Acceptable flow for Master Level Control (2000 gpm feed flow to the Reactor) is expected to be maintained down to the following approximate indications of power: 13% as indicated on the APRMs, or 100 MWe Generator Load, or 21/2 BPVs open This is intended to maintain Master Level Control operation through removing the main turbine/generator from the grid fo r better overall level control.

A. WHEN Reactor Feed Pump A(B, C) Discharge Flow (flow to Vessel) reaches approximately 5500 gpm during shutdown, THEN , this Reactor Feed Pumps' Minimum Flow Valve in Manual to achieve 3,500 gpm minimum flow IAW HC.OP-SO.AE-0001(Q), Feedwater System Operation. ____ Continued Next Page

5.1.18 (continued)

B. the following: 1. AC-HV-1041/42/43 (A,B,C)CROSS AROUND (1 push button) ____ 2. AF-HV-1373 A, B, C (FWH #3 SHELL SIDE)-

EXTR LINE DRAINS (3 push buttons) ____ 3. AF -HV-1388 A, B, C (FWH #3 SHELL SIDE)-

EXTR LINE DRAINS (3 push buttons) ____ 4. AF -HV-1355 A, B, C (FWH #4 SHELL SIDE)-

EXTR LINE DRAINS (3 push buttons) ____ 5. AF -HV-1377 A, B, C (FWH #4 SHELL SIDE)-

EXTR LINE DRAINS (3 push buttons) ____ 6. AF-HV-1387 A, B, C (FWH #5 SHELL SIDE)- X-AROUND STM LINE DRAIN (3 push buttons) ____ 7. AF-HV-1359 A, B, (FWH #6 SHELL SIDE)-

EXTR STM DRN VLVS (2 push buttons) ____ C.

the following valves auto open: 1. AB-HV-F033 CTMT INBD STM LNS/MN STM LINE AFT STOP V DRN HDR-DRN HDR OP DRN V. ____ 2. AB-HV-F069 STEAM LINE BEFORE STOP VALVE DRAINS-DRN HDR OP DRN VLV. ____ 5.1.19. the following valves:

A. AB-HV-1026 STM LEAD S/U (1 push button) ____ B. AC-HV-1013 A,B,C,D MN STM VLV BFR SEAT (1 push button) ____ C. AC-HV-1015 CONT VLV BFR SEAT (1 push button) ____ D. AC-HV-1017A/B STEAM LEAD 1&2 (1 push button) ____ E. Steam Lead Drain AC-HV-1018B Steam Lead 3 ____ 5.1.20.

Steam Lead Drain AC-HV-1018A Steam Lead 4 is open. ____

5.1.21.

the appropriate sections of HC.OP-FT.AC-0004(Q);

Main Turbine Functional Test - Refueling, as required:

A. IF required to implement R egular Maintenance Plan 14852, THEN applicable sections of the procedure IAW Outage scheduling requirements. ____

B. IF required to 24 month test frequency is NOT exceeded AND no maintenance work is scheduled to be performed on the Front Standard, THEN

POST and EOST Offline tests. ____

5.1.22.

theEOC Recirc Pump Trip System is BYPASSED as follows: A. RECIRC PUMP TRIP DISABLE SYSTEM "A", Switch C71A-S12A, to BYP. (10C609) ____

B. RECIRC PUMP TRIP BYPASS DISABLE SYSTEM "B", Switch C71A-S12B, to BYP. (10C611) ____ It is recommended to unload the Turbine-Generat or from 15% to 5% of rated load (190 to 63 MWe OR lower) and trip the turbine within a total time of 45 minutes. This will maintain the required low pressure turb ine temperature during shutdown and Turbine-Generator unloading.

5.1.23. the Main Turbine/Generator from the grid IAW HC.OP-SO.AC-0001(Q), Main Turbine Operation. ____ Acceptable flow for Master Level Control (2000 gpm feed flow to the Reactor) is expected to be maintained down to the following approximate indications of power: 13% as indicated on the APRMs, or 100 MWe Generator Load, or 21/2 BPVs open This is intended to maintain Master Level Control operation through removing the main turbine/generator from the grid fo r better overall level control.

5.1.24.

Feedwater Control from Master Level Control to Startup Valves IAW HC.OP-SO.AE-0001(Q). ____

5.1.25.

an IRM/APRM overlap at 10% power (average APRM reading ) as follows: [] ____ A. all IRM RANGE SELECT Switches to position 10. ____ All IRM RANGE SELECT Switches should be in RANGE 10 prior to IRM insertion B.

the IRM Detectors to the full in position IAW HC.OP-SO.SE-0001(Q), Nuclear Instrumentation System Operation. ____

C. that the IRM and APRM channels overlap for at least 1/2 decades by verifying the following: ____ indicate 50 on range 10 4% (Downscale Setpt). CHANNEL INITIAL CHANNEL INITIAL A A B B C C D D E E F F G H GETARS will be re-booted and placed in "SENTINEL MODIFIED" using Work File 14 to allow

for Data collection in SENTINEL while the Main Generator Output Breakers are open. This will allow for data collection following the manual Rx Scram later in this procedure. The

Turbine Trip Limit Check (as sensed by Turb ine Generator Output breaker position) is removed from Work File 14.

5.1.26. GETARS in "SENTINEL MODIFIED" using Work File 14. ____

5.1.27. AF-HV-1459 A,B,C HTRS 1 & 2/DC, S/U AND OPR VENTS ____ IF Reactor Engineering will collect control r od scram time data during manual scram in Step 5.1.29.D, Reactor pressure will be required to be 950 psig prior to the manual scram.

5.1.28. IF necessary to support plant testing, THEN Throttle Pressure Set as follows:

A. Control , Pressure Control ____

B. Throttle Pressure Set Ramp Rate AND as desired. ____

C. Throttle Pressure Set Setpoint AND desired setpoint OR Throttle Pressure Set, Manual Adj. Raise , Lower as needed. ____ Step 5.1.29 is to be performed only if di rected by the Operations Director; OTHERWISE Step 5.1.29 is to be N/A'd.

IF AB-HV-F020, AB-HV-F021, BG-HV-F034 and BG-HV-F035 will be required for Pressure and Level control, Step 5.1.

29 should not be performed.

5.1.29. IF directed by the Operations Dir ector to Lock the Mode Switch in Shutdown, THEN

the following:

A. IF control rod scram time data will be collected by Reactor Engineering, THEN Reactor pressure is 950 psig. ____

B. Attachment 2 AND

Attachment 10 prior to locking the Mode Switch to Shutdown. ____

SM/CRS C. IF it is desirable to continue reactor cooldown, Reactor Coolant System Cooldown rate, not to exceed 90 F/Hr, IAW Attachment 4 of this procedure and s of HC.OP-DL.ZZ-0026(Q), Surveillance Log.

[]. ____ (Continued on next page)

5.1.29 (continued)

Following a "Manual Reactor Scam", a Reactor water level 3 (12.5" RPV Lvl.) RPS signal is

expected to occur.

D.

the Mode Switch in Shutdown, AND HC.OP-AB.ZZ-0000(Q). ____ E. IF adjusted for testing, THEN Throttle Pressure Set to 905 as follows: 1. Control , Pressure Control ____

2. Throttle Pressure Set Ramp Rate AND as desired. ____

3. Throttle Pressure Set, Setpoint AND 905 psig OR

Throttle Pressure Set, Manual Adj Raise / Lower as needed. ____

F. IF it is desired to continue Reactor Cooldown AND Depressurization in preparation for Refueling activities, THEN Maintenance Department to remove Reactor Cavity Shield Plugs IAW HC.MD-FR.KE-0001(Q), Refuel Floor Shield and Pool Plugs Removal and

Replacement, OR HC.MD-FR.KE-0035(Q). ____ Post Trip Review should be commenced as soon as possible after the plant is stabilized to

prevent possible loss of Alarm Chronolog data.

G.

the STA to commence Post Trip Review IAW OP-HC-108-114-1001 and OP-AA-108-114. ____

H. in this procedure at Step 5.1.37. ____

The RPS Mode Switch SHALL be placed in STARTUP & HOT STBY IAW Steps 5.1.30 through 5.1.37 prior to APRM indication decr easing to the Downscale setpoint (4%).

