ML070790652
| ML070790652 | |
| Person / Time | |
|---|---|
| Site: | Salem  |
| Issue date: | 02/14/2007 |
| From: | Down R - No Known Affiliation |
| To: | Office of Nuclear Reactor Regulation |
| References | |
| CC-AA-309-1001, Rev 3, S-C-SF-MDC-1810, Rev 7 | |
| Download: ML070790652 (109) | |
Text
{{#Wiki_filter:CC-AA-309-1001 Revision 3 ATTACHMENT I Design Analysis Major RevIimon Cover Sh-eer Design Analysis (Major Revision) Last Page No. 6 Attachment 5 page 1 of 1 Analysis No.: I S-C-SF-MDC-1810 Revision: 2 7 3
Title:
Decay Heat-up Rates and Curves ECJECR No.: 4 80091615 M01RO Revision: 5 0 Station(s): Salem 1 Component(s): 14 Unit No.: 8 Units 1 & 2 Discipline: 9 Mechanical Descrip. Code/Keyword: 10 Safety/QA Class: " Safety Related System Code: 12 Spent Fuel (SF) Structure: 13 CONTROLLED DOCUMENT REFERENCES 35 Document No.: From/To Document No.: From/To S-C-SF-MCS-0113 From S-C-SF-MEE-1302 From SC.OM-AP.ZZ-0001 From S-C-SF-MDC-1780 From S-1-FHV-MDC-0705 From Is this Design Analysis Safeguards Information? 16 Yes [] No [ If yes, see SY-AA-101-106 Does this Design Analysis contain Unverified Assumptions? 17 Yes El No Z If yes, ATI/AR# This Design Analysis SUPERCEDES: NONE in its entirety. Description of Revision (list affected pages for partials): 19 The purpose of this calculation is to provide heat-up times, temperatures and curves for the Salem Generating Station (SGS) Unit I Spent Fuel Pool (SFP) for Refueling Outage IRI 8, as directed by Reference 4.7. The calculation is being revised to analyze the spent fuel pool temperature as a result of additional spent fuel transferred to the pool during the upcoming refueling outage. Affected pages - See page revision index on page 2. Preparer: 20 Robert Down Print Name
'--e'04AJ)- Sian Name 2/14/2007 Date Method of Review: 21 Detailed Review [ Alternate Calculations (attached) El Testing C1 Reviewer: 22 Kevin King /~~~~~
Print Name Sign Date Review Notes: 23 Independent review [ Peer review LI (For External Analyses Only) External Approver: 24 N/A N/A N/A Print Name Sign Name Date Exelon Reviewer: 25 N/A N/A N/A Print Name Sign Name Date Independent 3 Party Review Reqd? 26 Yes[:I No Exelon Approver: 27 Alan Johnson L S t) Print Name -Sign Name Date
(NC.DE-AP.ZZ-0002(Q), Rev. 11, Form 1) CALCULATION COVER SHEET Page Ia of 12 CALCULATION NUMBER: S-C-SF-MDC-1810 REVISION: 7 TITLE: Decay Heat-up Rates and Curves
#SHTS (CALC): 12 #ATT/#SHTS: 5/25 #IDV/50.59 SHTS: 2/0 (2) #TOTAL SHTS: 39 CHECK ONE:
[ FINAL D INTERIM (Proposed Plant Change) [] VOID D FINAL (Future Confirmation Req'd, enter tracking Notification number:) SALEM OR HOPE CREEK: 0 Q - LIST 0 IMPORTANT TO SAFETY E] NON-SAFETY RELATED HOPE CREEK ONLY: '-Q --Qs -]Qsh EIF FIR [] ARE STATION PROCEDURES IMPACTED? YES [] NO Z IF "YES", INTERFACE WITH THE SYSTEM ENGINEER & PROCEDURE SPONSOR. ALL IMPACTED PROCEDURES SHOULD BE IDENTIFIED IN A SECTION IN THE CALCULATION BODY [CRCA 70038194-0280]. INCLUDE AN SAP OPERATION FOR UPDATE AND LIST THE SAP ORDERS HERE AND WITHIN THE BODY OF THIS CALCULATION. [] CP and ADs/CDs INCORPORATED (IF ANY): N/A DESCRIPTION OF CALCULATION REVISION (If applicable.): The calculation is being revised to analyze the spent fuel pool temperature as a result of additional spent fuel transferred to the pool during the upcoming 1R1 8 refueling outage. PURPOSE: The purpose of this calculation is to provide heat-up times, temperatures and curves for the Salem Generating Station (SGS) Unit 1 Spent Fuel Pool (SFP) for Refueling Outage 1R1 8. CONCLUSIONS: Case 1: For crosstie operation with the Unit 1 SFP aligned to the Unit 2 SFHX, the isolated Unit 2 peak SFP temperature is below the licensing basis limit of 180'F wth one SFHX isolated, and thus no swapping of SFPs is required. The heatup rate for the Unit 1 and Unit 2 SFP's is approximately 7.8 0 F/hr and 1.30F/hr respectively. Case 2: For normal SFP cooling, the licensing basis limit of 149°F is exceeded for the bounding case with 99 0 F CC temperature. However, interpolating between the 80°F CC temperature case and bounding case results, a peak SFP temperature of 149'F is reached with a CC temperature of about 94°F, correlating to a maximum SW temperature of 85°F. Since this will be higher than the SW temperature at the time of core offload, the SFP temperature limit of 149°F will not be exceeded. Also, through linear interpolation of the results, the SFP high temperature alarm setpoint of 125°F will be reached with a CC temperature of 69°F for an offload start time of 121 hours. The heatup rate for the Unit 1 SFP is approximately 1.20 F/hr. Case 3: On a loss of cooling to the Unit 1 SFP, the maximum design limit of 180'F will be reached in a range of 2.2 hours to 4.9 hours after core offload is complete. The Unit 1 SFP will not boil ifcooling is not restored. The heat-up rate for Unit 1 is within a range of 9.4 0 F/hr to 10.1 °F/hr Case 4: Heat-up rates in the event of Unit 1 SFP loss of cooling post-outage. See curves for details.
(NC.DE-AP.ZZ-0002(Q), Rev. 11, Form 2) CALCULATION CONTINUATION SHEET SHEET: 2 of 12 CALC NO.: S-C-SF-MDC-1810 REV: 7 REF: CONT'D ON SHEET: ORIGINATOR: DATE: REVIEWER: DATE: VERIFIER: DATE: R. Down 2/21/07 K. King 2/22/07 K. King 2/22/07 REVISION HISTORY Revision Issue Date Revision Description 0 4/6/99 Initial Issue. 1 1/20/00 Revision 1 provides heat-up times for the Unit 1 Fuel Pool as of 12/31/99 to support heat exchanger service and valve repairs to provide realistic heat-up times based on the present conditions of the Fuel Pool. 2 3/2/01 The calculation is being revised to analyze the spent fuel pool temperature as a result of additional spent fuel transferred to the pool during the upcoming refueling outage. 3 9/6/02 The calculation is being revised to analyze the spent fuel pool temperature as a result of additional spent fuel transferred to the pool during the upcoming refueling outage. 4 10/2/02 The calculation is being revised to provide additional heat-up curves for Component Cooling (CC) temperatures of 70°F and 75 0 F, to better represent expected CC temperatures during the upcoming 1R1 5 refueling outage. This is to support LCR S02-03 to revise the minimum time from shutdown before fuel offload can begin from 168 hours to 100 hours. 5 2/24/04 The calculation is being revised to analyze the spent fuel pool temperature as a result of additional spent fuel transferred to the pool during the upcoming refueling outage. 6 9/8/05 The calculation is being revised to analyze the spent fuel pool temperature as a result of additional spent fuel transferred to the pool during the upcoming refueling outage. 7 See cover The calculation is being revised to analyze the spent fuel pool temperature as a sheet result of additional spent fuel transferred to the pool during the upcoming 1 R1 8 refueling outage. PAGE REVISION INDEX PAGE REV PAGE REV PAGE REV PAGE REV 1 7 Attachment 1 7 la 7 Attachment 2 7 2 7 Attachment 3 7 3 7 Attachment 4 7 4 7 Attachment 5 7 5 7 6 7 7 7 8 7 9 7 10 7 1 11 7 12 7
(NC.DE-AP.ZZ-0002(Q), Rev. 11, Form 2) CALCULATION CONTINUATION SHEET SHEET: 3 of 12 CALC NO.: S-C-SF-MDC-1810 REV: 7 REF: CONT'D ON SHEET: ORIGINATOR: DATE: REVIEWER: DATE: VERIFIER: DATE: R. Down 2/21/07 K. King 2/22/07 K. King 2/22/07 TABLE OF CONTENTS REVISION HISTORY ............................................................................................................................ 2 PAGE REVISION INDEX ...................................................................................................................... 2 TABLE OF CONTENTS ........................................................................................................................ 3 1.0 P URP O S E ................................................................................................................................. 4 2 .0 S C OP E ...................................................................................................................................... 4 3.0 ASSUMPTIONS / INPUTS / CONDITIONS ............................................................................ 4 4.0 REFE RENCES ........................................................................................................................... 5 5.0 ANALYSIS ........................................................................... ..............................
. ............ 6 5.1 Methodology ................................................................................................................... 6 5.2 Discussio n ...................................................................................................................... 6 5.3 SFP Inventory Data Files ............................................................................................ 7 5.4 SFP Water Volume .................................................................................................... 7 5.5 Discussion of Input Data File "Rfile" ........................................................................ 8 5.6 Parameters Inputted at Run Time ............................................................................ 9 5.7 100-hr Limiting Core Offload Time ............................................................................ 9 5.8 Run the Crosstie Program ........ .................................. 10 5.9 Import the Output Files ............................................................................................ 10
6.0 CONCLUSION
S ....................................................................................................................... 10 7.0 IMPACT TO STATION PROCEDURES: .............................................................................. 12 8.0 DOCUMENTS AFFECTED: ............................. ..................................................................... 12 9.0 DESIGN MARGIN: ................................................................................................................... 12 10.0 CROSS
REFERENCES:
.......................................................................................................... 12 ATTACHMENT 1 - 1R18 Schedule-"Executive Summary" ATTACHMENT 2 - Nuclear Fuels Letter NF0600209 "Salem 1 RFO18 Assembly Burnup Data for SFP Heat Load Analyses, Rev 0" ATTACHMENT 3 - SFP Heat-up Curves ATTACHMENT 4 - CROSSTIE Input and Output Files (Electronic files on CD)
ATTACHMENT 5 - Salem Verification of Decay Heat Removal for Core Off-load
(NC.DE-AP.ZZ-0002(Q), Rev. 11, Form 2) CALCULATION CONTINUATION SHEET SHEET: 4 of 12 CALC NO.: S-C-SF-MDC-1810 REV: 7 REF: CONT'D ON SHEET: ORIGINATOR: DATE: REVIEWER: DATE: VERIFIER: DATE: R. Down 2/21/07 K. King 2/22/07 K. King 2/22/07 1.0 PURPOSE The purpose of this calculation is to provide heat-up times, temperatures and curves for the Salem Generating Station (SGS) Unit 1 Spent Fuel Pool (SFP) for Refueling Outage 1R1 8, as directed by Reference 4.7. 2.0 SCOPE This calculation is being performed for the SFP Cooling System for SGS Unit 1. The following cases are analyzed: Case 1: Crosstie Operation with one SFHX unavailable Case 2: Normal SFP cooling with no crosstie operation Case 3: Loss of SFP cooling in Unit 1 Case 4: Loss of SFP cooling in Unit 1 - Post Outage 3.0 ASSUMPTIONS / INPUTS / CONDITIONS 3.1 The computer program CROSSTIE is used in this analysis to predict the SFP temperatures (when the unit 1 or unit 2 heat exchanger is out of service) and to evaluate the SFP heat-up rates and equilibrium temperatures without forced cooling. The CROSSTIE program is critical software as defined by ND.DE-AP.ZZ-0052(Q), designated CROSSTIE, Reference 4.1. 3.2 The Crosstie program does not use the first two digits of the year to specify the date (i.e., 1980 uses "80", 1995 uses "95"). In order to manipulate the program such that a "delta time" from initial spent fuel discharge through 1R18 discharge could be determined, year 2000 is represented as year "100", and years following are represented sequentially from "100". This format for the year is changed in the input files, and ".dcy" files that document spent fuel discharged to the SFP. 3.3 The Fuel Handling Building (FHB) ambient temperature is assumed to be the design value of 105 0 F (Reference 4.4, Section 9.4.3.1) for the "bounding" cases, and the estimated ambient temperature during the outage time period of 75 0 F for the "best estimate" cases. The FHB humidity is assumed to be the design value of 100% (Reference 4.4, Section 9.4.3.1). To maintain an ambient temperature of 105 0 F with the SFP temperature over 150°F, the Fuel Handling Ventilation (FHV) system must be operating. The basis for this assumption was analyzed in Reference 4.6, Attachment 8. 3.4 The Component Cooling (CC) supply temperature is assumed to be the maximum procedural limit of 99 0 F for the "bounding" cases, and the estimated values of 70°F, 75 0F, and 80°F during the outage time period for the "best estimate" cases. 3.5 The net water volume (55896 ft3) includes the SFP and transfer pool volumes (minus the volume displaced by the fuel assemblies and racks, see Section 5.4). The volume is assumed to be that for Unit 1, since this is the Unit of concern (CROSSTIE automatically applies the water volume to both unit pools). The net volume is valid for Refueling Outage (1R18) only, as the volume displaced by the fuel assemblies is dependent on the total number of fuel assemblies in the SFP. The post outage volume will be slightly greater due to the two-thirds core fuel assemblies being reloaded into the vessel. However, the difference would have a minor impact on the heat-up rate, and will conservatively be ignored. 3.6 The surface area of the SFP water volume includes the surface area of the transfer pool in the program code. 3.7 Service Water (SW) temperatures determine CC temperatures. For conservatism, the difference between the CC and SW temperatures is assumed to be 90F as determined in Reference 4.10, which is based on a higher SFP heat load. It is also assumed that only one CCHX is available, and there is no parallel SFHX operation (i.e., the Unit 1 SFP is not aligned to both SFHXs). Therefore, the corresponding SW temperatures for CC temperatures of 70°F, 75 0F, 80°F and 99 0F, are 61°F, 66 0 F, 71OF and 90°F respectively.
(NC.DE-AP.ZZ-0002(Q), Rev. 11, Form 2) CALCULATION CONTINUATION SHEET SHEET: 5 of 12 CALC NO.: S-C-SF-MDC-1810 REV: 7 REF: CONT'D ON SHEET: ORIGINATOR: DATE: REVIEWER: DATE: VERIFIER: DATE: R. Down 2/21/07 K. King 2/22/07 K. King 2/22/07 3.8 The Technical Specification states that fuel cannot be moved for a minimum of 100 hours from October 15 through May 15 through the year 2010 (Reference 4.3 Section 3.9.3). Since core offload begins on April 1, the core offload start of 121 hours (as stated in the current schedule included as Attachment 1) is acceptable. 3.9 The duration of core offload is assumed to be the current schedule of 41 hours (see Attachment 1). 3.10 The core reload is scheduled to start 116 hours after the offload is completed, and be completed in 45 hours (see Attachment 1). 3.11 Crosstie operation, with one SFHX unavailable, is assumed to begin immediately after core offload is completed. A nominal minimum value of 1 hour after core offload is used for the analysis. 3.12 The CC flow to the SF Heat Exchanger (SFHX) is assumed to be the design value of 3000 gpm (Reference 4.9). 3.13 The SF flow to the SFHX is assumed to be 2500 gpm (Reference 4.2). 3.14 The current scheduled reactor shutdown date and time for Unit 1 is 3/27/07 at 2000 hours (see Attachment 1). However, for the Crosstie "Rfile" Reactor shutdown date as stated in Section 5.5 Line 2, a shutdown date and time of 3/28/07 at 0000 hours will be used. 3.15 The current SFP inventory (i.e., pre-offload) for Units 1 and 2 are contained in files "Unit1 .dcy" and "Unit2.dcy" for Units 1 and 2, respectively. 3.16 The Unit 1 core parameters - fuel assembly burnups and average assembly Uranium weight - are included in Attachment 2. 3.17 The SFP cooling system maintains pool temperature at or below 1491F provided one SFP heat exchanger is available for each pool, and at 180°F if only one hx is available between both pools (Reference 4.4, Section 9.1.3.2). These design base limits are used as acceptance criteria in the model. 3.18 There are -1137 fuel assemblies in the Salem Unit 1 SFP prior to the start of the 1 R1 8 outage (per Salem Reactor Engineering and Reference 4.12), and is used to calculate the SFP net water volume after core offload. The actual number may vary slightly, but would not impact the calculation results. 3.19 The Core Rated Thermal Power has been increased from 3411 MWt to 3459 MWt (due to the 1.4% power uprate) per Reference 4.3 Section 1.25. This new value applies to all fuel assemblies transferred to the SFP after June 2001 (for both Units 1 & 2), and is conservative for the fuel assemblies that have been radiated at both power levels.
4.0 REFERENCES
4.1 Critical Software, S-C-SF-MCS-0113, "CROSSTIE" A. Sheet 1, Critical Software Document, Revision 1 B. Sheet 1, Software Media, Revision 0 4.2 Calculation S-C-SF-MDC-1 780, "Capability Of Salem SPENT Fuel Pool Heat Exchanger To Maintain 1491F Pool Temperature", Revision 0 4.3 Salem Technical Specifications 4.4 Salem Updated Final Safety Analysis Report (UFSAR) 4.5 Vendor Document, 316748, "Pool Layout - (Region I & II)for Spent Fuel Pool Storage Racks, Revision 1 4.6 Calculation S-1-FHV-MDC-0705, "FHV Sys Htg/Clg Load and Airflow Determination Calcs Unit 1", Revision 4 4.7 Administrative Procedure SC.OM-AP.ZZ-0001 (Q), "Shutdown Safety Management Program - Salem Annex", Revision 1 4.8 Engineering Evaluation S-C-SF-MEE-1 302, "Evaluation to Determine the Equilibrium Temperature for the SFP Without Forced Cooling", Revision 0 4.9 Westinghouse's Letter PSE-89-744 (11/8/89) to M. F. Metcalf (PSE&G), "Salem CCW Calculation Summaries" 4.10 Engineering Evaluation S-C-SF-MEE-1 679, "Spent Fuel Pool Cooling System Capability with Core Offload Starting 100 Hours After Shutdown", Revision 1 4.11 Exelon Procedure OU-AA-103, "Shutdown Safety Management Program", Revision 6 4.12 Nuclear Fuels Calculation DN2.6-0018, "Salem 2 Scoping Study", dated 5/17/2005 4.13 Alarm Response Procedure, S2.OP-AR.ZZ-0003, "Overhead Annunciators Window C, Alarm C-1 9", Rev. 13
(NC.DE-AP.ZZ-0002(Q), Rev. 11, Form 2) CALCULATION CONTINUATION SHEET SHEET: 6 of 12 CALC NO.: S-C-SF-MDC-1810 REV: 7 REF: CONT'D ON SHEET: ORIGINATOR: DATE: REVIEWER: DATE: VERIFIER: DATE: R. Down 2/21/07 K. King 2/22/07 K. King 2/22/07 5.0 ANALYSIS 5.1 Methodology The purpose .of this calculation is to provide heat-up times, temperatures and curves for the SGS Unit 1 Refueling Outage #18 (1R18). This analysis is required by Reference 4.7. The calculation is performed using Holtec's computer program CROSSTIE (Reference 4.1), for the following cases: Case 1: Crosstie Operation with one SFHX unavailable (see note 1) Case 2: Normal SFP cooling with no crosstie operation Case 3: Loss of SFP cooling in Unit 1 (see note 2) Case 4: Loss of SFP cooling in Unit 1 - Post-Outage Note I Removing a SFHX from service is a manual action, and would never be scheduled during or following a core offload. The one exception where this could occur is a tube leak in one of the SFHXs, and would be a management decision. This case predicts when swapovers between the two pools would be required with one available SFHX, if required, to maintain the design limit of 180'F for this potential but unlikely condition. Note 2 Due to the upgrade of the SF Cooling Systems, a loss of cooling due to a seismic event no longer needs to be postulated. However, for a shutdown condition, a loss of cooling is postulated as follows. During shutdown modes 5 and 6, an EDG can be removed from service with no LCO (Reference 4.3 Section 3.8.1.12). As such, on a loss of offsite power (LOOP), another single failure needs to be considered. Ifthe EDG out of service powers one of the SF pumps, and single failure occurs on the other pump, or its EDG, no SF pumps would remain. The loss of cooling case, then, is performed. The results would show how long operators would have to re-establish forced cooling prior to the pool reaching the design limit of 180'F. The following methodology was used to evaluate each of the above cases: APPROACH Step 1: Provide discussion of CROSSTIE program, and its application for each case Step 2: Establish SFP inventory data files. Step 3: Determine SFP net water volume. Step 4: Establish the outage related input data file. Step 5: Determine the remaining parameters inputted at run time Step 6: Run the program. Step 7: Import the Output file plot.dat into EXCEL and generate the heat-up curves. 5.2 Discussion The CROSSTIE code was designed to model crosstie operation during an outage with one SFHX unavailable. To model other scenarios, the code needs to be manipulated to get the meaningful results. The program requires the following inputs: (1) Pre-offload SFP inventory burnup data for both units. This is contained in files "Uniti .dcy" and "Unit2.dcy" for Units 1 and 2, respectively. The program requires these specific file names to be used. (2) A user defined input data file (called Rfile) related to the specific outage. This includes the core offload burnup data, outage start date, unit in the outage, core offload start time and duration, amongst other inputs. This is to allow the transient condition of offloading a "hot" core to be modeled. (3) Miscellaneous data inputted at run time, including the time to start crosstie operation and the pool temperature limit.
