ML022880098

Transmittal Information of Spent Fuel Pool Cooling Request for License Amendment, Refueling Operations - Fuel Decay Time Prior to Commencing Core Alterations or Movement of Irradiated Fuel

ML022880098
Person / Time
Site:	Salem
Issue date:	10/02/2002
From:	Garchow D Public Service Enterprise Group
To:	Document Control Desk, Office of Nuclear Reactor Regulation
References
LCR S02-03, LRN-02-0331
Download: ML022880098 (155)

Text

PSEG Nuclear LLC P 0 Box 236, Hancocks Bndge, New Jersey 08038-0236 OCT 0220020 PSEG Nuclear LLC LRN-02-0331 LCR S02-03 United States Nuclear Regulatory Commission ATTN: Document Control Desk Washington, DC 20555 Gentlemen:

ADDITIONAL INFORMATION - SPENT FUEL POOL COOLING REQUEST FOR LICENSE AMENDMENT REFUELING OPERATIONS - FUEL DECAY TIME PRIOR TO COMMENCING CORE ALTERATIONS OR MOVEMENT OF IRRADIATED FUEL SALEM GENERATING STATION, UNIT NOS. 1 AND 2 FACILITY OPERATING LICENSE NOS. DPR-70 AND DPR-75 DOCKET NOS. 50-272 AND 50-311 On October 1, 2002, PSEG Nuclear LLC (PSEG) met with Mr. R. Fretz and Mr.

S. Jones of the Nuclear Regulatory Commission (NRC) staff, to discuss the subject request for license amendment submitted by PSEG on June 28, 2002 (LR-N02-0231).

The purpose of the meeting was to discuss information in the submittal related to Spent Fuel Pool (SFP) temperature limits and the calculations performed to determine maximum SFP temperatures. Attachment 1 summarizes our resolution of the issues as discussed at the meeting, and includes regulatory commitments to be met as part of implementation of the proposed amendment. provides the non-proprietary portion of the Critical Software Document for the Crosstie computer code used for the Decay Heat Management Program calculations. Attachment 3 contains the proprietary portion of the Critical Software Document for Crosstie, and the affidavit to support its withholding from public disclosure pursuant to 10 CFR 2.790. Attachment 4 provides Calculation S-C-SF-MEE-1679 Revision 0, "SFP Cooling System Capability With Core Offload Starting 100 Hours After Shutdown," that is described in our amendment request dated June 28, 2002.

The information provided herein provides additional details regarding spent fuel cooling calculation methodology and operational controls, and it does not affect our determination of no significant hazards consideration contained in our June 28, 2002 amendment request.

This letter forwards proprietaryinformation in accordancewith IOCFR 2.790. The balance of this lettermay be considerednon-proprietaryupon removal of . kA 0 95-2168 REV 7/99

Document Control Desk LRN-02-0331 OCT 0 2 2002 Should you have any questions regarding this transmittal, please contact Mr. William McTigue at (856) 339-1033.

Sincerer, D. F. Gbrchow Vice President - Operations Attachments (4)

C Mr. H. J. Miller, Administrator - Region I U. S. Nuclear Regulatory Commission 475 Allendale Road King of Prussia, PA 19406 U. S. Nuclear Regulatory Commission Attn: Mr. R. Fretz Licensing Project Manager - Salem Mail Stop 08B2 Washington, D.C. 20555-0001 USNRC Senior Resident Inspector - Salem (X24)

Mr. K. Tosch, Manager IV Bureau of Nuclear Engineering P.O. Box 415 Trenton, NJ 08625 This letter forwardsproprietaryinformation in accordance with IOCFR 2.790. The balance of this letter may be considerednon-proprietaryupon removal of .

LR-N02-0331 Attachment I

1. Decay Heat Management Program Calculations The Decay Heat Management (DHM) program calculates peak SFP temperature for each refueling outage using the Crosstie computer program. The Critical Software Document for Crosstie is provided in Attachments 2 (non-proprietary) and 3 (proprietary). The results of the DHM program calculations performed for the twelfth refueling outage at Salem Unit 2 (2R12) showed substantial conservatism when predicted SFP temperatures were compared to actual SFP temperatures during the outage. During the October 1, 2002 meeting, actual recorded values of Component Cooling Water (CCW) supply temperatures were used as input to the DHM calculation for 2R12, resulting in calculated SFP temperatures that closely correlated with actual SFP temperatures; this provided validation of the ability of the DHM calculations to predict maximum SFP temperatures accurately.

As part of implementation of the requested amendment, PSEG commits to using the DHM program calculation methodology prior to each Salem refueling to:

1. Calculate that the SFP temperature will not exceed 149 degrees F following full core offload, using one and only one heat exchanger for each SFP and to provide to the Operations staff the required Component Cooling Water temperature to achieve such results.

2. Calculate that the SFP temperature will not exceed 180 degrees F following full core offload with one heat exchanger available for both SFP's and to provide to the Operations staff the required Component Cooling Water temperature to achieve such results.

2. Decay Heat Management Procedures The integrated operating procedures for movement of spent fuel, S1(2).OP-IO.ZZ-0010(Q), Spent Fuel Pool Manipulations, establish SFP cooling requirements using the Outage Risk Assessment Model (ORAM) logic and Outage Risk Assessment procedure NC.OM-AP.ZZ-0001 (Q).

These documents establish the administrative controls needed to validate the assumptions used in the DHM calculations.

1 of 2

LRN-02-0331 Attachment 1 As part of implementation of the requested amendment, prior to initiating core offload, PSEG commits to

1. Ensuring the availability of both SFP heat exchangers, each with an available spent fuel pit pump, to support spent fuel cooling for a full core offload; and

2. Verifying that actual CCW supply temperatures validate the DHM calculation input requirements.

3. Spent Fuel Pool (SFP) High Temperature Alarm vs. Predicted Peak Temperature The Salem Unit 1 and 2 SFP high temperature alarm setpoint is 125 degrees F. The alarm setpoint is also an entry condition for abnormal operating procedure SI (2).OP-AB.SF-0001 (Q), Loss of Spent Fuel Pool Cooling. Actions directed by this procedure include suspension of fuel movement into the SFP, periodic monitoring of SFP temperature, restoration or increase of SFP cooling, verification of SFP level, and operation of the Fuel Handling Building Ventilation System.

If peak SFP temperature, as predicted by the DHM program, exceeds 125 degrees F for a refueling outage, then exceeding the alarm setpoint is an expected condition, and the alarm would not be indicative of an actual loss or degradation of SFP cooling. Therefore, PSEG commits to maintain SFP high temperature alarm capability to alert the operators in the event that SFP temperature exceeds the peak temperature predicted by the DHM program for each refueling outage. This commitment will be met as part of implementation of the requested amendment.

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LRN-02-0331 Critical Software Document for CROSSTIE Non-Proprietary

CRITICAL SOFTWARE DOCUMENT FOR CROSSTIE 1 Added Note I to Section 3.0 to address PIRS K. King item 960125282, CRCA 2, regarding interface with FHB Ventilation System. Added Note 2 to A, /

Section 3.0 to address PIRS item 980701273, CRCA 1, regarding spent fuel storage rack and assembly volume. Revised Controlled Copy Holders and CPU information on Attachmentl.

Revised Index Form (misc. information).

0 First issue. K. King Revision Revision Summary Prenare CRITICAL SOFTWARE DOCUMENT S-C-SF-MCS-0 113, Sheet I CROSSTIE REV I PAGE I OF 8

TABLE OF CONTENTS SECTION DESCRIPTION PAGE Computer Software Critical Parameters 1.0 PURPOSE 4 2.0 SCOPE 4 3.0 SOFTWARE DESCRIPTION 4

4.0 REFERENCES

5 5.0 SOFTWARE CONTROLLED DISTRIBUTION 5 6.0 VALIDATION AND VERIFICATION 6 7.0 TRAINING REQUIREMENTS 7 8.0 USERS MANUAL 7 9.0 SOURCE CODE STORAGE 7 10.0 ERROR REPORTING 7 11.0 INPUT DATA VALIDATION 8 12.0 ATTACHMENTS 9 13.0 APPENDICES 9 CRITICAL SOFTWARE DOCUMENT S-C-SF-MCS-0 113, Sheet I CROSSTIE REV 1 PAGE 2 OF 8

Computer Software Index Form Software I.D. No: S-C-SF-MCS-0O 13, Sheet 2 Software Document Name: CROSSTIE Software Document Revision: 1 System Code: I SF Software Source: HOLTEC (If not PSE&G)

Description of Software Application & Intended Use:

CROSSTIE will be used to predict Spent Fuel Pool temperatures when either the Unit 1 or Unit 2 Spent Fuel Heat Exchanger is out of service, with the operational Heat Exchanger being cycled between the two pools, requiring operation of the Spent Fuel cross-tie. This program will also be used to evaluate Spent Fuel Pool heat-up rates and equilibrium temperatureswithoutforced cooling.

Development/Revision: CROSSTIE Version 1.0 Name of Sponsor: K C. King, EDME/SE Software User Designated

Contact:

K C. King Hardware System and Operating System/Software on which Software is run:

1. Current Software Name/Rev.: CROSSTIE Version 1.0

2. Valid Software Users Manual I.D./Rev.: S-C-SF-MCS-0113, Sheet 3, Rev 0 Valid Software Theoretical Manual I.D./Rev.: N/A

4. Operating System I.D./Rev.: MS-DOS

5. Hardware: IBM PCModel 286, or compatible, with Math Coprocessor Windows 95/NT Compatible Computer Programming Facilities Used: Holtec Computer Programming Language: Fortran Time of Day Program is Normally Run: No Restrictions Associated Vendor Manuals: None Other Information: Not Y2K compatible - upgrade required prior to 1/1/2000 CRITICAL SOFTWARE DOCUMENT S-C-SF-MCS-0 113, Sheet I CROSSTIE REV I PAGE 3 OF 8

1.0 PURPOSE The purpose of this procedure is to document all pertinent aspects of the installation, usage, and maintenance of the CROSSTIE software application in accordance with references 4.1, 4.2. and 4.3.

2.0 SCOPE This procedure is applicable to the usage of the CROSSTIE program for non-safety related, as well as for, safety related applications at the PSE&G Nuclear Department by PSE&G and Contractors performing work for the PSE&G Nuclear Department.

3.0 SOFTWARE DESCRIPTION The CROSSTIE program is a comprehensive software package for predicting Spent Fuel Pool temperatures, for use on personal computers. It was developed by HOLTEC in support of the Spent Fuel Pool Rerack Project.

The program will be able to predict Spent Fuel Pool temperatures when either the Unit I or Unit 2 Spent Fuel Heat Exchanger is out of service. During this time, the operational Heat Exchanger is being cycled between the two pools, meaning at any given time, one pool is being cooled and the other is heating up. Cooling for the other Unit's pool is provided by means of the Spent Fuel cross-tie. This program can be used to predict the rate of rise in temperature for the isolated pool and at what point it will reach the maximum limit, at which point cooling will be swapped to this pool or restored to both pools.

This program will be used primarily as a pre-outage. planning tool to determine if a Spent Fuel Heat Exchanger can be taken out of service after the core is unloaded, and if cycling between pools is feasible, amongst other considerations. It can also be used (1) any time taking a Spent Fuel Heat Exchanger out of service is being considered; (2) when a loss of SF cooling is being analyzed; and (3) as a design tool to determine if the Spent Fuel Cooling System will need to be upgraded in the future.

CROSSTIE is a custom software product which may be executed on any personal computer with a standard monitor and hard/floppy drive. Although a printer is not required, a printer is recommended to provide verifiable output.

CRITICAL SOFTWARE DOCUMENT S-C-SF-MCS-01 13, Sheet I CROSSTIE REV I PAGE 4 OF 8

NOTES:

(1) The CROSSTIE program does not interface with the Fuel Handling Building Ventilation System (FHBVS), and implicitly assumes that the system can maintain the input ambient conditions. Thus, when evaluating a loss of forced cooling scenario, a separate evaluation should be performed to determine if the FHBVS can support the heat removal and pool equilibrium temperature predicted by the CROSSTIE code, and to determine the FHBVS requirements. Refer to User's Guide for further information.

(2) The Spent Fuel Pool net water volume, which is a program input, is the total volume below a given elevation less the volume of the racks and fuel assemblies. The program validation (Appx. 2) was based on testing that was performed prior to the Spent Fuel Pool Rerack. Thus, the current rack volume is different then that determined in Appx. 2, Section 5.1. The number of fuel assemblies, of course, has also changed.

Prior to the rerack, the Spent Fuel Pool contained twelve Exxon racks. Since the rerack, the Spent Fuel Pool contains three of the original Exxon racks and nine new maximum density Holtec racks. The current rack volume, which should be used for future calculations, is as follows:

Exxon Racks:

Inputs

• 300 cells (Ref. 4.4)

• 338 lbs/cell (Appx. 2, Section 5.1) 0 p = 0.29 lbs/in 3 (Appx. 2, Section 5.1)

Volume = (300

• 338)/0.29 = 349655 in3 = 202.3 ft3 Holtec Racks:

Inputs 0 1332 cells (Ref 4.4) 0 136 lbs/cell (Ref 4.5, page 21) 0 p = 0.29 lbs/in3 (Ref. 4.5, page 22)

Volume = (1332

• 136)/0.29 = 624662 in3 = 361.5 ft3 Total Rack Volume = 202.3 + 361.5 = 563.8 ft3 say 564 ft3 CRITICAL SOFTWARE DOCUMENT S-C-SF-MCS-0 113, Sheet I CROSSTIE REV 1

'PAGE 5 OF 8

The volume of the fuel assemblies is determined based on the process provided in Appx. 2, Section 5.1. The volume of one fuel assembly is:

Volume = 1700 lb / 0.23 lbs/in 3 = 7391.3 in3 = 4.277 ft3 Thus the total volume of all assemblies = # assemblies

• 4.277 The net water volume in the Spent Fuel Pool, then, should be based on the above rack volume and assembly volume.

4.0 REFERENCES

4.1 NMEDED-0001, Rev. 0, "NME Computer Software Division Procedure" 4.2 ND.DE-AP.ZZ-0052(Q), Rev. 1, "Software Control" 4.3 NC.NA-AP.ZZ-0064(Q), Rev. 1, "Software Quality Assurance" 4.4 DCPs 1EC-3252 and 2EC-3254, Spent Fuel Pool Rerack 4.5 Design Calculation S-C-SF-MDC-1240, Rev. 0, "Spent Fuel Pool thermal Hydraulic Calculation 5.0 SOFTWARE CONTROLLED DISTRIBUTION This software shall be used only on authorized machines (platforms.) Each controlled copy holder must verify the installed software prior to initial usage, as well as, whenever the operating system, and/or the computer hardware is changed. Additionally, the verification shall be performed when the CROSSTIE software is updated.

Controlled copies of the executable code shall be installed and verified only on platforms identified on the controlled copy holder list. This list is provided as Attachment 1 to this procedure.

Notification of executable source code revision will be distributed through CCG along with floppy disks containing the executable source code revision, and with instructions for installation, updating, and validation.

Attachment 3 is a sample of NME - Critical Software Revision/Update Notice and instructions. A formal record of executable code update transmittal will be included in CRITICAL SOFTWARE DOCUMENT S-C-SF-MCS-0 113, Sheet I CROSSTIE REV I PAGE 6 OF 8

6.0 VALIDATION AND VERIFICATION The initial verification of this program was performed by Holtec, the developer of CROSSTIE. The Verification and Validation Documentation (Holtec document ID HI 931099) has been reviewed and accepted by PSE&G, and is included in Appendix 2.

Platform validation / verification will also be performed when the executable code is initially installed on a particular PC platform, as well as, when the operating system or hardware of that platform is changed. The results of the initial and subsequent platform verification / validations will be documented on the Validation / Verification Results form provided as Attachment 2 to this procedure. Completed platform validations which are as a result of executable code revision will be filed in Appendix 3 of this procedure.

Completed platform validations which are as a result of when the operating system or hardware of that platform is changed will be filed in Appendix 3 of the Software Sponsors copy of this procedure.

The controlled copy holder shall be responsible for the initial software and all subsequent software revision installations, as well as, for all associated platform validations. The controlled copy holder shall forward the completed verification / validation form to the program sponsor when the installation is complete and when the platform revision is complete. The controlled copy holder shall have the verification inputs and results verified by and independent party prior to transmittal to the program sponsor. Specific instructions, if required, for performing the validation will be provided with the installation package.

7.0 TRAINING REQUIREMENTS No specific training is required to use this application. The users manual contains all the necessary information. The program requires input of test data, including temperatures and flow rates, from the user and does not require any detailed interpretation of the output results by the user.

8.0 USERS MANUAL Users manuals for this application will be issued through the CCG as S-C-SF-MCS-01 13, Sheet 3, "CROSSTIE Software Manual". This manual has a limited controlled distribution. The HOLTEC published "USERS MANUAL FOR COMPUTER PROGRAM CROSSTIE" will be included in the S-C-SF-MCS-0113, sheet 3, "CROSSTIE Software Manual".