5.1.30. inserting control rods to reduce power to between 6 and 9%. ____

5.1.31. all Operational Condition 2 surveillance items in HC.OP-DL.ZZ-0026(Q) have been initiated. ____

5.1.32.

the RECORDER INPUT IRM A,B,C,D,E,F,G,H PB's to

transfer the IRM/RBM/APRM Recorders to the IRM indication. ____

5.1.33. IRM RANGE SELECT Switches are positioned so all IRM instruments read between 25 and 75 (on the 0 to 125 scale). ____

5.1.34. the IRM drawers are NOT INOP, (At panels 10C635 and 10C636, each IRM drawer for NO IRM trip condition or INOP light) OR if they are INOP that they are BYPASSED. ____

5.1.35. Attachment 1 AND Attachment 10 PRIOR to placing the Mode Switch to STARTUP & HOT STBY in the following step. ____

SM/CRS With the RPS mode switch in STARTUP & HO T STBY, an APRM rod block occurs at 11%

AND an APRM scram occurs at 14%.

5.1.36. the RPS MODE SWITCH to STARTUP & HOT STBY. ____

5.1.37.

the following: tagout on the Power Supply for AB-HV-F020 AND AB-HV-F021. ____

BG-HV-F031 RWCU FLOW ORIFICE BYPASS for blowdown operation as follows:

1.

an Operator to TAG open breaker 52-264042 (BG-HV-F031) ____ 2.

Electrical Maintenance to perform . [] ____ 3. WHEN Notified by Electrical Maintenance, an Operator to RELEASE breaker 52-264042. (BG-HV-F031) ____

5.1.38. IF a Hot Standby condition is to be maintained, THEN to HC.OP-IO.ZZ-0007(Q). ____

5.2 Control

rod insertion and cooldown of the R eactor Coolant System can be performed simultaneously. When this occurs, the cooldow n rate and neutron flux should be closely monitored for any sudden changes.

IF control rod insertion is stopped prior to a ll rods being inserted, re-criticality must be anticipated due to cooldown. The Reactor Operator shall NOT have any other concurrent duties during this evolution.

During plant Cooldown/Depressurization, simila r Rx water level instrumentation should be monitored for significant deviation, indica ting possible reference line de-gassing. Also, all maintenance activities which have the potential for draining the Rx vessel should be terminated.

5.2.1. to reduce Reactor power by inserting control rods. ____

5.2.2.

the IRM flux between 25 and 75 (on the 0 to 125 scale) by repositioning the IRM RANGE SELECT Switches. ____ 5.2.3. As required, a Circulating Water Pump IAW HC.OP-SO.DA-0001 (Z), Circulating Water System Operation. ____

5.2.4. Reactor

Coolant System C ooldown rate, not to exceed 90 F/Hr, IAW Attachment 4 of this procedure and s of HC.OP-DL.ZZ-0026(Q), Surveillance Log.

[]. ____ 5.2.5. As Reactor power decreases, the SRM count rate between 10 2 and 10 5 cps by inserting the SRM detectors IAW HC.OP-SO.SE-0001 (Q), Nuclear Instrumentation System Operation. ____

5.2.6. PRIOR

to Locking the Mode Switch in Shutdown in the following step, Attachment 2 AND Attachment 10. ____

SM/CRS The actions in Step 5.2.7 should be completed at that point in the plant shutdown where all control rods are fully inserted.

5.2.7. WHEN all control rods have been fully inserted, the following: The RHR Shutdown Cooling operability requirement s of T/S 3.4.9 shall be complied with.

The following step will result in a Reactor scram.

A. the RPS MODE SWITCH in SHUTDOWN. ____ B. Following the 10 second time delay, the scram IAW HC.OP-SO.SB-0001(Q),

Reactor Protection System Operation. ____

Steam Loads, Decay Heat, and Feed will directly affe ct Cooldown / Depressurization. Impact of these variables, regardless of DEHC C ontrol mode selected, MUST be continuously evaluated for impact on the cooldown.

At approximately 200 psig reactor pressure , the cooldown rate should be limited to approximately 30º F/hr to prevent ex cessive cavitation of the RWCU pump.

5.2.8.

a cooldown rate of 90°F/hr using Rx Cooldown mode, Pressure Control mode OR Bypass Valve Manual Jack as follows:

A. Establish as follows:

1. Control , Pressure Control ____

2. Throttle Pressure Set Ramp Rate AND desired rate. ____ 3. Throttle Pressure Set, Setpoint AND desired Pressure to match Throttle Press. ____

4.

expected valve response as Pressure Reference changes to match Pressure Setpoint. ____

5. IF desired to continue cooldown using Pressure Control mode, Ramp Rate and Pressure Setpoint as desired. ____

Continued next page

5.2.8 (Continued)

When Rx Cooldown mode is initiated with a bypass valve open, a minor Pressure Rise will occur. This pressure rise should be antic ipated when placing Rx Cooldown controller in service. Any cooldown that has occurred since the shutdow n must be considered prior to establishing Rx Cooldown mode in determining initial c ooldown so as NOT to exceed 90°F/hr.

Once Rx Cooldown mode is established, the INTENT is to remain on the Rx Cooldown controller for the duration of the Cooldown / D epressurization. An In-Progress Cooldown can be interrupted to support plant manipulations without exiting the Rx Cooldown mode by establishing the temperature Setpoint at the desired hold point on the Cooldown Controller.

B. IF desired, Establish as follows:

1. Control , RX Cooldown ____

2.

Ramp Rate AND desired rate not to exceed 90 deg F/hr. ____

3. Temperature AND

desired temperature. ____

4. Reactor Cooldown ON AND Rx Cooldown Controlling indication is observed. ____

5.

Throttle Pressure Set, Pressure Setpoint approximately 50-100 psig above Throttle Pressure not to exceed 905 psig. ____

6. IF desired to Interrupt Cooldown, THEN Cooldown Temperature AND Temp Setpt to match indicated "Calc Rx" temperature. ____

7. WHEN desired to Re-establish Cooldown, THEN Cooldown Temperature AND desired Temp Setpt. ____ Continued next page

5.2.8 (Continued)

Should it become necessary to transition from mode to mode with BPV's initially open, the following response should be anticipated: BPV's will immediately close due to the contro l logic resulting in a minor pressure rise. BPV's will then re-open to stabilize pressure after a short time delay.

Initially, when Rx Cooldown goes to off, "BPV Manual Jack in Control" will be displayed until

"Throttle Pressure Ref Controlling" takes control.

C. IF NECESSARY to transition from to Mode, THEN Establish Mode as follows:

1. Cooldown Temperature AND Temp Setpt to match indicated "Calc Rx" temperature

AND allow conditions to stabilize. ____ 2. Control , Pressure Control ____ 3. Throttle Pressure Set Setpoint AND desired Pressure to match Throttle Pressure. ____

4. WHEN Pressure Reference is equal to Pressure Setpoint THEN Control , RX Cooldown ____

5. Reactor Cooldown OFF ____

6. Control , Pressure Control AND BPV Control Status indicates "Throttle Pressure Ref Cont rolling" after a short time delay. ____

Continued next page

5.2.8 (Continued)

D. IF desired, Establish as follows while maintaining Pressure Setpoint 50-100 psi above actual Throttle

Pressure. Bypass Valve Manual Opening (Jack) Control, Manual Adj. Raise / Lowerresponse is based on Ramp Rate selected. Approximate response as follows: Ramp Rate % 10 20 30 40 50 60 70 80 90 100 BPV Jack Setpoint % 0.2 0.3 0.5 0.7 0.8 1.0 1.1 1.3 1.5 1.7

1.

Bypass Valve Manual Opening (Jack) Control Ramp Rate AND as desired. ____

2.