(NC.DE-AP.ZZ-0002(Q), Rev. 11, Form 2) CALCULATION CONTINUATION SHEET SHEET: 7 of 12 CALC NO.: S-C-SF-MDC-1810 REV: 7 REF: CONT'D ON SHEET: ORIGINATOR: DATE: REVIEWER: DATE: VERIFIER: DATE: R. Down 2/21/07 K. King 2/22/07 K. King 2/22/07 The CROSSTIE code automatically starts offloading the core to the unit specified at the time specified. At the time specified to start crosstie operation, the code automatically isolates cooling to the unit not in the outage first (since this has the lower heat load), and then swaps cooling to the isolated pool when it reaches the specified pool temperature limit. The cycle continues between the two pools until the specified end time is reached. To model the first case, crosstie operation, the inputs and program execution are straightforward. To model the second case, normal cooling with no crosstie operation, the time to start crosstie is simply set to the end time for the model run or greater, such that no swapping takes place. To model the third and fourth cases, loss of cooling, the opposite unit must be specified as the outage unit, since the program automatically isolates cooling to the unit not in the outage first at the specified time to start the crosstie. Thus, to model a loss of cooling for the Unit 1 SFP, Unit 2 needs to be specified. However, CROSSTIE also automatically applies the core offload data in "Rfile" to the unit specified as being in the outage. Thus, it would add the Unit 1 core to the Unit 2 pool, while isolating cooling to the Unit 1 pool. As such, the core offload data in "Rfile" is set to 0, and is included in the SFP inventory data file "Unit1 .dcy. The reactor shutdown date then has to be changed to coincide with the offload completion date, since the analysis starts with a full "hot" core offload already in the pool. Time t = 0 in this case, then, corresponds to the time offload is completed for both the crosstie and normal cooling cases. The initial pool temperature starting with the full core in the pool will tend to be slightly higher than the temperature corresponding to offload complete time for the crosstie and normal cooling cases. This temperature difference is minor and is conservative in nature; therefore, the analysis will be used as-is. 5.3 SFP Inventory Data Files The current (pre-offload) SFP inventory burnup data is contained within data files "Unit1 .dcy" and, "Unit2.dcy" for. Units 1 and 2, respectively. These files currently contain the inventory up through Cycle 17 for Unit 1 and Cycle 15 for Unit 2 (for Cases 1 & 2), as developed in Reference 4.8. The Cycle 18 offload burnup data was provided by Fuels per Attachment 2, including both the full core offload and the fuel to remain in the pool after core reload. For the first two cases, crosstie operation and normal cooling, "Unit1 .dcy" remains as is with the core offload data inputted under a separate input file discussed in Section 5.5 below. The current "Unitl.dcy" file is included as an electronic file on CD in Attachment 4 (listed as "unitl-17.dcy"). For the loss of cooling case (Case 3), as discussed in Section 5.2, "Uniti .dcy" is updated to include the Unit 1 full core offload for Cycle 18. The core offload data included with "Rfile", as discussed in Section 5.5, is inputted as three batches, with an average burnup for each batch. The updated "Uniti .dcy" file includes these three batches, as shown in Attachment 4 (included as an electronic file on CD, and listed as "unitl-18FC.dcy"). For the loss of cooling case post-outage (Case 4), "Unit1 .dcy" is updated to reflect the fuel assemblies permanently discharged to the SFP after the Reactor vessel is reloaded for Cycle 19. The updated "Unitl.dcy" file is shown in Attachment 4 (included as an electronic file on CD, and listed as "unitl-18-PO.dcy"). 5.4 SFP Water Volume The net water volume includes the SFP and transfer pool volumes at an elevation of 23 feet above the fuel assemblies, minus the volume displaced by the fuel assemblies and racks. The volume is calculated based on the methodology from Section 3.1 of Reference 4.1A.
(NC.DE-AP.ZZ-0002(Q), Rev. 11, Form 2) CALCULATION CONTINUATION SHEET SHEET: 8 of 12 CALC NO.: S-C-SF-MDC-1810 REV: 7 REF: CONT'D ON SHEET: ORIGINATOR: DATE: REVIEWER: DATE: VERIFIER: DATE: R. Down 2/21/07 K. King 2/22/07 K. King 2/22/07 SFP volume with transfer pool Total volume of SFP3 and transfer pool at 23 feet above the fuel assemblies: 62148 ft 3 Rack volume: 564 ft 3 Volume/fuel assembly: 4.277 ft
# fuel assemblies in Unit 1 pool after core offload: 11373 (Assumption 3.18) + 193 = 1330 assemblies Total fuel assembly volume: 1330
- 4.277 = 5688.4 ft Net water volume: 62148 - 564 - 5688.4 = 55896 ft3 (net SFP volume) 5.5 Discussion of Input Data File "Rfile" This file contains seven lines of input. The following provides a breakdown of the Input Data File:
Line 1: Description of job (freeform comments). Line 2: Reactor shutdown date.
- Cases 1 & 2: 3/28/07 (Actual shutdown @ 20:00 on 3/27/07; rounded forward to next day)
- Case 3: 4/3/07 (Actual shutdown @ 20:00 on 3/27/07; rounded forward to next day)
- Case 4: 5/1/07, 6/1/07, 7/1/07, 8/1/07, then quarterly beginning on 9/1/07 through 9/1/08 Note: For loss of cooling (case 3), this is actually the start date of Crosstie run. The total elapsed time is just 6 days and 18 hours; however, Crosstie cannot model time into the shutdown date, and an elapsed time of 6 days will conservatively be used.
Line 3: Unit in outage.
- Cases1&2: 1
- Cases3&4: 2 Line 4: CC flow, SF flow, SFP water volume: 3000,2500, 55896 (Design values for flows used -- see Assumptions 3.12 & 3.13)
Line 5: batch 1 # assemblies, batch 2 # assemblies, batch 3 # assemblies, decay time before fuel transfer, total transfer time for offload The first 3 numbers are the number of assemblies in each discharge batch. The average burnup for each batch is included in the next line. CROSSTIE has the core unload in 3 batches, with the assumption that about 1/3 has a one-cycle burnup, 1/3 has a two-cycle burnup and the remaining 1/3 has a three-cycle burnup (the 1/3 that will remain in the pool after reload). The 1R1 8 core doesn't quite fit those percentages. The batches are grouped based on the burnups rather than in 3 equal groups. From Attachment 2, the number of assemblies per batch, in descending order of fuel burnup, are taken as: 52, 68, 73 The 4th number is the decay time before fuel transfer. From Section 3.8, the value is 121 hours. The 5th value is the total transfer time for the offload. From Section 3.9, the current schedule is 41 hours. The input lines, then, are:
- Cases 1 & 2: 52,68,73,121,41
- Cases 3 & 4: 0,0,0,0,0 Line 6 It #: Reactor rated power: 3459 MW 2 nd #: Capacity Factor for last 4 months -- about 1.0 from fuels. Note: Since the individual capacity factors are also built into the burnup each individual assembly, 1.0 is entered.
(NC.DE-AP.ZZ-0002(Q), Rev. 11, Form 2) CALCULATION CONTINUATION SHEET SHEET: 9 of 12 CALC NO.: S-C-SF-MDC-1810 REV: 7 REF: CONT'D ON SHEET: ORIGINATOR: DATE: REVIEWER: DATE: VERIFIER: DATE: R. Down 2/21/07 K. King 2/22/07 K. King 2/22/07 3 rd, 4 th & 5 th #'s: Average burnups for 3 batches in line 4 - 51109, 45234, 24740 (based on Attachment 2) 6 th #: Average uranium weight - 456.3 (Attachment 2) The input lines, then, are:
- Cases 1 & 2: 3459,1.0,51109,45234,24740,456.3
- Cases 3 & 4: 3459,0,0,0,0,0 Line 7 Ambient air temperature and RH in Fuel Handling Building. From Section 3.3:
- Cases 1, 2, 3 & 4- bounding cases: 105,1.0
- Cases 1, 2, 3 & 4 - best estimate cases: 75,1.0 These files are included as electronic files on CD in Attachment 4, and are saved as the following input files:
- Cases 1 & 2- bounding case: 1R18clg1.dat
" Cases 1 & 2 - best estimate case: 1R18clg2.dat
- Cases 3 & 4 - bounding case: 1 R1 8locl.dat (case 4 will adjust the shutdown date in line 2)
- Cases 3 & 4 - best estimate case: 1 R1 81oc2.dat (case 4 will adjust the shutdown date in line 2) 5.6 Parameters Inputted at Run Time Input 1: Rfile (*.dat) from Section 5.5 Input 2: Time after shutdown to start crosstie (hrs):
- Case 1:163 (121+41+1) (Assumptions 3.8, 3.9, 3.11)
- Case 2: 500 (Section 5.2 - bounds core reload)
- Cases 3 & 4: 1 (Section 5.2 and Assumption 3.11)
Input 3: Pool water temperature limit for switchover:
- Case 1:180 (limit with one SFHX unavailable)
- Cases 2, 3 & 4: 210 (high enough to prevent swapover)
Input 4: CCW coolant temperature (Assumption 3.4):
- Cases 1, 2, 3 &4 - bounding cases: 99
- Cases 1 & 4- best estimate cases: 80
- Cases 2 & 3 - best estimate cases: 70, 75 and 80 Input 5: Ending time for integration:
- Cases 1 & 2: 500 (bounds core reload)
" Cases 3 &4: 250/500 (high enough to reach equilibrium temperature for Unit 1 pool or boiling)-
determined by trial and error These inputs are also shown on the "result.tem" files on CD in Attachment 4. 5.7 100-hr Limiting Core Offload Time To determine the maximum CC/SW temperature for the 100-hr limiting core offload time, the "Data" file for Case 2 is modified slightly such that "line 5" reflects the 100-hr decay time before fuel transfer (see Section 5.5). The model is run using Case 2 inputs from Section 5.6, and CCW temperatures of 801F and 991F. Internal calculations in the supporting spreadsheet perform an interpolation that determines the approximate CCW limiting temperature to ensure that the SFP will not exceed the licensing basis limit of 149 0 F. A second set of model runs are performed using CCW temperatures one degree above and below the temperature found in the previous step. The results of the second set of model runs are provided in Attachment 5.
(NC.DE-AP.ZZ-0002(Q), Rev. 11, Form 2) CALCULATION CONTINUATION SHEET SHEET: 10 of 12 CALC NO.: S-C-SF-MDC-1 810 REV: 7 REF: CONT'D ON SHEET: ORIGINATOR: DATE: REVIEWER: DATE: VERIFIER: DATE: R. Down 2/21/07 K. King 2/22/07 K. King 2/22/07 5.8 Run the Crosstie Program The model was run for the following scenarios. The output files "result.tem", "unit1 .htl" and "plot.dat" are included as electronic files on CD in Attachment 4 (bounding and best estimate sub-cases). Case 1: Crosstie Operation with one SFHX unavailable Case 2: Normal SFP cooling with no crosstie operation Case 3: Loss of SFP cooling in Unit 1 Case 4: Loss of SFP cooling in Unit 1 - Post-outage 5.9 Import the Output Files The PLOT.DAT file for each unit was imported into EXCEL, and a temperature vs. time graph was plotted for each unit. The graphs can be found in the Attachment 3.
6.0 CONCLUSION
S The Unit 1 SFP analysis for 1 R1 8 was performed for the following cases, with one heat exchanger available and Crosstie swapover at 180°F: Case 1- Crosstie Operation Case 2- Normal SFP Cooling to Unit 1 Case 3- Loss of Unit 1 SFP Cooling Case 4- Loss of Unit 1 SFP Cooling - Post Outage For each of these cases, a bounding case with design inputs and best estimate cases based on estimated input parameters for the current 1 R1 8 schedule were run. All cases were run with CC supply temperatures of 99°F and 80°F, which correlate to maximum SW temperatures of 90°F and 71'F, respectively. Cases 2 and 3 were also run with CC supply temperatures of 75°F and 700F1, which correlate to maximum SW temperatures of 66 0 F and 61°F, respectively. Plots showing "SFP temperature vs time" for each case are included in Attachment 3. A summary of the results is as follows: Case 1: For crosstie operation with the Unit 1 SFP aligned to the Unit 2 SFHX, a summary of the results is included in the table below. Bounding: The isolated Unit 2 peak SFP temperature hits the licensing basis limit of 180'F with one SFHX isolated, and swapping of SFPs is required. Best Estimate: The isolated Unit 2 peak SFP temperature is below the licensing basis limit of 180'F with one SFHX isolated, and thus no swapping of SFPs is required. In addition, the time at which the peak temperature is reached is after the scheduled core reload, by which time crosstie operation would most likely have been suspended and normal cooling restored. Case Unit Offload- Hx in Peak Temp (OF) Time to reach (hr) Heatup (°F/hr) Start (hr) Service 99 80 99 80 99 80 1 1 121 yes 180 135 4 (1) 50(2) 7.8 1.2 1 2 121 no 180 178 60_M 140k77 1.2 1.3 (1 Time from Crosstie swapover (2)Time from "start of off-load" The new CC temperature cases of 75°F and 70'F were not performed for Case 1, since the results for the previously performed temperature cases indicate the Unit 2 SFP never reaches 180'F. Thus no swapover between the SFPs occurs, and the Case 1 results for the Unit 1 SFP are the same as for Case 2. The new CC temperature cases of 75°F and 70°F were not performed for Case 4 since these are post-outage cases with low heat load conditions; also, the CC temperature would likely be set between 80'F and 99°F.
(NC.DE-AP.ZZ-0002(Q), Rev. 11, Form 2) CALCULATION CONTINUATION SHEET SHEET: 11 of 12 CALC NO.: S-C-SF-MDC-1810 REV: 7 REF: CONT'D ON SHEET: ORIGINATOR: DATE: REVIEWER: DATE: VERIFIER: DATE: R. Down 2/21/07 K. King 2/22/07 K. King 2/22/07 Case 2: A summary of the results is included in the table below. For normal SFP cooling, the licensing basis limit of 149 0 F is exceeded for the bounding case with 99°F CC temperature. However, interpolating between the 80°F CC temperature case and bounding case results, a peak SFP temperature of 149°F is reached with a CC temperature of about 94 0 F, correlating to a maximum SW temperature of 85 0 F. Since this will be higher than the SW temperature at the time of core offload, the SFP temperature limit of 149 0 F will not be exceeded. Case Offload Peak Temp (OF) Time to 125 0 F (hr)* Heatup (°F/hr) Start (hr) 99 80 75 70 99 80 75 70 99 80 75 70 2 121 154 135 131 126 20.5 36.1 40.3 47.4 1.2 1.2 1.2 1.2
*Time from "start of off-load" Also, through linear interpolation of the results, the SFP high temperature alarm setpoint of 125°F will be reached with a CC temperature of 69°F for an offload start time of 121 hours. Alarm Response Procedure S2.OP-AR.ZZ-0003 allows the setpoint to be increased to allow refueling activities to continue. Temporary alarm setpoints as a function of CC temperature, if required, are provided in the table below. The setpoints are set to a value 50 F higher than the calculated peak SFP temperature. This accounts for a 2.5 0 F instrument uncertainty (Reference SAP ICD screen for FLOC S2SF -2TIC651) plus provides a 2.5 0 F margin above the peak temperature."
CC temperature (F) Alarm Setpoint (OF) 80 140 75 136 70 131 Case 3: A summary of the results is included in the table below. On a loss of cooling to the Unit 1 SFP, the maximum design limit of 180'F will be reached in a range of 2.2 hours to 4.9 hours after core offload is complete. This is the time operators have to take contingency actions to re-establish forced cooling. The Unit 1 SFP will not boil if cooling is not restored. Case Offload Peak Temp (OF) Time to reach 180°F (hr)* Heatup (°F/hr) Start (hr) 99 80 75 70 99 180 75 70 99 80 75 70 3 121 205 205 205 205 2.2 4.0 4.5 4.9 9.4 9.9 10.0 10.1
*Time from "loss of cooling" after core offload complete Case 4: Heat-up rates in the event of Unit 1 SFP loss of cooling post-outage. See curves for details.
Case 5: Maximum CC/SW Temperatures for 100-hr Limiting Core Offload The maximum river temperature is based on a maximum CC supply temperature, to ensure that the SFP will not exceed the licensing basis limit of 149 0 F. For the Tech Spec minimum offload start time of 100 hours, the maximum SW and CC temperatures are 82°F and 91°F, respectively (see Attachment 5).
(NC.DE-AP.ZZ-0002(Q), Rev. 11, Form 2) CALCULATION CONTINUATION SHEET SHEET: 12 of 12 CALC NO.: S-C-SF-MDC-1810 REV: 7 REF: CONT'D ON SHEET: ORIGINATOR: DATE: REVIEWER: DATE: VERIFIER: DATE: R. Down 2/21/07 K. King 2/22/07 K. King 2/22/07 7.0 IMPACT TO STATION PROCEDURES: None 8.0 DOCUMENTS AFFECTED: None 9.0 DESIGN MARGIN: This calculation is used to determine heat-up rates for the SFP during refueling outages. It provides a planning tool for the Central Outage Group (COG) and Operations to plan fuel moves to ensure SFP temperature are manageable, and allows contingency planning in the event that a pump and/or heat exchanger is lost. Design margin is not applicable to this calculation. 10.0 CROSS
REFERENCES:
Cross-References - Critical Software S-C-SF-MCS-01 13 and Design calculation S-C-SF-MDC-1 780 were used as input for development of this calculation. There are no output documents resulting from this calculation.
2007 ActIft Oriq Total Early Early MIARAP ID Dur Float? Start Finish l.fl~ 116 l9~ 123 126 i 12 19 MkATMf`YaA0W 0 -"P-TRIPRREAC-TORP&_TURBINE 4j-( AME D 01 7, /. fe o 2N;U 2D aqO1f
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Start Date 03DEC06 00:00 Early1Bar 800 Sheet1 oIf 10 L2- LEVEL 2 WINDOW SCHEDULE Finish Date 31 MAY07 18:59 EG NUCLEAR LLC - SALEM 1*1 Shaded Areas: Data Date 25MAR07 00:00 Progress Bar 1st area - Midloop LEVEL 2 SCHEDULE 2nd area - Core Offload Run Date 21NOV06 08:18 Critical Activity 3rd area - Core Reload 4th area - Midloop © Primavera Systems, Inc.