CRITICAL SOFTWARE DOCUMENT S-C-SF-MCS-0 113, Sheet 1 CROSSTIE REV 1 PAGE 7 OF 8

9.0 SOURCE CODE STORAGE CCG has been designated as controlled copy holder number 001 for this software application. CCG will provide storage of this controlled copy in accordance with reference 4.2 10.0 ERROR REPORTING Error reporting will be in accordance with reference 4.1. Error Evaluations will be documented per reference 4.1, Attachment 1, form NMEDED-001-A1-2. Completed error evaluations will be documented within Appendix 1 of this procedure. A log of error notices will be maintained by the software sponsor. The error log will be documented per reference 4.1, Attachment 1, form NMEDED-001-AI-l. The software sponsor shall update this procedure bi-annually (minimum frequency) to add the current error log to Appendix 1. Additionally, the software sponsor will distribute all valid error notices to the controlled copy holders for inclusion into their "USERS MANUAL FOR COMPUTER PROGRAM CROSSTIE". Error notices will be removed from the Users Manual by the controlled copy holder only after the error has been corrected with a new source code release/revision as directed by the software sponsor.

11.0 INPUT DATA VALIDATION In accordance with reference 4.1, input data for this software application has been designated as "Uncontrolled Data". As such, the input data for this application will be required to be printed along with the calculation results. The input data will be verified along with the results by the verifier of the design calculation utilizing the results.

12.0 ATTACHMENTS

1. CONTROLLED COPY HOLDER LIST

2. VERIFICATION RESULTS FORM

3. SAMPLE - Critical Software Revision/Update Notice 13.0 APPENDICES

1. ERROR REPORTS AND NOTICE LOG

2. INITIAL BENCHMARK VERIFICATION

3. VERIFICATION / VALIDATION RESULTS CRITICAL SOFTWARE DOCUMENT S-C-SF-MCS-0 113, Sheet 1 CROSSTIE REV 1 PAGE 8 OF 8

ATTACHMENT 1 CONTROLLED COPY HOLDERS - SOFTWARE CONTROLLED NAME LOCATION CPU COPY (Group)

NUMBER 001 MASTER (CCG) CCG N/A 002 Software Sponsor NDAB Model - Dell OptiPlex GXa Kevin King MC N24 Dell S/N - EDP4S (EDME/SE) 003 Robert Down NDAB Model - Dell OptiPlex GXa (EDME/SE) MC N24 Dell S/N - E8KT2 004 Emin Ortalan NDAB Model - Dell OptiPlex GXa (EDME/NSS) MC N24 Dell S/N - G75N2 005 Glen Schwartz NDAB Model - Dell OptiPlex GXa

_ (EFU/SA) MC N20 Dell S/N - EFXY6 006 Ed Capper Salem Tech. Model - Dell OptiPlex GXa (ESSA/SRE) MC S02 Dell S/N - EF5HD 007 Unassigned Model Dell S/N 008 009 010 011 012 013 CRITICAL SOFTWARE DOCUMENT A-O-ZZ-MCS-01 13, Sheet I CROSSTIE ATTACHMENT I REV. I PAGE I OF 1

ATTACHMENT 2 Validation and Verification Results Form Transmittal Date : 12/09/95 Software Name & Copy Number: cccn),

Copy Holder Name & Company: wname>

Installation Location: 4location Installation Computer: ,(cpu>

Validation and Verification Results:

[ ] Success, the results of the test run match the sample output. Results attached.

[ ] Failure, the results of the test run do not match the sample output. Results attached.

Installer : Date:

Verifier : Date:

This section to be completed by Software Sponsor. A copy of the completed form will be returned to each copy holder for retention.

[ ] Installation and verification accepted.

Sponsor: Date:

CRITICAL SOFTWARE DOCUMENT A-O-ZZ-MCS-0113, Sheet 1 CROSSTIE ATTACHMENT 2 REV. 0RF K PAGE 1 OF 23

ATTACHMENT 2 I. Software Validation:

SOFTWARE VALIDATION SHALL BE PERFORMED WHENEVER THE EPG APPENDIX C PROGRAM IS INSTALLED, WHENEVER THE EPG APPENDIX C SOFTWARE IS REVISED, AND WHENEVER THE COMPUTER OPERATING SYSTEM OR COMPUTER CONFIGURATION IS REVISED OR CHANGED.

NOTES: (1) Read file "DATAFILE.TXT" -- also included as sheet 5A -- regarding data files included with the CROSSTIE program.

(2) These instructions assume the crosstie program is on the "c:" drive under the directory "crosstie". If the program is on a different drive and/or directory, substitute the actual path where it states "c:\crosstie".

Step 1: From DOS Editior, create a data file "case3.dat" as follows:

Program CROSSTIE Verification Case 3 10, 02, 93 1

2100, 2040, 60000 0, 0, 0, 240.0, 60.0 3411.0, 1.0, 0.0, 0.0, 0.0, 461.0 75.5, 0.60 Step 2: Save file "case3.dat" under directory "crosstie" Step 3. Run the CROSSTIE Program:

Open file: "c:\crosstie\crosstie.exe" Step 4. Screen will ask "INPUT THE NAME OF YOUR INPUT DATA FILE" Type: case3.dat Press: JEnter Key CRITICAL SOFTWARE DOCUMENT A-O-ZZ-MCS-O113, Sheet I CROSSTIE ATTACHMENT 2 REV. Op 'ý*k PAGE 2 OF 23

ATTACHMENT 2 Step 5: Screen will ask "INPUT TIME AFTER SHUTDOWN TO START CROSS-TIE" Type: 496.0 Press: JEnter Key Step 6: Screen will ask "INPUT POOL WATER TEMPERATURE LIMIT FOR SWITCH" Type: 140.3 Press: JEnter Key Step 7: Screen will ask "INPUT CCW COOLANT TEMPERATURE" Type: 77.0 Press: JEnter Key Step 8: Screen will ask "INPUT ENDING TIME FOR INTEGRATION" Type: 600.0 Press: JEnter Key CALCULATION IS COMPLETED Step 9: Create Sub-directory under Directory "crosstie" entitled "verif" Step 10: Move the following files from the Directory "crosstie" to the Sub directory "verif":

"case3.dat" "result.tem" "plot.dat" Step 11: Review results from DOS Editior:

Open File "c:\crosstie\verif\result.tem" Print File Open File "c:\crosstie\verif\plot.dat" Print File CRITICAL SOFTWARE DOCUMENT A-O-ZZ-MCS-0113, Sheet I CROSSTIE ATTACHMENT 2 REV. (*"ek PAGE 3 OF 23

ATTACHMENT 2 Step 12: Compare the printed results for "result.tem" and "plot.dat" to the sample output -- page 6 for "result.tem" and pages 7 to 23 for "plot.dat". If results are an exact match (time and date may vary) then installation and configuration validation was a success. Sign and date the appropriate line on the bottom of the cover form, then return the form, and printed output to the sponsor stated above.

II. If results vary in any manner other than date and time then the installation failed.

Step 1 Sign and date the appropriate line on the bottom of the form.

Step 2 Move all "save" and "output print" files to a separate directory to avoid losing them.

Step 3 Delete all of the remaining program and data files from the hard drive of the installation computer.

Step 4 Return the form, and printed output to the sponsor stated above.

III. Hardware Configuration Validation:

Step 1. At the C: Prompt:

Type: MSD Press: J Enter key Press: Alt-F (for File Menu)

Press: P (for Print)

Step 2. Using the Tab key to move and the space key to toggle the options select the options as shown below.

Press: J Enter key (for OK)

CRITICAL SOFTWARE DOCUMENT A-O-ZZ-MCS-0 113, Sheet 1 CROSSTIE ATTACHMENT 2 REV. 09' qL"*

PAGE 4 OF 23

ATTACHMENT 2

+ - - - - - - - - - - - - - - -- - - - - - - - - - - - - - - - --- --

Report Information

[ I Report ALL * [X] Mouse C] Memory Browser EX] Customer Information EX] Other Adapters [X] CONFIG.SYS CX] System Summary C] Disk Drives EX) AUTOEXEC.BAT

[X) Computer C] LPT Ports I[I WIN.INI

] Memory 3 C] COM Ports I[I SYSTEM.INI I Video IRQ Status C ] Network C] TSR Programs

[X] OS Version C] Device Drivers Print to:

(I) LPT1 ( ) COM1 ! C ) COM4

() LPT2 C )COM2 @ C ) FiLe: CREPORT.MSD .....................

() LPT3 C ) COM3 OK Cancel Step 3. Customer Information screen appears. Use the Tab key to move from field to field.

• Enter your name in the Name field.

• Enter your group in the Company Name field.

• Enter your work location including mail code in the Addressl field.

• Enter your Phone extension in the Phone field

• Enter your computers Brass tag number in the Comments field.

Press: J Enter key (for OK)

Step 4. Make two (2) copies of the resulting report from step 3. Provide one (1) with the Validation and Verification Results Form to the software sponsor and maintain the second copy in your CROSSTIE Users Manual for reference when the next Validation/Verification is required.

CRITICAL SOFTWARE DOCUMENT A-O-ZZ-MCS-0 113, Sheet 1 CROSSTIE ATTACHMENT 2 REV. 0W KeJ(

PAGE 5 OF 23

DATAFILE.TXT The data files included with the CROSSTIE computer program -

"unitl.dcy", "unit2.dcy" and "ilrll.dat." -- represent the spent fuel burnup data at the time the program was supplied by Holtec, as follows:

"unitl.dcy": U1 SF Pool inventory prior to IRI1 (cycles 1 -> 10)

"unit2.dcy": U2 SF Pool inventory at time of IRI1 (cycles 1 -> 7)

"Irll.dat"  : IRII specific data The verification performed by Holtec (see CSD A-0-MCS-0113, Sheet 1, Appx. 1) used these specific data files. As such, the verification to be performed by each Controlled Copy holder (see CSD A-0-MCS-0113, Sheet 1, Att. 2) is to also use these specific data files.

Upon successful completion of the verification, "unitl.dcy" and "unit2.dcy" should be updated to reflect the current SF Pool inventory for each unit.

NOTES:

(1) The CROSSTIE program specifically looks for the file names "unitl.dcy" and "unit2.dcy". If it is desired to maintain separate "*.dcy" files representing the updates from each cycle, they can be renamed for storage purposes. If it is desired to use these files at a future time, they must be changed back these specific file names prior to running the program. As an alternative, they can be stored under individual sub-directories, keeping the same file names.

(2) All data files -- "unitl.dcy", "unit2.dcy" and the outage specific data file -- must be included in the same directory as "crosstie.exe"I when running the program. If specific "*.dcyl" files stored under a separate sub-directory are desired, they must be moved or copied to the same directory as "crosstie.exe".

If there are any questions, please contact Kevin King at x1858.

47-?o- - ?_ -%1 5-0/ 23 Page 1 +7 . 2

141 r 7-T4 C liF_/t r- Z

• * * * *HOLTEC INTERNATIONA*****

• • • • • • • • COMPUTER CODE CROSSTIE********

$Revision: 1.0 $

$Date: 17 Dec 1993 23:30:18 $ 'r14F_

$Logfile: C:/RACKHEAT/CONTROL/CROSSTIE. FOV $ _ýlff-nz THIS PROGRAM WAS VERIFIED BY THE TEST PERFORMED DURING SALEM IRII OUTAGE, OCTOBER 1993 DESCRIPTION OF YOUR JOB Program CROSSTIE Verification Case 3 REACTOR SHUTDOWN DATE:

10 2 93 OUTAGE UNIT, TIME TO START CROSS-TIE (HR), AND TEMP LIMIT (F) 1 496.00 140.30 CCW FLOW(GPM),SFP FLOW(GPM),CCW TEMP(F),& NET WATER VOLUME (ftA 3) 2100.00 2040.00 77.00 60000.00 N1,N2,N3,Tao(HR),TaoS(HR) 0 0 0 240.00 60.00 RP(MW), CF, BP1(MWD/MTU), BP2, BP3, UW(Kg) 3411.0 1.0 .00 .00 .00 461.00 FH BUILDING AMBIENT TEMP(F), RELATIVE HUMIDITY(%)

75.50 .60 THE ENDING TIME(HR) 600.00 Heat Exchanger Temperature effectiveness p= .4489 UNIT 1 JNIT 2 (POOL) (HT-TO-HX) (HT-LOSS) (POOL) (HT-TO-HX) (HT-LOSS)

TIME T1 Q1 Qlsl HX1 T2 Q2 Qls2 HX2 (HR) (F) (BTU/HR) (BTU/HR) (F) (BTU/HR) (BTU/HR)

.00 81.8 .2215E+07 .57E+05 1 83.2 .2911E+07 .68E+05 1 495.50 81.8 .2215E+07 .57E+05 1 83.2 .2911E+07 .68E+05 0 580.31 81.8 .2215E+07 .57E+05 0 140.3 .1720E+07 13E+07 1 pA6-- 6

.00 81.75 83.24 7 2.00 81.75 83.25 4.00 81.75 83.25 7r, 6.00 81.75 83.25_____ _______

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537.00 81.75 114.50 537.50 81.75 114.85 538.00 81.75 115.19 538.50 81.75 115.54 539.00 81.75 115.88 539.50 81.75 116.23 540.00 81.75 116.57 540.50 81.75 116.91 541.00 81.75 117.25 541.50 81.75 117.59 542.00 81.75 117.93 542.50 81.75 118.27 543.00 81.75 118.61 543.50 81.75 118.94 544.00 81.75 119.28 544.50 81.75 119.61 545.00 81.75 119.94 545.50 81.75 120.28 546.00 81.75 120.61 546.50 81.75 120.94 547.00 81.75 121.26 547.50 81.75 121.59 548.00 81.75 121.92 548.50 81.75 122.24 549.00 81.75 122.57 549.50 81.75 122.89 550.00 81.75 123.21 550.50 81.75 123.53 551.00 81.75 123.85 551.50 81.75 124.17 552.00 81.75 124.49 552.50 81.75 124.80 553.00 81.75 125.12 553.50 81.75 125.43 554.00 81.75 125.74 554.50 81.75 126.05 555.00 81.75 126.36 555.50 81.75 126.67 556.00 81.75 126.98 556.50 81.75 127.28 557.00 81.75 127.59 557.50 81.75 127.89 558.00 81.75 128.19 558.50 81.75 128.49 559.00 81.75 128.79 559.50 81.75 129.09 560.00 81.75 129.39 560.50 81.75 129.68 561.00 81.75 129.98 561.50 81.75 130.27 562.00 81.75 130.56 562.40 81.75 130.80 562.79 81.75 131.02 563.17 81.75 131.24 563.54 81.75 131.46 563.91 81.75 131.67 564.26 81.75 131.87 564.61 81.75 132.07 564.95 81.75 132.26 565.28 81.75 132.45 p~ci-&-- 11 0/- -_

565 .60 81.75 132.63 14-(-,f c l-r. 7 565.92 81.75 132.81 566.23 81.75 132.98 566.53 81.75 133.15 566.83 81.75 133.31 567.12 81.75 133.48 567.40 81.75 133.63 567.68 81.75 133.79 567.95 81.75 133.93 568.21 81.75 134.08 568.47 81.75 134.22 568.73 81.75 134.36 568.97 81.75 134.49 569.22 81.75 134.63 569.45 81.75 134.75 569.68 81.75 134.88 569.91 81.75 135.00 570.13 81.75 135.12 570.35 81.75 135.24 570.56 81.75 135.35 570.77 81.75 135.46 570.97 81.75 135.57 571.17 81.75 135.68 571.37 81.75 135.78 571.56 81.75 135.88 571.75 81.75 135.98 571.93 81.75 136.08 572.11 81.75 136.17 572.28 81.75 136.27 572.45 81.75 136.36 572.62 81.75 136.44 572.79 81.75 136.53 572.95 81.75 136.61 573.10 81.75 136.70 573.26 81.75 136.78 573.41 81.75 136.85 573.56 81.75 136.93 573.70 81.75 137.01 573.84 81.75 137.08 573.98 81.75 137.15 574.12 81.75 137.22 574.25 81.75 137.29 574.38 81.75 137.36 574.51 81.75 137.42 574.64 81.75 137.49 574.76 81.75 137.55 574.88 81.75 137.61 575.00 81.75 137.67 575.11 81.75 137.73 575.22 81.75 137.79 575.33 81.75 137.84 575.44 81.75 137.90 575.55 81.75 137.95 575.65 81.75 138.00 575.75 81.75 138.06 575.85 81.75 138.11 575.95 81.75 138.15 576.05 81.75 138.20 576.14 81.75 138.25 576.23 81.75 138.30 - o

,,4r -4fCtA1r k2 576.32 81.75 138.34 576 .41 81.75 138.39 576 .50 81.75 138.43 576.58 81.75 138.47 576.67 81.75 138.51 576.75 81.75 138.55 576.83 81.75 138.59 576 .91 81.75 138.63 576.98 81.75 138.67 577.06 81.75 138.71 577.13 81.75 138.75 577.20 81.75 138.78 577.27 81.75 138 .82 577.34 81.75 138 .85 577.41 81.75 138 .88 577.48 81.75 138.92 577.54 81.75 138.95 577.61 81.75 138 .98 577.67 81.75 139.01 577.73 81.75 139.04 577.79 81.75 139.07 577.85 81.75 139.10 577.91 81.75 139.13 577.97 81.75 139.16 578.02 81.75 139.19 578 .08 81.75 139.21 578 .13 81.75 139.24 578 .18 81.75 139.27 578 .23 81.75 139 .29 578 .28 81.75 139.32 578 .33 81.75 139.34 578.38 81.75 139.36 578.43 81.75 139.39 578.48 81.75 139.41 578.52 81.75 139.43 578.57 81.75 139.45 578.61 81.75 139.47 578.65 81.75 139.50 578.69 81.75 139.52 578.74 81.75 139.54 578.78 81.75 139.56 578.82 81.75 139.58 578.86 81.75 139.59 578.89 81.75 139.61 578.93 81.75 139.63 578.97 81.75 139.65 579.00 81.75 139.67 579.04 81.75 139.68 579 .07 81.75 139.70 579 .11 81.75 139.72 579.14 81.75 139. 73 579.17 81.75 139.75 579.21 81.75 139 .76 579.24 81.75 139 .78 579.27 81.75 139.80 579.30 81.75 139 .81 579.33 81.75 139. 82 579.36 81.75 139.84 579.39 81.75 139.85 579.41 81.75 139.87 rit OF Z