Bypass Valve Manual Opening (Jack) Control Setpoint AND as desired OR Bypass Valve Manual Opening (Jack) Control Manual Adj. Raise / Lower as required. ____

3. Bypass Valve Jack is NO LONGER REQUIRED ,

BPV Jack Setpoint is lowered to (minus) -0.5%. ____ 5.2.9. IF it is desired to continue Reactor Cooldown AND Depressurization in preparation for Refueling activities, THEN Maintenance Department to remove Reactor Cavity Shield Plugs IAW HC.MD-FR.KE-0001(Q), Refuel Floor Shield and Pool Plugs Removal and Replacement,

OR HC.MD-FR.KE-0035(Q). ____

5.2.10. At approximatel y 500 psig, (approx 470 F) the following:

A. the remaining RFP's from service IAW HC.OP-SO.AE-0001(Q), Feedwater System Operation. ____

B. the second Secondary Condensate Pump A(B,C)P137 IAW HC.OP-SO.AD-0001(Q), Condensate

System Operation. ____ The preparation of the RHR System for Shut down Cooling Operation should be performed

while the plant cooldown is continuing.

RHR Loop B is preferred for Shutdown Cooling due to its Radwaste connection.

C. RHR Loop A or B for Shutdown Cooling Operation IAW HC.OP-SO.BC-0002(Q), Decay Heat Removal Operation. ____

D.

the RWCU system suction path to the Bottom Head Drain, IAW HC.OP-SO.BG-0001(Q), Section 5.13. ____ Attachment 9 is to be performed when an increase in the cooldown rate of the Main Turbine

Shell is desired, and used only during a "Controlled" shutdown (NOT following a scram), as a time-saving measure. The inferences to "C ooling/Cooldown", or "Warming", are dependent upon whether the direction is referring to the activity of cooling or the nomenclature on the

instrumentation/indications.

5.2.11. IF desired, THEN

Cooldown of the HP Turbine Shell using Attachment 9 - Main Turbine Shell Cooldown. ____

5.2.12.

I&C to adjust the CRD flow controller H1BF -1BFFIC-R600-C11 per the ICD card. ____

5.2.13. the HWCI System from service IAW HC.CH-SO.AX-0001 (Q). ____

5.2.14. Prior to reaching 300 psig, (approx 421 F) the following:

A.

the SJAE Steam Supply from Main Steam to Auxiliary Steam IAW HC.OP-SO.CG-0001(R)

OR , SJAE from service AND MVPs in service IAW HC.OP-SO.CG-0001(R). ____

B. IFthe SJAE is to remain in service on Auxiliary Steam, the Recombiner Preheater Steam Supply from Main Steam to Auxiliary Steam IAW HC.RW -SO.HA-0001(R), Gaseous Radwaste System Operation. ____ 5.2.15. At approximately 200 psig, cooldown rate to 30º F/hr or less to avoid cavitation of the RWCU pump 5.2.16. WHEN the PRESSURE is reduced to 150 psig, (approx 365 F) IF NOT required, a Primary and Secondary Condensate Pump IAW HC.OP-SO.AD-0001(Q), Condensate System Operation. ____ 5.2.17. At approximatel y 100 psig, (approx 328 F) the HPCI System isolates. ____ 5.2.18. At approximatel y 80 psig, (approx 323 F) the following:

A. RHR Loop A OR B has been prepared to be prewarmed (for Shutdown Cooling operation)

IAW HC.OP-SO.BC-0002(Q), Decay Heat Removal Operation. ____

B.

RHR Loop A OR B for Shutdown Cooling Operation IAW HC.OP-SO.BC-0002(Q). ____

5.2.19. that the RPS MODE SWITCH is Locked in SHUTDOWN. ____

5.2.20. IF Noble Metals Chemical Application (NMI) will be performed during plant shutdown, THEN the following IAW HC.DE-SP.ZZ-0001(Q),

Noble Metals Chemical Addition-Infrequently Performed Evolution (IPTE): ____

A.

both Reactor Recirc Pumps speed, as required by the IPTE. ____ RHR Loop B is the preferred loop to be placed in service. RHR Loop A may be placed in service if Loop B is unavailable OR if necessary to support outage scheduling.

B. RHR Loop B or A in Shutdown Cooling Operation, at the flowrate required by the IPTE,

IAW HC.OP-SO.BC-0002(Q), Decay Heat Removal Operation. ____ C. GO TO Step 5.2.25. ____

RCIC System isolates on a RPV Pressure of 64.5 psig after a 4 second TD.

RCIC Turbine trips on a Reactor level of 54".

If the RCIC System is still required for Level

/Pressure Control then Steps 5.2.21 - 5.2.25 should be performed prior to reducing pressure below 65 psig.

Main Turbine Sealing Steam will automatically trans fer from Main Steam to Auxiliary Steam at approximately 60 psig.

5.2.21.

the Reactor Recirculation System as follows:

A. IF RHR Loop A will be used for Shutdown Cooling, THEN both Reactor Recirc Pumps IAW HC.OP-SO.BB-0002(Q). ____

B. IF RHR Loop B will be used for Shutdown Cooling, AND it is NOT desired to maintain a Reactor Recirc Pump in service until rated Shutdown Cooling flow is established, THEN both Reactor Recirc Pumps IAW HC.OP-SO.BB-0002(Q). ____

C. IF RHR Loop B will be used for Shutdown Cooling, AND it is desired to maintain a Reactor Recirc Pump in service until rated Shutdown Cooling flow is established,

THEN BP201 Reactor Recirc Pump IAW HC.OP-SO.BB-0002(Q),

AND AP201 running to provide forced core flow. ____

Step 5.2.22 is to be performed only if Shutdown Cooling can NOT be placed in service AND Reactor Recirculation Pumps are NOT available; otherwise Step 5.

2.22 is to be disregarded and performance conti nued with Step 5.2.23.

5.2.22. IF Shutdown Cooling can NOT be placed in service AND Reactor Recirc Pumps are NOT available, THEN slowly Reactor Vessel level to 80 inches, Reactor level shutdown range, using temperature-compensated indication, (Vessel Level Instrumentation Temperature

Compensation Curves may be required), to allow for natural

circulation, WHILE monitoring Reactor Coolant System Temperature/Pressure Data IAW Attachment 4 so as NOT to exceed cooldown rate.[] ____ If Shutdown Cooling becomes unavailable, the plant may be placed in Alternate decay heat removal IAW Attachment 6.

5.2.23. Based on the decisi on made in Step 5.2.21, RHR Loop A or B in Shutdown Cooling Operation to maintain a cooldown rate 90°F/hr IAW HC.OP-SO.BC-0002(Q),

Decay Heat Removal Operation. ____

5.2.24. WHEN RHR is in Shutdown Cooling at rated flow (approximately 10,000 gpm), IF the AP201 Reactor Recirc Pump is in service, THEN the AP201 Reactor Recirc Pump IAW HC.OP-SO.BB-0002(Q). ____

5.2.25. plotting cooldown using the appropriate RHR Heat Exchanger inlet temperature. ____ 5.2.26. At approximatel y 64.5 psig, (approx 311 F) the RCIC System isolates. ____

5.2.27. At approximatel y 50 psig , (approx 298 F) AND when RHR cooling is established, the following: A. Reactor pressure setpoint matched to current throttle pressure. ____ B. Rx Cooldown Control OFF selected if utilized. ____ C. IF desired, Bypass Valve Manual (Jack) Control may be used to continue cooldown below 50 psig as follows:

1.

Bypass Valve Manual Opening (Jack) Control Ramp Rate AND as desired. ____

2.

Bypass Valve Manual Opening (Jack) Control Setpoint AND as desired OR Bypass Valve Manual Opening (Jack) Control Manual Adj. Raise / Lower required. ____ 3. Bypass Valve Jack is NO LONGER REQUIRED ,

BPV Jack Setpoint is lowered to (minus) -0.5%. ____ 5.2.28. At approximately 25 psig, IF open, THEN the Turbine Bypass Valves by BPV Jack Setpoint to (minus) -0.5%. ____ As the Reactor pressure approaches 0 psig, the RWCU System becomes susceptible to flashing and differential flow isolation and RWCU Pu mp trips. This condition can persist until reactor inventory becomes subcooled.

Flashing can be prevented by reducing RWCU System flow, by slowly reducing Reactor pre ssure and by preventing the RPV from reaching vacuum conditions.