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NUCLEAR FUELS TRANSMITTAL OF DES-SGN INFOR--MATON 0 SAFETY RELATED Originating Organization NF ID# NF0600209 E] NON-SAFETY RELATED [ Nuclear Fuels Revision 0 17 REGULATORY RELATED El Other (specify) Page 1 of 6 Station: Salem Unit: 1 Cycle: 19 Generic: To: Alan Johnson (Design Engineering Supervisor)
Subject:
Salem 1 RFO1 8 Assembly Burnup Data for SFP Heat Load Analyses, Rev 0 Joe Dascanio 8/2/06 Prepared by (igna(re Date Keith Robinson 8/2/06 Reviewed by Signature Date Bob Tsai Aw-) ~ (W ~ ~ 4 812/06~ Approved by Signature Date Status of Information: [ Verfied El Unverified C] Engineering Judgement Action Tracking # for Method and Schedule of Verification for Unverified DESIGN INFORMATION: Purpose.of Information.- The atta.c.hed data is being provided as requested input into the CROSSTIE computer code (or equivalent) used in performing decay heat analyses. Description of Information: The data Is based upon a conservative EOCI8 bumup assumption of 19,900 MWD/MTU. That bumup covers a capacity factor of over 100% from 08/01/2006 to the start of the IR18 refueling outage on 3/27/2007. The following pages contain the assembly identifiers (IDs), projected assembly burnups at EOC1 8, the assemblies to be discharged to the spent fuel pool at EOC1 8, and the assemblies to be reinserted into Cycle 19 from the spent fuel pool. In addition, the total core loading and core average assembly loading are provided. Source of Information: Westinghouse Letter, NF-PSE-06-34, "Salem Unit 1 Core Follow Models for The Cycle 19 Design," July 5, 2005. Output File: 12slci8 ancqcjrev.3837.out Supplemental Distribution: EM-0l: Hard Cop R. Down Salem Records Management I
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NUCLEAR FUELS NF0600209, Revision 0 TRANSMITTAL OF DESIGN INFORMATION Page 2 of 6 Assumed Capacity Factor of 100% from 08/01/06 to 3/27/07. Total Core Loading = 88.058 MTU. Therefore, Core Average Assembly Loading is 456.3 KgU/Assembly. The following is the list of Cycle 18 assembly identifiers (total of 193), their corresponding EOC18 projected assembly burnups in units of MWD/MTU, and the assemblies discharged to the SFP at E0C18, where EOC18 = 19,900 MWD/MTU. EOC18 Assemblies Assembly Discharged Assembly ID Burnup (MWD/MTU) to SFP AF01 44799 x AF02 44773 x AF03 44773 x AF04 44799 x AF05 44799 X* AF06 44773 AF07 44773 x AF08 44799 x AF10 51035 x AF1 1 50304 x AF12 50844 x AF17 50391 x AF1 8 50844 x AF1 9 51035 x AF20 50304 x AF21 51035 x AF24 52638 x AF25 50391 x AF26 52638 x AF27 50844 x AF28 50381 x AF31 50844 x AF32 50391 x X AF33 51035 x AF39 50391 x AF41 50381 x AF47 52638 x AF50 50381 x AF51 50304 x AF52 50381 x AF54 52638 x AF56 50304 x AF57 52588 x AF58 52464 x AF59 52464 x AF60 52588 x AF61 52588 x AF62 52464 x AF63 52464 x S-C--F -I),)bc- /9/0 -7 /8/0,eev S~c<F,~1bC / 9 Tq-?67-ýq Z -, 2'q
NUCLEAR FUELS NF0600209, Revision 0 TRANSMITTAL OF DESIGN INFORMATION Page 3 of 6 EOC18 Assemblies Assembly Discharged Assembly ID BurnuD (MWD/MTU) to SFP AF64 52588 x AF65 51714 x AF66 54741 x AF67 54741 x AF68 51714 x AF69 51714 x AF70 54741 x AF71 54741 x AF72 51714 x AGO1 47296 x AG02 47296 x AG03 46867 x AG04 48756 x AG05 46826 x AGO6 46867 x AG07 48756 x AGO8 46826 x AG09 47296 x AG10 46826- x AG11 48756 x AG12 47304 x AG13 47304 x AG14 48756 x AG15 47304 x AG16 46867 x AG17 47304 x AG18 46867 x AG19 46826 x AG20 47296 x AG21 43511 AG22 43496 AG23 43496 AG24 43511 AG25 43511 AG26 43496 AG27 43496 AG28 43511 AG29 45038 AG30 44096 AG31 45118 AG32 45118 AG33 44044 AG34 45038 AG35 44096 AG36 44096 AG37 44096 AG38 44044 AG39. 45118
- sp-MiDc- /8 /0 W,5v ' ~v'7 I~TT,9C,141L~>J7 Z 3 ~y2 ~
NUCLEAR FUELS NF0600209, Revision 0 TRANSMITTAL OF DESIGN INFORMATION Page 4 of 6 EOC18 Assemblies Assembly Discharged Assembly ID Burnup (MWD/MTU) to SFP AG40 45038 AG41 45118 AG42 44044 AG43 44044 AG44 45038 AG45 40739 AG46 46860 X AG47 46278 AG48 46278 AG49 45059 AG50 45059 AG51 46278 AG52 46870 X AG53 46860 X AG54 40739 AG55 40739 AG56 45059 AG57 46278 AG58 46870 X AG59 49280 X AG60 49280 X AG61 49280 X AG62 46860 X AG63 45059 AG64 46860 X AG65 49280 X AG66 49280 X AG67 49280 X AG68 49280 X AG69 49280 X AG70 40739 AG71 46870 X AG72 46870 X AH01 25030 AH02 26700 AH03 26319 AH04 26700 AH05 26319 AH06 26777 AH07 26706 AH08 26266 AH09 26282 AH10 26700 AH11 25180 AH12 26389 AH13 26700 AH14 25180 AH15 26266 S- - - - /n;D d- / b, /0 /Ls v '7 /q'7-/T, C /Y Z-?,4. AOY I "i
NUCLEAR FUELS NF0600209, Revision 0 TRANSMITTAL OF DESIGN INFORMATION Page 5 of 6 EOC18 Assemblies Assembly Discharged Assembly ID Burnup (MWD/MTU) to SFP AH16 26706 AH17 26706 AH18 26282 AH19 26389 AH20 26266 AH21 26389 AH22 26706 AH23 25180 AH24 26073 AH25 26266 AH26 25180 AH27 26073 AH28 26792 AH29 26777 AH30 26792 AH31 26389 AH32 26792 AH33 26073 AH34 26073 AH35 26777 AH36 26282 AH37 26282 AH38 26319 AH39 26792 AH40 26319 AH41 26777 AH42 20807 AH43 22669 AH44 20790 AH45 20790 AH46 22622 AH47 20807 AH48 22669 AH49 22669 AH50 22669 AH51 22622 AH52 20790 AH53 20807 AH54 20790 AH55 22622 AH56 22622 AH57 20807 AH58 23309 AH59 24062 AH60 24062 AH61 23309 AH62 24117 AH63 23309
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NUCLEAR FUELS NF0600209, Revision 0 TRANSMITTAL OF DESIGN INFORMATION Page 6 of 6 EOC18 Assemblies Assembly Discharged Assembly ID Burnup (MWD/MTU) to SFP AH64 23395 AH65 23395 AH66 24062 AH67 24117 AH68 23309 AH69 23395 AH70 24062 AH71 24117 AH72 24117 AH73 23395 No Assemblies will be reinserted from the Spent Fuel Pool into Cycle 19. 5--C - -5117' - In b C - /8 /0 pEv -7 1,-- A -/ -i /0--Y 4 " - 4ý
Design Calculation S-C-SF-MDC-1810, Rev. 7 Final Attachment 3 page / of Unit I - Crosstie Operation Swapover at 180TF (CCW 3000 gpm, SF 2500 gpm) Plant Shutdown (3/27/07) - 20:00 hr, Offload Start - 121 hr, Offload Complete - 162 hr 200 180 160 CL 140 E I-- U-
-* 120 100 80 60 4-100 150 200 250 300 350 400 450 500 time (hrs)
Case 1a - Bounding
Design Calculation S-C-SF-MDC-1810, Rev. 7 Final Attachment 3 page I- of 11 Unit I - Crosstie Operation Swapover at 180°F (CCW 3000 gprn, SF 2500 gpm) Plant Shutdown (3127107) - 20:00 hr, Offload Start - 121 hr, Offload Complete - 162 hr 200D, unit 2 - SFP ambient air @7 °F and 100% RH, 80F CCW a-ar~d 71F SW 180 -__ 160 _
': 140 E
U.
-L 120.
unit 1 - SFP mbient air@7 Fand 100% RH, 80F CCW and 71F SW 100 80 60 100 150 200 250 300 350 400 450 500 time (hrs) Case lb - Best estimate
- Design Calculation S-C-SF-MDC-1810, Rev. 7 Final i 57 Attachment 3 page 3 of Unit I - Normal Cooling - No Crosstie (CCW 3000 gpm, SF 2500 gpm)
Plant Shutdown (3/27107) - 20:00 hr, Offload Start - 121 hr, Offload Complete - 162 hr 160 unit 1-SFP ambient ir @105F a nd 100% RV-I, 99F CCW and 90F SW 150 140 130
,, 120 E
- a. 110 LL unit 1- SFP ambient ai @75°F an2 100% RH, 80F COW and 71F SW
-7 100 unit 1 - SFI ambient aii @75°F and 100% RH, 75F CCW and 66F SW 90 unit 1 - SFI ambient aiý @75°F and 100% RH, 70F CCW a d 61F SW 80
_ _1 70 60 0 50 100 150 200 250 300 350 400 450 500 time (hrs) Case 2
Design Calculation S-C-SF-MDC-1 810, Rev. 7 Final Attachment 3 page I of /*5 Unit 1 - Loss of Cooling Based on 4/3107 Start Date at 00:00 hr - With Core Offloaded 210 I I I unit 1- SFP ambient air @105'Fad apd 100%oRl 10 R 99FCCWand W 90FSV\/V I 200 1 1901-1/ x unit 1 - SF ambient air @750F and 100% RH, 80F CCW aj id 71F SW 180 unit 1 - SFP ambient ai1 @75'F anc 100% RH, 75F CCW ai id 66F SW 170 -j A-I---,' 4 4 unit 1 - SEP ambient ai1 @75°F and 100% RH, 70F CCW ai id 61F SW I-- 0. S160 150 140 I ____________________ ____________________ ___________________ 130 -f 4 -b 120 F- F- F- 4 -F F- F-0.0 0O 2.00 4.00 6.00 8.00 10.00 12.00 14.00 16.00 18.00 20.00 time (hrs) Case 3
Design Calculation S-C-SF-MDC-1810, Rev. 7 Final Attachment 3 page cf /5-Unit 1- Loss of Cooling Based on loss of cooling on 5/11/07 210 - 200 190 180 170
- 0. 160 E
CL I-150 U. 140
-1, 130 120 110 100 90 80 0 10 20 30 40 50 60 time (hrs) case 4a
Design Calculation S-C-SF-MDC-1 810, Rev. 7 Final Attachment 3 page ýf , S-Unit 1- Loss of Cooling Based on loss of cooling on 6/1/07 210 200 190 180 170 160 C. 0.. 150 I-L- (n 140 130 120 110 100 90 80 0 10 20 30 40 50 60 70 80 time (hrs) case 4b
.Design Calculation S-C-SF-MDC-1810, Rev. 7 Final Attachment 3 page 7 of I Unit 1- Loss of Cooling Based on loss of cooling on 7/1/07 210 200 190 180 170 160 E 150 q-U- Cl) 140 130 120 110 100 90 80 0 10 20 30 40 50 60 70 80 90 100 time (hrs) case 4c
Design Calculation S-C-SF-MDC-1810, Rev. 7 Final Attachment 3 page 8 of 15 Unit 1- Loss of Cooling Based on loss of cooling on 8/1/07 200 190 180 _____ 170 _ _ _ _ _ ____ 160 h C. 150 /_ _ _ _ E 105'F SFP ambient air @ 100% RH - 99TF CCW C- 140 - CC
- 1 0...... 75TF SFP ambient air @ 100% RH - 80TF CCW 7- 130 -_ _ __/
120 --- _ 110 -, 100 -- 90" 80 2 4 i , 0 10 20 30 40 50 60 70 80 90 100 time (hrs) case 4d
.Design Calculation S-C-SF-MDC-1810, Rev. 7 Final Attachment 3 page 9 of 1 Unit 1- Loss of Cooling Based on loss of cooling on 9/1/07 200 190 180 170 160 15O E 140 (L1 U. 130 120 110 100 90 80 0 10 20 30 40 50 60 70 80 90 100 time (hrs) case 4e
Design Calculation S-C-SF-MDC-1810, Rev. 7 Final Attachment 3 page 10 of 1 Unit 1- Loss of Cooling Based on loss of cooling on 12/1/07 190 180 170 160 150 E 140 I-- 1.1 LL Cn 130 120 . 110 100 / 90 80 0 20 40 60 80 100 120 time (hrs) case 4f
I Final Attachment 3 page 11 of 1 ,*
- Design Calculation S-C-SF-MDC-1810, Rev. 7 Unit 1- Loss of Cooling Based on loss of cooling on 3/1/08 190 180 170 160 150 C.
E 140 a) U. U-130 120 110 100 90 80 0 20 40 60 80 100 120 140 time (hrs) case 4g
IDesign Calculation S-C-SF-MDC-1810, Rev. 7 Final Attachment 3 page 12 of 15 Unit 1- Loss of Cooling Based on loss of cooling on 611/08 190 180 170 160 150 iZZ E 140 I-a. i 130 120 110 100 90 80 0 20 40 60 80 100 120 140 160 time (hrs) case 4h
,Design Calculation S-C-SF-MDC-1810, Rev. 7 Final Attachment 3 page 13 of 1 Unit 1- Loss of Cooling Based on loss of cooling on 9/1108 190 180 I -i- - 170 160 150 U-E 140 I-
//
.nLL. 105TF SFP ambient air @ 100% RH - 99TF CCW U,. 130 75 // SFP ambient air @ 100% RH - 80 0 F CCW
.75TF CO 120 110 100 90
/
80 1* I t -1 0 20 40 60 80 100 120 140 160 time (hrs) case 4i
Design Calculation S-C-SF-MDC-1 810, Rev. 7 Final Attachment 3 page 14 of Unit 1 - Normal Cooling - No Crosstie (CCW 3000 gpm, SF 2500 gpm) Plant Shutdown (3/27/07) - 20:00 hr, Offload Start - 100 hr, Offload Complete - 141 hr 160-140 130 -___ _--- ._______ G 120 - C. un t 1 - SFP ambient air @ 75°F and 10 0% RH, 80F CCW and 1F SW E 110 IL U. S100 90 80-70 60 0.0 50.0 100.0 150.0 200.0 250.0 300.0 350.0 400.0 450.0 500.0 time (hrs) Case 5a - 100-hr limiting
Design Calculation S-C-SF-MDC-1 810, Rev. 7 Final Attachment 3 page 15 of 1 i Unit I - Normal Cooling - No Crosstie (CCW 3000 gpm, SF 2500 gpm) Plant Shutdown (3127107) - 20:00 hr, Offload Start - 100 hr, Offload Complete - 141 hr 150 r T - 149
/
mit 1 - SFP bmbient air 75°F and 00% RH, 9,F CCW andt 83F SW I i ?O 140 130 Unit 1 - SFP ambient air @W5°F and 109:% RH, 91F ICOWand ZF~& 0. E I-a. U. 120
/
(1-110 100 1 90 0.00 50.00 100.00 150.00 200.00 250.00 300.00 350.00 400.00 450.00 500.00 time (hrs) Case 5b - 100-hr limiiting
S-C-SF-MDC- 1810 Revision 7 Attachment 4 pa.-e I of I Attachment 4 CROSSTIE Input and Output Files (Electronic files on CD) Input files SFP inventory data files
" Unitl-17.dcy
- Unitl-18FC.dcy 0 Unitl-18POST.dcy
" Unit2.dcy Input Data File "Rfile".
- 1R18clgl(2).dat
- 1R181ocl(2).dat
& 5-1-07a(b).dat
- 6-1-07a(b).dat 0 7-1-07a(b).dat
- 8-1-07a(b).dat
- 9-1-07a(b).dat
- 12-1-07a(b).dat 0 3-1-08a(b).dat 0 6-1-08a(b).dat
. . ................---.... g---1--08a(b)-dat-- -.- - Output files - The output files are included for each case for both the bounding and best-estimate sub-case with the following hierarchy: Cases 1, 2, 3 and 5
> Bounding
> Plot.dat
> Result.temp
> Unitl.htl
> Unit2.htl
> Best estimate
> Plot.dat
> Result.temp
> Unitl.htl
> Unit2.htl Case 4
> 5-1-07 (typical for each date)
> Bounding
> Plot.dat
, Result.temp
> Unitl.htl
> Unit2.htl
> Best estimate
> Plot.dat
> Result.temp
> Unitl.htl
> Unit2.htl
SC.OM-AP.ZZ-0001(Q) REVISION I ATTACHMENT 5 SALEM VERIFICATION OF DECAY HEAT REMOVAL FOR CORE OFF-LOAD Ref Order/Notification: Calculation Number: S-C-SF-MDC-1810. Rev. 7 Attachments: Calculation cover sheet, conclusion and heatup curves Required maximum river inlet temperature to support core off load at 100 hours after sub-criticality is 82 0F The minimum time in which core off load could be conducted and adequate decay heat removal would exist in the spent fuel pool is 100 hours after sub criticality. Engineer: Robert Down Date: 2/14/07 Engineering Supervisor: Alan Johnson k \4\ A Date: z]l'!7 IJ Copy to: Control Room Supervisor Outage Control Center - Shift Outage Manager RAT Notes:
- 1. Max river temperature based on a max CC supply temperature of 91TF, to ensure SFP will not exceed licensing basis limit of 149_°F, For current scheduled offload start of 121 hours, max SW and CC temperatures are 85TF and 94TF, respectively.
- 2. SFP high temperature alarm setpoint of 125°F will be reached with a CC temperature 69 0F, for the scheduled offload start of 121 hours (corresponding.SW temperature m60TF). Ifthe alarm setpoint is reached, Alarm Response Procedure S2.OP-AR.ZZ-0003 (Alarm C-1 9) allows the setpoint to be temporarily increased to allow refueling activities to continue. Ifnecessary to reset, the temporary alarm setpoint is 140 0F, conservatively based on a CC temperature of 800F.
- 3. Calculation results assume one SFHX and one CCHX in service.
..S-xCwSF--MDC-i-181-0 Revision Attachment 5 Page 1 of I
USER RESPONSIBLE FOR VERIFYING REVISION, STATUS AND CHANGES PRINTED 20070305 PSEG Internal Use Only Page 1 of I PSEG NUCLEAR L.L.C. SC.OM-AP.ZZ-0001(Q) - REV. 1 SHUTDOWN SAFETY MANAGEMENT PROGRAM - SALEM ANNEX SPONSOR ORGANIZATION: Outage Management REVISION
SUMMARY
- Biennial Review Required: Yes __ No 4i The following changes are;
- 1. Changed the definition of "Available" to be consistent with OU-AA-103, Shutdown Safety Management Program.
- 2. Editorial changes to text throughout for preferential wording as instructed by sponsor.
Implementation Requirements Effective Date: 5- ]( h~lGW-At wihCU. 03, Shutdevn Safety MmangacmoPfgnt Przgzdrue Paid roplaezs NC.M- .A,p Zz . 0. , Rz- , auLagr kwtao ýen
- uocde.,e... "V W*.r16*-'4 Approved:
Outage Manager Date
USER RESPONSIBLE FOR VERIFYING REVISION, STATUS AND CHANGES PRINTED 20070305 SC.OM-AP.ZZ-0001(Q) SHUTDOWN SAFETY MANAGEMENT PROGRAM - SALEM ANNEX TABLE OF CONTENTS Section Title Page 1.0 PURPOSE ........................................................................................................................ 3 2.0 SCOPE .............................................................................................................................. 3 3.0 RESPONSIBILITIES ................................................................................................... 3
4.0 BACKGROUND
......................................................................................................... 3 5.0 PROCEDURE ............................................................................................................. 4 5.1 Outage Risk Assessment Process ........................................................................ 4 5.2 ORAM Contingency Planning ............................................................................. 5 5.3 Outage Risk Assessment and Management (ORAM) Software ........................ 5 5.4 ORAM Software and M odel Changes ................................................................ 6 5.5 Risk Assessment during Outage Execution ....................................................... 6 5.6 Forced Outages .................................................................................................. 8 5.7 Salem Integrated Decay M anagement ................................................................. 8 6.0 RECORDS ........................................................................................................................ 9 7.0 DEFINITIONS ............................................................................................................ 9
8.0 REFERENCES
......................................................................................................... 10 Page 1 of 46 Rev. I
USER RESPONSIBLE FOR VERIFYING REVISION, STATUS AND CHANGES PRINTED 20070305 SC.OM-AP.ZZ-OOO1(Q) TABLE OF CONTENTS (Continued) Section T itle ATTACHMENT S Attachment 1 Salem Risk Assessment Review Guidelines ............................................... 12 Attachment 2 Salem Shutdown Safety Considerations .................................................... 15 Attachment 3 ORAM Contingency Planning ................................................................... 25 Attachment 4 ORAM Contingency Plan Content ............................................................. 27 Attachment 5 Salem Verification of Decay Heat Removal for Core off Load .................. 30 Attachment 6 Decay Heat Load and Heat up Curves/Tables - Development Time Line ....... 31 Attachment 7 Salem Shutdown Safety Assessment Worksheet ........................................ 34 Attachment 8 Salem Shutdown Risk Status Sheet ........................................................... 39 Attachment 9 Transitional Modes Guidelines (TMG) - Salem I and 2 ........................... 41 Attachment 10 Cross-Unit Heightened Awareness Equipment List ................................... 47 Page 2 of 46 Rev. 1
USER RESPONSIBLE FOR VERIFYING REVISION, STATUS AND CHANGES PRINTED 20070305 SC.OM-AP.ZZ-0001(Q) 1.0 PURPOSE 1.1 This manual defines the PWR key safety functions and provides guidance for the deterministic status assessment for each function. 1.2 On a case by case basis, the station may dictate a specified set of component(s) resulting in a different assessment than prescribed in this manual. If a variation from the guidance specified in this manual is deemed appropriate, the rationale and the Shutdown Safety Review Board (SSRB) approval shall be documented on form OU-AA-103 Attachment 1, Shutdown Safety Approval form. The SSRB has the final authority to determine what constitutes compliance specified herein. 2.0 SCOPE NOTE This procedure should be implemented in conjunction with OU-AA-103, Shutdown Safety Management Program 2.1 This procedure applies to the planning, scheduling, and execution of work on a unit already in or expected to be in a shutdown mode of operation. The actual periods of applicability are determined on a site-specific basis in conjunction with the online risk process. 2.2 This procedure does not apply to units that are permanently shutdown. 3.0 RESPONSIBILITIES 3.1 Responsibilities are described in OU-AA-103, Shutdown Safety Management Program.