579.44 81.75 139.88 579.47 81.75 139.89 579.49 81.75 139.90 579.52 81.75 139.92 579.54 81.75 139.93 579.57 81.75 139.94 579.59 81.75 139.95 579.62 81.75 139.96 579.64 81.75 139.98 579.66 81.75 139.99 579.69 81.75 140.00 579.71 81.75 140.01 579.73 81.75 140.02 579.75 81.75 140.03 579.77 81.75 140.04 579.79 81.75 140.05 579.81 81.75 140.06 579.83 81.75 140.07 579.85 81.75 140.08 579.87 81.75 140.09 579.89 81.75 140.10 579.91 81.75 140.10 579.93 81.75 140.11 579.94 81.75 140.12 579.96 81.75 140.13 579.98 81.75 140.14 579.99 81.75 140.15 580.01 81.75 140.15 580.03 81.75 140.16 580.04 81.75 140.17 580.06 81.75 140.18 580.07 81.75 140.18 580.09 81.75 140.19 580.10 81.75 140.20 580.11 81.75 140.20 580.13 81.75 140.21 580.14 81.75 140.22 580.16 81.75 140.22 580.17 81.75 140.23 580.18 81.75 140.24 580.19 81.75 140.24 580.21 81.75 140.25 580.22 81.75 140.25 580.23 81.75 140.26 580.24 81.75 140.26 580.25 81.75 140.27 580.26 81.75 140.28 580.28 81.75 140.28 580.29 81.75 140.29 580.30 81.75 140.29 580.31 81.75 140.30 580.32 81.75 140.29 580.33 81.76 140.21 580.34 81.77 140.10 580.36 81.78 139.96 580.38 81.79 139.78 580.41 81.81 139.54 580.45 81.84 139.24 580.51 81.87 138.84 580.58 81.91 138.33

2.-

580.66 81.97 137.67 580.78 82.04 136.83 580.93 82.13 135.77 581.12 82.25 134.43 581.36 82.40 132.77 581.67 82.59 130.74 582.07 82.83 128.30 582.57 83.14 125.38 583.07 83.44 122.65 583.57 83.75 120.10 584.07 84.05 117.72 584.57 84.36 115.50 585.07 84.67 113.42 585.57 84.97 111.47 586.07 85.27 109.66 586.57 85.58 107.96 587.07 85.88 106.37 587.57 86.19 104.88 588.07 86.49 103.50 588.57 86.79 102.20 589.07 87.09 100.98 589.57 87.39 99.84 590.07 87.70 98.78 590.57 88.00 97.78 591.07 88.30 96.85 591.57 88.60 95.98 592.07 88.90 95.17 592.57 89.20 94.40 593.07 89.50 93.69 593.57 89.80 93.02 594.07 90.09 92.40 594.57 90.39 91.81 595.07 90.69 91.26 595.57 90.99 90.75 596.07 91.28 90.27 596.57 91.58 89.82 597.07 91.88 89.40 597.57 92.17 89.01 598.07 92.47 88.64 598.57 92.76 88.30 599.07 93.05 87.97 599.57 93.35 87.67 600.07 93.64 87.39 P66-op 01c 57

mmmmE HOLTEC i+-7. /

I NTER NAT ION AL VERIFICATION AND VALIDATION DOCUMENTATION FOR COMPUTER PROGRAM CROSSTIE for PUBLIC SERVICE ELECTRIC AND GAS COMPANY by Yu Wang, Ph.D.

Holtec International Holtec Project 20890 Holtec Report HI-931099 Safety Related Report Category: A

. :COMPANY PRIVATE Ti.i

.e,.:is.:p. .- o rea " l .-.

Thi dcumntisproritay andi te property ofHoltec nternationl and its

,Client..It is to be used 6nly in connection with, the performan.ce of work by Holtec Iternational'o~ri its. desgn~ subcontract~ors.RFe~prod'uctio~n, :"publication or

-. *pe,.nttio*;:' inhole .o.rin.part, for any other'"purpose by -any party other than the Clieft"is"iepressly forbidden. ...

NOOSE HOLTEC INTERNATIONAL REVIEW AND CERTIFICATION LOG DOCUMENT NAME: Verification and Validation Documentation for Computer Program CROSSTIE HOLTEC DOCUMENT I.D. NUMBER: HI-931099 HOLTEC PROJECT NUMBER: 20890 CUSTOMERJCLIENT: Public Service Electric and Gas Company REVISION BLOCK ISSUE AUTHOR & REVIEWER & QA APPROVED' NUMBER DATE DATE MANAGER- & DATE

& DATE ORIGINAL * ~ 7 REVISION 1 REVISION 2 REVISION 3 REVISION 4 REVISION 5 REVISION 6 This document conforms to the requirements of the design specification and the applicable sections of the governing codes.

Note: Signatures and printed names are required in the review block.

• Must be Project Manager or his designee.

TABLE OF CONTENTS 1.0 REQUIREMENTS .................. 1-2 S 2.0 DESIGN .................... 2-1 2.1 Heat Exchanger Data ....... 2-2 2.2 Decay Heat Calculations .... ................ Q 2-2 2.3 Heat Loss Calculation ...... ................

2-4 2.4 Inital Temperatures ........ 2-5 2.5 Exchanger Isolation Condition 2-5 2.6 Program Acceptance Criteria 2-6 3.0 IMPLEMENTATION .............................. 3-1 4.0 EXPERIMENT ............... ...........

4-1 4.1 Test Setup ............... ...........

4-1 4.2 Test Results ............. ...........

4-8 5.0 PROGRAM VALIDATION ...... 5-1 5.1 Input Data .............. 5-1 5.2 Program Verification ....... 5-8 Appendix A: Test Data (224 pages)

Appendix B: Test Instrument Calibration (19 pages)

1.0 REQUIREMENTS Program CROSSTIE is developed for Salem Generating Station Units 1 and 2. This report provides the necessary verification and validation for the program in compliance with the applicable Holtec QA requirements. The prime verification approach will be the experimentation. The program will be verified against the site measurements obtained during Salem 1R1l outage.

The principal objectives of this program are summarized in four points:

(i) Predict the refueling cycle when the cross-tie operation is not possible with 120TF maximum operating temperature limit.

(ii) Develop a predictive tool which enables PSE&G to adjust the system variab]e to extend the cross-tie operation further into the future without upgrading the system.

(iii) Provide the plant reactor engineering a user friendly code to predict spent fuel pool water temperatures and to manage the cross-tie operation with minimum input data requirements.

(iv) Provide the plant mechanical engineering group with the necessary software capability to quantify system changes which might be required in the future to deal with cross-tie operation.

This computer program will have the following features:

(i) The heat load from the fuel presently stored in both pools computed precisely using the actual fuel burnups for will be able to be each fuel assembly.

When the user enters the reactor shutdown date and time outage, the heat load from the stored inventory will automaticallyfor a specific be updated.

(ii) The SFP and CCW flow rates will be set at design values to which the program will default unless new values are supplied.

temperature, Fuel Handling Building ambient air temperature The CCW inlet and relative humidity ratio will be entered as constants.

(iii) The manner of fuel discharge to the pool (number of bundles per day) and fuel specific power which will occur in the immediate future will be inputted.

(iv) The threshold temperature to initiate a switchover will be set at 120TF unless the user overrides it with another number.

(v) The program will output the maximum time between cross-tie switchovers and bulk pool temperature profiles. The user can study the effect of changing the threshold temperature to arrive at the best cross-tie operation strategy.

1-1

2.0 DESIGN To meet the requirements described in Section 1.0, the following analytical components need to be developed for the program:

a) Decay heat calculations.

b) Heat loss calculations.

c) Heat exchanger performance.

d) Numerical algorithm for the non-linear differential equations.

The bulk spent fuel pool temperature evaluation is performed by constructing a heat balance model with the heat generation source term derived from the stored fuel assemblies, and the heat removal term equal to the heat duty of the heat exchanger expressed in terms of its temperature effectiveness.

Referring to the spent fuel pool/cooler system, the governing differential equation can be written by utilizing conservation of energy:

dT C =QL" QHx (2-1) dr QL = Pco, + Q(ir., ",) - QEV (T, tQ) where:

C: Thermal capacitance of the pool, BtutF QL: Heat load to the heat exchanger, Btu/hr Q(@0o,,',) Heat generation rate from freshly discharged fuel, which is a specified function of time after reactor shutdown r, and reactor operating time r.r, Btu/hr 2-1

P~ons = *Po: Heat generation rate from the inventory spent fuel storage background heat load, Btu/hr Average specific power of fuel assemblies, Btu/hr Qvo: Heat removal rate by the heat exchanger, Btu/hr Qwv (T~t.): Heat loss to the surroundings, which is a function of pool temperature T and ambient temperature t., Btu/hr 2.1 Heat Exchanger Data QHx is a non-linear function of time. It can, however, be written in terms of effectiveness p as follows:

QHx = Wt Ct p (T- t) (2-2) where:

Wt: Coolant flow rate, lb/hr Ct: Coolant specific heat, Btu/fb - *F p: Temperature effectiveness T: Pool water temperature, *F t,: Coolant inlet temperature, *F The temperature effectiveness p is defined as

t. - ti P =  ?. (2-3)

T-t where tO is the coolant outlet temperature. Based on fuel pool heat exchanger data, p can be obtained.

2.2 Decay Heat Calculations Q(, 0 ,r,) is specified according to the provisions of USNRC Branch Technical Position ASB9-2, "Residual Decay Energy for Light Water Reactors for Long Term Cooling", Rev.

2, July, 1981.

2-2

For finite reactor operating time (r.) the fraction of operating power, P/P0 (701, '), to be used for the fission product decay power at a time -., after shutdown may be calculated as follows:

P 1 n=11 (0,0) 1 A.exp (-a.-,) (2-4)

Po 200 n=1 P P P

- (or,) =(1 + K)- (o, ;)- -(o, r, + r.) (2-5)

Po Po Po where:

Po = operating power per fuel assembly S = P/Po = fraction of operating power T = cumulative reactor operating time, seconds I, = time after shutdown, seconds K = uncertainty factor; NRC recommends 0.2 for o <_ ', < 10' and 0.1 for 103 < r, _<107 and will be determined based on the test calibration.

A*, a, = fit coefficients having the following values:

n a. (sec) 1 0.5980 1.772 x 100 2 1.6500 5.774 x 10"1 3 3.1000 6.743 x 10.2 4 3.8700 6.214 x 10.'

5 2.3300 4.739 x 10.4 6 1.2900 4.810 x 10.'

7 0.4620 5.344 x 10-6 8 0.3280 5.716 x 10-7 9 0.1700 1.036 x 10.7 10 0.0865 2.959 x 10" 11 0.1140 7.585 x 10.10 2-3

Reactor operating time, r. is determined from the assumed burnup and -, is determined upon the actual discharge dates.

Q(r, ,) is a function of time after reactor shutdown, number of assemblies, and reactor operating time r. During the fuel transfer, the heat load in the pool will increase with respect to the rate of fuel transfer and equals Q(r0', j) after the fuel transfer.

2.3 Heat Loss Calculation Qzv is a non-linear function of pool temperature and ambient temperature. QEV contains the heat evaporation loss through the pool surface, natural convection from the pool surface and heat radiation from the pool surface. The conduction heat loss through the spent fuel pool walls is not included in the analytical model; however, it is considered in the correction factor c, which will be determined in the test calibration. The heat loss can be expressed as:

Qv (m A, + hc A. 0 +Ec oA, (T4 - t,)) (2-6) where:

m: Mass evaporation rate, lb/hr - ft2

)L: Latent heat of pool water, BtuIlb A,: Pool surface area, ft2 he: Convection heat transfer coefficient at pool surface, Btu/hr - ft2 - *F 0 = T-t.: The temperature difference between pool water and ambient air, OF 6: Emissivity of water = 0.94 a: 0.1713 x 10-1 Btu/hr - ft 2 - OF' a: Correction factor to be determined by the experiment.

2-4

The mass evaporation rate m can be obtained as a non-linear function of 0. We, therefore, have m = hD (0) (WP - Wj) (2-7) where:

WP,: Humidity ratio of saturated moist air at pool water surface temperature T, lbs. of water vapor per lb. of dry air.

WS: Humidity ratio of moist air at ambient temperature t,, lbs. of water vapor per lb. of dry air.

hD(o): Mass transfer coefficient at pool water surface. hD is a non-linear function of 0, lb/hr - ft'.

2.4 Initial Temperatures Equation (2-1) is solved as an initial value problem, by noting that the pool is at a steady state condition such that the cooler heat removal rate must equal the heat generation rate from previously discharged assemblies. Hence:

Wt C, P (Tin - ti) = P*.. - QEv where:

T.i: Coincident pool water temperature (initial value before beginning of discharge)

The above equation yields:

T,. = + ti Wt Ct p 2.5 Exchanger Isolation Condition When heat exchanger is isolated, the governing enthalpy balance equation for this condition can be written as dT C = Pco, + Q (T)-QOv (2-8)

"dr 2-5

2.6 Program Acceptance Criteria Program CROSSTIE will be validated by comparing runs with experimental data.

Experimental data is to be collected from both Salem spent fuel pools and Unit 2 spent fuel pool cooling system during the scheduled 1R1l outage. Calculated water temperatures from program CROSSTIE will be compared to experimental data water temperatures over a defined time period. The acceptability of the program will be determined through engineering judgement based on the percent deviation between the measurements and the calculated results.

2-6

3.0 IMPLEMENTATION The program CROSSTIE is developed using FORTRAN. The Salem specific heat exchanger parameters and the spent fuel pool geometry are built into the program. The program requires the following input data files:

UNIT1.DCY: Burnup and discharge data for the Unit 1 spent fuel inventory.

UNIT2.DCY: Burnup and discharge data for the Unit 2 spent fuel inventory.

RFILE: Name will be specified by the user. The file contains input parameters for a specific outage.

Note: RFILE contains data for a specific outage and the fuel inventory data inputted in files UNIT1.DCY AND UNIT2.DCY should cover all prior discharges.

A sample of the input files is attached.

The program also requires the following inputs during the execution of the program:

RFILE Name TAVC: Time after-reactor-shutdown to start crosstie, hrs.

TL: Pool water temperature limit for switchover, *F TI: CCW coolant temperature, *F TEND: Ending time for integration, hrs.

CROSSTIE will generate the following output files:

RESULT.TEM: Hard copy of input and output results.

PLOT.DAT: Containing time coordinates, Units 1 and 2 pool temperatures for plotting.

UNIT1.HTL: Decay heat results from Unit 1 spent fuel inventory.

UNIT2.HTL: Decay heat results from Unit 2 spent fuel inventory.

3-1

INPUT FILE #1 & #2 BURNUP FOR THE FUEL IN THE SFP SALEM UNIT 1 & 2 File Name: "Unitl.dcy" "Unit2.dcy" Cycle Discharge Batch No. of Wt. Assy Exposure Power No. Date No. FAS KgU NMWD/MTU MW(t)

Note: Batch Group of assemblies having same burnup. It is numbered sequantialy from 1 for each cycle.

Power Reactor power in MWN(t).

Discharge Date- Enter in the format of MM,DD,YY. Example: 09,22,93 3-2

EXAMPLE DATA FILE:

1 01,21,83 1 56 461.0 18400 3411 1 01,21,83 2 12 461.0 19700 3411 2 10,04,84 1 09 461.0 20700 3411 2 10,04,84 2 52 461.0 23900 3411 2 10,04,84 3 07 461.0 21600 3411 3 10,02,86 1 53 461.0 33400 3411 3 10,02,86 2 04 461.0 21600 3411 4 08,31,88 1 02 461.0 32200 3411 4 08,31,88 2 30 461.0 37000 3411 4 08,31,88 3 42 461.0 37400 3411 4 08,31,88 4 03 461.0 38500 3411 5 03,31,90 1 09 461.0 36000 3411 5 03,31,90 2 08 461.0 29700 3411 5 03,31,90 3 12 461.0 41500 3411 5 03,31,90 4 01 461.0 25300 3411 5 03,31,90 5 45 461.0 36500 3411 6 11,09,91 1 33 461.0 42400 3411 6 11,09,91 2 28 461.0 36400 3411 6 11,09,91 3 ,08 461.0 32300 3411 7 03,16,93 1 08 461.0 34800 3411 7 03,16,93 2 08 461.0 39600 3411 7 03,16,93 3 01 461.0 29300 3411 7 03,16,93 4 01 461.0 43100 3411 7 03,16,93 5 08 461.0 39900 3411 7 03,16,93 6, 39 461.0 36400 3411 7 03,16,93 7 04 461.0 30400 3411 3-3

INPUT FILE #3 SALEM UNITS 1&2 CROSS-TIE EVALUATION INPUT INSTRUCTION LINE 1:

FIN: DESCRIPTION OF YOUR JOB LINE 2:

MTD(3): REACTOR SHUTDOWN DATE. MTD(1)=MONTH, MTD(2)=DAY, MTD(3) = YEAR.