5.2.29. WHEN Reactor Pressure is in the range of 10 to 50 psig, THEN MSIVs IAW HC.OP-SO.AB-0001(Q), Main Steam System Operation. ____ 5.2.30. At approximately 5 psig THEN AB-HV-F016 CTMT INBD STM LINE DRAIN HDR ISLN INBOARD. ____

5.2.31. HP Turbine Shell Cooldown using Attachment 9 - Main Turbine Shell Cooldown. ____

5.2.32. tags AND breakers ready for the following valves per SM/CRS direction: A. BB-HV-F001 Reactor Head Vent. ____

B. BB-HV-F002 Reactor Head Vent. ____ C. AE-HV-F011A, B Inboard Feedwater Isolation.[] ____ 5.2.33. WHEN the Reactor coolant temperature is < 212 °F, THEN the following (10C651C):

A. the second RWCU Pump in service at approximately 90 gpm IAW HC.OP-SO.BG-0001(Q),

Reactor Water Cleanup System Operation. ____

B.

Chemistry to place 2nd RWCU Demineralizer in service at approximately 90 gpm. ____

C. the following (MAIN STEAM LINE DRAINS AND VENTS) valves are closed: 1. AB-HV-F019 CTMT INBD STM LINE DRAIN HDR ISLN OUTBOARD. ____ 2. AB-HV-F016 CTMT INBD STM LINE DRAIN HDR ISLN INBOARD. ____ 3. AB-HV-F021 CTMT INBD STM LNS/MN STM LINE AFT STOP V DRN HDR-DRN HDR S/U DRN V. ____ 4. AB-HV-F033 CTMT INBD STM LNS/MN STM LINE AFT STOP V DRN HDR-DRN HDR OP DRN V. ____ 5. AB-HV-F072 STEAM LINE BEFORE STOP VALVE DRAINS-DRN HDR S/U DRN V. ____ 6. AB-HV-F069 STEAM LINE BEFORE STOP VALVE DRAINS-DRN HDR OP DRN V. ____

D. AC-HV-1013 A/B/C/D TURBINE SEALING STEAM AND DRAINS STEAM LINE DRAINS-MN STM VLV BFR SEAT is closed. ____

Continued next page

5.2.33 (continued)

E.

the following MAIN STEAM LINE DRAINS AND VENTS:

1. BB-HV-F005 REACTOR HEAD VENT, STM LINE A. ____ 2. BB-HV-F001 REACTOR HEAD VENT, CRW INBD ISLN. ____ 3. BB-HV-F002 REACTOR HEAD VENT,

CRW OTBD ISLN. ____

F. IF necessary, Reactor Vessel level with the RWCU System IAW HC.OP-SO.BG-0001(Q). ____ 5.2.34. PRIOR to reaching a Reactor coolant temperature of 200°F, all Operational Condition 4 surveillance items in HC.OP-DL.ZZ-0026(Q) are initiated. ____

SM/CRS 5.2.35. Attachment 3 AND

Attachment 10 PRIOR to reducing Reactor Coolant Temperature to < 200°F. ____

SM/CRS The unit will be in Cold Shutdown (OPERATIONAL CONDITION 4) WHEN Reactor Coolant temperature is < 200°F WITH the RPS MODE SWITCH in SHUTDOWN.

5.2.36. the cooldown to < 200 °F AND in Control Room Log(s) the time the unit enters Cold Shutdown. ____

Completion of the following step will allow for nat ural circulation in the event that forced circulation is subsequently lost.

Main Steam Isolation Valves require closing at 90 inches.

Main Steam Line flooding occurs at 118 inches.

If a degraded Shutdown Cooling condition occurs or if there is indication that the RHR Heat Exchanger inlet temperature may NOT be representative of average Reactor Coolant temperature, HC.OP-AB.R PV-0009(Q), Shutdown Cooling, should be referred to.

During performance of the following step, Ve ssel metal temperatures above and below the water level and Reactor Coolant System Temperat ure/Pressure Data should be monitored to ensure the TS Cooldown lim its are not exceeded.

5.2.37.

Reactor Vessel level to 80 inches, Reactor level shutdown range, usi ng temperature-compensated indication, (Vessel Level Instrumentation Temperature Compensation Curves may be required), WHILE continuing in this section.[] ____ RWCU Regen Hx Bypass can only be opened once Cold Shutdown has been attained.

5.2.38. At the discreti on of the Shift Manager, the RWCU System in Regenerative Heat Exchanger bypass operation IAW HC.OP-SO.BG-0001(Q) Section 5.

9, throttling 1-ED-V035 RWCU NRHX RACS RTN PLUG VLV as necessary to maintain RWCU Demineralizer inlet temp. < 120°F

AND RWCU System outlet temp. 79°F CRIDS Point A215). ____

5.2.39. IF the Containment (Drywell/Torus) is to be opened, THEN the following: The purge alignment requirements of ODCM 3.11.2.8. shall be observed.

A. that a Release Permit has been obtained from the RP Dept. AND the applicability of CPCS requirements reviewed. ____

SM/CRS B. IF required, THEN

Containment Pre-purge Cleanup IAW HC.OP-SO.GS-0001(Q), Co ntainment Atmosphere Control System Operation. ____ C. WHEN atmospheric radioactivity levels are within the limits specified by RP AND by radiological effluent Tech Specs, THEN Containment Pre-purge Cleanup. [] ____ The Primary Containment Air Lock operability requirements of T/S 4.6.1.3.c (and its associated note) shall be observed.

D.

the Containment (Drywell/Torus) IAW HC.OP-SO.GS-0001(Q), Containment Atmosphere Control System Operation ( T/S 3.6.1.8). ____

5.2.40. WHEN Reactor coolant temperature reaches 150°F, RWCU System Demineralizer flow to 150 gpm per Demin Vessel. ____ An administrative temperature range of 90°F - 110°F should be maintained. Other

temperature(s) within Technica l Specification limits may be us ed to support specific plant operations, as necessary.

The Reactor Vessel and Head Flange temperature lim its of Technical Spec ification 3.4.6.1.d shall be complied with.

5.2.41. the cooldown UNTIL the desired final Reactor coolant temperature is reached. ____

5.2.42. AFTER ensuring that the temperat ure readings at the final desired temperature are to the ri ght of limit line of Technical Specification Fi gure 3.4.6.1-2, plotting the Reactor Coolant Cooldown rate. ____ 5.2.43.

I&C to remove Reactor Vessel Level Purge from service IAW HC.IC-GP.ZZ-0119(0120, 0121, 0122)(Q), Filter Replacement and Flow Adjustment Procedure - Backfill Station -

RPV Channel A(B, C, D), removing one channel at a time AND initialing for each channel. RPV Channel A ____ RPV Channel B ____ RPV Channel C ____ RPV Channel D ____

5.2.44. IF desired to break Main Condenser Vacuum, THEN the following:

A. DIVISION 1 and 2 and 3 and 4 CONDENSER LOW VACUUM BYPASS Switches to BYP (Control Room Panels 10C609 and 10C611). ____

B.

Chemistry that the Condensate Drain Tank is no longer available to receive drains,

AND to align the drains from the Turbine Building Sample Station IAW HC.CH-SA.RC-0001. ____ (continued on next page)

5.2.44 (continued)

Main Condenser vacuum should NOT be broken before Main Turbine speed decreases to less than 1200 rpm EXCEPT in emergency conditions such as high vibration, which require the Main Turbine to be slowed down as fast as possible.

C.

the Condenser Air Removal System IAW HC.OP-SO.CG-0001(R), Condenser Air Removal System Operation. ____

D.

the Gaseous Radwaste System IAW HC.RW-SO.HA-0001(R), Gaseous Radwaste System Operation. ____

E.

Chemistry to shut down the Offgas Vial Sampling Panel 10C335 IAW HC.CH-SA.HA-0001(R).[] ____ To prevent pulling in cold air along the Turbi ne Rotor, there should be no vacuum prior to removing the Main Turbine Steam Seals. It may be desirable to open the Vacuum Breakers to assure this.

F. the Main Turbine Steam Seals from service IAW

HC.OP-SO.CA-0001(Z), Main and RFP Turbine Sealing Steam System Operation. ____ The Condensate System should be left in service if needed for Reactor Pressure Vessel floodup in HC.OP-IO.ZZ-0005(Q).