4.0 BACKGROUND
4.1 Shutdown Safety Management Programs (SSMP)
- The SSMP uses as its basis the philosophy and recommendations stated in NUMARC 91-06, "Guidelines for Industry Actions to Assess Shutdown Management".
- The SSMP is also designed to meet the applicability requirements of 10CFR50.65a(4) and NUMARC 93-01, "Industry Guidance for Monitoring the Effectiveness of Maintenance at Nuclear Power Plants".
- This procedure is not intended to meet the requirements specified in the SAR or Technical Specifications.
Page 3 of 46 Rev. 1
USER RESPONSIBLE FOR VERIFYING REVISION, STATUS AND CHANGES PRINTED 20070305 SC.OM-AP.ZZ-0001(Q) 5.0 PROCEDURE 5.1 Shutdown Safety Review Process 5.1.1 Outage risk assessments in preparation for planned/refueling outage execution are performed by analyzing the outage schedule in the ORAM software or manually using hard copies of the Safety Function Assessment Trees (SFAT) in the modes where the ORAM computer model is applicable as well as Attachments 1 and 2. During transition modes, the applicable guidance of this procedure is used. 5.1.2 The SSRB should assess the adequacy of the DEFENSE-IN-DEPTH provided for the duration of the outage. This assessment should also include a detailed examination of the outage schedule, including system interactions, support system availability, logic ties and the impact of temporarily installed equipment. When entry into Mode 5 is planned, the ORAM software should be used to assist in the risk assessment. The SSRB members should first use paper copies of the SFAT logic in order to study the schedule and to form an opinion about the resulting risk, and then the ORAM computer tool should be used to validate the results. Discrepancies will either be the result of improper schedule coding or human error in the risk assessment. Once these discrepancies are resolved, there will be a high degree of confidence in the accuracy of the risk assessment. The assessment should include consideration of the guidance in Attachments 1 & 2 - Salem Risk Assessment Review Guidelines & Salem Shutdown Safety Considerations. 5.1.3 For planned/refueling outages, the SSRB membership quorum requires 3 members to conduct business. The Shutdown Safety Manager chairs the SSRB. One member must hold a current operating license. One member must be a nuclear engineer (NE)/Reactor Engineer (RE) for issues involving reactivity Control Key Safety Function. Other members could also include knowledgeable representatives from the following departments:
" Operations
" Engineering
" Maintenance
" Radiation Protection
- Chemistry a Work Management
" Training 5.1.4 Continuity of SSRB membership is highly desirable and consideration should be given to not changing team members once assigned.
5.1.5 The SSRB should not include those directly involved in the development of the outage schedule. 5.1.6 The Shutdown Safety Manager shall document all reviews, recommendations, approvals, and other actions taken by the SSRB using OU-AA-103, Attachment 1. Page 4 of 46 Rev. 1
USER RESPONSIBLE FOR VERIFYING REVISION, STATUS AND CHANGES PRINTED 20070305 SC.OM-AP.ZZ-0001(Q) 5.1.7 The SSRB chair is responsible to deliver a pre-outage Risk Assessment to the Outage Manager. The report should include the following elements;
- Overall risk profile for the outage
" Planned entries into Yellow/Orange/Red conditions
- Contingency plans
- Controls in place to prevent inadvertent entry into a decrease defense in depth color change 5.1.8 The final SSRB pre-outage assessment report and associated contingency plans should be reviewed and approved by the SORC.
5.1.9 The risk assessment should be validated after final schedule issuance (integrated reviews). The Shutdown Safety Manager is responsible for coordinating this validation to ensure safety function logic ties are maintained by reviewing the Level 2 schedule. Any problems identified from the validation would be presented to the Outage Manager for resolution. 1700159841 5.2 ORAM Continaency Plannin[ 5.2.1 Any required ORAM contingency plans shall be developed and processed in accordance with Attachment 3, ORAM Contingency Planning, and Attachment 4, ORAM Contingency Plan Content. The use of PRA and ORAM information should be considered when developing ORAM contingency plans. 5.2.2 ORAM contingency plans are returned to the Outage Manager for development of the post-outage critique. Retention of ORAM contingency plans is not required. 5.3 Outage Risk Assessment and Management (ORAM) Model 5.3.1 ORAM determines the level of "DEFENSE-IN-DEPTH" for the following KEY SAFETY FUNCTIONS:
" Shutdown Cooling
- Containment
" Electrical Power Availability
- Service Water
" Inventory Control
- Swgr Penetration Area Ventilation
" Reactivity Control
- CREACS
" Spent Fuel Pool Cooling 5.3.2 Outage schedules are analyzed by the ORAM software and colors assigned to designate the risk level to each of the safety functions. The risk levels (colors) assigned represent the DEFENSE-IN DEPTH that is AVAILABLE for each safety function. ORAM determines the appropriate risk level by analyzing the availability of selected individual pieces of equipment that are necessary to support the safety function.
Page 5 of 46 Rev. 1
USER RESPONSIBLE FOR VERIFYING REVISION, STATUS AND CHANGES PRINTED 20070305 SC.OM-AP.ZZ-0001(Q) 5.3.3 The following plant conditions correspond to the ORAM assigned risk levels (colors): GREEN Based on the combination of available pathways and activity types, a failure or error could be easily mitigated without presenting a significant challenge in that Key Safety Function. YELLOW Based on the combination of available pathways and activity types, a failure or error can still be mitigated but might present a challenge in that Key Safety Function. ORANGE Based on the combination of available pathways and activity types, a failure or error would potentially lead to the loss of the Key Safety Function. RED Based on the combination of available pathways and activity types, the Key Safety Function is potentially not maintained. 5.3.4 As a backup, printed copies of the SFATs and Fault Trees from the ORAM model can be used to perform manual risk assessment. 5.4 ORAM Software and Model Changes 5.4.1 The ORAM Program Administrator maintains configuration management of the ORAM Software and Model. The Operations Manager, or licensed designee provides guidance on revisions or enhancements to the ORAM model, in accordance with the ORAM Software and Model Quality Assurance Plan. 5.4.2 Proposed revisions to the ORAM Model should be processed using the Notification process in SAP to create a NUTS Order. The Notification should be assigned to the ORAM Program Administrator. 5.4.3 When necessary due to ORAM model changes, the ORAM Program Administrator shall issue a revision request (Notification) to Operations to update the Safety Assessment Worksheet, Attachment 7. 5.4.4 The ORAM Program Administrator shall retain a record of all implemented ORAM model changes. 5.4.5 The Operations Manager shall assign an actively-licensed SRO to verify that changes have been programmed correctly. When a change has been satisfactorily validated, the SRO shall sign, date and return the change request to the ORAM Program Administrator, who should issue the change in accordance with the Software Quality Assurance Plan. At this point, the ORAM model is considered updated. 5.5 Risk Assessment during Outage Execution 5.5.1 Outage risk assessments during outage execution are performed by analyzing the outage schedule in the ORAM software or manually using hard copies of the SFATs in the modes where the ORAM computer model is applicable. During transition modes, the applicable guidance of this procedure is used. 5.5.2 The Shutdown Safety Manager or designee should perform a risk assessment once per shift following each schedule update, as well as when significant changes in plant conditions or outage schedule occur which could potentially impact the risk assessment. Page 6 of 46 Rev. I
USER RESPONSIBLE FOR VERIFYING REVISION. STATUS AND CHANGES PRINTED 20070305 SC.OM-AP.ZZ-0001(Q) 5.5.3 Assessments should be performed for significant change in plant configuration prior to a planned evolution or scope change which has the potential to impact one of the KEY SAFETY FUNCTIONS, and for which it has not been previously analyzed. 5.5.4 During cold shutdown and refueling modes, a Shutdown Risk Status Sheet Attachment 8 or similar, provided the similar forms contain the same required information, should be generated that includes the following as a minimum: A. Designated ORAM color code for each KEY SAFETY FUNCTION. B. Overall shutdown risk potential (color). (Overall safety is equal to the "worst" color of any Key Safety Function.) C. Important protected systems, equipment, instrument channels and/or protected areas. D. Active ORAM contingency plans. E. Time to boil/design limit for fuel located in reactor vessel and/or Spent Fuel Pool.
" Should track time to boil for reactor vessel whenever fuel is in the vessel
" Track time to boil when core is completely offloaded.
F. Shutdown condition (i.e., mid-loop, loops not full, loops full, etc.). 5.5.5 Emergent high-risk activities should not start until a risk assessment takes place. However, any activities that require immediate attention may commence prior to the completion of the assessment. 5.5.6 If the unplanned need to establish containment integrity is identified during a forced or refueling outage, the SOM may consider the following:
- Actions needed to assist Operations in implementing required operating procedures and abnormal operating procedures, e.g., S l/S2.OP-AB.CONT-0001(Q)
" Closing larger containment penetrations first
" Utilizing the outage schedule and SAP to identify containment breeches
" Performing walk downs to determine actual penetration status
" Utilizing containment coordinators or shop personnel to obtain work status and identify closure options.
Page 7 of 46 Rev. I
USER RESPONSIBLE FOR VERIFYING REVISION, STATUS AND CHANGES PRINTED 20070305 SC.OM-AP.ZZ-0001(Q) 5.5.7 Manual risk assessment when required is to be conducted as follows: A. Obtain a copy of the outage schedule B. Review what the change is; system and/or equipment, start/stop time, duration C. Determine if the schedule meets the requirements for shutdown safety D. Determine if the system/equipment affects the risk assessment. Compare the Plant Configuration Definitions with the system/equipment being changed. E. If the selected item affects risk assessment, review the SFAT to determine the risk. F. Trace the logic tree to determine if a color (risk) change has occurred for each Key Safety Function. G. Notify the SM/CRS and the SOM if any changes affect the risk assessment. 5.5.8 During outages there should be a heightened awareness on cross-unit equipment that supports the DEFENSE-IN-DEPTH of the outage unit. See Attachment 10 for a list of cross-unit equipment. 5.6 Forced Outa2es 5.6.1 For planned and unplanned forced outages, the initial risk assessment shall be performed as soon as the outage schedule is available. 5.6.2 The risk assessment for a forced or unplanned outage must consider the impacts of on-line work in progress in addition to the work that will be performed during the outage. 5.6.3 While in the transition mode, the risk assessment should be performed in accordance with Attachment 9 once per shift or whenever a configuration change occurs. 5.6.4 The Forced Outage Risk Assessment Report may be submitted to SORC as determined by the Forced Outage Manager, based upon outage complexity and length. 5.6.5 With the plant in the transition mode the Forced Outage Manager may convene a SSRB to address specific issues, configurations or evolutions. 5.7 Salem Intearated Decay Heat Manaeement 5.7.1 Prior to each refueling a calculation shall be performed to ensure that the decay heat load expected and river water temperature are adequate to meet the required heat removal capability for core offload after the reactor has been subcritical for 100 hours. Completion of this calculation should be documented on Attachment 5 and provided to the CRS as documentation that the Tech Spec Hold time is valid for the refueling. [Ref. LCR S02-03] 5.7.2 Decay heat load management should be accomplished by reactor vessel and spent fuel pool decay heat load and heat-up calculations which are essential input to effectively assessing outage risk during schedule development. This information is required to establish appropriate controls over spent fuel pool heat exchanger operation in the cross connected mode of operation to assure that both Units' Spent Fuel Storage Pools are adequately cooled. Calculations also provide the means to identify the proper time frames for taking major systems out of service for maintenance (i.e., CCW, SW, Electric Power, Other Unit's FPC Heat Exchanger, CCW, SW systems). Page 8 of 46 Rev. 1
USER RESPONSIBLE FOR VERIFYING REVISION, STATUS AND CHANGES PRINTED 20070305 SC.OM-AP.ZZ-0001(Q) 5.7.3 Attachment 6 provides the necessary timeline for developing decay heat loads and heat-up curves/tables to allow the stations to assess the time available to implement SFP heat exchanger cross connect operations, time available for other mitigating actions. Finalized decay heat load and heat-up curves shall be provided to the on-duty Ops Shift Manager prior to the start of refueling outages by the Outage Superintendent. 6.0 RECORDS 6.1 The calculation and supporting documentation for Attachment 5 are to be kept as life of plant records IAW the records management process. 6,2 The Risk Assessment Report is not a quality record and need only be retained as long as necessary to support Post-Outage Critiques. 7.0 DEFINITIONS 7.1 Available - A system, structure, or component along with its necessary auxiliary systems, controls, instrumentation, and power supplies is capable of performing its intended function and can be placed in service by manual or automatic means. Recognizing that in this condition all applicable technical specification requirements or licensing/design basis assumptions may not be maintained. This does not infer the system or component is OPERABLE in accordance with Technical Specifications, or available as defined in SH.SE-DG.ZZ-0017(Z), Unavailability Log Keeping Guidelines. [80027168-0010] NOTE An ECCS injection source may be considered available irrespective of the Reactor Cavity being flooded, when its breaker is racked down and Cleared & Tagged (C/T) for the SM/CRS 7.2 Key Safety Functions - For shutdown and refuel conditions, these are functions that provide Shutdown Cooling (Decay Heat Removal), Fuel Pool Cooling, Vessel and Spent Fuel Pool inventory control, electrical power availability, reactivity control, vital support systems: Containment, Control Air, Service Water, and CREACS. 7.3 Reduced Inventory is the condition of the reactor coolant system when fuel is in the reactor vessel and the RCS level has been drained to less than or equal to 101 ft elevation. 7.4 Risk Management - Process of assessing and reducing the likelihood and/or consequence of an adverse event. 7.5 Protected Equipment - Minimum amount of equipment required to maintain planned DEFENSE-IN-DEPTH. No work should be allowed on or around this equipment, without, operations approval. Operating personnel would install/remove barricades, flags, etc. to protected equipment or areas as specified in the contingency plans and operating procedures. 7.6 Transitional Modes - Modes 3, and 4. 7.7 ORAM Translation Matrix - The software table that relates the schedule to the ORAM software. Page 9 of 46 Rev. 1
USER RESPONSIBLE FOR VERIFYING REVISION, STATUS AND CHANGES PRINTED 20070305 SC.OM-AP.ZZ-0001(Q)
8.0 REFERENCES
+ INPO 97-005, Guidelines for the Management of Planned Outages at Nuclear Power Stations
+ INPO AP-925, Outage Process Description
+ NUMARC 91-006, Guidelines for Industry Actions to Assess Shutdown Management, December 1991
+ NUREG-1449, Shutdown and Low-Power Operations at Commercial Nuclear Power Plants in the United States
+ NUREG-1410, Loss of Vital AC Power and the Residual Heat Removal System during Mid-loop Operations at Vogtle Unit # 1
+ NSAC-173, Survey of BWR Plant Personnel on Shutdown Safety Practices and Risk Management Needs
+ SOER 98-1, Safety System Status Control
+ Salem Technical Specifications
+ NRC Bulletin 93-03, Resolution of Issues Related to Reactor Vessel Water Level Instrumentation in BWRs
+ Salem LCR S02-03
+ 1 OCFR50.65(a)(4) 8.1 Cross-References
- NC.LR-DG.ZZ-0007, Desk Top Guide for ORAM
- OU-AA- 103, Shutdown Safety Management Program 8.2 Commitments
- None Page 10 of 46 Rev. I
USER RESPONSIBLE FOR VERIFYING REVISION, STATUS AND CHANGES PRINTED 20070305 SC.OM-AP.ZZ-.0001(Q) ATTACHMENT 1 SALEM RISK ASSESSMENT REVIEW GUIDELINES Page 1 of 4 1.0 The purpose of the initial risk assessment review by the SSRB is to perform an independent assessment of the outage schedule and ORAM contingency plans with the intent of ensuring the proper scheduling and availability of the safety systems needed at various stages of the outage. This review should ensure that the schedule:
" Clearly identifies High Risk and required ORAM Contingency Plans
" Identifies Infrequently Performed Tests or Evolutions and the required additional management oversight and controls
- Identifies the systems, structures and components needed to provide DEFENSE-IN-DEPTH for Key Safety Functions for the different plant conditions that will be experienced during the outage
" Sequence outage activity such as integrated testing to ensure continued operability of required systems per the Mode.
1.1 Review the outage schedule and associated logic ties for activities or combinations of activities that could possibly present a challenge to the shutdown safety functions. 1.2 Review industry experience relative to shutdown events. NOTE Shutdown Safety Considerations are provided as a reference for the SSRB Review (Attachment 2). 1.3 As a minimum, the following shall be reviewed when evaluating risk level in the outage schedule. 1.3.1 Reactivity Control
" Number of charging pumps available
" RWST/BAST availability, including the ability to cross-tie BASTs between units
" Number of source range instruments available
" Periods with fuel movement
" Fuel Status (prior to refueling, defueled, post refueling).
Page 11 of 46 Rev. I
USER RESPONSIBLE FOR VERIFYING REVISION, STATUS AND CHANGES PRINTED 20070305 SC.OM-AP.ZZ-0001(Q) ATTACHMENT 1 SALEM RISK ASSESSMENT REVIEW GUIDELINES Page 2 of 4 1.3.2 Shutdown Cooling
- RCS condition (i.e. loops full, loops not full, mid-loop, cavity full, etc.)
" Number of steam generator loops available with associated Aux Feedwater pumps (applies in Transition Modes ONLY)
" Number of RHR loops available
- Number of Service Water headers available
" Number of ECCS injection pumps available
" Equipment required available to support decay heat removal, including backup and alternate methods.
1.3.3 Spent Fuel Pool heat removal capability
" Number of SFP pumps w/heat exchanger(s) available to maintain 1491F pool temperature under normal conditions, and 180OF under abnormal conditions
" Component cooling water availability
" FHB Ventilation exhaust fan availability
" Activities during high decay heat periods
" Availability of SFP make up systems
" Opposite unit SF cooling available for cross-tie.
1.3.4 Inventory Control
- Periods when RCS inventory is less than 10% in the pressurizer but greater than 101'.
" Periods at Reduced Inventory (less than 101').
" Periods at mid-loop (less than 99').
" Number of level instruments available.
" Availability of ECCS Inventory Makeup systems.
" Fuel Status (prior to refueling, defueled, post refueling).
Page 12 of 46 Rev. 1
USER RESPONSIBLE FOR VERIFYING REVISION, STATUS AND CHANGES PRINTED 20070305 SC.OM-AP.ZZ-0001(Q) ATTACHMENT 1 SALEM RISK ASSESSMENT REVIEW GUIDELINES Page 3 of 4 1.3.5 Electrical Availability
" Availability of Emergency Diesel Generators.
" Number of available offsite power sources.
" Maintenance work in the switchyard.
" Vital AC Bus status.
" DC battery/status.
- Fuel Oil transfer pumps.
" Gas turbine.
" Availability of power sources required to support Key Safety functions
" Likelihood of maintaining offsite sources, i.e., weather conditions, ice, snow, temperature, etc. to the extent practical (Daily review, not pre-outage).
1.3.6 Control Air System Availability
" Number of station air compressors.
" Number of Emergency Air Compressors.
" Control Air headers.
" Station blackout compressor available.
" Temporary Air Compressors 1.3.7 Containment Requirements for
" Periods of containment status where integrity or modified integrity is required.
" Periods when Equipment Hatch/Outage Equipment Hatch is required.
" Containment closure be achieved prior to fuel damage.
" Core alteration requirements.
" CFCUs available Page 13 of 46 Rev. I
USER RESPONSIBLE FOR VERIFYING REVISION, STATUS AND CHANGES PRINTED 20070305 SC.OM-AP.ZZ-0001(Q) ATTACHMENT 1 SALEM RISK ASSESSMENT REVIEW GUIDELINES Page 4 of 4 1.3.8 Control Room Emergency Requirements
" Number of CREACS available.