LINE 3:

ND: UNIT WHICH IS GOING TO BE IN OUTAGE LINE 4:

Wt: SPENTFUEL POOL COOLING HEAT EXCHANGER COOLANT (CCW) FLOW RATE, GPM (DESIGN=3000 GPM)

Ws: SPENT FUEL POOL WATER FLOW RATE, GPM (DESIGN= 2280GPM; 1Rl MEASUREMENT: UNIT1=2400 GPM, UNIT2=2000 GPM)

V: NET WATER VOLUME IN THE SPENT FUEL POOL AND THE TRANSFER POOL LINE 5:

Ni: NO OF FAS IN THE BATCH 1 OF THE DISCHARGE N2: NO OF FAS IN THE BATCH 2 OF THE DISCHARGE N3: NO OF FAS IN THE BATCH 3 OF THE DISCHARGE TAO: DECAY TIME BEFORE TRANSFER FOR THE 1st DISCHARGE,HRS TAOS: TOTAL FUEL TRANSFER TIME FOR THE DISCHARGE, HRS 3-4

RP: REACTOR RATED POWER, MW(t)

CF: CAPACITY FACTOR OF THE LAST 4 MONTHS BEFORE THE LATEST SHUTDOWN BP1: AVERAGE BURNUP FOR THE ASSEMBLIES IN BATCH 1 BP2: AVERAGE BURNUP FOR THE ASSEMBLIES IN BATCH 2 BP3: AVERAGE BURNUP FOR THE ASSEMBLIES IN BATCH 3 UW: ASSEMBLY AVERAGE URANIUM WEIGHT 7.

TDRY: AMBIENT AIR TEMPERATURE (DRY BULB) IN THE FUEL HANDLING BUILDING, F WR: RELATIVE HUMIDITY IN THE FUEL HANDLING BUILDING, %

3-5

EXAMPLE INPUT:

Salem cross-tie for Unit 1 outage, 10/2/93 10,02,93 1

2100, 2040, 59000 0,0,0,240.,60.

3411., 1.0,40300.,43200.,40000.,461.0 74., 0.33 3-6

4.0 EXPERIMENT An experimental program to calibrate program CROSSTIE was jointly undertaken by PSE&G and Holtec International. Data was collected at both Salem spent fuel pools and Unit 2 spent fuel pool cooling system during the scheduled 1R1l outage in the timeframe October-November, 1993. The test instruments were installed by the Research and Testing Laboratory of PSE&G. The instrument calibration data is attached in Appendix B of this report.

4.1 Test Setup Measurements are connected to four data acquisition computers. The corresponding data collected are named "DB1", "DB2", "DB3", and "DB4", respectively. DBI contains all the measurement channels from the Unit 1 pool; DB2 contains all the measurement channels from Unit 2 pool; DB3 contains measurement channels for the Unit 2 SFHX; and DB4 contains the measurement channels for both Unit 1 and Unit 2 spent fuel water flows. The readings on the ambient air temperature and the relative humidity of the Fuel Handling Building are done manually. The instrument locations in the spent fuel pool and the associated data acquisition channels are described below and shown in Figures 4.1 to 4.4.

The instrument locations for the fuel pool cooling systems are shown in Figure 4.5a.

a. Unit 1 Spent Fuel Pool Temperature

• Suction from pool (DB1, CH 1,2)

• Discharge to pool (DB1, CH 3,4)

• Grid location CC-14, 5' above fuel rack (DB1, CH 5,6)

• Grid location CC-14, 2' below water surface (DB1, CH 7,8)

• Grid location A-35, 5' above fuel rack (DB1, CH 9,10)

° Grid location A-35, 2' below water surface (DB1, CH 11,12)

• Near existing station monitor probe (TIC-651) (DB1, CH 13) 4-1

b. Unit 2 Spent Fuel Pool Temperature

• Suction from pool (DB2, CH 1, 2)

• Discharge to pool (DB2, CH 3,4)

• Grid location D-36, 5' above fuel rack (DB2, CH 5,6)

• Grid location D-36, 2' below water surface (DB2, CH 7,8)

• Grid location DD-10, 5' above fuel rack (DB2, CH 9,10)

• Grid location DD-10, 2' below water surface (DB2, CH 11,12)

• Near existing station monitor probe (TIC-651) (DB2, CH 13)

• Air near water surface (DB2, CH 14)

c. Spent Fuel Temperature at Inlet of Unit 2 SFHX

• Used a surface style thermocouple with a range of 32-200°F near inlet to SFHX. (DB3, CH 1)

d. Spent Fuel Temperature at Outlet of Unit 2 SFHX

• Temporarily removed local temperature indicator at instrument location TI 653. Installed a thermocouple with a range of 32-200'F. (DB3, CH 7)

• Installed a surface style thermocouple with a range of 32-200'F near outlet from SFHX (DB3, CH 2)

e. Unit 1 Spent Fuel Flow

• Installed a Panametric System Flow Meter, or equivalent, on 8" line 1-SF-50 (DB4, CH 1)

f. Unit 2 Spent Fuel Flow

• Installed a Panametric System Flow Meter, or equivalent, on 8" line 2-SF-59 (DB4, CH 2)

(Performed a zero flow calibration on the flow meter.)

g. Component Cooling Inlet Temperature to Unit 2 SFHX 4-2

1) 21CCHX Outlet Temperature Installed a thermocouple with a range of 32-200'F, at location TA9286, along with existing instrumentation (DB3, CH 5)

2) 22CCHX Outlet Temperature XK *jg q

0 Installed a thermocouple with a range of 32-200'F, at location TA;264, along with existing instrumentation (DB3, CH 6)

3) Installed a surface style thermocouple with a range of 32-2000 near inlet to SFHX (DB3, CH 3)

h. Component Cooling Outlet Temperature from Unit 2 SFHX

• Temporarily removed local temperature indicator at instrument location TI 604. Installed a thermocouple with a range of 32-200'F.

• Installed a surface style thermocouple with a range of 32-200'F near outlet from SFHX (DB3, CH 4).

i. Component Cooling Flow through SFHX

• Installed a Differential Pressure Transmitter, 4-20 mA, in parallel with FE 603, at outlet of Unit 2 SFHX (DB3, CH 9,10)

j. Service Water Temperature

• Installed a thermocouple with a range of 32-200'F, at location TT-14726 (DB3, CH 19,20)

k. Unit 1 Fuel Handling Building Temperature/Relative Humidity

• Installed three Solomat temperature/relative humidity probes, or equivalent, with a range of 32-200'F/10-90%, in the area around the Unit 1 Spent Fuel Pool (manual reading)

1. Unit 2 Fuel Handling Building Temperature/Relative Humidity

• Installed three Solomat temperature/relative humidity probes, or equivalent, with a range of 32-200°F/10-90%, in the area around the Unit 2 Spent Fuel Pool (manual reading) 4-3

N4ý l

FIGURE 4-1 ISOMETRIC VIEW OF TEMPERATURE PROBE LOCATIONS SALEM UJNITl, I

N4 STATION IPROBE 3,/1 y

13 0

9,10, 11, 12 I)ISCIIARGE LINE I

SUCTION STRAINER 5,6,7,8 FIGURE 4.2 PLAN VIEW OF TEMPERATURE PROBE LOCATIONS SAILEhA UNIT I

FIGURE 4-3 ISOMETRIC VIEW OF TEMPERATURE PROBE LOCATIONS SAIE'M UNI'1' P

N -qq

-J FIGURE 4.4 PLAN VIEIY OF TEMPERATURE PROBE LOCATIONS SALEM UNIT 2

FIGURE 4.5o. INSTRUMENT CHANNEL LOCATIONS FOR DI13 4.2 Test Results The original plan for the test was to measure the pool temperatures, and other parameters, when the Unit 1 SFHX was out of service and the Unit 2 SFHX was cycled between the Unit 1 and Unit 2 pools, utilizing the cross-tie for the IRl outage. Heatup and cooldown data could then be obtained for both pools. However, due to a high heat load with the Unit 1 core unload, the Unit 1 pool could not be isolated from cooling. Therefore, during the cross-tie the Unit 1 SFHX was out of service, the Unit I pool was being cooled by Unit 2 SFHX via the cross-tie, and the Unit 2 pool was heating up. The Unit 2 pool was allowed to heat up to 1400 F, at which time the Unit 1 SFHX was returned to service, and normal cooling restored to both pools.

The theorywas developed based on the overall average bulk temperature, and the operation was controlled based on the station monitor probe reading. Observations were made on the pool water temperature distribution during different phases of the cross-tie, namely, steady state cooling, heating up without cooling, and cooling down when resuming cooling.

A close perusal of the pool temperature data shows that (i) during the heating up the maximum temperature difference between different measurement points in spent fuel pool is less than I°F; (ii) during the steady state cooling, the temperature differences between different locations are less than I°F.

After resuming forced cooling, the Unit 2 pool temperature cooled down sharply from 140'F to 80'F, at about 2.5°F/hr. As a result: (i) the maximum temperature difference between the bulk water temperature and the station probe is increased to about 5°F; (ii) the temperature difference between water surface and the bottom is less than 25 0 F. The surface temperature is higher, since coolant has higher density and tends to "sink".

4-8

The SFP bulk temperatures were calculated individually by taking nodal averages of temperature:

A T '

n where:

n - number of the measurement points T= temperature value at the measurement point i (i = 1,2,3...n except for the discharge location)

Average pool surface/bottom/station temperature was calculated by m

where:

m = number of local measurement points Tj = test value at surface/bottom/station location i,i = 1,2...m Hard copy of all raw test data and the processed bulk temperatures at surface, bottom, and station probe locations are attached in Appendix A.

The bulk temperatures for both Units 1 and Unit 2 are plotted vs. time after reactor shutdown in Figures 4.5 through 4.12. The fuel handling building ambient air temperature and relative humidity ratio during the cross-tie are also plotted vs. time in Figures 4.13 to 4.16.

4-9

HOLTEC INTERNATIONAL SALEM OUTAGE 1Rll, UNIT 1 FUEL POOL WATER TEMPERATURE 120 I.

w I

I w

I 01 00 a-_~1I m

TIME AFTER REACTOR SHUTDOWN. HRS 1R(ve0P- 4.5-~

ITOLTEC INTERNATIONAL SALEM OUTAGE IRI1, UNIT 1 POOL BULK WATER/DISCIIARGE FLOW TEMPERATUPS (During Cross-Tie Initiation) 120 Lt Lj w

tjLK a

Of w 100 V

1C 3>1sc',P I

WIA 49 I I 498 500 502 504 506 TIME AFTER REACTOR SHUTDOWN. HRS

'R40

HOLTEC INTERNATIONAL SALEM OUTAGE 1Ri 1. UNIT 1 POOL WATER/UNIT 2 CCW INLET TEMPERATURES (Dur Lng Cross- t to) 120 LL

< 100 I

r--

I.d w

6-

< cc v 8 0 - i I I I I I I I I I I i i I I I TIME AFTER REACTOR SHUTDOWN, HRS

'F I&10It& 4. 17

WATER TEMPFERATURE, F 0

C:

> -4 z

20 z

-nm__ _ _ _I 0 _ _ _ _ _ _ _ _ _ _co CC

-4F-4 P1 ZR) -4

>1 2

TIOLTEC INTERNATIONAL SALEM OUTAGE IRIl, UNIT 2 POOL BULK WATER TEMPERATURE (During Cross-Tie Heating Up and After Cross-Tie Cooling Down) 130 LL F-I <

LL 110 h_

I 3--

Q:

W h

H 90 70 TINE AFTER REACTOR SHUTDOWN. HRS F 161u(~

U 4c-+

IIOT.TEC INTERNATIONAL SALEM OUTAGE IR]., UNIT 2 POOL BULK VATER/CCW TEMPERATURES (Post Cross-Tie Cooling Down and Steady State) 130 LL ui - _ ___

<110 n

I I

W 90

-L 7 0 -- i I I i i I i i i , J 626 676 TIME AFTER REACTOR SHUTDOWN, HRS IF-qo z- 4. 10

HOLTEC INTERNATIONAL SALEM OUTAGE 1Ri1. UNIT 2 POOL BULK WATER/CCW TEMPERATURES (STEADY STATE) 85 w "80 w - __

Hy M 2

352 372 39 4112 TIME AFTER REACTOR SHUTDOWN, HRS FV-telo ,4..,

HOLTEC INTERNATIONAL SALEM OUTAGE IRl, UNIT 2 POOL BULK WATER/CCW TEMPERATURES (STEADY STATE) 78 LL~

w 0j It H

I w

w w 74 7

70 664 704 TIME AFTER REACTOR SHUTDOWN, HRS FI 0RE- +.IZ-

HOLTEC INTERNATIONAL SALEM UNIT 1 FUEL HANDLING BUILDING RELATIVE HUMIDITY RATIO 80.00 70.00 S60.00 H

0

50.00 40.00 0.00 10.00 20.00 30.00 40.00 50.00 60.00 70.00 80.00 10-22-93 TIME 18:00 , , ,.,

II'It.L 4. 13

HOLTEC INTERNATIONAL SALEM UNIT 1 FUEL HANDLING BUILDING AMBIENT AIR TEMPERATURE 70.00 69.00 D* 68 .00

ý,I CL Y

S67.00 66.00 65.00 '

0.00 10.00 20.00 30.00 40.00 50.00 60.00 70.00 80.00 10-22-93 TIME 18:00 FIGURE 4.14

HOLTEC INTERNATIONAL SALEM UNIT 2 FUEL HANDLING BUILDING RELATIVE HUMIDITY RATIO 90.00 80.00_

tR70.00 I z S60.00 0

-E 50.00 40.00 30.00 0.00 10.00 20.00 30.00 40.00 50.00 60.00 70.00 80.00 10-22-93 TIME 18:00 FIGURE 4.15

HOLTEC INTERNATIONAL SALEM UNIT 2 FUEL HANDLING BUILDING AMBIENT AIR TEMPERATURE 78.00 77.00 W

H-42 76.00 H--

75.00-7 4 .0 0 11111 11 111111111 11.1111-1 11 111 1 1' 0.00 10.00 20.00 30.00 40.00 50.00 60.00 70.00 80.00 10-22-93 TIME 18:00 FIGURE 4.16

5.0 PROGRAM VALIDATION 5.1 Input Data Net water volume The 3-D schematic view of the Salem Units 1 and 2 pool is shown in Figure 4.1 and 4.3, respectively. The rectangular dimensions of the pool are 28.5' x 37' from Elevation 89'-6" (floor surface) to 105'-6" and 28.5' x 39' from 107'-6" to 130' (top of the pool). The normal water level is EL 128'-8" which is 38.17' from the floor. The transfer pool, which is connecting to the spent fuel pool through a 4' channel, is 16' x 28.5' from EL 130' to 89'-6"

'and 12' x 28.5' from EL 89'-6" to EL 84'-6".

Volume below water surface, V V = (28.5') (39') (128.67' - 107.5')

"+ (1/2) (107.5' - 105.5') (39' + 37') (28.5') J spent fuel pool

"+ (28.5') (37') (105.5' - 89.5')

"+ (28.5') (16') (128.67' - 89.5')

"+ (28.5') (12') (89.5' - 84.5') 1 transfer pool

= 62,140 ft3 Volume of Fuel Bundles, V.

Consider 483 assemblies in the Unit 2 pool and 656 assemblies in the Unit 1 pool.

Single assembly wt: 1700 lbs (with control rods).

Wf2 = 483 x 1700

= 821,000 lbs WA = 656 x 1700 = 1,115,200 lbs 5-1

A fuel assembly consists of U0 2, Zinc, S/S 3

P UO = 0.39 #/in 3

p,,A== 0.23 #/in P, I = 0.294#/in 3 Use p = 0.23 for maximum volume V N 821,000 f2 0.23 0.23 in 3 = 2066 ft"3

= 3,570,000 W 1,115,200 0.23 0.23

= 4,849,000 in 3 = 2806 fr3 Volume of racks, V, Rack wt. = 338 lbs/cell (Exxon Rack)

Total 1170 cells.

W, = 1170 x 338 lbs

= 395,460 Stainless steel density, p, = 0.29 lbs/in 3. Therefore V - W, = 395,460 S

P, 0.29

= 1,363,655 in3 = 789 ft 3 V= Vr + Vf = 789 + 2066 3

= 2855 ft VA, = V, + VfI = 789 + 2806"

= 3595 .f3 5-2

Net water volume, V.

Unit 2:

V =V-V

= 62148 ft 3 - 2855 ft 3

= 59285 ft 3 Unit 1:

V- = V-Vh,

= 62148 - 3595

= 58,553 ft 3 If one considers some credit for water in the pipe, storage tank, and the heat exchanger, use 59,000 ft3 for Unit 1 pool verification and 60,000 ft3 for Unit 2 pool verification.