G. from feeding the Reactor Vessel with the Condensate System IAW HC.OP-SO.AE-0001(Q),

Feedwater System Operation. ____

H. Condensate Drain Tank Level Control in Manual

AND lower output signal to 0 %. ____ I.

the Condensate System IAW HC.OP-SO.AD-0001(Q), Condensate System Operation. ____ (Continued on next page)

5.2.44 (continued)

The Main Turbine should remain on the turning gear if Turbine restart is expected soon OR until turbine metal temperatures are < 300 F.

J. the Main Turbine from Turning Gear operation IAW HC.OP-SO.AC-0001(Q),Main Turbine Operation. ____

K. the remaining Circulating Water Pumps IAW HC.OP-SO.DA-0001(Z), Circulating Water System Operation. ____

6.1 the entire procedure IAW R M-AA-101, Records Management Program. 7.1

HC.OP-IO.ZZ-0007(Q), Operations from Hot Standby 7.2 HC.OP-SO.AB-0001(Q), Main Steam System Operation HC.OP-SO.AC-0001(Q), Main Turbine Operation HC.OP-SO.AD-0001(Q), Condensate System Operation HC.OP-SO.AE-0001(Q), Feedwat er System Operation HC.OP-SO.BB-0002(Q), Reactor Reci rculation System Operation HC.OP-SO.BC-0001(Q), Residual Heat Removal System Operation HC.OP-SO.BC-0002(Q), Decay Heat Removal Operation HC.OP-SO.BG-0001(Q), Reactor Wa ter Cleanup System Operation HC.OP-SO.CA-0001(Z), Main and RFP Turbi ne Sealing Steam System Operation HC.OP-SO.CG-0001(R), Condenser Air Removal System Operation HC.OP-SO.DA-0001(Z), Circulating Water System Operation HC.OP-SO.GS-0001(Q), Containment Atmo sphere Control System Operation HC.OP-SO.SB-0001(Q), Reactor Pr otection System Operation HC.OP-SO.SE-0001(Q), Nuclear Inst rumentation System Operation

7.3 HC.RW-SO.HA-0001(R), Gaseous Radwaste System Operation HC.RE-RA.ZZ-0011(Q), Cr ossflow Operations CD-015B, GE SIL 254 CD-019Y, FSAR 11.3.2.2.1 CD-049X, FSAR 5.3.3.6 CD-066X, FSAR 5.4.7.2.6 CD-251C, INPO SE 85-83 CD-393B, INPO SOER 84-02R03 CD-523B, NRC IE INFO NOTICE 83-75 CD-693A, INPO SOER 82-2 CD-786D, GE AID 48-78 CD-973B, GE SIL 357 HC.OP-DL.ZZ-0027(Z), Tempor ary Reading Log, Rev. 0 CD-953B CD-249E CD-101E CD-174E, Power Ascension Walk through Aug. 85 CD-491Y, FSAR ACRS-1 HC.OP-DL.ZZ-0026(Q) CD-354F NRC Bulletin 88-07 CD-573F NRC GEN LTR 92-04 NRC Bulletin 93-03 CD-609G NHO LET 4EC3411 Technical Specifications 3.6.6.2, 4.3.1.1, 4.3.6, 4.3.7.6, 4.9.2 CD-454H PR 960326238, LER 354/95-033-05 CD-448H PR 960326107, LER 96-012 PR 960508151 GE SIL 541, Rev 2 Nuclear Fuels Memo NFS96-416 CD-781A (GE SIL 203 and 203 Supp. 1) CD-210E INPO SOER 85-4 CR 981117261 Loss of Feedwater Flow During Plant Cooldown HC.MD-FR.KE-0035(Q), Reactor Pressure Vessel Disassembly HC.MD-FR.KE-0001(Q), Refuel Floor Shield and Pool Plugs Removal and Replacement 80048294, Electro Hydraulic Control (EHC) digital upgrade 80048295, Main Turbine Retrofit 80065875, OPRM trips to RPS.

The following checks may be performed in any order.

1.1 OP-HC-108-115-1001 forms to ensure t he equipment required to enter Condition 2 is available. Any shutdown LCO's which will not be exited prior to changing modes have been assessed IAW Tech Spec 3.0.4.b and

OP-HC-108-115-1001. __________________________________

SM/CRS __________________

Date/Time 1.2 all current notifications are screened fo r operability prior to mode change.

__________________________________

SM/CRS __________________

Date/Time

1.3 PRIOR

to taking the RPS MODE SWITCH to STARTUP & HOT STBY, the following:

1.3.1. WCM "Current Operating Mode" from 1 to 2, the Mode Dependent Tagging/Current Mode/Change function. ____ The Components in the Off - Normal Position Report will indicate all components NOT in the required position for STARTUP.

1.3.2. a Components In Off - Normal Position Report the WCM Reports/Off Normal Report function. ____

1.3.3. all components as required. ____

1.3.4. WCM using the Mode Dependent Tag/Current Positions/Change Function. ____ 1.3.5. The above items have been completed with all equipment required for going into STARTUP available. __________________________________

SM/CRS __________________

Date/Time

1.4 System

requirements and surveillances r equired for entering Operational Condition 2 are completed. __________________________________

I&C __________________

Date/Time

__________________________________

Operations __________________

Date/Time All department system requirements, above, for entering Operational Condition 2 are satisfied. __________________________________

SM/CRS __________________

Date/Time

The following checks may be performed in any order.

1.1 OP-HC-108-115-1001 forms to ensure t he equipment required to enter Condition 3 is available. Any shutdown LCO's which will not be exited prior to changing modes have been assessed IAW Tech Spec 3.0.4.b and

OP-HC-108-115-1001. __________________________________

SM/CRS __________________

Date/Time 1.2 all current notifications are screened fo r operability prior to mode change.

__________________________________

SM/CRS __________________

Date/Time

1.3 PRIOR

to taking the RPS MODE SWITCH to SHUTDOWN, the following: ____

1.3.1. WCM "Current Operating Mode" to 3 using the Mode Dependent Tagging/Current Mode/Change function. ____ The Components in the Off-Normal Position Report will indicate all components NOT in the required position for HOT SHUTDOWN.

1.3.2. a In Off - Normal Position Report the WCM Reports/Off Normal Report function. ____

1.3.3. all components as required. ____

1.3.4. WCM using the Mode Dependent Tag/Current Positions/Change Function. ____ 1.3.5. The above items have been completed with all equipment required for going into HOT SHUTDOWN available. __________________________________

SM/CRS __________________

Date/Time

1.4 System

requirements and surveillances r equired for entering Operational Condition 3 are completed. __________________________________

I&C __________________

Date/Time __________________________________

Operations __________________

Date/Time All department system requirements, above, for entering Operational Condition 3 are satisfied.

__________________________________

SM/CRS __________________

Date/Time

The following checks may be performed in any order 1.1 OP-HC-108-115-1001 forms to ensure t he equipment required to enter Condition 4 is available. __________________________________

SM/CRS __________________

Date/Time 1.2 all current notifications are screened fo r operability prior to mode change.

__________________________________

SM/CRS __________________

Date/Time

1.3 PRIOR

to reaching a Reactor Coolant temperature of 200°F, the following:

1.3.1. WCM "Current Operating Mode" from 3 to 4 using the Mode Dependent Tagging/Current Mode/Change function. ____ The Components in the Off - Normal Position Report will indicate all components NOT in the required position for HOT SHUTDOWN.

1.3.2. a Components In Off - Normal Position Report the WCM Reports/Off Normal Report function. ____

1.3.3. all components as required. ____

1.3.4. WCM using the Mode Dependent Tag/Current

Positions/Change Function. ____ 1.3.5. The above items have been completed with all equipment required for going into COLD SHUTDOWN available. __________________________________

SM/CRS __________________

Date/Time

1.4 System

requirements and surveillances r equired for entering Operational Condition 4 are completed. __________________________________

Maintenance __________________

Date/Time __________________________________

I&C __________________

Date/Time __________________________________

Operations __________________

Date/Time All department system requirements, above, for entering Operational Condition 4 are satisfied.

__________________________________

SM/CRS __________________

Date/Time

1.0 Reactor

Coolant System Temperat ure on this attachment every 30 minutes. ____ Only points which have past the element should be used.