" Number of CAACS available.
" Number of chillers available,
" Number of chilled water pumps available.
" CAV alignment and the opposite units ability to support maintenance mode.
" Power availability to CAV components (460VAC, 230VAC and 125VDC).
1.3.9 Switchgear Penetration Area Ventilation (SPAV)
" Switchgear return and exhaust fans.
" Electrical penetration exhaust fans.
" Switchgear supply fans.
" Emergency diesel generator backup for SPAV fans.
1.3.10 Service Water
- Number of Service Water headers available.
- Number of Service Water pumps available fed from different power supplies.
e The decay heat load of the reactor. 1.3.11 Potential Fire Hazards
" Review of all scheduled outage hot work
- Review focused on hot work in areas that contain equipment required for decay heat removal.
Page 14 of 46 Rev. 1
USER RESPONSIBLE FOR VERIFYING REVISION, STATUS AND CHANGES PRINTED 20070305 SC.OM-AP.ZZ-0001(Q) ATTACHMENT 2 SALEM SHUT DOWN SAFETY CONSIDERATIONS Page 1 of 9 1.0 STATION PHILOSOPHIES 1.1 Electrical Power Vital AC and DC power to KEY SAFETY FUNCTIONS is required during shutdown conditions to maintain cooling to the reactor core and spent fuel pool, to transfer heat to the ultimate heat sink, to restore containment integrity, if required, and to support other important safety functions. The outage schedule should be reviewed to ensure that periods of plant vulnerability do not coincide with periods where significant sources of Vital power are not available.
" All sources of Offsite power should remain AVAILABLE if possible during HIGHER RISK EVOLUTIONS. If a source of Offsite power is made unavailable (due to maintenance, for example) the remaining available sources of Offsite power should be protected during HIGHER RISK EVOLUTIONS.
In MODE 5 or 6, if only one Offsite AC power source is OPERABLE during HIGHER RISK EVOLUTIONS, three diesel generators should be maintained AVAILABLE with two maintained OPERABLE as defined in Technical Specifications. Exceptions to this shall be approved by the (Station) Operations Manager and OSM.
" Electrical power availability during non-higher risk evolutions should be consistent with Technical Specifications and at no time should a planned removal of all Offsite power sources be scheduled.
" Plant personnel should be kept aware of the status of safety-related electrical systems and unusual configurations/electrical lineups (i.e., buss cross-ties, emergency breakers closed) created due to outage work requirements.
" Control should be maintained over switchyard work to minimize the possibility of personnel error causing a loss of power. Particular care should be taken to ensure that work that could potentially affect the availability of Offsite Power is reviewed for its effect on the DEFENSE-IN-DEPTH of the electrical power system during critical plant evolutions or during periods of high decay heat load.
" There should be no delays in returning critical electrical equipment to service.
1.2 Diesel Generators Equipment relied upon for DEFENSE-IN-DEPTH should be evaluated for Emergency Power Supply Requirements. Attention should be given to structures, systems and components requiring Diesel Generator backup for OPERABILITY. (Examples include Charging Pumps, Fuel Handling Building Exhaust Fans and CREACS fans, and Source Range Nuclear Instruments.) Page 15 of 46 Rev. 1
USER RESPONSIBLE FOR VERIFYING REVISION, STATUS AND CHANGES PRINTED 20070305 SC.OM-AP.ZZ-0001(Q) ATTACHMENT 2 SALEM SHUT DOWN SAFETY CONSIDERATIONS Page 2 of 9 1.3 Shutdown Cooling 1.3.1 Maintaining core/reactor cavity flooding capability and providing an alternate means of decay heat removal is a key safety function during shutdown conditions. 1.3.2 Emergency Core Cooling Systems (ECCS) and decay heat removal systems AVAILABILITY should be maximized during periods of high decay heat or minimum coolant inventory. Equipment outages which impact shutdown cooling systems should be scheduled during periods of low decay heat and/or maximum coolant inventory (i.e., while the Reactor Cavity is flooded) or while the core is off-loaded, whenever practical). 1.3.3 The suitability of a system as a decay heat removal system does not directly relate to its Tech Spec operability status. Since decay heat generation will vary with core power history, the decay heat removal capability required will also vary. During defueled conditions the spent fuel cooling safety function is utilized. A system is considered to be AVAILABLE if it can be used to maintain core temperature below the temperature limit imposed by Operating Procedures. DEFENSE-IN-DEPTH is met for the Shutdown Cooling Safety Function when a primary and a backup decay heat removal system is AVAILABLE, each being capable of removing decay heat and the ECCS Tech Specs are met. The systems available to meet this object are:
" RHR Loop - 11 (21) RHR Pump and RHR Heat Exchanger, 11 (21) Service Water Header, 11 (21) Component Cooling Header and support components (associated control air, electrical power supplies, etc.).
" RHR Loop - 12 (22) RHR Pump and RHR Heat Exchanger, 12 (22) Service Water Headers, 12 (22) Component Cooling Header and support components (associated control air electrical power supplies, etc.).
The ability to cross-tie these trains exists and is effective in providing for adequate shutdown cooling. Page 16 of 46 Rev. 1
USER RESPONSIBLE FOR VERIFYING REVISION, STATUS AND CHANGES PRINTED 20070305 SC.OM-AP.ZZ-0001(Q) ATTACHMENT 2 SALEM SHUT DOWN SAFETY CONSIDERATIONS Page 3 of 9 1.3.4 The following Core Cooling measures are provided by Operating procedures S 1(2).OP-AB.RHR-0001 (Q), Loss of RHR; or S 1(2).OP-AB.RHR-0002(Q), Loss of RHR at Reduced Inventory:
" Hot Leg Injection using the designated AVAILABLE Safety Injection Pump.
" Cold Leg Injection using the designated AVAILABLE Charging Pump.
- Spent Fuel Pool Cooling (only when Reactor Head is removed).
- Steam Generator Reflux Cooling with Gravity Feed from RWST (applies when RCS is depressurized and Reactor Coolant Pumps are not available).
" Cold Leg Recirculation.
1.4 Inventory Control 1.4.1 Control of reactor coolant system inventory is essential in maintaining the overall decay heat removal function. During reduced inventory operations, boiling and potential fuel damage can occur in a relatively short time period if decay heat removal is not restored. 1.4.2 Regardless of the amount of ECCS Equipment available to support this safety function, the ORAM risk color shall be determined as no better than ORANGE'with the RCS at reduced-inventory conditions. This is due to the reduced time to core boiling with reduced mass in the RCS. 1.4.3 The reactor coolant boundary expands during shutdown periods to include the RHR piping, Spent Fuel Pool, Refueling Cavity and other connected support systems. 1.4.4. Special plant configurations during outages increase the possibility of a valve misalignment or other plant problem, which could cause a loss of reactor vessel or Refueling Cavity inventory. Plant configurations where a single active failure or personnel error can result in a rapid loss of reactor water should be identified and minimized to the greatest extent possible. 1.4.5 Reduced Inventory is the condition of the reactor coolant system when fuel is in the reactor vessel and the RCS level has been drained to less than or equal to 101 ft elevation. 1.4.6 No changes should be made in the refueling cavity or reactor vessel level without adequate level instrumentation. During any reactor vessel level changes, proper instrumentation response should be verified. Where possible, redundant level instrumentation shall be used. Page 17 of 46 Rev. 1
USER RESPONSIBLE FOR VERIFYING REVISION, STATUS AND CHANGES PRINTED 20070305 SC.OM-AP.ZZ-0001(Q) ATTACHMENT 2 SALEM SHUT DOWN SAFETY CONSIDERATIONS Page 4 of 9 1.4.7 The following additional measures should be considered to maintain RCS inventory control capabilities:
- Hot side Steam Generator manways should be removed first and installed last.
" Cold side Steam Generator nozzle dams should be installed first and removed last.
1.4.8 Containment closure should be maintained or monitored at all times during reduced inventory conditions with fuel in the vessel, or during fuel movement activities. Schedules should ensure that when the RCS level is less than 104 feet but greater than 101 feet in the RCS, that either the Equipment Hatch remains INSTALLED or the Outage Equipment Hatch is instafled and the ability to close it in a timely manner exists. 1.4.9 When the RCS is not intact, both containment sump pumps and a flowpath to the in-service waste holdup tank should be maintained AVAILABLE to the maximum extent possible. Local Leak Rate Testing of the sump lines should be performed as efficiently as possible and the system returned to service expeditiously. 1.4.10 A source of makeup water adequate to provide makeup to the Spent Fuel Pool with a peak heat load from the decay heat of the full core at the end of a fuel cycle, plus the remaining decay heat of the spent fuel, should be AVAILABLE. 1.4.11 Systems AVAILABLE FOR Spent Fuel Pool makeup are:
" Refueling Water Storage Tank (RWST)
" Primary Water Storage Tank (PWST)
" Demineralized Water System
" CVCS Holdup Tanks
" Emergency fill from RWST (Portable Pump)
" Emergency Fill from PWST (Portable Pump)
Page 18 of 46 Rev. 1
USER RESPONSIBLE FOR VERIFYING REVISION, STATUS AND CHANGES PRINTED 20070305 SC.OM-AP.ZZ-0001(Q) ATTACHMENT 2 SALEM SHUT DOWN SAFETY CONSIDERATIONS (Page 5 of 9) 1.4.12 DEFENSE-IN-DEPTH is met for the Inventory Control Safety Function when a primary and a backup core flooding system is AVAILABLE and the Technical Specifications for the ECCS are met. At least one of the systems must be capable of being powered from its emergency power source. A system is considered to be AVAILABLE when it is capable of providing makeup flow commensurate with the present plant conditions and activities. When plant conditions are being controlled such that a potential to drain the vessel does not exist, the required makeup capability can be significantly reduced. The primary or the backup system must have an AVAILABLE source of emergency power. 1.5 Spent Fuel Pool Cooling 1.5.1 When the core has been off-loaded to the Spent Fuel Pool, the guidelines that apply to core cooling also apply to the Spent Fuel Pool. 1.5.2 A primary and a backup means of cooling the Spent Fuel Pool should be AVAILABLE. At least one of these systems, and its required support systems, must be powered from an emergency power supply. Each system must be capable of maintaining Spent Fuel Pool temperature at the Design/Licensing basis limits under the worst anticipated heat load. Primary means must be capable of maintaining the Spent Fuel Pool temperature at 1497F or less under the worst anticipated heat load with two SFPC pumps and two SFPC heat exchangers in parallel operation. The duration of parallel/single Spent Fuel Pool Cooling Heat Exchanger operation is limited to maintain Spent Fuel Pool temperature limits at both units. Page 19 of 46 Rev. I
USER RESPONSIBLE FOR VERIFYING REVISION, STATUS AND CHANGES PRINTED 20070305 SC.OM-AP.ZZ-0001(Q) ATTACHMENT 2 SALEM SHUT DOWN SAFETY CONSIDERATIONS (Page 6 of 9) 1.5.4 Systems available for Spent Fuel Pool Cooling are:
- 11 (21) Spent Fuel Pool Pump
- 12 (22) Spent Fuel Pool Pump
- 11 (21) Fuel Handling Building Exhaust Fan
- 12 (22) Fuel Handling Building Exhaust Fan
- Spent Fuel Pool cross-connect from opposite unit
- Spent Fuel Pool Pit Heat Exchangers and associated pumps with component cooling and service water available for Heat Sink 1.5.5 Caution must be exercised when utilizing alternate feeds to power the Spent Fuel Pool Pumps and fuel Handling Building Ventilation Fans to ensure that electrical separation criteria are met.
1.5.6 DEFENSE-IN-DEPTH is met for Spent Fuel Pool Cooling when a Primary and a Backup means of Spent Fuel Pool Cooling is AVAILABLE. Each system must be capable of maintaining the spent Fuel Pool temperature at the Design/Licensing basis limits under the worst anticipated heat load. Backup means must be capable of maintaining the Spent Fuel Pool temperature at 180*F or less under worst anticipated heat load. 1.6 Reactivity Control 1.6.1 An important element of shutdown safety is maintaining reactivity control. Boron dilution events during shutdown conditions have resulted in reductions in reactivity shutdown margin. Uncontrolled or inadvertent criticalities that occur while the plant is shut down could lead to unplanned radiation exposure of plant personnel and possible fuel damage. 1.6.2 Sufficient instrumentation should be available to monitor critical parameters. Page 20 of 46 Rev. 1
USER RESPONSIBLE FOR VERIFYING REVISION, STATUS AND CHANGES PRINTED 20070305 SC.OM-AP.ZZ-0001(Q) ATTACHMENT 2 SALEM SHUT DOWN SAFETY CONSIDERATIONS (Page 7 of 9) 1.7 Containment Control 1.7.1 Containment closure is a preliminary action that immediately and effectively reduces the likelihood of a release while providing flexibility to have the containment building open under appropriate conditions. The containment building provides the last integral barrier to the release of radioactive material to the general public. 1.7.2 The ability to establish containment integrity or closure should be maintained during HIGHER RISK EVOLUTIONS. 1.7.3 Containment closure should be maintained at all times during reduced inventory conditions with fuel in the vessel, or during fuel movement activities. Schedules should ensure that either the Equipment Hatch remains INSTALLED or the Outage Equipment Hatch is installed and the ability to close it in a timely manner exists. 1.7.4 Activities planned and scheduled during periods requiring containment closure (i.e., during fuel movement) should be carefully reviewed to ensure that neither the work activity nor its tagout will cause a breach of containment integrity:
" Example: Opening a high point vent (inside) and a low point drain (outside around a containment isolation valve).
" Prior to opening systems inside containment, ensure that containment breaches are considered that are not readily apparent (e.g., steam generator secondary side manway removed with an associated main steam safety valve removed).
" With fuel in the reactor vessel, any penetration with maintenance or testing in progress should have the redundant isolation valve closed; or for penetrations without redundant isolation valves (including electrical penetrations), appropriate plugs or sealing material should be installed, unless specifically controlled by approved procedure(s).
" Expedited containment closure capability, including staging of required tools, should be maintained when there is fuel in the reactor vessel. This should include contingencies for the loss of AC power.
Page 21 of 46 Rev. I
USER RESPONSIBLE FOR VERIFYING REVISION, STATUS AND CHANGES PRINTED 20070305 SC.OM-AP.ZZ-0001(Q) ATTACHMENT 2 SALEM SHUT DOWN SAFETY CONSIDERATIONS (Page 8 of 9) 1.8 Service Water 1.8.1 The Service Water System serves as the ultimate heat sink for various key safety-related heat loads which are vital for safe operation of the plant. The Service Water System must operate during all phases of plant operations including startup normal operation, shutdown, safety injection and recirculation phases. 1.8.2 Two independent Service Water loops are required OPERABLE in MODE 4. 1.8.3 Removal from service and return to service of a Nuclear Header should be accomplished in accordance with approved Operations procedures to preclude the loss of any Key Safety Functions supplied by service water. 1.8.4 With one Nuclear Header out of service, work activities which have the potential to affect the operable Service water Header should be deferred. 1.8.5 Ensure that work activities on Service Water piping do not create an unmonitored flooding path to an operable Service Water Bay. 1.8.6 Systems/equipment that provide support functions for other activities should be maintained AVAILABLE with their associated electrical/water/air sources available commensurate with the activity being supported. Among those providing support functions are:
" Auxiliary Demineralizer Transfer Pump available
- Primary Water Storage Tank
- Demineralized Water Pumps
" Fresh Water Storage Tank & Pumps Page 22 of 46 Rev. 1
USER RESPONSIBLE FOR VERIFYING REVISION, STATUS AND CHANGES PRINTED 20070305 SC.OM-AP.ZZ-0001(Q) ATTACHMENT 2 SALEM SHUT DOWN SAFETY CONSIDERATIONS (Page 9 of 9) 1.9 Control Air 1.9.1 In order to ensure an adequate air supply to vital safety-related equipment and to support plant outage activity needs, only one of the below listed compressors should be out of service at a time:
- Station Air Compressors
" Emergency Control Air Compressors
- Station Blackout Compressor 1.9.2 Air Headers associated with the above should be maintained commensurate with plant conditions and needs.
1.10 Fire Protection 1.10.1 All scheduled outage hot work should be reviewed to ensure that Fire Protection measures are adequate. 1.10.2 Special emphasis should be given to hot work in areas that contain equipment required for maintenance of Key Safety Functions (i.e., decay heat removal). 1.10.3 At least one Diesel Driven Fire Pump should be maintained AVAILABLE. 1.10.4 Fire Pump cross-connect should be maintained AVAILABLE. Page 23 of 46 Rev. 1
USER RESPONSIBLE FOR VERIFYING REVISION, STATUS AND CHANGES PRINTED 20070305 SC.OM-AP.ZZ-0001(Q) ATTACHMENT 3 ORAM CONTINGENCY PLANNING (Page I of 2) These guidelines are to help ensure appropriate considerations are used when developing Contingency Plans. It is not intended to be all inclusive. 1.0 Additional monitoring required to minimize the potential for unplanned equipment unavailability should be considered.
" Additional operator rounds or need to perform walk downs once per shift
" Additional log taking
" Dedicated operator
" Other additional measures 2.0 Protection of monitoring equipment/minimum essential equipment; location and placement should be specified.
" Use of barricades (include Protected Equipment/Protected Areas)
" Caution flags/roped off (include Protected Equipment/Protected Areas)
" Other controls/measures (include Protected Equipment/Protected Areas) 3.0 Alternate Equipment/power supplies AVAILABLE?
4.0 Temporary equipment/power supplies AVAILABLE? 5.0 Special procedure required? Should Infrequently Performed Tests or Evolution's briefings be applied? Are 50.59 evaluations required? 6.0 Reference mitigating procedures (existing procedures with compensatory action) 7.0 Any additional limits needed? ... Pressure, temperature, etc. 8.0 Actions to minimize time in the condition requiring the ORAM contingency plan. 9.0 Applicability of abort criteria 10.0 Required actions to restore KEY SAFETY FUNCTIONS. Walk through required?
" Additional training required?
" Personnel on station?
Page 24 of 46 Rev. 1
USER RESPONSIBLE FOR VERIFYING REVISION, STATUS AND CHANGES PRINTED 20070305 SC.OM-AP.ZZ-0001(Q) ATTACHMENT 3 ORAM CONTINGENCY PLANNING (Page 2 of 2) 11.0 Identification of who is required to take action
" Qualification requirements of personnel
" Familiarize personnel with required actions
" Notification of personnel supervision
- Personnel should be adequately trained and prepared to take actions 12.0 If personnel off site are required to implement the plan, this should be documented in the plan.
13.0 Determine what equipment is necessary (if any) to complete compensatory actions
" Have equipment staged
" Ensure equipment is clearly marked to prevent removal
" Equipment is tested and ready to be used 14.0 Any restrictions needed on plant conditions or other activities? Are other activities on-going or planned which could further degrade the plant's DEFENSE-IN-DEPTH?
15.0 Determine briefing requirements factoring in lessons learned and industry experience. 16.0 Compare with other Contingency Plans for conflicts/consistency. 17.0 Consider scheduled work planned or in progress which could possibly affect key safety system power supplies during a HIGH RISK EVOLUTION. 18.0 If freeze seals are used, consider providing protection for surrounding equipment in the event the seal fails. Immediate actions to correct the failed seal should be identified. 19.0 If the Contingency Plan cannot provide comparable equipment /measures to provide the original DEFENSE-IN-DEPTH, then the evolution should be identified as HIGH RISK. 20.0 Consider time to saturation or temperature limits under the following conditions:
" Spent Fuel Pool Cooling lost
" Total loss of RHR (Shutdown Cooling) 21.0 Identify when protective measures specified in 2.0 may be discontinued/removed.
22.0 Use risk informed tools, when possible, for help in deciding on ORAM contingency plans. Page 25 of 46 Rev. 1
USER RESPONSIBLE FOR VERIFYING REVISION, STATUS AND CHANGES PRINTED 20070305 SC.OM-AP.ZZ-0001(Q) ATTACHMENT 4 ORAM CONTINGENCY PLAN CONTENT (Page 1 of 3) Contingency plans are required for ORAM-ORANGE conditions. They may also be utilized during transitional modes when minimum DEFENSE IN DEPTH is identified. Cover Sheet - Cover sheet which includes the title of the ORAM contingency plan, scheduled implementation date and time, scheduled completion date and time, planned duration and the following required signatures: PRE-SORC Required Signatures:
" Initiator
" Shutdown Safety Manager
" A person cognizant of ORAM from the Nuclear Safety and Licensing Group
" Outage Manager or designee
" Operations Manager or designee Entry into the ORAM ORANGE condition required signatures:
" Lead shop owner
" SM/CRS authorization Entry into the ORAM RED condition required signatures:
- Shift Manager Authorization
" Plant Manager Approval
" Site Vice President Notification Scope: A brief description of the scope of work to be undertaken and objective during the period covered by the ORAM contingency plan.