Water surface area Spent Fuel Pool As1 = 39' x 28.5' = 1111.5 fe Transfer Pool As 2 = 16' x 28.5' = 456 fte Total As = As1 + As2 = 1567.5 ft2 5-3

Heat Exchanger Performance Data The heat exchanger performance data was referenced from Report HI-92942 (PSE&G Report S-C-SF-MDC-1240). The design basis parameters are summarized as follows:

Heat exchanger type: One shell pass, 4-tube passes, U-tube shell-and tube SFP water flow rate, lb/hr: 1.14 x 106 CCW flow rate, lb/hr: 1.49 x 106 Film coefficient tubeside, Btu/ft2,F: 2215.64 Film coefficient shellside, Btu/ft2 ,F: 1500.64 Overall heat transfer surface area, fte: 2320 Tube outside diameter, in.: 0.75 Tube inside diameter, in.: 0.650 Tube wall thickness/conductivity, Btu/fte,F: 5.0 x 54E-4 Fouling factor shellside, Btu/ft2 -F 0.0002 Fouling factor tubeside, Btu/ft?-F 0.0002 The fouling factors were calibrated by the test on the Unit 2 heat exchangers.

5-4

SALEM UNIT 1 SPENT FUEL POOL SPENT FUEL INVENTORY BURNUP DATA pc£o ivs r-" ' ,Y CYCLE DISCHARGE BATCH #FAs U-kg MWD/MTU R-POWER DATE MW (t) 04,03,79 1 28 461.0 17000 3338 04,03,79 2 04 461.0 17500 3338 04,03,79 3 06 461.0 12000 3338 09,19,80 1 36 461.0 24100 3338 09,19,80 2 28 461.0 26100 3338 01,01,82 1 32 461.0 34100 3338 01,01,82 2 24 461.0 32400 3338 10,15,82 1 33 461.0 36200 3338 10,15,82 1 01 461.0 23600 3338 02,20,84 1 01 461.0 26800 3338 02,20,84 2 39 461.0 32500 3338 02,20,84 3 28 461.0 26700 3338 02,20,84 4 03 461.0 26600 3338 02,20,84 5 01 461.0 17200 3338 02,20,84 6 01 461.0 12200 3338 03,21,86 1 09 461.0 30300 3338 03,21,86 2 04 461.0 34500 3338 03,21,86 3 01 461.0 41900 3338 03,21,86 4 25 461.0 33500 3338 03,21,86 5 28 461.0 31100 3338 03,21,86 6 02 461.0 36900 3338 10,02,87 1 01 461.0 43600 3411 10,02,87 2 09 461.0 36400 3411 10,02,87 3 08 461.0 40500 3411 10,02,87 4 41 461.0 36600 3411 10,02,87 5 11 461.0 36900 3411 03,23,89 1 08 461.0 39800 3411 03,23,89 2 04 461.0 37900 3411 03,23,89 3 01 461.0 40600 3411 03,23,89 4 04 461.0 34900 3411 03,23,89 5 47 461.0 35000 3411 03,23,89 6 08 461.0 37800 3411 03,23,89 7 02 461.0 16300 3411 02,09,91 1 07 461.0 31500 3411 02,09,91 2 01 461.0 31800 3411 02,09,91 3 04 461.0 37000 3411 02,09,91 4 29 461.0 42700 3411 02,09,91 5 41 461.0 33500 3411 02,09,91 6 01 461.0 39300 3411 02,09,91 7 02 461.0 19500 3411

461.0 26600 3411 10 04,03,92 1 08 461.0 33200 3411 10 04,03,92 2 07 461.0 36600 3411 10 04,03,92 3 04 4 02 461.0 41100 3411 10 04,03,92 10 04,03,92 5 08 461.0 39500 3411 6 15 461.0 40100 3411 10 04,03,92 3411 7 25 461.0 39300 10 04,03,92 01 461.0 25700 3411 10 04,03,92 8 10 04,03,92 9 08 461.0 38600 3411 10 04,03,92 10 08 461.0 48100 3411 10 04,03,92 11 01 461.0 43700 3411 10 04,03,92 12 01 461.0 31500 3411 10 04,03,92 13 01 461.0 48700 3411 14 01 461.0 49400 3411 10 04,03,92 3411 04,03,92 15 01 461.0 43600 10 3411 04,03,92 16 01 461.0 42200 10 3411 04,03,92 17 01 461.0 48200 10 5-6

SALEM UNIT 2 SPENT FUEL POOL SPENT FUEL INVENTORY BURNUP DATA C P£O*v-O D By P$*-5<61)

CYCLE DISCHARGE BATCH #FAs U-kg MWD/MTU R-POWER DATE MW (t) 01,21,83 1 56 461.0 18400 3411 01,21,83 2 12 461.0 19700 3411 10,04,84 1 09 461.0 20700 3411 10,04,84 2 52 461.0 23900 3411 10,04,84 3 07 461.0 21600 3411 10,02,86 1 53 461.0 33400 3411 10,02,86 2 04 461.0 21600 3411 08,31,88 1 02 461.0 32200 3411 08,31,88 2 30 461.0 37000 3411 08,31,88 3 42 461.0 37400 3411 08,31,88 4 03 461.0 38500 3411 03,31,90 1 09 461.0 36000 3411 03,31,90 2 08 461.0 29700 3411 03,31,90 3 12 461.0 41500 3411 03,31,90 4 01 461.0 25300 3411 03,31,90 5 45 461.0 36500 3411 11,09,91 1 33 461.0 42400 3411 11,09,91 2 28 461.0 36400 3411 11,09,91 3 08 461.0 32300 3411 03,16,93 1 08 461.0 34800 3411 03,16,93 2 08 461.0 39600 3411 03,16,93 3 01 461.0 29300 3411 03,16,93 4 01 461.0 43100 3411 03,16,93 5 08 461.0 39900 3411 03,16,93 6 39 461.0 36400 3411 03,16,93 7 04 461.0 46800 3411 5-7

5.2 Program Verification Input from Measurement:

Verification Case 1 - Unit 2 Steady State Before Cross-Tie CCW inlet temperature, 'F: 77 CCW flow rate, gpm: 2100 SFP flow, gpm: 2040 Time after reactor shutdown, hr.: 336-408 (about 15-17 days after reactor shutdown)

Ambient air temperature, 'F: 74 Relative humidity ratio, %: 33 Verification Case 2 - Unit 2 Steady State After Cross-Tie CCW inlet temperature, °" 73.2 CCW flow rate, gpm: 3050 SFP flow, gpm: 2040 Time after shutdown, hr.: 624-744 Ambient air temperature, *F: 75.5 Relative humidity ratio, %: 50 Verification Case 3 - Unit 2 Pool Heating Up During 1Rl Cross-Tie Before HX isolation:

CCW inlet temperature, 'F: 77 CCW flow rate, gpm: 2100 SFP flow rate, gpm: 2040 TARS' to start exchanger isolation, hrs.: 496 Temperature limit, 'F: 140.3 Ambient air temperature, °F: 75.5 Relative humidity ratio, %: 60 Verification Case 4 - Unit 2 Pool Cooling Down After 1Rl Cross-Tie CCW inlet temperature, 'F: 73 CCW flow rate, gpm: 3050 SFP flow rate, gpm: 2040 TARS to re-initiate exchanger, hrs.: 580.27 Initial temperature, °F: 140.3 Ambient temperature, 'F: 75 Relative humidity ratio, %: 60

" TARS = Time After Reactor Shutdown 5-8

Verification Case 5 - Unit 1 Pool Cooled by Unit 2 Exchanger During IRll Outage CCW inlet temperature, 'F: 77.5 CCW flow rate, gpm: 2100 SFP flow rate, gpm: 2460 TARS to start cooling, hrs.: 505 Initial bulk water temperature, 'F: 123.6 TARS to end, hrs.: 514.6 Average burnup of 61 assemblies, MWD/MTU 41,300 Average burnup of 64 assemblies, MWD/MTU 27,760 Average burnup of 68 assemblies, MWD/MTU 15,240 Capacity factor: 0.9 Ambient air temperature, °F: 68.5 Relative humidity, %: 55 Verification Case 6 - Unit 1 Pool Cooled by Unit 2 Exchanger During 1R1l Outage CCW inlet temperature, 'F: 70 CCW flow rate, gpm: 3050 SFP flow rate, gpm: 2460 TARS to start cooling, hrs.: 515.13 Initial bulk water temperature, 'F: 118.9 TARS to end, hrs.: 579.7 Average burnup of 61 assemblies, MWD/ITU 41,300 Average burnup of 64 assemblies, MWD/MITU 27,760 Average burnup of 68 assemblies, MWD/MTU 15,240 Capacity factor: 0.9 Ambient air temperature, 'F: 67.5 Relative humidity, %: 55 For each verification case, the calculated temperature is plotted with the measurement (see Figures 5.1 to 5.6). It is shown that the maximum mean error in all cases is less than 1.0%

of the measured values, which is well within the experimental error.

Hard copy of all cases of runs are attached.

5-9

HOLTEC INTERNATIONAL PROGRAM CROSSTIE VERIFICATION CASE 1 UNIT 2 POOL STEADY STATE CONDITION BEFORE IRIl CROSS-TIE 90 E5 LL~ CALWtLATMV -j-5$f ( wJL~K P06L iWAgjrl TEMP.)

I w

Iw - pgT{,"t* lLl n, "40 352 372 392 TIME AFTER REACTOR SHUTDOWN, HRS Fl60RP 5-1

HOLTEC INTERNATIONAL PROGRAM CROSSTIE VERIFICATION CASE 2 UNIT 2 POOL STEADY STATE CONDITION AFTER IRlI CROSS-TIE 80 78- CALCLATa"P Li F- 76 ILI- L (c w t LULET) w 74 TIME AFTER REACTOR SHUTDOWN, HRS R40(2&5 7--

HOLTEC INTERNATIONAL PROGRAM CROSSTIE VERIFICATION CASE 3

UNIT 2 POOL HEATING UP DURING IR1l CROSS-TIE 150 1j 130 w

H 110 Ii 520 560 '

TIME AFTER REACTOR SHUTDOWN, HRS F(@OriE 5.35

HOLTEC INTERNATIONAL PROGRAM CROSSTIE VERIFICATION CASE 4 UNIT 2 POOL COOLING DOWN DURING IRil CROSS-TIE

-I 125 LL CALC0LAtheCP (VW La]

Crj 105 w

a-(oL) 853 ItW T-r c. )

65 Sii i I I I I6 i i I6 i I I I 6 I 6I I I I 576 586 596 606 616 TIME AFTER REACTOR SHUTDOWN, HRS FIO40 -r- 5.4 H5-/ rH 1 VE", p r T ryel I, .E PF R/id OF 7"TAtE iyCLiC,- "rN Fo,4 )EA E-5')/& IAL-VtHIA-1A, £/ t ?t rfra o.A A4P', 4Oo'" " ! ArOUtA.

0 -FH? " Z £O* ' t Po"E

• L1 l -,..

I" HOLTEC INTERNATIONAL PROGRAM CROSSTIE VERIFICATION CASES 5 AND 6 UNIT I POOL TEMPERATURE DURING IRl CROSSTIE

-r -

CUt-5) (aAL0V'JLAiT&P 120 LL CAsr(o I (_ALC*LAt*&"P (pooL)

I 01 0_1 100 0

0 D twUT s0-I I I "7T-C T 6C i11 i i I I i iI i i iI I I 1 f 1i 1i iI1 i i i i i i i i i i i i ii i - - -i 504 514 524 534 544 TIME AFTER REACTOR SHUTDOWN, HRS V16CIOR-C 57- 5

INPUT DATA FILES FOR ALL VERIFICATION CASES CASEI.DAT PROGRAM CROSSTIE VERIFICATION CASE 1 10,02,93 1

2100, 2040, 60000 0,0,0,240.,60.

3411.,1.0,0.,0.,0.,461.0 74., 0.33 CASE2.DAT PROGRAM CROSSTIE VERIFICATION CASE 2 10,02,93 1

3050, 2040, 60000 0,0,0,240.,60.

3411.,1.0,0.,0.,0.,461.0 75.5, 0.50 CASE3.DAT PROGRAM CROSSTIE VERIFICATION CASE 3 10,02,93 1

2100, 2040, 60000 0,0,0,240.,60.

3411.,1.0,0.,0.,0.,461.0 75.5, 0.60 CASE4.DAT PROGRAM CROSSTIE VERIFICATION CASE 4 10,02,93 1

3050, 2040, 60000 0,0,0,240.,60.

3411.,1.0,0.,0.,0.,461.0 75., 0.60 5-15

CASE5.DAT PROGRAM CROSSTIE VERIFICATION CASE 5 10,02,93 1

2100, 2460, 59000 61,64,68,240.,60.

3411.,0.9,41300.,27760.,15240.,461.0 68.5, 0.55 CASE6.DAT PROGRAM CROSSTIE VERIFICATION CASE 6 10,02,93 1

3050, 2460, 59000 61,64,68,240.,60.

3411.,0.9,41300.,27760.,15240.,461.0 67.5, 0.55 5-16

OUTPUT FILES FOR ALL VERIFICATION CASES FILE: CASEI.TEM

• • • • • HOLTEC INTERNATIONAL*****

• • • • • • • • COMPUTER CODE CROSSTIE********

$Revision: 1.0 $

$Date: 17 Dec 1993 23:30:18 $

$Logfile: C:/RACKHEAT/CONTROL/CROSSTIE.FOV $

THIS PROGRAM WAS VERIFIED BY THE TEST PERFORMED DURING SALEM IRi1 OUTAGE, OCTOBER 1993 DESCRIPTION OF YOUR JOB PROGRAM CROSSTIE VERIFICATION CASE 1 REACTOR SHUTDOWN DATE:

10 2 93 OUTAGE UNIT, TIME TO START CROSS-TIE (HR), AND TEMP LIMIT (F) 1 412.00 140.30 CCW FLOW(GPM),SFP FLOW(GPM),CCW TEMP(F),& NET WATER VOLUME (ft^3) 2100.00 2040.00 77.00 60000.00 N1,N2,N3,Tao(HR),TaoS (HR) 0 0 0 240.00 60.00 RP(MW), CF, BP1(MWD/MTU), BP2, BP3, UW(Kg) 3411.0 1.0 .00 .00 .00 461.00 FH BUILDING AMBIENT TEMP(F), RELATIVE HUMIDITY(%)'

74.00 .33 THE ENDING TIME(HR) 412.00 Heat Exchanger Temperature effectiveness p= .4489 UNIT 1 FNIT 2 (POOL) (HT-TO-HX) (HT-LOSS) ,(POOL) (HT-TO-HX) (HT-LOSS)

TIME T1 Q1 Qlsl HX1 T2 Q2 Qls2 HX2 (HR) (F) (BTU/HR) (BTU/HR) (F) (BTU/HR) (BTU/HR)

.00 81.7 .2196E+07 .83E+05 1 83.2 .2912E+07 .96E+05 1 411.50 81.7 .2196E+07 . 83E+05 1 83.2 .2912E+07 .96E+05 0 5-17 "XavrE

" rH7-T Of jervt&LY PQ'NV7- OC/TV ri6qFA~ AdA THI: 414 = .3 3

$t" wX UT or- ý4541'eZE ei-yp)

FILE: CASE2.TEM

• • • • • HOLTEC INTERNATIONAL*****

• • • • • • • • COMPUTER CODE CROSSTIE********

$Revision: 1.0 $

$Date: 17 Dec 1993 23:30:18 $

$Logfile: C:/RACKHEAT/CONTROL/CROSSTIE.FOV $

THIS PROGRAM WAS VERIFIED BY THE TEST PERFORMED DURING SALEM iRiI OUTAGE, OCTOBER 1993 DESCRIPTION OF YOUR JOB PROGRAM CROSSTIE VERIFICATION CASE 2 REACTOR SHUTDOWN DATE:

10 2 93 OUTAGE UNIT, TIME TO START CROSS-TIE (HR), AND TEMP LIMIT (F) 1 744.00 140.30 VOLUME (ft^3)

CCW FLOW(GPM),SFP FLOW(GPM),CCW TEMP(F),& NET WATER 3050.00 2040.00 73.20 60000.00 N1,N2,N3,Tao(HR),TaoS(HR) 0 0 0 240.00 60.00 RP(MW), CF, BP1(MWD/MTU), BP2, BP3, UW(Kg) 3411.0 1.0 .00 .00 .00 461.00 FH BUILDING AMBIENT TEMP(F), RELATIVE HUMIDITY(%)

75.50 .50 THE ENDING TIME(HR) 744.00 Heat Exchanger Temperature effectiveness p= .3473 UNIT 1 UNIT 2 (POOL) (HT-TO-HX) (HT-LOSS) (POOL)(HT-TO-HX) (HT-LOSS)

TIME TI Qi Qlsl HX1 T2 Q2 Qls2 HX2 (F) (BTU/HR) (BTU/HR) (F) (BTU/HR) (BTU/HR)

(HR)

.00 77.4 .2219E+07 .29E+05 1 78.7 .2860E+07 .39E+05 1 743.50 77.4 -. 2219E+07 .29E+05 1 78.7 .2860E+07 .39E+05 0 5-18

FILE: CASE3.TEM

• • • • • HOLTEC INTERNATIONAL*****

• • • • • • • • COMPUTER CODE CROSSTIE********

$Revision: 1.0 $

$Date: 17 Dec 1993 23:30:18 $

$Logfile: C:/RACKHEAT/CONTROL/CROSSTIE.FOV $

THIS PROGRAM WAS VERIFIED BY THE TEST PERFORMED DURING SALEM 1R1. OUTAGE, OCTOBER 1993 DESCRIPTION OF YOUR JOB PROGRAM CROSSTIE VERIFICATION CASE 3 REACTOR SHUTDOWN DATE:

10 2 93 OUTAGE UNIT, TIME TO START CROSS-TIE (HR), AND TEMP LIMIT (F) 1 496.00 140.30 CCW FLOW(GPM),SFP FLOW(GPM),CCW TEMP(F),& NET WATER VOLUME (ft^3) 2100.00 2040.00 77.00 60000.00 N1,N2,N3,Tao(HR),TaoS(HR) 0 0 0 240.00 60.00 RP(MW), CF, BP1(MWD/MTU), BP2, BP3, UW(Kg) 3411.0 1.0 .00 .00 .00 461.00 FH BUILDING AMBIENT TEMP(F), RELATIVE HUMIDITY(%)

75.50 .60 THE ENDING TIME(HR) 600.00 Heat Exchanger Temperature effectiveness p= .4489 UNIT 1 UNIT 2 (POOL) (HT-TO-HX) (HT-LOSS) (POOL)(HT-TO-HX) (HT-LOSS)

TIME TI Ql Qlsl HXI T2 Q2 Qls2 HX2 (HR) (F) (BTU/HR) (BTU/HR) (F) (BTU/HR) (BTU/HR)

.00 81.8 .2215E+07 .57E+05 1 83.2 .2911E+07 ".68E+05 1 495.50 81.8 .2215E+07 .57E+05 1 83.2 .2911E+07 ".68E+05 0 580.31 81.8 .2215E+07 .57E+05 0 140.3 .1720E+07 .13E+07 1 5-19

FILE: CASE4.TEM

• • • • • HOLTEC INTERNATIONAL*****

• • • • • • • • COMPUTER CODE CROSSTIE********

$Revision: 1.0 $

$Date: 17 Dec 1993 23:30:18 $

$Logfile: C:/RACKHEAT/CONTROL/CROSSTIE.FOV $

THIS PROGRAM WAS VERIFIED BY THE TEST PERFORMED DURING SALEM IRI1 OUTAGE, OCTOBER 1993 DESCRIPTION OF YOUR JOB PROGRAM CROSSTIE VERIFICATION CASE 4 REACTOR SHUTDOWN DATE:

10 2 93 OUTAGE UNIT, TIME TO START CROSS-TIE (HR), AND TEMP LIMIT (F) 1 490.30 140.30 VOLUME (ft^3)

CCW FLOW(GPM),SFP FLOW(GPM),CCW TEMP(F),& NET WATER 3050.00 2040.00 73.00 60000.00 N1,N2,N3,Tao(HR),TaoS(HR) 0 0 0 240.00 60.00 RP(MW), CF, BP1(MWD/ITU), BP2, BP3, UW(Kg) 3411.0 1.0 .00 .00 .00 461.00 FH BUILDING AMBIENT TEMP(F), RELATIVE HUMIDITY(%)

75.00 .60 THE ENDING TIME(HR) 624.00 Heat Exchanger Temperature effectiveness p= .3473 UNIT 1 UNIT 2 (POOL) (HT-TO-HX) (HT-LOSS) (POOL)(HT-TO-HX) (HT-LOSS)

Q1 Qlsl HXI T2 Q2 Qls2 HX2 TIME Tl (F) (BTU/HR) (BTU/HR) (F) (BTU/HR) (BTU/HR)

(HR)

.27E+05 1 78.6 .2945E+07 .37E+05 1

.00 77.3 .2245E+07

.27E+05 1 78.6 .2945E+07 .37E+05 0 490.00 77.3 .2245E+07 77.3 .2245E+07 .27E+05 0 140.3 .1716E+07 .13E+07 1 580.58 5-20

FILE: CASE5.TEM

• • • • • HOLTEC INTERNATIONAL*****

• • • • • • • • COMPUTER CODE CROSSTIE********

$Revision: 1.0 $

$Date: 17 Dec 1993 23:30:18 $

$Logfile: C:/RACKHEAT/CONTROL/CROSSTIE.FOV $

THIS PROGRAM WAS VERIFIED BY THE TEST PERFORMED DURING SALEM IRli OUTAGE, OCTOBER 1993 DESCRIPTION OF YOUR JOB PROGRAM CROSSTIE VERIFICATION CASE 5 REACTOR SHUTDOWN DATE:

10 2 93 OUTAGE UNIT, TIME TO START CROSS-TIE (HR), AND TEMP LIMIT (F) 1 448.50 123.50 CCW FLOW(GPM),SFP FLOW(GPM),CCW TEMP(F),& NET WATER VOLUME (ft^3) 2100.00 2460.00 77.50 59000.00 N1,N2,N3,Tao(HR),TaoS(HR) 61 64 68 240.00 60.00 RP(MW), CF, BP1(MWD/MTU), BP2, BP3, UW(Kg) 3411.0 .9 41300.00 27760.00 15240.00 461.00 FH BUILDING AMBIENT TEMP(F), RELATIVE HUMIDITY(%)

68.50 .55 THE ENDING TIME(HR) 520.00 Heat Exchanger Temperature effectiveness p= .4824 UNIT 1 UNIT 2 (POOL) (HT-TO-HX) (HT-LOSS) (POOL) (HT-TO-HX) (HT-LOSS)

TIME TI Qi Qlsl HX1 T2 Q2 Qls2 HX2 (HR) (F) (BTU/HR) (BTU/HR) (F) (BTU/HR) (BTU/HR)

.00 81.8 .2173E+07 ".10E+06 1 83.2 .2881E+07 ".11E+06 1 240.00 81.8 .2173E+07 ".10E+06 1 83.3 .2881E+07 ".11E+06 1 240.50 81.9 .2358E+07 ".10E+06 1 83.3 .2881E+07 ".11E+06 1 317.46 120.4 .2148E+08 .67E+06 1 83.3 .2881E+07 ".11E+06 1 317.59 120.4 .2148E+08 .67E+06 1 83.3 .2881E+07 ".11E+06 1 317.72 120.4 .2148E+08 .67E+06 1 83.3 .2881E+07 ".11E+06 1 448.39 115.5 .1892E+08 .56E+06 1 83.3 .2881E+07 ".11E+06 0 502.95 113.8 .1808E+08 .53E+06 0 123.5 .2244E+07 .75E+06 1 504.98 123.7 .1781E+08 .76E+06 1 113.5 .2476E+07 .52E+06 0 519.91 114.4 .1781E+08 .54E+06 0 123.5 .2244E+07 .75E+06 1 5-21

FILE: CASE6.TEM

• • • • • HOLTEC INTERNATIONAL*****

• • • • • • • • COMPUTER CODE CROSSTIE********

$Revision: 1.0 $

$Date: 17 Dec 1993 23:30:18 $

SLogfile: C:/RACKHEAT/CONTROL/CROSSTIE.FOV $

THIS PROGRAM WAS VERIFIED BY THE TEST PERFORMED DURING SALEM IR1I OUTAGE, OCTOBER 1993 DESCRIPTION OF YOUR JOB PROGRAM CROSSTIE VERIFICATION CASE 6 REACTOR SHUTDOWN DATE:

10 2 93 OUTAGE UNIT, TIME TO START CROSS-TIE (HR), AND TEMP LIMIT(F) 1 454.50 118.90 CCW FLOW(GPM),SFP FLOW(GPM),CCW TEMP(F),& NET WATER VOLUME (ft^3) 3050.00 2460.00 70.00 59000.00 NI,N2,N3,Tao(HR),TaoS(HR) 61 64 68 240.00 60.00 RP(MW), CF, BP1(MWD/MTU), BP2, BP3, UW(Kg) 3411.0 .9 41300.00 27760.00 15240.00 461.00 FH BUILDING AMBIENT TEMP(F), RELATIVE HUMIDITY(%)

67.50 .55 THE ENDING TIME(HR) 600.00 Heat Exchanger Temperature e:fectiveness p= .3773 UNIT 1 UNIT 2 (POOL) (HT-TO-HX) (HT-LOSS) (POOL) (HT-TO-HX) (HT-LOSS)

T1 Q1 Qlsl HX1 T2 Q2 Qls2 HX2 TIME (HR) (F) (BTU/HR) (BTU/HR) (F) (BTU/HR) (BTU/HR)

.00 73.9 .2225E+07 .50E+05 1 75.2 .2935E+07 .58E+05 1

.50E+05 1 75.2 .2935E+07 .58E+05 1 240.00 73.9 .2225E+07 240.50 73.9 .2410E+07 .50E+05 1 75.2 .2935E+07 .58E+05 1 108.3 .2177E+08 .43E+06 1 75.2 .2935E+07 .58E+05 1 315.50 108.3 .2176E+08 .43E+06 1 75.2 .2935E+07 .58E+05 1 316.00 108.3 .2175E+08 .43E+06 1 75.2 .2935E+07 .58E+05 1 316.50 454.00 103.6 .1903E+08 .36E+06 1 75.2 .2935E+07 .58E+05 0 102.0 .1814E+08 .33E+06 0 118.9 .2349E+07 .64E+06 1 511.87 118.9 .1777E+08 ".64E+06 1 100.5 .2683E+07 .31E+06 0 515.19

.1772E+08 ".32E+06 0 118.9 .2349E+07 .64E+06 1 540.88 101.6 118.9 .1735E+08 ".64E+06 1 99.8 .2693E+07 .30E+06 0 544.37 100.8 .1732E+08 "3.1E+06 0 118.9 .2349E+07 .64E+06 1 570.99

.1694E+08 ".64E+06 1 98.9 .2705E+07 .29E+06 0 574.72 118.8 5-22

LRN-02-0331 Salem Generating Station Units 1 and 2 Calculation S-C-SF-MEE-1679, Rev. 0

OUTSTANDING CHANGES MUST BE ATTACHED FOR WORKING COPY 20021001 Salem Generating Station, Units 1 and 2 S-C-SF-MEE-1679, Rev. 0 SFP Cooling System Capability With Core Offload Starting 100-hours After Shutdown Page I of 12

OUTSTANDING CHANGES MUST BE ATTACHED FOR WORKING COPY 20021001 EE No.: S-C-SF-MEE-1679 Rev. No.: 0 Date: 6/14/02 TITLE: SFP System Cooling Capability with Core Offload Starting 100-hours After Shutdown Periodic Review Required: Yes No X Order No.: N/A TABLE OF CONTENTS TABLE OF CONTENTS ............................................................................................................................................ 2 REVISION

SUMMARY

............................................................................................................................................. 2 1.0 PURPOSE .................................................................................................................................................... 3 2.0 SC OPE ........................................................................................................................................................... 3 3.0 DISCUSSION ................................................................................................................................................. 3 3.1 Background .............................................................................................................. 4 3.2 Assumptions/Initial Conditions ................................................................................. 4 3.3 Basic Parameters ....................................................................................................... 5 3.4 Methodology .............................................................................................................. 5 3.5 Inherent Conservatisms ............................................................................................. 5 3.6 Evaluation ....................................................................................................................... 6

4.0 CONCLUSION

/RECOMMENDATION ................................................................................................ 10

5.0 REFERENCES

............................................................................................................................................. 11 6.0 EFFECTS ON OTHER TECHNICAL DOCUMENTS .......................................................................... 11 7.0 SIGNATUURES ............................................................................................................................................. 12 ATTACHMENT A - Decay Heat Spread Sheets (5 Pages)

ATrACHMENT B - River Water Temperature Analysis (8 Pages)

ATTACHMENT C - Heat Exchanger Data Sheets at Various Temperatures (II Pages)

ATTACHMENT D - Pool Evaporative Heat Losses (1 Page)

ATTACHMENT E - Reference Documents (4 pages)

ATTACHMENT F - CC Temperature Assumption Validation (6 pages)

REVISION

SUMMARY

Revision # I Description 0 Original Issue Page 2 of 12

OUTSTANDING CHANGES MUST BE ATTACHED FOR WORKING COPY 20021001 1 EE No.: S-C-SF-MEE-1679 Rev. No.: 0 Date: 5/10/02 1.0 PURPOSE This document evaluates SFP Cooling Capabilities with 100-hours of in-vessel decay, rather than the 168 hour0.00194 days <br />0.0467 hours <br />2.777778e-4 weeks <br />6.3924e-5 months <br /> delay currently required by technical specifications. As such, this evaluation is intended to provide a technical basis for a Licensing Change Request (S02-03) to the USNRC to revise the technical specifica tions of both Salem Unit I and Salem Unit 2. This evaluation, along with the LCR, covers the time period up through 2010.

2.0 SCOPE This evaluation applies to both Salem Unit 1 and Salem Unit 2, and covers the period from October 15 th through May 15 th, annually. During the remainder of the year (May 16th through October l 4 th), the current 168-hour technical specification requirement will remain intact. This evaluation deals only with decay heat resulting from the radioactive decay of fuel rods loaded into the Spent Fuel Pools. It does not address radiological dose issues associated with fuel transfer to the SFP. Radiological dose issues are handled separately.

3.0 DISCUSSION The Salem UFSAR, Section 9.1.3.1 makes the following statements:

"The Spent Fuel Pool Cooling System maintains pool temperature at or below 149"F, provided both SFP heat exchangers are available. If only one heat exchanger is available, pool temperature is limited to 1800F."

Later, in Section 9.1.3.2, the UFSAR makes additional clarifying statements:

"In 1998, additional spent fuel pool heat removal analyses were performed. The analyses addressed potential full-core off-loads during upcoming refueling outages as well as end of plant life. These analyses concluded one pump and one heat exchanger can maintain pool temperature below 149 0 F under all combinations of decay time and CCW temperature except minimum decay times and very high cooling water temperatures. Under these later conditions, in vessel decay-time would be extended or parallel heat exchanger operation would be used to maintain pool temperature below 140°F. I" In view of the above statements, the questions to be resolved in this evaluation are:

I. With in-vessel delay time reduced from 168-hours to 100-hours, during the period from October 15 th to May 15 th (when CCW temperatures are relatively low), can the SFP cooling system maintain pool temperatures at or below 149°F with both heat exchangers available and below 180'F with one heat exchanger available? If so, is there a time limit on this activity based upon background heatwithin the Spent Fuel Pool?

2. If pool temperature is predicted to rise above 1491F, can temperatures of both pools be maintained below 1497F by employing parallel heat exchanger operations? If so, with what frequency are the heat exchangers shifted between pools to maintain 149°F?

'UFSAR Section 9.1.3.2 states parallel heat exchanger operation would be used in high heat, high CCW temperature conditions to maintain pool temperatures below 140°F. This is a discrepancy since both of the previous statements, and the Salem design calculations are based upon maintaining pool temperatures below 149*F when two heat exchangers are available. Notification 20100275 has bccn written to address this discrepancy Page 3 of 12

OUTSTANDING CHANGES MUST BE ATTACHED FOR WORKING COPY 20021001 LEE No.: S-C-SF-MEE-1 679 Rev. No.: 0 Date: 5/10/02,]

3.1 Background

There are two reasons why in-vessel decay is required before moving a fresh, hot core into the SFP, the po tential for radiation doses and pool cooling requirements. With regard to pool cooling, decay heat rates from previously irradiated fuel elements constantly decrease as the fission products and heavy elements de cay. Therefore, the longer the elements are allowed to decay within the reactor vessel, the less heat duty is transferred to the SFP.

The 168-hour limit is based upon the capability of the SFP cooling system when River temperatures, and the consequent CCWY temperatures, are at their highest. These analyses considered the River temperature to be at 90 0 F, with CCW at 99*F. This condition has never occurred at Salem, but if it did, it would occur in late July or early August, when River temperatures typically peak. While the 168-hour delay conservatively covers the entire year, it imposes an unnecessary penalty on plant operators in the cooler months, when re fuelings are typically scheduled.

In view of the above, this evaluation considers SFP cooling capabilities if a 100-hour delay is imposed prior to defueling, and is restricted to defuelings that occur between October 15 th and May 15th.

3.2 Assumptions/Initial Conditions I. Both SFPC heat exchangers will be assumed to have 6% of the tubes plugged. This is a conservative assumption because the highest tube current tube plugging is 4% (Assumption 5.0.c of Reference 5.1),

and there are no reasons to expect additional plugging in these pure water to treated water exchangers.

2. SFPC (one pump) flow to the heat exchanger is 2500 gpm (Reference 5.1, paragraph 6.2). When two heat exchangers are aligned to a single pool, 2 pumps will be assumed running with an average flow rate of approximately 1500 gpm per heat exchanger (Reference 5.1, paragraph 4.0.e).

3. CCW flow to the SFP heat exchanger is 3000 gpm (Assumption 5.0.a of Reference 5.1).

4. SFP heat exchanger fouling factor will be conservatively held equal to or greater than its design basis value (0.001075). The heat exchanger data sheet is shown in Reference 5.7.

5. Reactor power is conservatively considered to be 3479 MWt (1.02 x 3411 MWt). This envelopes the current 3459MWt based upon the 1.4% power uprate (Reference 5.3, Input 3.19).

6. Based on current refueling programs, fuel assemblies while in the reactor vessel will be assumed to be expended in accordance with the following (Reference 5.2):

• 76 assemblies with 17 months of effective full power operation 0 76 assemblies with 34 months of effective full power operation a 41 assemblies with 51 months of effective full power operation.