2.0 WHEN temperature is 212 F, Reactor Coolant System Temperature as follows: 2.1 On TR-R650-B31 (10C650C) RECIRC PUMP SUCTION - LOOP A TEMP ____ RECIRC PUMP SUCTION - LOOP B TEMP ____ 2.2 Recirc Loop Temperature, usi ng the following Computer Points: A221, RECIRC LOOP A INLET TEMP 1 ____ A222, RECIRC LOOP A INLET TEMP 2 ____ A223, RECIRC LOOP B INLET TEMP 1 ____ A224, RECIRC LOOP B INLET TEMP 2 ____ B2042, RECIRC LOOP A AVG INLET TEMP ____ B2043, RECIRC LOOP B AVG INLET TEMP ____ 2.3 RHR Hx Inlet Temperature usi ng the following computer points: A2380, RHR A Hx Inlet Temperature ____ A2382, RHR B Hx Inlet Temperature ____ 2.4 RWCU Bottom Head Drain Temperature from Computer Point A2942. ____

3.0 WHEN temperature is 212 f, data can be obtained by converting Reactor Steam Dome pressure to saturated temperature using steam tables. ____

4.0 the cooldown rate is 90 f/hr, AND delta-t for the 30 minute interval below the Reactor Coolant System Temperature plot on the space provided. ____

5.0 the RCS temperature and pressure are to the right of the limit line of Technical Specification Figure 3.4.6.

1-2 (if reactor is NOT critical) or Figure 3.4.6.1-3 (if reactor is critical), every 30 minutes, AND on Attachment 3s of HC.OP-DL.ZZ-0026 (Q), Surveillance Log. ____

____ DATE to Saturated Tem

p. PSIA / Steam Table Hi ghest Recirc Suction Tem

p. or RHR Hx Inlet Pressure converted Reactor Steam Dome RPV Press + 14.7 = PSIA Saturation Tem perature or RWCU Bottom Head Drain VI.2 Delta T __ __ __ __ __ __ __ __ __ __ __ __ __

__ __ __ __ __

Note: 1.

completed Attachment 4 sheets with the on going procedure HC.OP-IO.ZZ-0004(Q).

2. temperatures in conjunction with HC.OP-DL.ZZ-0026(Q), Attachment 3s AND operation to the right of the applicable curve in Tech S pec 3.4.6.1 as well as HC.OP-DL.ZZ-0026(Q), Attachment 3s. 3. Below 212°F water temperature must be read directly. The point s are listed in order of preference (highest Recirc suction temperature, RHR Hx Inlet, RWCU Bottom Head Drain). 4. There must be forced flow past the temperature elem ent in order to obtain a valid temperature reading. 5. Above 212°F Reactor Steam Dome pressure should be used to obtai n the saturation temperature from the Steam Tables. This tem perature should then be plotted.

1.0 System

Engineering has performed analysis to determine Alternate Decay Heat Removal method []:

Time after shutdown at which the specified

alternate heat removal configuration can be used.

Max SACS and/or RACS temperature for which

the specified alternate decay heat removal

configuration is valid.

Recirculation requirements (e.g., single Recirc Pp at minimum speed, RHR

Pp in shutdown cooling lineup, natural circulation)

Decay heat removal requirements (e.g., FPCC and RWCU, RWCU cooled by RACS, RWCU cooled by Chilled Water, other.)

__________________________________

System Engineer __________________

Date/Time

2.0 The SM has been informed that the Alter nate Decay Heat Removal method will adequately remove decay heat for the system lineup specified by system engineering, IAW Technical

Specification 3/4.9.11. __________________________________

SM __________________

Date/Time 3.0 RWCU in the Regenerative Heat Exc hanger Bypass Mode of operation IAW HC.OP-SO.BG-0001(Q), if required. __________________________________

SM/CRS/RO __________________

Date/Time

4.0 one, or both, Fuel Pool Cooling Heat Exchangers in service IAW HC.OP-SO.EC-0001(Q). __________________________________

SM/CRS/RO __________________

Date/Time

5.0 flow through the core WITH either one Recirculation Pump, (IAW HC.OP-SO.BB-0002(Q)), OR one RHR Pump aligned for shutdown cooling WITH the heat exchanger by passed, IAW HC.OP-SO.BC-0002(Q). __________________________________

SM/CRS/RO __________________

Date/Time

6.0 C RHR Pump has been placed in service for Alternate Decay Heat Removal IAW HC.OP-AB.RPV-0009(Q). [

] __________________________________

SM/CRS/RO __________________

Date/Time

7.0 D RHR Pump has been placed in service for Alternate Decay Heat Removal IAW HC.OP-AB.RPV-0009(Q). [

] __________________________________

SM/CRS/RO __________________

Date/Time

The following should be performed by Qualified Maintenance Personnel.

An independent verification shall be performed for the following steps.

1.0 Upon notification that 52-264042 (BG-HV-F031 RWCU FLOW ORIFICE BYPASS) is tagged open, the following: 1.1 At Breaker 52-264042 the following:

1.1.1. Breaker

52-264042 BG-HV-F031 RWCU FLOW ORIFICE BYPASS is open. ____

1.1.2. Job Information Tag for the Breaker Overloads. ____ 1.1.3.

the Breaker Overloads for Breaker 52-264042 BG-HV-F031 RWCU FLOW ORIFICE BYPASS ( H1022 (LO)) ____

1.2 the Main Control room to release breaker 52-264042. ____

2.1 The Main Control Room that the Breaker Overloads have been installed. ____

Attachment 9 is to be performed when an increase in the cooldown rate of the Main Turbine Shell is desired, and used only during a "Controlled" shutdown (NOT following a scram), as a time-saving measur

e. The inferences to "Cooling/Cooldown", or "Warming", are dependent upon whether the direction is referring to the activity of cooling

or the nomenclature on the in strumentation/indications.

This attachment cannot be used simultaneous ly with the cooldown controller.

The cooldown controller is interlocked such t hat a Main Turbine trip signal must exist.

Performing the turbine shell cooldown require s the turbine trip signal to be reset.

1.0 Start

HP Turbine Shell Cooldown by performing the following steps:

1.1 Diag Reset S1 , Diag Reset P1 , Master Reset P1 ____

1.2 Control

, Valve Limiters ____

1.3 the following: Valve Position Limiter, VPL Setpoint: 100% ____ Max Combined Flow Limit, Setpoint: 130% ____ 1.4 Control , Pre-Warming ____ 1.5 the following:

Chest Warming: OFF ____ Shell Warming: OFF ____ 1.6 Control , Speed - Load ____ 1.7 the following:

Turbine Trip Status: Reset ____ Turbine Control Status: Valves Closed Controlling ____ Load Setpoint: 0% ____ 1.8 Speed Control , Acceleration RPM/Min Fast (180) . ____

1.9 the following valves: 1.9.1. AC-HV-1013A, B, C and D STEAM LINE DRAINS - MN STM VLV BFR SEAT. ____ 1.9.2. AC-HV-1015 STEAM LINE DRAINS - CONT VLV BFR SEAT. ____

1.9.3. AC-HV-1041/42/

43/A/B/C STEAM LINE DRAINS - CROSS AROUND. ____ 1.9.4. AC-HV-1018B STEAM LINE DRAINS - LEAD 3. ____

1.9.5. AC-HV-1360A, B and C FWH #5A, B and C SHELL SIDE MOIST SEP B DRN. ____ 1.9.6. AC-HV-1361A, B and C FWH #5A, B and C SHELL SIDE MOIST SEP A DRN. ____ 1.9.7. AC-HV-1362A,B and C FWH # 5A,B and C SHELL SIDE CROSS AROUND STM ISLN. ____ 1.9.8. AC-HV-1751A, B and C RFPT A, B and C LO PRESS STM ISLN VLV. ____

While in Shell Cooldown, the temperature lim its of Attachment 2 of HC.OP-SO.AC-0001(Q) should be referred to.