Justification: Justification for entering the condition including a discussion of the overall benefit to plant safety that will be achieved by performing the maintenance. Page 26 of 46 Rev. 1
USER RESPONSIBLE FOR VERIFYING REVISION, STATUS AND CHANGES PRINTED 20070305 SC.OM-AP.ZZ-0001(Q) ATTACHMENT 4 ORAM CONTINGENCY PLAN CONTENT (Page 2 of 3) Prerequisites: A listing of prerequisite activities for entering the ORAM-ORANGE condition:
- Applicable work packages in a "task ready" status.
- Troubleshooting plans complete and approved by Operations.
- Pre-fabrication work complete.
- Pre-installation testing complete.
- Procedures prepared and approved.
- Plant temporary modification documentation prepared and approved.
- Specific Radiation Work Permit prepared when required.
- ALARA planning complete and shielding installed when practicable.
- Scaffolding installed when possible without affecting operability.
- Containments installed when possible without affecting operability.
- Calibration and staging of special tools and test equipment complete.
" Verification and staging of parts and consumable materials complete.
" Personnel briefings on procedures and associated risks complete.
" Special training requirements complete.
" Procedures verified and dry runs by assigned crews complete.
" Testing of alternate (redundant counterpart) systems complete.
High Risk Evolution Work Activities - A brief description of activities to be performed from the time the condition is entered until the activity ceases to be high risk evolution, i.e., ORAM-YELLOW or GREEN:
" Tagout and system draining and venting activities.
" Temporary modification installation and removal.
" Support activities including installation of shielding, scaffolding and containments which could not be installed as prerequisite activities.
" Maintenance and surveillance activities in appropriate work steps.
" Inspections and non-destructive tests.
- Post-maintenance tests including time estimates.
" Activities to restore and return components and systems to service.
" Operations retests other than surveillances.
" Surveillances required to demonstrate operability.
" Removal of containments, shielding and scaffolding.
" Area cleanup and decontamination.
Page 27 of 46 Rev. I
USER RESPONSIBLE FOR VERIFYING REVISION, STATUS AND CHANGES PRINTED 20070305 SC.OM-AP.ZZ-0001(Q) ATTACHMENT 4 ORAM CONTINGENCY PLAN CONTENT (Page 3 of 3) Estimates - Estimated activity durations and the number of personnel by discipline required to perform each activity and approved by the performing department. Fragnet - A resource-loaded, logically sequenced Fragnet of all activities to be performed during the ORAM-ORANGE condition, including restoration activities and appropriate prerequisites, which clearly show the critical activities to restore from an ORAM-ORANGE condition. ORAM Contingency and Compensatory Measures - A description of ORAM contingency and compensatory measures included in the plan activities using guidelines in Attachment 5. Risk Assessment - Risk Assessment methodology used as documented by the Outage Risk Assessment plan. Contacts - Names and telephone numbers of the key personnel to contact as applicable. Briefings and Critiques - Appropriate check sheets and signoff sheets for the plan implementation briefings and post-work critique. Page 28 of 46 Rev. 1
USER RESPONSIBLE FOR VERIFYING REVISION, STATUS AND CHANGES PRINTED 20070305 SC.OM-AP.ZZ-0001(Q) ATTACHMENT 5 SALEM VERIFICATION OF DECAY HEAT REMOVAL FOR CORE OFF-LOAD Page I of 1 Ref Order/Notification: Calculation Number: Attachments: Required maximum river inlet temperature to support core off load at 100 hours after sub-criticality is The minimum time in which core off load could be conducted and adequate decay heat removal would exist in the spent fuel pool is __ hours after sub criticality. Engineer: Date: Engineering Supervisor: Date: Copy to: Control Room Supervisor Outage Control Center - Shift Outage Manager Shutdown Safety Manager Page 29 of 46 Rev. I
USER RESPONSIBLE FOR VERIFYING REVISION, STATUS AND CHANGES PRINTED 20070305 SC.OM-AP.ZZ-0001(Q) ATTACHMENT 6 DECAY HEAT LOAD AND HEAT UP CURVES/TABLES DEVELOPMENT TIME LINE (Page 1 of 3) TIME PRIOR TO RESPONSIBLE THE OUTAGE ACTION GROUP 12 Months Provide initial data for development of decay heat Engineering loads to Nuclear Fuels; bundles unloaded to fuel pool vs. time and schedule. 12 Months Provide probable system lineups for development Engineering of heat up curves to Design Engineering. Loss of all cooling for different volumes associated with refueling configurations. 11.5 Months Provide preliminary fuel burnup data to Design Nuclear Fuels Engineering Analysis for decay heat calculations 10 Months Determine heat up curves/tables to Engineering Design Engineering determine heat removal capability versus time after shutdown at appropriate cooling water system temperatures and transmit to the Outage Manager. 8 Months Verify the adequacy of shutdown cooling and Outage Management alternate shutdown cooling for all periods during the refueling outage by performing a review of the Level 1 schedule, using the data presented in the letter from Engineering above. Page 30 of 46 Rev. 1
USER RESPONSIBLE FOR VERIFYING REVISION, STATUS AND CHANGES PRINTED 20070305 SC.OM-AP.ZZ-ODO1(Q) ATTACHMENT 6 DECAY HEAT LOAD AND HEAT UP CURVES/TABLES DEVELOPMENT TIME LINE (Page 2 of 3) TIME PRIOR TO RESPONSIBLE THE OUTAGE ACTION GROUP 7 Months Perform an independent verification of the Operations adequacy of shutdown cooling and alternate Department shutdown cooling for all periods during the refueling outage by performing a review of the Level I schedule. 6 Months Issue the backbone Level 2 schedule. Outage Group 6 Months Validate data provided at 12 months. Provide Engineering changes to Fuels as necessary. 6 Months Validate and provide data as necessary for Engineering refinement of previously developed heat up curves (loss of all cooling), and development of new heat up curves for ORAM contingency planning purposes to Design Engineering. 5.5 Months Validate previously calculated fuel burnup data and Nuclear Fuels provide to Design Engineering Analysis for decay heat calculations Determine core decay heat load curves for reactor vessel contents during fuel reload versus time after shutdown and provide data to Outage Management Page 31 of 46 Rev. 1
USER RESPONSIBLE FOR VERIFYING REVISION, STATUS AND CHANGES PRINTED 20070305 SC.OM-AP.ZZ-0001(Q) ATTACHMENT 6 DECAY HEAT LOAD AND HEAT UP CURVES/TABLES DEVELOPMENT TIME LINE (Page 3 of 3) TIME PRIOR TO RESPONSIBLE THE OUTAGE ACTION - GROUP 4 Months Provide and validate design calculations and Design Engineering references for refined heat up curves/tables, including new heat up data necessary to support development of ORAM contingency plans. Determine heat removal capability versus time after shutdown at appropriate cooling water system temperatures. Issue letter to Outage Group providing final data for decay heat loads and heat up curves/tables based on coastdown and core exposure. 3 Months Issue the Level 3 schedule. Outage Management 2 Months Validate previously calculated fuel burnup data Nuclear Fuels and provide Design Engineering Analysis for decay heat calculations 1 Month Validate previously determined heat up Design Engineering curves/tables against current data. Provide results and changes to Engineering. Determine heat removal capability versus time after shutdown at appropriate cooling water system temperatures. Perform final validation of data based on coastdown and core exposure. 2 Weeks Provide results of final validation to the Outage Engineering Management. Page 32 of 46 Rev. I
USER RESPONSIBLE FOR VERIFYING REVISION, STATUS AND CHANGES PRINTED 20070305 SC.OM-AP.ZZ-0001(Q) ATTACHMENT 7 SALEM SHUTDOWN SAFETY ASSESSMENT WORKSHEET (MODES 5,6 OR DEFUELED) (Page 1 of 5) 1.0 INSTRUCTIONS 1.1 PERFORM an equipment available walkdown verification in accordance with the Shutdown Safety Assessment Worksheet. (Safety functions not required for the current mode are "N/A") 1.2 EVALUATE each Safety Function in accordance with the current Mode and determine the appropriate ORAM color assignment. 1.3 OBTAIN the SSA/STA's review and signature. 1.4 RECORD the most severe ORAM color condition on Attachment 11. Page 33 of 46 Rev. 1
USER RESPONSIBLE FOR VERIFYING REVISION, STATUS AND CHANGES PRINTED 20070305 SC.OM-AP.ZZ-0001(Q) ATTACHMENT 7 SALEM SHUTDOWN SAFETY ASSESSMENT WORKSHEET (MODES 5,6 OR DEFUELED) (Page 2 of 5) UNIT 1 MODE: DATE: TIME: SAFETY FUNCTION EOUIPMENT AVAILABLE NOMENCLATURE/STATUS AC Electric Power " Off-Site Circuits i] 5021 [-] 5024 [-- 5037 O[GDYoDR Li 13SPT-] 14SPT
" EDG with associated electrical power subsystems I lA FL 1B L IC
" Gas Turbine Li Available " Unavailable
- Fuel Oil Transfer Pumps Li11 Li 12 Containment Integrity Established Li-Yes E No EGDYDODR Closure Established D Yes ]No Open LiYes No Equipment Hatch li on D off Movement In-progress Li Head ' Fuel Li Internals Time to RCS boiling = minutes Time for Containment closure minutes CFCUs available Control Room Emeruencv CREACS Fans (shutdown unit only) Lii11 j12
[:]GDYDoDR CAACS Fans Li II Li 12 L 13 Chilled Water Pumps L]11 j--12 Chillers i 11 i] 12 L 13 Maintenance Mode Li Yes I-INo [F-1 I F]112 Fuel Pool Cooling Spent Fuel Pumps D- GDF- YL-] o [ R Fuel Handling Ventilation E IIExh E-] 12 Exh L-i Supply Cross-Tie Li Available D Unavailable Li In-Service Heat Exchanger (With Heat Sink) Li Available D- Unavailable Inventory Control Charging Pumps ]i11 ll2 F-] GD[D] YDO-] 0
- R Safety Injection Pumps L-11 [_-]12 RVLIS [:JDCh A D1ChB Midloop Indication L 11 F-" 13 Page 34 of 46 Rev. I
USER RESPONSIBLE FOR VERIFYING REVISION, STATUS AND CHANGES PRINTED 20070305 SC.OM-AP.ZZ-0001 (Q) ATTACHMENT 7 SALEM SHUTDOWN SAFETY ASSESSMENT WORKSHEET (MODES 5,6 OR DEFUELED) (Page 3 of 5) UNIT 1 MODE: DATE: TIME: SAFETY FUNCTION EOUIPMENT AVAILABLE
-- l[ .........
NOMENCLATURE/STATUS Reactivity Control 0 Boric Acid Pumps Available II D0 12
* .Boric Acid Tanks Available E 11 [--12
- Charging Pumps Available 13
- RWST E Available E Unavailable 0 Source Range NIS E N31 F-1 N32
- Boric Acid Stg Tk Cross-Tie 7 Available - Unavailable Shutdown Cooline a RHR Loops DF1 D12 EGEYDOE]R E- 13
- SG Loops Available El 11 E 12 14 []-- None
- CCW Pumps F 11 12 E 13 E- Ji
- CET F-1 D12 [7]H4 E K12
- RHR Hx Inlet Temperature [-]l11 F-1 12 0 11 SW Loop E Available E Unavailable
- 12 SW Loop E Available E Unavailable I13 Switche-ear Pen Cooling
- SWGR Supply E 11 r-]12 EGE]YDOER
- SWGR Exh l11 E12
- Elec Pen Exh F-ll E--]12 Service Water
- SW Pumps El]I Ei12 7 13 EGY[1YDo0 R E 14 E-15 M"16
- Nuclear SW Headers [] 11 12 Completed by:
E- The current configuration is consistent with the Risk Assessment Plan. The ORAM contingency plan in effect for any Orange or Red Safety Function is the most recent revision U and the review with the shift crew is complete. Verified by (STA): Verified and Approved by (SM/CRS): Page 35 of 46 Rev. 1
USER RESPONSIBLE FOR VERIFYING REVISION, STATUS AND CHANGES PRINTED 20070305 SC.OM-AP.ZZ-0001(Q) ATTACHMENT 7 SALEM SHUTDOWN SAFETY ASSESSMENT WORKSHEET (MODES 5,6 OR DEFUELED) (Page 4 of 5) UNIT2 MODE: DATE: TIME: SAFETY FUNCTION EQUIPMENT AVAILABLE NOMENCLATURE/STATUS AC Electric Power
- Off-Site Circuits r- 5021 [:] 5024 D 5037 SY - 0 Fo R E-] 23 SPT D24 SPT
- EDG with associated electrical power subsystems -" 2A -- 2B D2C
- Gas Turbine F-1 Available D] Unavailable
- Fuel Oil Transfer Pumps 11 21 D 22 Containment " Integrity Established 11 Yes ] No F IGoY-0F-1DR
- Closure Established D Yes EJ No
" Open DYes ]No
" Equipment Hatch D-]lOn L-iOff
" Movement In-progress SHead E Fuel 0 Internals
" Time to RCS boiling = minutes
- Time for Containment closure = minutes
" CFCUs available Control Room Emereency
- CREACS Fans(shutdown unit only) 11 21 1122 F-- GD-] Y D-1 0 FQ ] R
- CAACS Fans D- 21 D] 22 1 23
- Chilled Water Pumps Dý 21 D' 22
" Chillers 1121 F-122 E 23
" Maintenance Mode D Yes 11 No Fuel Pool Cooling
- Spent Fuel Pumps 1121 1"- 22
-- G ] Y [7 OF-] R
- Fuel Handling Ventilation E] 21Ex D 22Exh F] Supply
" Cross-Tie -- ] Available F-1 Unavailable D-7 In-Service 0 Heat Exchanger (With Heat Sink) 17 Available 11 Unavailable Inventory Control 0 Charging Pumps 1121 D22
[E] GL--] Y [] OF-] R 0 Safety Injection Pumps 1 21 1122 0 RVLIS F-1 ChA D CbB 0 Midloop Indication [ 21 1 23 Page 36 of 46 Rev. 1
USER RESPONSIBLE FOR VERIFYING REVISION, STATUS AND CHANGES PRINTED 20070305 SC.OM-AP.ZZ-0001(Q) ATTACHMENT 7 SALEM SHUTDOWN SAFETY ASSESSMENT WORKSHEET (MODES 5,6 OR DEFUELED) (Page 5 of 5) UNIT2 MODE: DATE: TIME: SAFETY FUNCTION EQUIPMENT AVAILABLE NOMENCLATURE/STATUS Reactivity Control 0 Boric Acid Pumps Available [-21 E- 22 E] G - Y [7 o R S Boric Acid Tanks Available El 21 E 22 0 Charging Pumps Available [-']21 E-22 [- 23 S RWST D Available D Unavailable S Source Range NIS El N31 El N32 S Boric Acid Stg Tk Cross-Tie -] Available - Unavailable Shutdown Cooling F-I G7[-IY7[0]o7-IR
" RHR Loops El-21 [-- 22
- SQ Loops Available E] 21 E-] 22 -123 [E 24 [:: None
- CCW Pumps El21 El 22 23
[1 J1
" CET El D12 [-] H4 F1K12
- RHR Hx Inlet Temperature E-l 21 E-l 22
" 21 SW Loop El Available D Unavailable
" 22 SW Loop F-" Available El Unavailable Switchaear Pen Cooline DGDYLZODR
- SWGR Supply E 21 [11 22 -] 23
" SWGR Exh l],21 [1 22
- Elec Pen Exh El'21 F-]22 Service Water
- SW Pumps El-21 [-] 22 [1-23
[--] DG-- Y--E o -DR El 24 -25 [E 26
" Nuclear SW Headers E-- 21 E-- 22 Completed by:
E-l The current configuration is consistent with the Outage Risk Assessment plan. The ORAM contingency plan in effect for any Orange or Red Safety Function is the most recent revision and the review with the shift crew is complete. Verified by (STA): I Verified and Approved by (SM/CRS): Page 37 of 46 Rev. I
USER RESPONSIBLE FOR VERIFYING REVISION, STATUS AND CHANGES PRINTED 20070305 SC.OM-AP.ZZ-0001(Q) ATTACHMENT 8 SALEM SHUTDOWN RISK STATUS SHEET (Page 1 of 2) This form available on LAN at: M:\Shared\Operations\Forms\SALEMISHUTDOWN_RISKSTATUS.doc Page 38 of 46 Rev. I
USER RESPONSIBLE FOR VERIFYING REVISION, STATUS AND CHANGES PRINTED 20070305 SC.OM-AP.ZZ-000I(Q) ATTACHMENT 8 SALEM SHUTDOWN RISK STATUS SHEET (Page 2 of 2) This form available on LAN at: M:\Shared\Operations\Forms\SALEM2_SHUTDOWN_RISKSTATUS.doc Page 39 of 46 Rev. I
USER RESPONSIBLE FOR VERIFYING REVISION, STATUS AND CHANGES PRINTED 20070305 SC.OM-AP.ZZ-0001(Q) ATTACHMENT 9 TRANSITIONAL MODES GUIDELINES (TMG) - SALEM 1 AND 2 (Page 1 of 6) The main risk assessment tools for Salem are EOOS (Equipment Out of Service), for Modes 1 and 2, and ORAM (Outage Risk Assessment and management), for Modes 5, 6 and 7 (De-fueled). For Transition modes, 3 and 4, the main risk assessment tool would be based on the Salem Transitional Modes Guidelines (TMG), shown below in Table STMG-1. However, based on circumstances the users may deviate from these guidelines if they judge that the recommended method is either too conservative or not conservative enough. Factors, such as Tech Specs limitations, duration of restoration of the affected components or systems, whether the affected components or equipment is modeled in risk assessment codes like EOOS and ORAM, and time away from upper or lower modes could be used to decide if EOOS, ORAM or other assessment means are applicable. The guidelines set below do keep these issues in mind, but it is possible that they may not cover all circumstances. These guidelines are related to single unit outages, and situations where significant equipment is unavailable for a maximum of one week. If there is dual unit shutdown or equipment is expected to be unavailable for more than one week, PSA support may be needed. These guidelines address how to evaluate risk for each of the nine Key Safety Functions (KSF) used in the Salem Unit 1 and 2 ORAM models. These KSFs can be found in the ORAM Model for Salem Unit 1 and 2, S-1-ZZ-RZZ-0033 and S-2-ZZ-RZZ-0034, respectively. The most limiting color for a KSF would be the overall plant risk. Table STMG Salem Transitional Modes Guidelines KSF For Mode 3, use: For Mode 4, use: AC Power (AC) EOOS ORAM, SFAT No. AC-2 Containment (CON) Table STMG-2 Table STMG-3 Control Room Emergency (CRE) EOOS ORAM, SFAT No. CRE-1 Fuel Pool Cooling (FPC) ORAM, SFAT No. FPC-1 ORAM, SFAT No. FPC-1 Inventory Control (IC) EOOS ORAM, SFAT No. IC-3 Reactivity Control (RC) EOOS ORAM, SFAT No. RC-2 Shutdown Cooling (SDC) Table STMG-4 Table STMG-5 Switchgear Penetration Cooling (SPV) EOOS ORAM, SFAT No. SPV-1 Service Water (SW) EOOS Table STMG-6 Page 40 of 46 Rev. I
SC.OM-AP.ZZ-OOO1(Q) ATTACHMENT 9 TRANSITIONAL MODES GUIDELINES (rMG) - SALEM 1 AND 2 (Page 2 of 6) 0F Table STMG Risk Matrix for Salem Units 1 & 2 Containment (CON) KSF - Mode 3 0 Is RWST Available? How many How many RHR How many CFCUs RISK Level Containment Spray pumps are available are available? Pumps are along with available? containment sump and at least one C)ý RHR HX? Yes 2 2 =>3 _GREEN .... 2 YELLOW 1 ORANGE 0 RED 1=>3 YELLOW 2 ORANGE
<2 RED 1 2 =>4 GREEN 3 YELLOW
- 2. ORANGE
<2 RED 1=>4 YELLOW 3 ORANGE
<3 RED 0 =>1 =>4 YELLOW
<4 RED Any 0 Any RED No RED Page 41 of 46 Rev. 1
-U C)ý
;:U SC.OM-AP.ZZ-0001 (Q)
--- C/)
0 ATTACHMENT 9 0 Cf) TRANSITIONAL MODES GUIDELINES (TMG) - SALEM 1 AND 2 (Page 3 of 6) U'l -9 0 Table STMG Risk Matrix for Salem Units 1 & 2 Containment (CON) KSF - Mode 4 Is RWST Available? How many How many RHR How many CFCUs RISK Level Containment Spray pumps are available are available? ;0 Pumps are along with available? containment sump and at least one RHR HX? Yes 2 C!)