7. Defueling of 193 assemblies will be assumed to require 46 hours5.324074e-4 days <br />0.0128 hours <br />7.60582e-5 weeks <br />1.7503e-5 months <br /> as per current scheduling (Reference 5.3, Input 3.9). Therefore defueling is complete 146 hours0.00169 days <br />0.0406 hours <br />2.414021e-4 weeks <br />5.5553e-5 months <br /> (6.08 days) after shutdown (100 hours0.00116 days <br />0.0278 hours <br />1.653439e-4 weeks <br />3.805e-5 months <br /> + 46 hours5.324074e-4 days <br />0.0128 hours <br />7.60582e-5 weeks <br />1.7503e-5 months <br />). Actual defueling times for the past 4 Salem outages are listed below (Reference 5.4):

a. IR13 60 hours6.944444e-4 days <br />0.0167 hours <br />9.920635e-5 weeks <br />2.283e-5 months <br />

b. IR14 53 hours6.134259e-4 days <br />0.0147 hours <br />8.763227e-5 weeks <br />2.01665e-5 months <br />

c. 2R10 58hours

d. 2RI1 53hours

8. There are currently 920 fuel assemblies in the Unit I SFP (as of 1R14 in April 2001) and 812 elements in the Unit 2 pool (as of 2R12 in April 2002). (Reference 5.4).

Page 4 of 12

OUTSTANDING CHANGES MUST BE ATTACHED FOR WORKING COPY 20021001 J EENo.: S-C-SF-MEE-1679 Rev. No.: 0 Date: 5/10/02

9. Background heat in the Unit I SFP was 2.31 x 106 Btu/hour prior to outage 1R13 in 1999 (Refcrence 5.1).

10. Background heat in the Unit 1 SFP at end of life (i.e. with a full pool) is 8.46 x 106 Btufhr (Reference 5.5).

I1. The maximum number of fuel elements that can be loaded into a Salem SFP is 1632 (Reference 5.11).

12. Background heat in the pool at any given refueling between the present and end of life (or full pool) is assumed to be a straight line between 2.31 x 106 Btu/hour (Input #9) and 8.46 x 106 Btu/hour (Input

1. 10).

13. Net thermal capacity of SFP water at the end of life with all fuel racks filled (thereby minimizing avail able water volume) is 1.96 x 106 Btu/°F, as shown on page 22 of Reference 5.5.

3.3 Basic Parameters The basic parameters that are used throughout the remainder of this evaluation are reiterated below:

1. Refueling operations are conducted during the period from October 15 to May 15.

2. All 193 fuel assemblies are off-loaded to the Spent Fuel Pool (full core offload).

3. In addition to the fresh 193 assemblies, the background heat (old assemblies) in the pool represents the background heat that will exist in the year 2010.

4. River temperatures are determined from 30 years of historical data.

5. Defueling begins 100 hours0.00116 days <br />0.0278 hours <br />1.653439e-4 weeks <br />3.805e-5 months <br /> after reactor shutdown.

6. All SFP heat removal is via the Spent Fuel Pool Cooling System. No credit is taken for heat transfer via evaporative cooling or to the SFP (concrete) structure.

3.4 Methodology

1. Determine the decay heat rate from the off-loaded core using USNRC Branch Technical Position ASB 9-2 (Reference 5.6).

2. Determine background heat that will exist in the SFP in the year 2010.

3. Evaluate Delaware River temperatures during the period from October through May.

4. Benchmark the SFP heat exchanger design basis parameters against the Joseph Oats (Manufacturer's) data sheet, using the HTC-STX heat exchanger design computer program.

5. Using the benchmarked heat exchanger model in the HTC-STX heat exchanger computer program, de termine heat duties with various SFP temperatures and CCW temperature appropriate for the time pe riod.

6. Evaluate the ability of the SFP Cooling System to maintain pool temperature limits.

3.5 Inherent Conservatisms This analysis considers heat removal from the Salem Spent Fuel Pools using only forced cooling provided by the SFPC heat exchangers. By relying exclusively on the SFPC heat exchangers, the analysis contains several substantial conservatisms as described below. These conservatisms could be quantified and credited in this calculation. However, at this time they will be left as providing additional temperature margins.

I No credit is taken for evaporative cooling, i.e. pool bulk temperature cooling resulting from evapo ration at the surface of the SFP. Reference 5.5 indicates that evaporative cooling contributes 0.86 x Page 5 of 12

OUTSTANDING CHANGES MUST BE ATTACHED FOR WORKING COPY 20021001 1 EE No.: S-C-SF-MEE-1679 Rev. No.: 0 Date: 5/10/02 106 Btu/hour at 150°F and 3.87 x 106 Btu/hour at 180°F. Consequently, if the pool reaches 180°F, evaporative cooling amounts to about 10% of the peak heat load in the SFP.

2 No credit is taken for cooling through the concrete structure of the pool. Heat is conducted through the pool steel liner, concrete structure, and ultimately to the cooler environment beyond the struc ture. The higher the pool water temperature, the more heat transmitted through the structure.

3 RHR cooling continues to provide forced cooling to the SFP with all fuel elements removed to the SFP as long as the refueling canal remains flooded and the transfer gate is open. The cooler water in the reactor vessel and refueling canal will transfer to the SFP via natural circulation through the transfer gate. This potential cooling source is never credited in any analysis or procedure.

3.6 Evaluation Core Decay Heat Decay heat from the newly discharged core is determined using the USNRC Branch Technical Position ASB 9-2, Residual Decay Heat for Light-Water Reactors for Long-Term Cooling (Reference 5.6). This is the most conservative of the various computer codes accepted by the USNRC for calculating fuel element decay heat, and is conservatively used here without scaling factors or other adjustments.

As shown in Attachment A, pages Al through A4, the residual heat from the 193 assembly offload to the SFP is shown to be 3.72 x 107 Btu/hr as summarized in the following table. The 146 hours0.00169 days <br />0.0406 hours <br />2.414021e-4 weeks <br />5.5553e-5 months <br /> after shutdown includes the I00-hour delay plus an additional 46 hours5.324074e-4 days <br />0.0128 hours <br />7.60582e-5 weeks <br />1.7503e-5 months <br /> to offload the 193 assemblies. A 10% uncertainty factor is included per the BTP.

This is the highest heat load in the pool from the newly discharged core, and it exists only at the moment that the final assembly is moved into the pool. After that time, the heat load continuously decays to lower values. Nonetheless, this value is used throughout this evaluation as the heat in the SFP.

Number or Reactor Power Time to Off-Load Effective Full Calculated Decay Assemblies After Shutdown Power Hours of Heat Burnup 76 3479 MWt 6.08 days (146 hrs) 12,410(17 mos.) 1.31 x 10" Btu/hr 76 3479 MWt 6.08 days (146 hrs. 24,820 (34 mos.) 1.36 x 10' Btu/hr 41 3479 MWt 6.08 days (146 hrs ) 37,230 (51 mos.) 7.43 x 10' Btu/hr Heavy Elements 3479 MWt 6.08 days (146 hrs ) Same as above 3.03 x 10' Btu/hr (all assemblies)

Core Total 3.72 x 10" Btu/hr Background 6.8 x 10' Btu/hr Heat2 Peak Pool Heat 4.4 x 10 Btu/hr Load in 2010 2 Derivation of background heat is discussed in the next paragraph.

Page 6 of 12

OUTSTANDING CHANGES MUST BE ATTACHED FOR WORKING COPY 20021001 EE No.: S-C-SF-MEE-1679 Rev. No.: 0 Date: 5/10/02-1 Background Heat Background heat in the SFP is based on Unit 1, since the Unit I SFP contains more spent fuel rods than does the Unit 2 pool (based on approximately 3 more refueling outages by Unit 1). With 965 elements in the pool as of October 2002 (I Ri5) and at the current rate of 76 elements being permanently discharged per refueling cycle, the Unit I pool will be full in approximately 8 more refueling cycles (e.g. by the year 2014).

Page A5 of Attachment A shows a straight-line graph of decay heat from 2002 to 2014, based upon In put/Assumptions Nos. 8 through 12.

As seen on page A5 of Attachment A, the background decay heat in the Unit I SFP in year 2010 is 6.8 x 106 Btu/hour. When added to the freshly offloaded core, the peak total pool heat in the year 2010 is 4.4 x 107 Btulhr.

Delaware River/CCW Temperature As shown in Attachment B, pages BI through B8, the average monthly temperature in the Delaware River (measured at Reedy Island) between the months of October and May are 63 0 F and below. These tempera tures are based upon 30 years of weekly data recorded at Reedy Island, a location just upstream of Salem and Hope Creek. These pages also show that on average, inlet temperatures at the plant run 3°F higher than Reedy Island. Even though there have been measurements of plant temperatures as much as 5°F higher (and as low as 1iF) than Reedy Island, the 3°F average is considered conservative in a condition where one of the two Salem Units is shutdown. The Salem Units account for nearly all of the output heat in the River.

Hope Creek has a cooling tower, through which most waste heat is released to the environment. Therefore, with one of the two Salem Units shutdown (and only discharging waste reactor heat), historical average dif ferentials between the plant and Reedy Island are conservative.

The CC supply temperature is determined in Attachment F, based on a Service Water inlet temperature of 66°F, with both one and two CCW heat exchangers aligned. Using both CCW heat exchangers during the few days that SFP heat loads are at their peak would have beneficial effects with regards to minimizing the CC supply temperature. However, since both CCW heat exchangers may not be available when fuel is moved, this analysis is being performed assuming only one CCHX is available. From Attachment F, the CCW supply temperature is 71°F with one CCHX and a Service Water inlet temperature of 66°F (with two SFHXs). '

Temperature Description 0

63 F Delaware River historical data 30 F Reedy Island to plant intake 0

66 F Service Water Inlet Temperature 71°F CCW Temperature Based on 66*F SW Inlet, as shown in Attachment F Use of 71°F for this analysis is considered appropriate for two reasons:

I This evaluation provides a technical basis for reducing the in-vessel decay time for defueling from 168 hours0.00194 days <br />0.0467 hours <br />2.777778e-4 weeks <br />6.3924e-5 months <br /> to 100-hours during the months from mid-October to mid-May. Before fuel is actually trans ferred, the Salem Decay Heat Management Program is implemented in accordance with Outage Risk Management procedures (Reference 5.10) for the actual conditions in existence at outage time. In the case of a particularly mild winter or particularly hot summer where River temperatures might be well Page 7 of 12

OUTSTANDING CHANGES MUST BE ATTACHED FOR WORKING COPY 20021001 1EE No.: S-C-SF-MEE-1679 Rev. No.: 0 Date: 5/10102 above predicted temperatures, fuel would not be transferred until the Decay Heat Manaogement Program indicated pool temperature limits would be achieved. 3 2 The inherent conservatisms in this analysis (i.e. evaporative cooling, structure cooling, RHR cooling) are of sufficient magnitude to account for any foreseeable changes in river temperatures or other poten tially non-conservative assumptions. Hence, this calculation is considered to be sufficiently conserva tive.

Both SFPC Heat Exchangers As shown in Attachment C, pages CI through C3, the HTC-STX heat exchanger computer code, Version 3.6, is benchmarked against the original Joseph Oats data sheet from Reference 5.7. It should be noted that the HTC-STX data sheet says that SFPC surface area is over-designed by 7.55%. This is consistent with HOLTEC International's analysis of this same heat exchanger (Reference 5.5). In Reference 5.5. HOLTEC concluded that the SFPC heat exchanger was over-designed by 7.04%. Based on their analysis, HOLTEC concluded that the design basis heat duty should have been 12.78 x 106 Btu/hour rather than the 11.94 x 106 Btu/hour of the Joseph Oats data sheet.

The same heat exchanger model that produced the benchmarked data sheet was then changed to incorporate 6% tube plugging and to revise shell-side (CCW) inlet temperature to 71*F. Using this model, heat duties were calculated for various spent fuel pool temperatures. As shown on Attachment C, page C4, the peak spent fuel pool heat rate of 4.4 x 107 Btuthour is removed at a pool temperature of 1610 F and a fouling fac tor of 0.00105. This says that if only one SFP heat exchanger and one SFP pump are used on the hot pool, pool temperature will eventually rise to 16 MF.

Technically, 161°F is not a limitation on the spent fuel pool as it is qualified to at least 180*F. A 161*F temperature in the pool can cause moisture and humidity on the refueling floor due to evaporative losses from the pool, but does not violate the design basis. Placing the heat exchanger from the non-refueling unit in parallel with the hot pool unit (crosstie mode), temperature in the hot pool can be reduced to 128.5°F, since that is the temperature that results from a heat duty of 2.2 x 107 Btu/hr per heat exchanger (see page C5) 4 . Since the non-refueling pool will contain only background heat, it will begin to heat up at 3.5°F per hour (6.8 x 106 Btu/hr/1.96 Btu/°F). At this rate, it would take 14 hours1.62037e-4 days <br />0.00389 hours <br />2.314815e-5 weeks <br />5.327e-6 months <br /> to heat from 100°F to 1497F 5

. In the meantime, during that same 14 hours1.62037e-4 days <br />0.00389 hours <br />2.314815e-5 weeks <br />5.327e-6 months <br />, the hot core will have decayed by I x 106 Btu/hr, or approxi mately 2% of its peak value.

At the end of 14 hours1.62037e-4 days <br />0.00389 hours <br />2.314815e-5 weeks <br />5.327e-6 months <br />, the non refueling pool will have to be cooled by realigning it back to it's unit's heat exchanger. At this time, the hot pool is at 128.5 0 F and with one heat exchanger still assigned to that pool, the heat up rate is 6.6°F per hour between 128.5°F and 149'F. This allows 3.7 hours8.101852e-5 days <br />0.00194 hours <br />1.157407e-5 weeks <br />2.6635e-6 months <br /> to cool down the non refueling pool before both heat exchangers need to be shifted back to the hot pool. In 3.7 hours8.101852e-5 days <br />0.00194 hours <br />1.157407e-5 weeks <br />2.6635e-6 months <br />, the non 3 in October. river temperatures are highest on the day the fuel is offloaded and river temperatures slowly decrease thereafter. With re sidual heat from the fuel also decaying. SFP cooling capabilities become more conservative with each passing day. In May, however, river temperatures can be expected to slowly increase (typically 2.2'F to 2.3*F per week) after the fuel has been offloaded. This is not conservative, although the decaying residual heat would offset the temperature increases. Nonetheless, to assure that temperature in creases after fuel offload do not adversely impact the results of the pre-outage analyses. refueling in May (with 100-hour in-vessel delay) has been limited to May 15th. This assures that the fuel inthe pool will be well decayed as River temperatures rise into the month of June.

4 This analysis assumes that pre-outage assessments of SFPC capabilities is ill result in operators placing the 2nd heat exchanger in paral lel with the hot pool at some time either before or shortly after the pool heats above 128 5*F.

5During the period when the non-refueling pool is not being cooled because its heat exchanger is being used on the hot-pool, the tem perature inthe non-refueling pool could be taken above 1491F. However, this evaluation is based upon activities necessary to maintain both Salem SFPs at temperatures of 1491F or below.

Page 8 of 12

OUTSTANDING CHANGES MUST BE ATTACHED FOR WORKING COPY 20021001 1 EE No.: S-C-SF-MEE-1679 Rev. No.: 0 Date: 5/10/02 refueling pool can be cooled to I IIF. In the final analybis, both pools will be maintained at 149TF or be low as summarized below:

Pool No. Average Heat Added Differential Heat up or Initial Final Time to HXs Heat Re- (Btu/hr) (Btu/lhr Cool down Temp. Temp. Switch moval Rate HX (Btuthr)

Refueling 2 4.4 x 10' 4.4 x 10C 0 0 128.5 0F 128.5 0 F 14 hrs.

Non-Refuel 0 0 6.8 x 106 + 6.8 x 106 +3.5 0 F/hr 100 0F 1490 F 14 hrs.

Refueling I 3.2 x 10' 4.3 x 10" 0 0

+1.3 x I0C +5.6 F/hr 128.5 F 149°F 3.7 hrs.

Non-Refuel I 2.7 x 107 6.8 x 106 -2.3 x 107 -10.3OF/hr 1490F 111 F 3.7 hrs.

Refueling 2 5.2 x 10' 4.3 x 107 -9 x 106 -4.6*F/hr 149*F 128.507 10.8 hrs.

"Non-Refuel 0 0 6.8 x lb +6.8 x 106 +3.5-F 111°F 149OF 10.8 hrs.

As can be seen above, the non refueling pool heat exchanger can be shifted between its own pool and the hot pool on a 3.7 hours8.101852e-5 days <br />0.00194 hours <br />1.157407e-5 weeks <br />2.6635e-6 months <br /> on, 10.8 hours9.259259e-5 days <br />0.00222 hours <br />1.322751e-5 weeks <br />3.044e-6 months <br /> off basis 6 as long as necessary to maintain both SFPs below 149TF.

In actual practice, it may be desirable to hold both SFPs below some lower temperature (i.e. less than 149°F) in order to minimize the humidity and moisture in theFuel Handling Building, both from the stand point of conducting refueling operations and to minimize affects on the FHB ventilation system. The fre quency of shifting heat exchangers to maintain some lower pool temperature (135°F for example), will be determined on an outage-by-outage basis via the Decay Heat Management Program pre-outage assessment of SFP heat loads. Judgments can made at just prior to the outage as to what temperatures to control (1491F or lower) and how often, under the specific circumstances of the outage, to swap heat exchangers.

One SFPC Heat Exchanger Attachment C gives the heat duties through one SFPC heat exchanger with SFP temperatures of 160'F, 161T, 170'F, and 180T, as summarized in the following table.

Pool CCW Heat Capacity Page Temperature Temperature No.