1.10 Control , Pre-Warming ____

1.11 Shell Warming , ON AND the following: 1.11.1. All Control Valves open fully, after a time delay. ____ 1.11.2. All Intermediate Stop Valves (ISV) go closed. ____ 1.11.3. All Intercept Valves (IV) remain closed. ____

1.11.4. All Main Stop Valves (MSV) remain closed. ____

IF the turbine should roll off the turning gear, it may be necessary to remove lift pumps from service. Alternate lift pump operations should be performed by referring to Attachment 6 of HC.OP-SO.AC-0001(Q).

1.12 HP Turbine Shell to a pressure which will allow for a 50°F difference between steam temperatur e and 1st Stage Shell Lower Inner Surface temperature as follows:

(

to Steam Tables for initial desired pressure/temperature) ____ Chest temperature changes should be obser ved as an indication of steam flow.

1.12.1. To establish cooldown steam, Adjust MSV2 Position Ramp Rate AND desired Ramp Rate. ____ 1.12.2. Intermittently Adjust MSV2 Position , Manual Adj.

Raise UNTIL flow is established through MSV-2. ____ 1.12.3. (STEAM LEAD DRAINS) - LEAD 1 & 2 AC-HV-1017A/B to maintain the 50°F temperature difference described in Step 1.12. ____

1.12.4. IF the turbine rolls off the turning gear, THEN Shell Warming- OFF ____

2.0 Stop HP Turbine Shell Cooldown by performing the following steps:

2.1 Adjust

MSV2 Position , Manual Adj Lower UNTIL Position indication is at zero PERCENT. ____ 2.2 the following valves: 2.2.1. AC-HV-1013A, B, C and D STEAM LINE DRAINS - MN STM VLV BFR SEAT. ____ 2.2.2. AC-HV-1015 STEAM LINE DRAINS - CONT VLV BFR SEAT. ____

2.2.3. AC-HV-1041/42/43/A/B/

C STM LINE DRAINS - CROSS AROUND. ____ 2.2.4. AC-HV-1018B STEAM LINE DRAINS - LEAD 3. ____ 2.2.5. AC-HV-1360A, B and C FWH #5A, B and C SHELL SIDE MOIST SEP B DRN. ____ 2.2.6. AC-HV-1361A, B and C FWH #5A, B and C SHELL SIDE MOIST SEP A DRN. ____ 2.2.7. AC-HV-1362A, B and C FWH #5A, B and C SHELL SIDE CROSS AROUND STM ISOL. ____ 2.2.8. AC-HV-1751A, B and C RFPT A, B and C LO PRESS STM ISLN VLV. ____

2.3 (STEAM LEAD DRAINS)-LEAD 1&2 AC-HV-1017A/B. ____ Overhead alarm D3-D5 - EHC PANEL 10C363 TROUBLE will come in (CRIDS Point D2031 MN TRB FAST CLOSE INTRCPT VLVS in alarm), IF cross-around pressure is still above 43 psig when the Shell Warming-OFF is selected.

2.4 After

Cross-around pressure drops below 43 psig, Shell Warming OFF ____

2.5 all Control, Stop and Intercept Valves close. ____

IRMS, Neutron Flux High Table 4.3.1.1-1

Function 1.a 2,3,4,5 2,3,4,5 2,3,4,5 Channel Check

Channel Functional Test

Channel Calibration Shiftly Weekly Refueling OC 2: Action 1 OCs 3,4: Action 2 OC 5: Action 3 IRM's, Inoperative Table 4.3.1.1-1

Function 1.b 2,3,4,5 Channel Functional Test Weekly OC 2: Action 1 OCs 3,4: Action 2 OC 5: Action 3 APRM's, Neutron Flux,

Upscale, Setdown Table 4.3.1.1-1

Function 2.a 2,3,4,5 2,3,4,5 2,3,4,5 Channel Check

Channel Functional Test

Channel Calibration Shiftly Weekly Semi-annually OC 2: Action 1 OCs 3,4: Action 2 OC 5: Action 3 APRM's, Inoperative Table 4.3.1.1-1

Function 2.d 2,3,4,5 Channel Functional Test Quarterly OC 2: Action 1 OCs 3,4: Action 2 OC 5: Action 3 APRM's, Inoperative Table 4.3.6.-1

Function 2.b 2,5 Channel Functional Test Quarterly OCs 2,5: Action 61 APRM's, Neutron Flux,

Upscale, Startup Table 4.3.6.-1

Function 2.d 2,5 2,5 Channel Functional Test

Channel Calibration Quarterly Semi-annually OCs 2,5: Action 61

SRMs, Detector Not

Full In Table 4.3.6.-1

Function 3.a 2,5 Channel Functional Test Weekly OCs 2,5: Action 61

SRMs, Upscale Table 4.3.6-1

Function 3.b 2,5 2,5 Channel Functional Test

Channel Calibration Weekly Refueling OCs 2,5: Action 61

SRMs, Inoperative Table 4.3.6-1

Function 3.c 2,5 Channel Functional Test Weekly OCs 2,5: Action 61

SRMs, Downscale Table 4.3.6-1

Function 3.d 2,5 2,5 Channel Functional Test

Channel Calibration Weekly Refueling OCs 2,5: Action 61

IRMs, Detector Not

Full In Table 4.3.6-1

Function 4.a 2,5 Channel Functional Test Weekly OCs 2,5: Action 61

IRMs, Upscale Table 4.3.6-1

Function 4.b 2,5 2,5 Channel Functional Test

Channel Calibration Weekly Refueling OCs 2,5: Action 61

IRMs, Inoperative Table 4.3.6-1

Function 4.c 2,5 Channel Functional Test Weekly OCs 2,5: Action 61

IRMs, Downscale Table 4.3.6-1

Function 4.d 2,5 2,5 Channel Functional Test

Channel Calibration Weekly Refueling OCs 2,5: Action 61

SRMs 4.3.7.6.a.1.a 2 Channel Check Shiftly Action 3.3.7.6.a SRMs 4.3.7.6.a.1.b 3,4 Channel Check Daily Action 3.3.7.6.b SRMs 4.3.7.6.a.2 2,3,4 Channel Calibration Refueling OC 2: Action 3.3.7.6.a OCs 3,4: Action 3.3.7.6.b SRMs 4.3.7.6.b 2,3,4 Channel Functional Test Monthly OC 2: Action 3.3.7.6.a OCs 3,4: Action 3.3.7.6.b SRMs 4.9.2.a.1 5 Channel Check Shiftly Action 3.9.2 SRMs 4.9.2.a.2 5 Verification That Detectors are Fully Inserted Shiftly Action 3.9.2 SRMs 4.9.2.b 5 Channel Functional Test Weekly Action 3.9.2 SRMs 4.9.2.c.3 5 Verification That Channel Count Rate is >

3 cps Daily (1) Action 3.9.2 (1) AND prior to control rod withdrawal OR Core Alterations

Actions Required if Technical Specification Surveillance Requirements Not Satisfied 3.3.1-1 Actions: 1: Be in at least Hot Shutdown within 12 hours1.388889e-4 days <br />0.00333 hours <br />1.984127e-5 weeks <br />4.566e-6 months <br /> 2: Verify all insertable control rods to be inserted in t he core and lock the reactor mode switch in the Shutdown position within one hour. 3: Suspend all operations involving CORE ALTERATION S* and insert all insertable control rods within one hour. *Except replacement of LPRM strings provided SRM in strumentation is OPERABLE per Specification 3.9.2.

3.3.6-1 Actions 61: With the number of Operable Channels:

a. One less than required by the Minimum Operable C hannels per Trip Function requirement, restore the inoperable channel to Operable status within 7 days or place the inoperable channel in the tripped condition within the next hour. b. Two or more less than required by the Minimum Operable Channels per Trip Function requirement, place at least one inoperable channel in the tripped condition within one hour. 3.3.7.6.a: In Operational condition 2 with one of the above required source range monitor channels inoperable, restore at least 3 source range monitor channels to an Operable status within 4 hours4.62963e-5 days <br />0.00111 hours <br />6.613757e-6 weeks <br />1.522e-6 months <br /> or be in at least Hot Shutdown within the next 12 hours1.388889e-4 days <br />0.00333 hours <br />1.984127e-5 weeks <br />4.566e-6 months <br />. 3.3.7.6.b: In Operational condition 3 or 4 with one of the above required source range monitor channels inoperable, verify all insertable control rods to be inserted in the core and lock the r eactor mode switch in the Shutdown position within one hour. 3.9.2: With the requirements of the above specification not satisfied, immediately suspend all operations involving Core Altera tions and insert all insertable control rods.