=>1 =>3 GREEN 2 YELLOW I ORANGE ý-I 0 RED CA 1 =>1 =>4 GREEN 3 YELLOW 2 ORANGE C/)
<2 RED M
0 =>1 =>4 YELLOW
--<3 RED CA Any 0 Any RED No RED Page 42 of 46 Rev. I
-0 C:
TIuCf) M SC.OM-AP.ZZ-O001 (Q) ATTACHMENT 9 TRANSITIONAL MODES GUIDELINES (TMG) - SALEM 1 AND 2 C-) (Page 4 of 6) 00 Table STMG Risk Matrix for Salem Units 1 & 2 Shutdown Cooling (SDC) KSF - Mode 3 0 - Is TDAFP Available? No. of Motor No. of RHR loops No. of SW headers Risk Level driven Aux FW operable with at available for CnM Pumps available least one backed operating RHR & by an EDG EDGs -q Yes =>1 2 2 GREEN 1 YELLOW
-T1 0 RED-1 2 YELLOW 1 ORANGE 0 RED U)
O 2 2 YELLOW CIO 1 ORANGE 0 RED Cf) 1 =>I ORANGE Z: 0 RED C!) No =>1 2 2 YELLOW 1 ORANGE 0 RED 1 2 ORANGE
=<I RED 0 2 2 ORANGE
=-<I RED
=<I_. Any RED Page 43 of 46 Rev. I
SC.OM-AP.ZZ-0001(Q) ATTACHMENT 9 TRANSITIONAL MODES GUIDELINES (TMG) - SALEM 1 AND 2 0 U) (Page 5 of 6) 0 F Table STMG Risk Matrix for Salem Units 1 & 2 Shutdown Cooling (SDC) KSF - Mode 4 0 No. of Motor driven Aux FW Pumps No. of RHR loops No. of SW headers Risk Level available operable with at available for -T1 least one backed operating RHR & by an EDG EDGs 2 2 2 GREEN I YELLOW 0 RED U) 1 2 YELLOW 1 ORANGE 0 RED 1 2 2 YELLOW I ORANGE 0 RED 1 ->1 ORANGE 0 RED 0 2 2 YELLOW 1 ORANGE 0 RED 1 2 ORANGE
=<1 RED Page 44 of 46 Rev. 1
UE SC.OM-AP.ZZ-OOO1(Q) ATTACHMENT 9 jZ TRANSITIONAL MODES GUIDELINES (TMG) - SALEM 1 AND 2 (Page 6 of 6) f-T Table STMG Risk Matrix for Salem Units I & 2 Service Water (SW) KSF - Mode 4 No. of SW Headers No. of SW Bays No. of SW pumps No. of available SW Risk Level Available available available pumps backed by any operable EDG
=>4 GREEN 0 2 2=>4 3 YELLOW 2 ORANGE
<2 RED 3 3 YELLOW 2 ORANGE
<2 RED z 2 2 ORANGE 1 RED 0 M
1 Any RED 3 3 YELLOW C/) 2 ORANGE
<2 RED 2 2 ORANGE
<2 RED 2=>4 =>4 YELLOW 3 ORANGE
<3 RED 3 3 ORANGE
<3 Any RED 1 3 3 ORANGE
<3 Any RED 0 RED Page 45 of 46 Rev. I
Ucz SC.OM-AP.ZZ-0001 (Q) ATTACHMENT 10 CD 0
'j Z CROSS-UNIT HEIGHTENED AWARENESS EQUIPMENT LIST Unit in Forced or Refueling Outage: 0r IMI-1-51,M-NI-01"M 0 U Salem Unit 3 Unit 2 Service Water Pumps Unit I Service Water Pumps Fire Protection Island Cross-Tie Unit 2 Service Water Headers Unit 1 Service Water Headers M Salem Switchyard Service Water Bays 2 and 4 Service Water Bays 1 and 3 X 0 5037 Line Service Water Test Line Service Water Test Line S Salem 500 kV Breakers 2-10 and #2 ECAC #1 ECAC 9-10 Station Air Compressors Station Air Compressors LI)
Diesel-driven Fire pumps Diesel-driven Fire pumps 0 Unit 2 Chillers Unit I Chillers Unit 2 Chilled Water Pumps Unit I Chilled Water Pumps Unit 2 CREACS Fans Unit I CREACS Fans (-I) Unit 2 CAACS Fans Unit 1 CAACS Fans Unit 2 SFP Pumps and Heat Exchanger Unit 1 SFP Pumps and Heat Exchanger Temporary Air Compressors Temporary Air Compressors Station Blackout Air Compressor Station Blackout Air Compressor Demineralized Water Tanks Demineralized Water Tanks Fresh Water Tanks (I) Fresh Water Tanks 23 Charging Pump 13 Charging Pump Hope Creek-Salem Fire Protection Cross-Tie Hope Creek-Salem Fire Protection Cross-Tie 5037 Line 5037 Line Hope Creek Switchyard Hope Creek Switchyard Hope Creek 500 kV Breakers 2-4, 3-4, and 2-6 Hope Creek 500 kV Breakers 2-4, 3-4, and 2-6 Notes
- 1. This list represents the minimum equipment desired to support cross-unit outages.
- 2. This list assumes SINGLE UNIT OUTAGE ONLY. Multi-unit outages should be evaluated separately and on a case basis.
- 3. Schedulers and Planners should maximize the availability of equipment in each column during forced or planned outages to enhance defense-in-depth management strategies for the outage unit.
- 4. During high risk activities (i.e., Midloop at Salem), serious consideration should be given to protecting equipment.
Page 46 of 46 Rev. I
USER RESPONSIBLE FOR VERIFYING REVISION, STATUS AND CHANGES PRINTED 20070305 PSEG Internal Use Only Page I of I PSEG NUCLEAR L.L.C. SALEM/OPERATIONS S2.OP-IO.ZZ-0007(Q) REV. 12 COLD SHUTDOWN TO REFUELING USE CATEGORY: I
- Biennial Review Performed: Yes _ No -/
- Packages and Affected Document Numbers incorporated into this revision: None
- The following OTSCs were incorporated into this revision: None REVISION
SUMMARY
- The following changes are a result of Technical Specification Amendment 275/257: [800876881 Attachment 2, Step 1.4.1 changed to read "SC.IC-FT.NIS-001 1(Q) and SC.tC-FT.NIS-0012(Q), for N31 & N32 Source Range Channels completed at least once per 7 days (as required by T/S 4.9.2)". Functional Testing of the Source Range Neutron Flux Monitors is no longer required within 8 hours prior to the initial start of CORE ALTERATIONS.
IMPLEMENTATION REQUIREMENTS Effective Date: 110 14 1 t, Technical Specification Amendment 275/257 APPROVED: Operations Director - Salem Date
USER RESPONSIBLE FOR VERIFYING REVISION, STATUS AND CHANGES PRINTED 20070305 S.OP-IO.ZZ-0007(Q) COLD SHUTDOWN TO REFUELING TABLE OF CONTENTS SECTION TITLE PAGE 1.0 PURPOSE ........ ........................................... .... 2 2.0 PR EREQ U ISITES .......................................................... 2 3.0 PRECAUTIONS AND LIMITATIONS ....... ......... . ............... 2 4.0 EQUIPMENT/MATERIAL REQUIRED ........................................ 3 5.0 PR O CED UR E ............................................................. 4 5.1 M ode 5 To M ode 6 Operations ........................................ 4 5.2 Core Alteration Operations (No Fuel Movement) ........................... 7 5.3 Core Alteration Operations (Fuel Movement) .............................. 8 5.4 Com pletion And Review .............................................. 9 6.0 R E C O RD S ............................................................... 10 7.0 REFEREN CES ........................................................... 10 ATTACHMENTS Attachment I Requirements and Reviews For Mode 6 ................................. 13 Attachment 2 Requirements And Review For Core Alterations (No Fuel M ovem ent) ................................................ 15 Attachment 3 Requirements And Review For Core Alterations - Movement Of Irradiated Fuel In The Reactor Pressure Vessel ................ 17 Attachment 4 Completion Sign-Off Sheet ........................................... 21 SajeM2 Page 1 of 23 Rev. 12
USER RESPONSIBLE FOR VERIFYING REVISION, STATUS AND CHANGES PRINTED 20070305 s 2 .oP-Io.zz-ooo0(Q) 1.0 PURPOSE 1.1 To provide the instructions necessary to: 1.1.1. Transition from Cold Shutdown (Mode 5) to Refueling (Mode 6). 1.1.2. Prepare for CORE ALTERATIONS. 1.1.3. Prepare for Movement of Irradiated Fuel in the Reactor Pressure Vessel. 2.0 PREREQUISITES 2.1 ENSURE a RCS vent path is established JAW S2.OP-SO.PZR-0006(Q), RCS Venting. 2.2 ENSURE RCS is degassed lAW S2.OP-SO.CVC-001 1(Q), RCS Degassification, OR Chemistry RCS sample analysis is acceptable to allow opening the RCS. 2.3 ENSURE Refueling Canal Drain Flange is installed. (Refueling Canal Drain Flange Drain Valve 2WL221 is located on this flange) 3.0 PRECAUTIONS AND LIMITATIONS 3.1 Procedure Use and Adherence Policy as found in NC.NA-AP.ZZ-0001(Q), Nuclear Procedure System, is applicable to this procedure. 3.2 A maximum of one Safety Injection Pump OR one Centrifugal Charging Pump shall be OPERABLE IAW T/S Surveillances 4.5.3.2.a or 4.5.3.2.b while in Mode 5 or 6 when the head is on the Reactor Pressure Vessel. 3.3 23 Charging Pump flow path shall be aligned to Unit 1 OR the pump shall be C/T when 21 or 22 Safety Injection Pump is capable of injection into the core with RCS temperature
- 312'F with the Reactor Vessel Head installed. 1C05651 3.4 S2.OP-IO.ZZ-0010(Q), Spent Fuel Pool Manipulations, is to remain active while in Section 5.3 of this procedure.
3.5 Maintaining RCS Activity at
- 0.05 uci/ml prior to filling the Refueling Cavity will result in lowered dose rates, *2.5 mr/hr, at the refueling water surface during refueling evolutions.
3.6 Technical Specification 3.9.3.a is valid through the year 2010. (T/S Amendment 251/232 Safety Evaluation) Salem 2 Page 2 of 23 Rev. 12
USER RESPONSIBLE FOR VERIFYING REVISION, STATUS AND CHANGES PRINTED 20070305 s 2 .OP-1O.ZZ-0007(Q) 3.7 Direct communications shall be maintained between the Control Room and personnel at the refueling station during CORE ALTERATIONS (T/S 3.9.5). [700276101 3.8 CORE ALTERATION shall be the movement of any fuel, sources, or reactivity control components, within the reactor vessel with the vessel head removed and fuel in the vessel. Suspension of CORE ALTERATION shall not preclude completion of movement of a component to a safe conservative location. [700276101 3.9 The boron concentration of the Refueling Canal, Fuel Storage Pool, and the Refueling Cavity shall be maintained within the limits of the COLR when connected to the Reactor Coolant System to comply with T/S 3.9.1 in Mode 6. 3.10 2R1 IA, 2R12A, and 2R12B are NOT required by Technical Specifications in Mode 6. The monitors will be blocked to defeat the Containment Ventilation Isolation control function in Mode 6. The block switches may be placed in NORMAL on an intermittent basis to support testing of the associated monitors. The test switches shall NOT be placed in NORMAL when all of the following conditions exist:
- Movement of irradiated fuel within the containment is in progress, and
- Containment Equipment Hatch is removed, and
+ Containment Purge To Plant Vent is relied upon as the required ventilation flow path in Attachment 3, Step 1.8.1.
4.0 EQUIPMENT/MATERIAL REQUIRED None Salem2 Page 3 of 23 Rev. 12
USER RESPONSIBLE FOR VERIFYING REVISION, STATUS AND CHANGES PRINTED 20070305 s 2 .OP-bo.zZ-0007(Q) 5.0 PROCEDURE 5.1 Mode 5 To Mode 6 Operations 5.1.1. INITIATE the following:
+ S2.OP-IO.ZZ-0107(Q), Cold Shutdown To Refueling Administrative Requirements
+ Attachment 1, Requirements and Reviews for Mode 6.
- S2.OP-IO.ZZ-0010(Q), Spent Fuel Pool Manipulations.
- S2.OP-ST.CAN-0007(Q), Refueling Operations - Containment Closure.
5.1.2. When Attachment 1 is completed with Mode 6 entry authorized: A. ENSURE the RCS is drained to <104 ft elevation JAW S2.OP-SO.RC-0005(Q), Draining the Reactor Coolant System to _>101 Ft Elevation OR S2.OP-SO.RC-0006(Q), Draining the Reactor Coolant System to <101 Ft Elevation With Fuel In The Vessel. B. NOTIFY Outage Management that the requirements to initiate Reactor Pressure Vessel Head detensioning are satisfied. 5.1.3. When the first stud (during first pass of Reactor Head deteasioning process) is detensioned:
+ RECORD time Mode 6 is entered in the Control Room Narrative Log.
- UPDATE WCM to Mode 6.
Salem2 Page 4 of 23 Rev. 12
USER RESPONSIBLE FOR VERIFYING REVISION, STATUS AND CHANGES PRINTED 20070305 s.OP--o.Zz-0007(Q) NOTE 2R1 1A control function is normally blocked in Modes 1-6. 2R12A and 2R12B control functions are normally blocked in Mode 6. 5.1.4. PERFORM the following at 2RP I:
- ENSURE 2R1 1A block switches Train A & B are in BLOCK position
_
- PLACE 2Rt 2A block switches Train A & B in BLOCK position
- PLACE 2R12B block switches Train A & B in BLOCK position
*' Direct a second Operator to PERFORM Independent Verification that 2R11 A, 2R12A, and 2R12B block switches are in the BLOCK position:
IV Signature Date 5.1.5. ENSURE the following valves arc CLOSED:
- 2WL2, FU]EL XFER CANAL DRAIN
- 2WL3, FUEL XFER CANAL DRAIN 5.1.6. CLOSE the following valves:
_ 2RC2, REACTOR HEAD INNER SEAL LEAK-OFF VALVE
- 2RC3, REACTOR HEAD OUTER SEAL LEAK-OFF VALVE 5.1.7. Direct a second Operator to PERFORM Independent Verification that 2RC2 and 2RC3 are CLOSED:
IV Signature Date SaIeM2 Page 5 of 23 Rev. 12
USER RESPONSIBLE FOR VERIFYING REVISION, STATUS AND CHANGES PRINTED 20070305 2 .OP-1o.ZZ-0007(Q) CAUTION The Containment Noble Gas Monitor (2R12A) could momentarily exceed the alarm setpoint during preparations for, and during the Reactor Pressure Vessel Head Lift. This event is considered a pre-planned sequence during reactor disassembly which is NOT reportable lAW the ECG. (PR990411101] 5.1.8. Prior to initiating the Reactor Pressure Vessel Head Lift, NOTIFY Outage Management that the Reactor Pressure Vessel Head Lift is authorized. 5.1.9. When the Reactor Pressure Vessel Head is removed FILL the Refueling Cavity as follows: A. IF surveillances are to be performed at this time, THEN COORDINATE with Outage Scheduling any of the following surveillances that are to be completed during the Refueling Cavity filling: S2.OP-ST.CS-0005(Q), In Service Testing Containment Spray Valves - Mode 6
- S2.OP-ST.RHR-0005(Q), In Service Testing Residual Heat Removal Valves and Orifices
- S2.OP-ST.SJ-0006(Q), In Service Testing Safety Injection Valves - Mode 6 NOTE The boron concentration of the Refueling Canal, Fuel Storage Pool, and the Refueling Cavity shall be maintained within the limits of the COLR when connected to the
.Reactor Coolant System to comply with T/S 3.9.1 in Mode 6.
B. FILL the Refueling Cavity lAW S2.OP-SO.SF-0003(Q), Filling the Refueling Cavity. 5.1.10. INITIATE S2.OP-SO.SF-0009(Q), Refueling Operations. 5.1.11. INITIATE Attachment 2, Requirements and Review for Core Alterations (No Fuel Movement). Salem2 Page 6 of 23 Rev. 12
USER RESPONSIBLE FOR VERIFYING REVISION, STATUS AND CHANGES PRINTED 20070305 s2.OP-IO.ZZ-0007(Q) 5.2 Core Alteration Operations (No Fuel Movement) 5.2.1. When Attachment 2 is completed and CORE ALTERATIONS with no movement of irradiated fuel in the reactor pressure vessel is authorized: A. ENSURE 22CA330 is OPEN AND PLACE Information Only tag on 22CA330 bezel:
"Air supply to the refueling manipulator crane".
B. ENSURE direct communications between the Control Room and personnel at the refueling station are established within 1 hour prior to start of Core Alterations AND verified at least once every 12 hours during Core Alterations using an Additional Reading/Operator Action Log JAW SH.OP-AP.ZZ-01 10(Q), Use and Development of Operating Logs. (T/S 4.9.5) C. NOTIFY personnel in the Control Room AND the refueling station that CRS permission is required to discontinue direct communications between the Control Room and personnel at the refueling station. 1700276101 D. IF any control rod is to be moved within the reactor pressure vessel, THEN ENSURE the last recorded Reactor Pressure Vessel Water Level in S2.OP-DL.ZZ-0002(Q) is within 2 hours of control rod movement (T/S 4.9.10). E. NOTIFY Outage Management that (except for fuel movement) the requirements to commence CORE ALTERATIONS are satisfied. 5.2.2. INITIATE Attachment 3, Requirements and Review for Core Alterations - Movement of Irradiated Fuel in the Reactor Pressure Vessel. SaleM2 Page 7 of 23 Rev. 12
USER RESPONSIBLE FOR VERIFYING REVISION, STATUS AND CHANGES PRINTED 20070305 S2.OP-lO.ZZ-0007(Q) 5.3 Core Alteration Operations (Fuel Movement) 5.3.1. When Attachment 3, Requirements and Review for Core Alterations - Movement of Irradiated Fuel in the Reactor Pressure Vessel is completed and movement of irradiated fuel in the reactor pressure vessel is authorized: A. ENSURE direct communications between the Control Room and personnel at the refueling station are established within I hour prior to start of Core Alterations AND verified at least once every 12 hours during Core Alterations using an Additional Reading/Operator Action Log lAW SH.OP-AP.ZZ-01 10(Q), Use and Development of Operating Logs. (T/S 4.9.5) B. NOTIFY personnel in the Control Room AND the refueling station that CRS permission is required to discontinue direct communications between the Control Room and personnel at the refueling station. 170027610] C. ENSURE Fuel Handling Area administrative and Technical Specifications requirements are satisfied lAW S2.OP-IO.ZZ-0010(Q), Spent Fuel Pool Manipulations. D. ENSURE the last recorded Reactor Pressure Vessel Water Level in S2.OP-DL.ZZ-0002(Q) is within 2 hours of fuel movement (T/S 4.9.10). E. NOTIFY Outage Management that the requirements to commence fuel movement are satisfied. F. When the Reactor Pressure Vessel is defueled:
- REMOVE Information Only tag from 22CA330 bezel.
__ MAINTAIN Unit 2 JAW S2.OP-DL.ZZ-0002(Q), Control Room Log - Mode 5, 6 and Defueled.
- MAINTAIN Salem Integrated Decay Heat Management requirements specified in NC.OM-AP.ZZ-0001 (Q), Outage Risk Assessment.
Salem2 Page 8 of 23 Rev. 12
USER RESPONSIBLE FOR VERIFYING REVISION, STATUS AND CHANGES PRINTED 20070305 s2.oP-Io.ZZ-0007(Q) 5.4 Completion And Review 5.4.1. COMPLETE Attachment 4, Sections 1.0 and 2.0, AND FORWARD this procedure to the SM/CRS for review and approval. 5.4.2. SM/CRS PERFORM the following: A. REVIEW this procedure with Attachments 1-4 for completeness and accuracy. B. COMPLETE Attachment 4, Section 3.0. C. IF this procedure is terminated prior to completion, THEN:
- 1. PROVIDE the reason, date and time of termination on Attachment 4, Section 1.0.
- 2. MAINTAIN with last completed lOP in Control Room file.
D. IF this procedure is completed successfully, THEN: I. ATTACH completed S2.OP-IO.ZZ-0107(Q), Cold Shutdown To Refueling Administrative Requirements to this procedure.
- 2. FORWARD last completed IOP from Control Room file to Operations Staff.
- 3. PLACE this procedure in Control Room File.