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ENSURE THE FOLLOWING 5.1.1 A.System Operator has been notified ofthe Shutdown. B.Reactor Engineer has been notified ofthe Shutdown. C.Steam Lead drain #4 is in AUTO.(AC-HV-1018A) Load Setpoint at 100%,OR as directed by CRS Rx PWR W/ Recirc 3rd RFP in Recirc. Crossflowto Not AppliedRecirc. Is in Indiv.Manual@ approx 45% - 50% speed Throttle Pressure Set

@ 905 psigRodsRecirc to MinimumWHEN the #4 CONTROL VALVE is Full Closed, the #4 STEAM LEAD DRAIN(HV-1018A) is OPENRFP with lowest discharge flow to vessel (2 nd RFP) in Recirc IAW HC.OP-SO.AE-0001(Q)one SCP A(B,C)P137 IAWHC.OP-SO.AD-0001(Q) Single Element Control RBM Bypass On MSLRM to Norm. Condenser backpressure less than orequal to 1.5" Hg Abs., using one or both of thefollowing procedures:HC.OP-SO.CG-0001(R), Condenser Air RemovalSystem OperationHC.OP-SO.DA-0001(Z), Circulating Water SystemOperation. I&C and Radiation Protection torestore MSL Rad Monitor Trip and Alarm Setpoints to normalone PCP A(B,C)P102 IAWHC.OP-SO.AD-0001(Q)one RFP previously placed inRecirc has been removed from svc IAWHC.OP-SO.AE-0001(Q)1 RFP in svc and 1 RFP in Recircpotential for RWCU pump trips as follows:Chemistry to reduce RWCU flow to .A or B RWCU Pump from svcIF not performed in previous 92 days,Main Turbine Lift Pump test IF directed to lock Mode Switch in Shutdown between 30% & 20% power with Main Turbine on line, the following:

an Operator for local observation andFeedwater Startup Level Control Valves.lube oil temp. controller, power system stabilizer IAW HC.OP-SO.AC-0001(Q)the following at current power level:(Intermediate Steps N/A)

5.1.22 for EOC RPT System5.1.27 for opening the FWH 1&2 vents5.1.28 for Pressure Setpoint Adjustment 5.1.29 for performing a Manual Scram the HWCI System from service RFP Min Flow Vlv controller in MAN AND Min flow to achieve 3500 gpm.<90 gpm with one Demin in svc while monitoring suction flow.Main Turbine Oil Pumps in svc, H1CA -CA-HV-1991 is open

Turb/Gen from GridWillScram Time Testing beperformed GETARS in "Sentinel" Mode FWHTR's #1 & 2Operating Vents Throttle Pressure Set, as directed by CRS.IRM/APRMOverlapMode Switch toShutdown Now ?MODE SW.TO S/DMODE SW.TO S/U RxPress > 950 psig Att. 2& Review Att. 10 the Mode Sw. inShutdownThrottle Pressure Setto 905,if necessary ReactorCavity Shield PlugRemoval, If not donePreviouslyInserting Rodsto 6 and 9%

Power OPCON2Surveillance Initiated Recordersto IRM's IRM'sBetween 25 - 75W/ Range Sw.IRM'sOperable Att. 1 AND Att. 10 Mode Sw.In S/U-H/STBY the following: to HC.OP-IO.ZZ-0007(Q)If Maintaining Hot Stby Drain Valves Drain ValvesSteam Lead Drain Open the appropriate sections ofHC.OP-FT.AC-0004(Q) EOC-RPTIF required to implement Regular Maintenance Plan14852, THENentire procedure.IF not performed in the last 12 months,ANDno work is to be done on the front standard,THEN POST and EOST. Tags for F020 & F021BG-HV-F034 AND BG-HV-F035for blowdown operation.In SVC RFP's Min Flow Vlv inManual to achieve 3,500 gpm min flowIAW HC.OP-SO.AE-0001(Q). Feedwater Control from Master Lvl Control to Startup Vlvs IAW HC.OP-SO.AE-0001(Q)If Refueling, Reactor CavityShield Plug RemovalLPSP on RWMLPAP AND Operation of RWMPost Trip Review.Cooldown Rate

ReducingPWR by InsertingControl Rods IRM'sBetween 25 -75W/ Range Sw. Circ. Water Pp's as Required COOLDOWN /DEPRESSURIZATION Cooldown Rate SRM's 10 2-10 5 byInserting SRM's Att. 2& Att. 10ALL RODS FULLYINSERTED Mode Sw.in Shutdown Scram SCRAM RESET Reactor Cavity Shield PlugRemoval If not done PreviouslyApprox.500 psig Remaining RFP's from Service 2nd SCP from Service If desiredHP Turb Shell, Att. 9RHR forShutdown CoolingApprox.300 psig SJAE to Aux. Stm.

OR MVP's In-Service Recombiner to Aux.

SteamPressure Set hasreached 150 psigApprox.100 psigHPCI IsolatesCONTINUE / MAINTAIN a cooldown rate of< 90°F/hr by opening a Turbine Bypass Valveusing either of the following:Rx Cooldown ModePressure Control ModeBypass Valve Manual JackIFNot Required STOP (1) SCP AND (1) PCPI&C to adjust CRDflow controllerApprox.200 PSIG Cooldown rate to 30 F/hror less to avoid cavitation. the HWCI System from service.

Aprox.80 psig the mode Sw.is Locked in shutdown RHRShutdown Clg. BothRecirc. Pump"s "B"Recirc PpAND "A"Recirc Pp I/S BothRecirc. Pump"s RPV LVL> 80" ANDMon. RPV Temp/PressAtt. 4 "A" or "B"RHRin Shutdown CoolingWHEN RHR is atrated flow (10,000gpm) "A" Recirc Pp. Plotting Cooldown Bypass ValvesAprox. 64.5 psig RCIC IsolatesAprox. 10 -25 psig the MSIV'sAprox. 5 psigAB-HV-F016 Turbine Shell Cooldown Tagson BB-HV-F001 Tags onBB-HV-F002Tags onAE-HV-F011A & BAprox. 50 psig PressureControl setpoint at 50# Rx Cooldown ControlOFF desired, USE BPV JackAprox. 25 psig BothRecirc. Pumps'speed per IPTE "B" OR "A"RHR in ShutdownCooling @ flowratespecified in IPTERx pressure setpoint matched tothrottle pressure IF

200 o FCooldown to < 200 o F Slowly LVL> 80"At SM discretion RWCU RHXin Bypass Operation NO YESRelease Permit hasbeen obtainedIf required Containment Pre-PurgeWhen Cont. Atm. Rad Levels arewithin Limits, Pre-PurgeDe-inerting the ContainmentRPV Coolant Temp. 150 o FEND NO YES RWCU Flow to150 gpm/ Demin Vessel Cooldown toDesired TemperatureAFTER Desired Temp. is ReachedAND W/in T.S.Limit Plotting Cooldown Rate I&C to RemoveVessel Level Purgefrom ServiceChemistry that the Condensate Drain Tkis no longer available,AND to the Turbine Sample Sink drainsIAW HC.CH-SA.RC-0001 CondenserAir Removal Sys. Gaseous RadWaste Sys. Chemistry toS/D Off Gas Vial SamplingSealing Steam Sys. from serviceFeeding the RPV from Condensate Cond. Drn. Tk.LVL Control in MAN.

AND Output to 0%The Condensate Sys.The Main Turbinefrom the Turning GearRemaining Circ. Water Pumps Main Cond. Low VacuumSwitches in BypassWhen RPV Temp. < 212 o F2nd RWCU PpI/S @ approx. 90 gpm.Chemistry toPlace 2nd RWCUF/D in-serviceTurb Sealing Stm Drains are ClosedMSL Vents and Drain are ClosedStm. Line DrainsOp Con 4surveillance's InitiatedAtt. 3 ANDAtt. 10IF necessary, Reactor Vessel level with the RWCU System