END OF PROCEDURE SECTION Sallem2 Page 9 of 23 Rev. 12
USER RESPONSIBLE FOR VERIFYING REVISION, STATUS AND CHANGES PRINTED 20070305 s2.oP-IO.ZZ-0007(Q) 6.0 RECORDS 6.1 Retain the following lAW RM-AA-101, Records Management Program: This Entire Procedure
7.0 REFERENCES
7.1 Updated Final Safety Analysis Report 7.1.1. Section 4, Reactor 7.1.2. Section 5, Reactor Coolant System and Connected Systems 7.1.3. Section 7, Instrumentation and Controls 7.1.4. Section 8, Electrical Systems 7.1.5. Section 9, Auxiliary Systems 7.1.6. Section 15.1, Condition 1, Nonnal Operation and Operational Transients 7.2 Technical Specifications - Unit 2 7.2.1. 3.1.2.3, Charging Pump - Shutdown 7.2.2. 3.3.3.1, Radiation Monitoring Instrumentation 7.2.3. 3.8.1.1, A.C. Sources - Operating 7.2.4. 3.8.1.2, A.C. Sources - Shutdown 7.2.5. 3.8.2.1, A.C. Distribution - Operating 7.2.6. 3.8.2.2, A.C. Distribution - Shutdown 7.2.7. 3.8.2.3, 125-Volt D.C. Distribution - Operating 7.2.8. 3.8.2.4, 125-Volt D.C. Distribution - Shutdown 7.2.9. 3.8.2.5, 28-Volt D.C. Distribution - Operating 7.2.10. 3.8.2.6, 28-Volt D.C. Distribution - Shutdown 7.2.11. 4.9.12, Fuel Handling Area Ventilation System 7.3 Procedures 7.3.1. NC.OM-AP.ZZ-000I (Q), Outage Risk Assessment 7.3.2. SH.OP-AP.ZZ-0108(Q), Removal and Return of Nuclear Safety Equipment 7.3.3. S2.OP-IO.ZZ-0001 (Q), Refueling to Cold Shutdown 7.3.4. S2.OP-IO.ZZ-0107(Q), Cold Shutdown To Refueling Administrative Requirements 7.3.5. S2.OP-SO.SF-0001(Q), Fill and Transfer of the Spent Fuel Pool Salein2 Page 10 of 23 Rev. 12
USER RESPONSIBLE FOR VERIFYING REVISION, STATUS AND CHANGES PRINTED 20070305 s.OP-iO.ZZ-0007(Q) 7.4 Drawings 7.4.1. 205301, No. 2 Unit Reactor Coolant System 7.4.2. 205328, No. 2 Unit Chemical and Volume Control System 7.4.3. 205331, No. 2 Unit Component Cooling System 7.4.4. 205332, No. 2 Unit Residual Heat Removal 7.4.5. 205333, No. 2 Unit Spent Fuel Cooling 7.4.6, 205334, No. 2 Unit Safety Injection System 7.4.7. 205340, No. 2 Unit Waste Disposal - Gas 7.5 Others 7.5.1. DCP 80029150 And 80029155, Unit CVCS Cross-Tie 7.5.2. INPO SOER 88-3, Losses of RHR With Reduced Reactor Vessel Water Level 7.5.3. NRC INFO 87-23, Loss Of Decay Heat Removal Function At PWRs With Partially Drained Reactor Coolant Systems 7.5.4. PIR #990411101, ESF Actuation During Reactor Head Lift (2R12A) 7.5.5. SC-R200-MSE-0738-1, Mid-Loop Operation, 10/10/88 7.5.6. Westinghouse Owners Group Abnormal Response Guideline WOG-ARG-1, Loss of RHR While Operating at Mid-Loop Conditions 7.5.7. 80050653, Technical Specification Amendment 263/245, Refueling Operations, Relaxation of T/S Requirements Applicable During The Movement Of Irradiated Fuel. 7.6 Cross-References 7.6.1. Technical Specifications - Unit 2: A. 3.4.10.3, RCS Overpressure Protection Systems B. 3.9.2, Instrumentation C. 4.9.8.2, Low Water Level (RHR Loops) D. 4.1.2.1.a. I & 2, Boration Systems Flowpath - Shutdown E. 4.1.2.1.b, Boration Systems Flowpath - Shutdown F. 4.1.2.3, Charging Pump - Shutdown G. 4.1.2.5.a & b, Borated Water Sources - Shutdown H. 4.5.3.2.a & b, ECCS Subsystems - TAVG <350'F I. 4.8.1.2, Electrical Power Systems - Shutdown J. 4.8.2.4.1, 125VDC Distribution System - Shutdown K. 4.8.2.6.1, 28VDC Distribution System - Shutdown L. 4.9.1, Boron Concentration M. 4.9.2.a, Instrumentation N. 4.9.3, Decay Time
- 0. 4.9.4.1, Containment Building Penetrations P. 4.9.4.2, Containment Equipment Hatch Q. 4.9.4.3, Containment Purge Isolation R. 4.9.5, Communications S. 4.9.8.1, Refueling Operations - All Water Levels (Coolant Circulation)
T. 4.9.10, Reactor Pressure Vessel Water Level Salem 2 Page 11 of 23 Rev. 12
USER RESPONSIBLE FOR VERIFYING REVISION, STATUS AND CHANGES PRINTED 20070305 s 2 .oP-Io.zz-0007(Q)
7.6.2. Procedures
A. NC.NA-AP.ZZ-0001(Q), Nuclear Procedure System B. RM-AA- 101, Records Management Program C. NC.OM-AP.ZZ-0001(Q), Outage Risk Assessment D. SC.CH-AP.ZZ-1 165(Q), Salem Chemistry Mode Change Requirements E. SC.CH-FR.ZZ-1 160(Q), Refueling and Fuel Movement Analytical Surveillances F. SC.CH-TI.ZZ-01 80(Q), Sampling Schedule and Chemistry Specifications G. S2.OP-DL.ZZ-0001(Q), Control Room Logs H. S2.OP-DL.ZZ-0002(Q), Control Room Log - Mode 5, 6 and Defueled
- 1. S2.OP-IO.ZZ-0010(Q), Spent Fuel Pool Manipulations J. S2.OP-SO.CBV-0002(Q), Containment Pressure-Vacuum Relief K. S2.OP-SO.RHR-0001(Q), Initiating RHR L. S2.OP-SO.SF-0003(Q), Filling the Refueling Cavity M. S2.OP-SO.WG-0006(Q), Containment Purge to Plant Vent 7.7 Commitments 7.7.1. C0565 - NLR N94229, POPS Setpoint Nonconservatism 7.7.2. C0636 - NSO AMEN 131/110, LER 88-06, Letter NLR-N91094 Salem2 Page 12 of 23 Rev. 12
USER RESPONSIBLE FOR VERIFYING REVISION, STATUS AND CHANGES PRINTED 20070305 S.OP-IO.ZZ-0007(Q) ATTACHMENT I (Page 1 of 2) REQUIREMENTS AND REVIEWS FOR MODE 6 1.0 TECHNICAL SPECIFICATIONS REQUIREMENTS 1.1 The Reactor Coolant System boron concentration shall be maintained within the limits specified in the Core Operating Limits Report (COLR) and shall be verified at least once every 72 hours in Mode 6 (T/S 3.9.1). 1.2 Two Source Range neutron flux monitors are OPERABLE each with continuous visual indication in the Control Room and one with audible indication in the Containment and Control Room (T/S 3.9.2), including the following:
+ OHA E-13, SR HI FLUX AT S/D
+ Audio Count Rate Channel
- 1.3 2N31 & 2N32 Channel Checks performed and recorded in S2.OP-DL.ZZ-0002(Q),
Control Room Log - Mode 5, 6 and Defueled, at least once per 12 hours (T/S 4.9.2.a.)
- 1.4 At least one RHR Loop is in service lAW S2.OP-SO.RHR-0001(Q), Initiating RHR, maintaining RCS temperature >50'F and !<140'F with coolant circulation Ž11000 gpm as verified at least once per 12 hours JAW S2.OP-DL.ZZ-0002(Q) (T/S 4.9.8.1).
1.5 Two independent RHR Loops are OPERABLE when RCS Level will be <23 feet above the top of the Reactor Pressure Vessel flange (T/S 3.9.8.2). 1.6 Unit Logs: Control Room, Primary and Secondary. Required Technical Specification readings must be current for Mode 6 and unsatisfactory conditions resolved. There are no T/S Action Statements which would prohibit entry into Mode 6. 1.7 S2.OP-DL.ZZ-0001(Q), Control Room review is complete for conditions that could prevent entering Mode 6. 1.8 Control Room Area Ventilation is aligned IAW SI/S2.OP-SO.CAV-0001(Q), Control Area Ventilation Operation, as required by T/S 3.7.6.1 (3.7.6) for Modes 5 and 6. SaleM2 Page 13 of 23 Rev. 12
USER RESPONSIBLE FOR VERIFYING REVISION, STATUS AND CHANGES PRINTED 20070305 s 2 .OP-IO.ZZ-0007(Q) ATTACHMENT 1 (Page 2 of 2) REQUIREMENTS AND REVIEWS FOR MODE 6 2.0 ADMINISTRATIVE REOUIREMENTS 2.1 SM/CRS REVIEW the following: 2.1.1 ORAM status supports Mode 6 entry. 2.1.2 Technical Specification Action Statement Log to ensure entry into Mode 6 is permitted by Technical Specifications. 2.2 WCM Update And Review: 2.2.1 CHANGE WCM to Operating Mode 6. 2.2.2 GENERATE a Components in Off Normal Position Report (Off-Normal and Off Normal Tagged). 2.2.3 POSITION all components -asrequired for Mode 6 OR DOCUMENT on the report the reason a component cannot be positioned and the justification for entering Mode 6. 2.2.4 UPDATE WCM for the new positions. 2.2.5 GENERATE an Unavailable Equipment Report and review for equipment required for entering Mode 6. 2.2.6 ATTACH the Components Off Normal Position Report (Off Normal and Off Normal Tagged) and the Unavailable Equipment Report to this procedure. 2.2.7 ENSURE all equipment required for Mode 6 is OPERABLE OR a justification is provided. 3.0 FINAL REVIEW [C06361
*, Administrative and Technical Specifications requirements for Mode 6 entry (Sections 1.0 - 2.0) are completed.
+ S2.OP-IO.ZZ-0107(Q), Cold Shutdown To Refueling Administrative Requirements has been reviewed. Systems - Technical Assessment Open or Required Action Items have been resolved and permission to proceed to Mode 6 has been granted:
Control Room Supervisor Date / Time Shift Manager Date / Time SaIeM2 Page 14 of 23 Rev. 12
USER RESPONSIBLE FOR VERIFYING REVISION, STATUS AND CHANGES PRINTED 20070305 s 2 .OP-IO.ZZ-0007(Q) ATTACHMENT 2 (Page I of 2) REQUIREMENTS AND REVIEW FOR CORE ALTERATIONS (NO FUEL MOVEMENT) 1.0 TECHNICAL SPECIFICATION SURVEILLANCE REQUIREMENTS 1.1 Attachment 1, Sections 1.0 - 3.0, are completed and in periodicity. 1.2 The boron concentration of the Refueling Canal, Fuel Storage Pool, and the Refueling Cavity shall be maintained within the limits of the COLR when connected to the Reactor Coolant System. The boron concentration of each volume shall be verified at least once every 72 hours while in Mode 6 (T/S 3.9.1). 1.3 The Reactor is subcritical for >96 hours (Ž4 days as required by Technical Specification Amendment 251/232 Safety Evaluation) Reactor Subcritical Date/Time: / Present Date/Time: / 1.4 The following Technical Specification Surveillance Requirements are completed and within periodicity: 1.4.1 SC.IC-FT.NIS-00 11(Q) and SC.IC-FT.NIS-00 12(Q), for N3 1 & N32 Source Range Channels completed at least once per 7 days (as required by T/S 4.9.2): Date/Time Surveillance completed: / Present Date/Time: / 1.4.2 IF any control rod is to be moved within the reactor pressure vessel, THEN S2.OP-DL.ZZ-0002(Q), Reactor Pressure Vessel Water Level, is performed within 2 hours prior to the start of movement of any control rod in the reactor pressure vessel (as required by T/S 4.9.10): Date/Time Surveillance completed: / Present Date/Time: / Salem2 Page 15 of 23 Rev. 12
USER RESPONSIBLE FOR VERIFYING REVISION, STATUS AND CHANGES PRINTED 20070305 s2.OP-IO.ZZ-0007(Q) ATTACHMENT 2 (Page 2 of 2) REQUIREMENTS AND REVIEW FOR CORE ALTERATIONS (NO FUEL MOVEMENT) 2.0 FINAL REVIEW [C06361
- Administrative and Technical Specifications requirements for CORE ALTERATIONS (Section 1.0) is completed.
- S2.OP-IO.ZZ-0107(Q), Cold Shutdown To Refueling Administrative Requirements has been reviewed. Systems - Technical Assessment Open or Required Action Items have been resolved and permission to proceed with CORE ALTERATIONS:
CRS_ Date Time/ CRS Date / Time SM Date / Time Salem2 Page 16 of 23 Rev. 12
USER RESPONSIBLE FOR VERIFYING REVISION, STATUS AND CHANGES PRINTED 20070305 8 2 .OP-IO.ZZ-0007(Q) ATTACHMENT 3 (Page 1 of 4) REQUIREMENTS AND REVIEW FOR CORE ALTERATIONS - MOVEMENT OF IRRADIATED FUEL IN THE REACTOR PRESSURE VESSEL 1.0 TECHNICAL SPECIFICATION SURVEILLANCE REQUIREMENTS 1.1 Attachment 1, Sections 1.0 - 3.0, are completed and in periodicity. 1.2 Attachment 2, Sections 1.0 - 2.0, are completed and in periodicity.
- 1.3 IF movement of irradiated fuel is to occur between October 15th through May 15th AND is to occur before January 1,2011, THEN the Reactor is subcritical for Ž! 100 hours
(>_4 days and 4 hours as required by TSAS 3.9.3.a): Reactor Suberitical Date/Time: Present Date/Time:
- 1.4 IF movement of irradiated fuel is to occur between May 16th through October 14th, OR is to occur any time after December 31, 2010, THEN the Reactor is subcritical for _Ž 168 hours
(Ž-7 consecutive days as required by TSAS 3.9.3.b): Reactor Subcritical Date/Time: Present Date/Time: 1.5 S2.OP-DL.ZZ-0002(Q), the water level shall be determined to be at least 23 feet over the reactor pressure vessel flange within 2 hours prior to the start of movement of fuel assemblies or control rods (T/S 4.9.10): Date/Time Surveillance completed: / Present Date/Time: / Salem2 Page 17 of 23 Rev. 12
USER RESPONSIBLE FOR VERIFYING REVISION, STATUS AND CHANGES PRINTED 20070305 s 2 .OP-bO.ZZ-0007(Q) ATTACHMENT 3 (Page 2 of 4) REQUIREMENTS AND REVIEW FOR CORE ALTERATIONS - MOVEMENT OF IRRADIATED FUEL IN THE REACTOR PRESSURE VESSEL 1.6 IF movement of irradiated fuel is to occur with the Reactor subcritical for <168 hours, (between October l5th through May 15th and before January 1, 2011) THEN:
- Obtain CCW refueling temperature requirement from Outage Management (Prior to each refueling outage an Engineering Calculation will be performed to ensure that the decay heat load expected and the river water temperature are adequate to meet the required heat removal capability for core offload after the reactor has been subcritical for at least 100 hours).
CCW refueling temperature requirement: OF Date/Time:
+ Slowly adjust CCHX Controller setpoint(s) as required to maintain the CCW refueling temperature requirement.
- Ensure both Unit I and Unit 2 Spent Fuel Heat Exchangers are available with Component Cooling Water flow capability of at least 3,000 gpm each.
Date/Time: / 1.7 The following Technical Specification Surveillance Requirements are completed and within periodicity: 1.7.1 S2.OP-ST.CAN-0007(Q), each of the required containment building penetrations shall be determined to be either in its required condition or capable of being closed by a manual or automatic containment isolation valve at least once per 7 days (T/S 4.9.4.1): Date/Time Surveillance completed: / Present Date/Time: / 1.7.2 S2.OP-ST.CBV-0004(Q), verify, once per 18 months, each required containment purge isolation valve actuates to the isolation position on a manual actuation signal (T/S 4.9.4.3): Date/Time Surveillance completed: / Present Date/Time: / Salem2 Page 18 of 23 Rev. 12
USER RESPONSIBLE FOR VERIFYING REVISION, STATUS AND CHANGES PRINTED 20070305 s2.OP-IO.ZZ-0007(Q) ATTACHMENT 3 (Page 3 of 4) REQUIREMENTS AND REVIEW FOR CORE ALTERATIONS - MOVEMENT OF IRRADIATED FUEL IN THE REACTOR PRESSURE VESSEL NOTE In the event of a Fuel Handling Accident (FHA) the following ventilation requirements will ensure that airflow will be into containment allowing continuous monitoring of the containment atmosphere until containment closure is accomplished. If for any reason, these ventilation requirements can not be met, movement of fuel assemblies within the containment building shall be discontinued until the ventilation flow path(s) can be restored or the equipment hatch and personnel airlocks are closed. (T/S Bases 3/4.7.7, T/S Amendment 263/245 Safety Evaluation) 1.8 IF movement of irradiated fuel within the containment building is planned with the Containment Equipment Hatch OPEN, THEN: 1.8.1 At least one of the following ventilation flow path(s) shall be established:
- Containment Purge To Plant Vent lAW S2.OP-SO.WG-0006(Q), or
- Auxiliary Building Ventilation System in operation IAW S2.OP-SO.ABV-0001 (Q) with both of the following conditions satisfied:
- a suction flowpath through at least one Containment Airlock, and
- 2R41A and 2R41D radiation monitors are OPERABLE.
1.8.2 2R11A, 2R12A, and 2R12B block switches are in the BLOCK position (Refer to Step 3.10). 1.8.3 SC.MD-FR.CAN-0001(Q), once per refueling prior to the start of movement of irradiated fuel assemblies, verify the capability to install, within 1 hour, the equipment hatch (T/S 4.9.4.2): Date/Time Surveillance completed: Present Date/Time: ________________ / ________________ 1.8.4 A periodic verification (once per shift) of the ventilation flow path(s) established in Step 1.8.1 is required TAW S2.OP-DL.ZZ-0002(Q) (T/S Amendment 263/245 Safety Evaluation). Salem2 Page 19 of 23 Rev. 12
USER RESPONSIBLE FOR VERIFYING REVISION, STATUS AND CHANGES PRINTED 20070305
.OP-IO.ZZ-0007(Q)
ATTACHMENT 3 (Page 4 of 4) REQUIREMENTS AND REVIEW FOR CORE ALTERATIONS - MOVEMENT OF IRRADIATED FUEL IN THE REACTOR PRESSURE VESSEL 2.0 FINAL REVIEW [C0636]
- Administrative and Technical Specifications requirements for movement of irradiated fuel in the reactor pressure vessel (Section 1.0) is completed.
- S2.OP-IO.ZZ-0107(Q), Cold Shutdown To Refueling Administrative Requirements has been reviewed. Systems - Technical Assessment Open or Required Action Items have been resolved and permission to proceed with the movement of irradiated fuel in the reactor pressure vessel:
CRS _ Date_____/ CRS Date / Time SM Date / Time Salem2 Page 20 of 23 Rev. 12
USER RESPONSIBLE FOR VERIFYING REVISION, STATUS AND CHANGES PRINTED 20070305 s.OP-IO.ZZ-0007(Q) ATTACHMENT 4 (Page 1 of 3) COM PLETION SIGN-OFF SHEET 1.0 COMMENTS (INCLUDE procedure deficiencies and corrective actions.) Salem2 Page 21 of 23 Rev. 12
USER RESPONSIBLE FOR VERIFYING REVISION, STATUS AND CHANGES PRINTED 20070305 s.OP-Io.ZZ-0007(Q) ATTACHMENT 4 (Page 2 of 3) COMPLETION SIGN-OFF SHEET 2.0 SIGNATURES Print Initials Signature Date Sa1em2 Page 22 of 23 Rev. 12
USER RESPONSIBLE FOR VERIFYING REVISION, STATUS AND CHANGES PRINTED 20070305 s 2 .OP-1o.ZZ-0007(Q) ATTACHMENT 4 (Page 3 of 3) COMPLETION SIGN-OFF SHEET 2.0 SIGNATURES (continued) Print Initials Signature Date 3.0 SM/CRS FINAL REVIEW AND APPROVAL This procedure with Attachments 1-4 is reviewed for completeness and accuracy. All deficiencies, including corrective actions, are clearly recorded in the Comments Section of Attachment 5. Technical Specification compliance and procedure compliance is evaluated. Signature: Date: SM/CRS Salem 2 Page 23 of 23 Rev. 12 A}}