161OF 71WF 4.40 x 107 Btulhour C4 170°F 71°F 4.80 x 10" Btu/hour C7 180°F 71 °F 5.31 x 10' Btu/hour C8 Average 4.84 x 10 Btu/hour The above table indicates that the average heat removal rate between a pool temperature of 180*F and 161IF is 4.84 x l07 Btu/hour. With a heat input rate of 4.4 x 107 Btu/hour, this leaves an average of 4.4 x 106 Btu/hour available for water temperature cool-down. With the thermal capacity of the water being 1.96 x 106 Btu/PF (see Assumption/Initial Conditions 413), this indicates that a single SFP heat exchanger with a single pump can cool the pool water from 180'F to 161'F in approximately 8.5 hours5.787037e-5 days <br />0.00139 hours <br />8.267196e-6 weeks <br />1.9025e-6 months <br /> as shown below.

6 Cycles times will increase as the SFP heat rate continues to decay. These values are the bounding values Page 9 of 12

OUTSTANDING CHANGES MUST BE ATTACHED FOR WORKING COPY 20021001 I EE No.: S-C-SF-MEE-1679 Rev. No.: 0 Date: 5/10/02 1.96 x 106 Btu/'F/4.4 x 106 Btulhour = 0.445 hour0.00515 days <br />0.124 hours <br />7.357804e-4 weeks <br />1.693225e-4 months <br />,/°F x 19'F = 8.5 hours5.787037e-5 days <br />0.00139 hours <br />8.267196e-6 weeks <br />1.9025e-6 months <br /> If only one SFPC heat exchanger was available to cool both pools, then the pool with only background heat would slowly heat up while the hot pool is being cooled. With a potential background heat of 6.8 x 106 0

Btu/hour in the year 2010, this pool will heat up at a rate of approximately 3.5 F/hour (6.8 x 106 Btu/hour/1.96 x 106 Bt1i3°F). The available heat exchanger would be shifted to the background pool when it heats to 180'F. In this case, the hot pool will heat from 161'F to 180°F in approximately 51 minutes (1.96 x 106 Btu/°F/44.0 x 106 Btu/hour = .0445 hour0.00515 days <br />0.124 hours <br />7.357804e-4 weeks <br />1.693225e-4 months <br />s/0 F x 19°F = 0.85 hours9.837963e-4 days <br />0.0236 hours <br />1.405423e-4 weeks <br />3.23425e-5 months <br /> = 51 minutes), at which time the heat exchanger would be returned to the hot pool. During the hot-pool heat up, the background pool would cool from 180°F to approximately 1591F. The cycle would continue, as shown in the following table, until either the 2nd heat exchanger is returned to service or the core is reloaded into the refueling Unit.

Refueling Time Dura- Heat up/ Initial Final Background Heat up/ Initial Final ltion Prior to Cooldown Temp Temp Pool Cooldown Temp Temp HX Transfer Rate Rate HX Aligned 17.1 hours1.157407e-5 days <br />2.777778e-4 hours <br />1.653439e-6 weeks <br />3.805e-7 months <br /> Holding 161OF 161 0 F NoHX +3.5F/hr 120OF 180OF No HX 0.85 hours9.837963e-4 days <br />0.0236 hours <br />1.405423e-4 weeks <br />3.23425e-5 months <br /> +21.4 0 F/hr 161OF 18D°F HX Aligned .24.7 0 F/hr 180OF 159OF HX Aligned 6 0 hours0 days <br />0 hours <br />0 weeks <br />0 months <br /> -3.26°F/hr 0 180-F 161-F No HX +3.5 F/hr 159-F 180OF No HX 0.85 hours9.837963e-4 days <br />0.0236 hours <br />1.405423e-4 weeks <br />3.23425e-5 months <br /> +21.4°F/hr 161 0F 180 0F HX Aligned -24.7 0F/hr 180-F 159-F As shown in the above table, the 6.0-hour (hot pool)/0.85 hour9.837963e-4 days <br />0.0236 hours <br />1.405423e-4 weeks <br />3.23425e-5 months <br /> (background pool) cycle can be continued as long as would be necessary to either restore the unavailable heat exchanger or begin transferring hot fuel back into the vessel of the refueling unit. In reality, a 51 minute duration between heat-exchanger shifting may be impractical due to the time it takes to manually swap the exchangers. However, it should be noted that the time durations would increase every time a new cycle begins because the hot fuel in the refueling pool would be rapidly decaying and reducing the amount of heat being transferred to its pool. In addition, the cooling times available prior to each heat exchanger shift are considered to be very conservative and in reality, are expected to be longer. This is the case because (1) the BTP heat loads are very conservative and the heat from the hot pool is expected to be 5% to 10% lower (2) there would be considerable evaporative cooling from the pools at these elevated temperatures (as much as 4 x 106 Btu/hr at 180'F) and (3) the con crete structure would also act to buffer the temperature changes. Furthermore, a more accurate assessment of the pool heat up times will be done prior to a refueling outage, as part of the Decay Heat Management Program, so proper planning on when and if to remove a SFHX from service can be performed.

4.0 CONCLUSION

/RECOMMENDATION During the period from October 15 th through May 15 th up to and including the year 2010, a fully radiated 193 element core can be off-loaded to a Spent Fuel Pool with a 100-hour in-vessel decay, rather than a 168 hour0.00194 days <br />0.0467 hours <br />2.777778e-4 weeks <br />6.3924e-5 months <br /> decay, because the SFPC system is capable of(l) maintaining both Salem pools below 1497F with two SFPC heat exchangers available and (2) maintaining both pools below 180'F with only one heat exchanger available. While this capability meets the requirements of UFSAR Chapter 9.1.3.1,a Technical Specifica tion change will be required because a 168-hour delay is currently required regardless of the time of year or cooling water temperatures.

This conclusion is justified because (i) the Salem Outage Risk Management Program, which includes a pre outage assessment of the SFP heat loads and heat up rates will assure available SFPC capability prior actu-Page 10 of 12

OUTSTANDING CHANGES MUST BE ATTACHED FOR WORKING COPY 20021001 I EE No.: S-C-SF-MEE-1679 Rev. No.: 0 Datc: 5/10/02 all) offloading fuel and (2) the inherent conservatisms in this calculation provide for additional cooling sources that are not credited herein. In order to maintain both pools below the required temperature limits, the SFPC heat exchangers may be required to operate in the crosstie mode (i.e. in parallel) for a period of time. as determined by the pre-outage assessment.

The evaluation does not address various practical considerations associated with pool temperatures above 140'F, namely that it is expected that humidity and moisture on the refueling floor will be high and may re quire special precautions for operations, refueling, or maintenance personnel. Long-term impact on the FHB ventilation system may also become a consideration.

5.0 REFERENCES

5.1 S-C-SF-MDC-1780, Revision 0, Capability of Salem Spent Fuel Pool Heat Exchangers to Maintain 149 0 F Pool Temperature 5.2 Phone Call with Glenn Schwartz, Salem Fuels of 5/2/02 (see Attachment E) 5.3 S-C-SF-MDC-1800, Revision 2, Decay Heat-up Rates and Curves 5.4 Phone call with Glenn Schwartz, Salem Fuels Department, on 5-3-02 (see Attachment E) 5.5 S-C-SF-MDC-1240, Revision 0, SFP Thermal-Hydraulic Calculation (HOLTEC International) 5.6 BTP ASB 9-2 Revision 2 of July 1981, USNRC Standard Review Plan 9.2.5, Ultimate Heat Sink, NUREG 0800 5.7 PSBP 301110, Revision 10, Westinghouse Instruction Manual, Auxiliary Heat Exchangers 5 8 Phone call with Kevin King, PSE&G Engineering, on 5/6/02 (see Attachment E) 5.9 LCR S02-03 5.10 NC.OM-AP.ZZ-0001, Revision 3, Outage Risk Assessment 5.11 T/S 5.6.3, Fuel Storage Capacity 6.0 EFFECTS ON OTHER TECHNICAL DOCUMENTS The following procedure changes are required upon NRC approval of the LCR:

I. The appropriate operating procedures (SI(2).OP-SO.SF-0009 &/or 0002 &/or S1(2).OP-IO.ZZ-000I) will need to be revised to ensure the maximum Service Water flow is established through the CCHXs during refueling outages when the core is offloaded into the Spent Fuel Pool, in order to minimize CC temperature and maximize SFHX heat load. This would also be applicable to the non-outage unit if the SF cross-tie were utilized. This can be accomplished by any of the following means, depending on the conditions existing at the time:

A. Set the CC temperature for the available CCHXs in accordance with SI(2).OP-SO.CC-0002 to the minimum allowable for the applicable mode. This will result in the SW flow going to the flow control setpoint of 10,000 gpm. OR B. Place the CCHX in the flow control mode. Flow will be maintained at the flow control setpoint of 10,000 gpm. OR C. If a SW header on the outage unit is removed from service rendering its CCHX unavailable, and the available CCHX control valves are gagged in position, the position should be set to ensure at least 10,000 gpm SW flow, but less than the maximum limit of 12,500 gpm.

Page IIof 12

OUTSTANDING CHANGES MUST BE ATTACHED FOR WORKING COPY 20021001 EE No.: S-C-SF-MEE-1679 Rev. No.: 0 Date: 6/14/02 7.0 SIGNATURES Page 12 of 12

OUTSTANDING CHANGES MUST BE ATTACHED FOR WORKING COPY 20021001 SFP Decay Heat 76 Assemblies with 24820 LFPH (34 months)

Days after Reactor Shutdown when Refueling Begins: 4.1667 No n An an S.D. ts P/Po Po P Elem. Bt/hr Infinite Core Fit Coeff. Fit Coeff. (days) (seconds) Power Fr. Full Power (Btu/hr) 1 0.5980 1.772E+00 6.08 5.26E+05 0 ODE+00 6.15E+07 0.OOE + 00 2 1.6500 5.774E-01 6.08 5.26E+05 O.00E + 00 6.15E+07 0.OOE + 00 3 3.1000 6.743E-02 6.08 5.26E+05 0.O0E+00 6.15E+07 0.0OE+00 4 3.8700 6.214E-03 6.08 5.26E + 05 0.OOE + 00 6.15E+07 0.OOE+00 5 2.3300 4.739E-04 6.08 5.26E + 05 7.77E-1 11 6.15E+07 4.78E-103 6 1.2900 4.81 OE-05 6.08 5.26E + 05 6.76E-1 4 6.15E÷07 4.16E-06 7 0.4620 5.344E-06 6.08 5.26E+05 1.39E-04 6.15E+07 8.57E+03 8 03280 5.716E-07 6.08 5.26E+05 1.21 E-03 6.15E+07 7.4 7E+ 04 9 0.1700 1.036E-07 6.08 5.26E+05 8.05E-04 6.1 SE+07 4.95E+04 10 0.0865 2.959E-08 6.08 5.26E+05 4.26E-04 6.15E+07 2.62E+04 11 0.1140 7.585E-10 6.08 5.26E + 05 5.70E-04 6.15E+07 3.51E+04 3.15E-03 6.15E+07 1.94E + 05 76 1.47E+07 n An an Op.Time to + ts P/Po Po P l/3Core-2Cycles Fit Coeff. Fit Coeff. (days) (seconds Power Fr. Full Power (Btu/hr) 1 0.5980 1.772E+00 1040 8.99E +07 O.OOE+00 6.15E+07 0.OOE + 00 2 1.6500 5.774E-01 1040 8.99E+07 0.0OE÷+0 6.15E+07 0.OOE+00 3 3.1000 6.743E-02 1040 8.99E+07 0.OOE+ 00 6.1SE+07 0.OOE + 00 4 3.8700 6.214E-03 1040 8.99E+07 O.OOE + D0 6.15E4 07 0.OOE + 00 5 2.3300 4.739E-04 1040 8.99E +07 O.OOE + 00 6.15E+07 O.OOE + 00 6 1.2900 4.81 OE.05 1040 8.99E+07 0.OOE + 00 6.15E+07 O.OOE + 00 7 0.4620 5.344E-06 1040 8.99E +07 5.88E-212 6.15E+07 3.62E-204 8 0.3280 5.716E-07 1040 8.99E +07 8.01 E-26 6.15E+07 4.93E-18 9 0.1700 1.036E-07 1040 8.99E+07 7.68E-08 6.15E+07 4.73E 4.00 10 0.0865 2.959E-08 1040 8.99E+07 3.03E-05 6.1SE+07 1.86E+03 11 0.1140 7.585E-10 1040 8.99E+07 5.32E-04 6.15E+07 3.28E +04 5.63E-04 6.15E+07 3.46E +04 76 2.63E+06 D.H. Rate 2.91 E-03 6.15E+07 1.79E+05 76 1.36E+07 Elements 2010 total In Pool in Pool 1/3 for 2 cycles (76 assemblies) 1.36E+07 1/3 for 3 cycles (41 assemblies) 7.43E+06 1/3 for I cycle (76 assemblies) 1.31E+07 Background 956 1412 6 BOE+06 Heavy Elements 3.03E + 06 TOTAL 4.39E+07 S-C-SF-MEE-1 679 Rev. 0 Attachment A Page Al

OUTSTANDING CHANGES MUST BE ATTACHED FOR WORKING COPY 20021001 SFP Decay Heat 76 Assemblies with 12410 FFPH (17 months)

Days after Reactor Shutdown when Refueling Begins: 4.1667 No.

n An an S.D. tS P/Po PO P Elem 8/hr Infinite Core Fit Coeff. Fit Coeff. (days) (seconds) Power Fr. Full Power (Btu/hr) 1 0.5980 1.772E + 00 6.08 5.26E+05 O.OOE + 00 6.15E+07 O.OOE + 00 2 1.6500 5.774E-01 6.08 5.26E + 05 O.OOE + 00 6.15E+07 O.OOE + 00 3 3.1000 6.743E-02 6.08 5.26E + 05 0.OOE + 00 6.15E+07 O.OOE + 00 4 3.8700 6.214E-03 6.08 5.26E+05 0.OOE + 00 6.15E+07 O.OOE + 00 5 2.3300 4.739E-04 6.08 5.26E+05 7.77E-l 11 6.15E+07 4.78E-1 03 6 1.2900 4.81 0E-05 6.08 5.26E+05 6.76E-14 6.15E+07 4.16E-06 7 0.4620 5.344E-06 608 5.26E + 05 1.39E-04 6.15E+07 8.57E + 03 8 0.3280 5.716E-07 6.08 5.26E+05 1.21E-03 6.15E+07 7.47E+04 9 0.1700 1.036E-07 6.08 5.26E+05 8.05E-04 6.15E+07 4.95E +04 10 00865 2.959E-08 6 08 5.26E + 05 4.26E-04 6.15E+07 2.62E+04 11 0.1140 7.585E-1 0 6.08 5.26E+05 5.70E-04 6.15E+07 3.51E+04 3.15E-03 6.15E+07 1.94E + 05 76 1.47E+07 n An an Op.Time to + Is P/Po Po P 1/3Core-lCycle Fit Coeff. Fit Coeff. (days) (seconds Power Fr. Full Power (Btk/hr) 0.5980 1.772E+00 523 4.52E + 07 00OE+00 6.15E+07 0.00E+00 2 1.6500 5.774E-01 523 4.52E+07 O.OOE + 00 6.15E+07 0.OOE + 00 3 3.1000 6.743E-02 523 4.52E+07 0 OOE+00 6.15E+07 O.OOE+O0 4 3.8700 6.214E-03 523 4.52E+07 O.OOE + 00 6.15E+07 OOOE + 00 5 2.3300 4.739E-04 523 4.52E+07 0O0E+00 6.15E+07 O.OOE +00 6 1.2900 4.8 10E-05 523 4.52E+07 O.OOE + 00 6.15E+07 O.OOE +00 7 0.4620 5.344E-06 523 4.52E+07 2.86E-108 6.15E+07 1.76E-1 00 8 0.3280 5.716E-07 523 4.52E +07 9.86E-1 5 6.15E+07 6.07E-07 9 0.1700 1.036E-07 523 4.52E +07 7.86E-06 6.15E+07 4 84E-02 10 0.0865 2.959E-08 523 4.52E +07 1.14E-04 6.15E+07 6.99E + 03 11 0.1140 7.585E-1 0 523 4 52E+07 5.51 E-04 6.15E+07 3.39E+04 6.72E-04 6.15E+07 4.14E+04 76 3.14E+06 D.H. Rate 2.80E-03 6.15E+07 1.72E + 05 76 1.31E+07 5-C-SF-MEE-1679 Rev. 0 Attachment A Page A2

OUTSTANDING CHANGES MUST BE ATTACHED FOR WORKING COPY 20021001 SFP Decay Heat 41 Assemblies with 37,230 EFPH (51 months)

Days after Reactor Shutdown when Refueling Begins 4.1667 No.

n An an S.D. ts P/Po Po P Elem. BVhr Infinite Core Fit Coeff. Fit Coeff (days) (seconds) Power Fr. Full Power (Btuhr) 1 0.`5980 1.772E+00 6.08 5.26E + 05 0OOE+00 6.15E+07 0 OE + 00 2 1.6500 5.774E-01 6.08 5.26E+05 0 COE + 00 6.15E+07 000E +00 3 3.1000 6.743E-02 6.08 5.26E + 05 OOOE + 00 6.15E+07 O.OOE + 00 4 38700 6.214E-03 608 5.26E + 05 O.OOE + 00 6.15E+07 O.OOE + 00 5 2.330O 4.739E-04 608 5.26E+05 7.77E-1 II 6.15E+07 4.78E-103 6 1.2900 4.8 1OE-05 608 5.261 + 05 6.76E-1 4 6 15E+07 4.1 6E-06 7 04620 5 344E-06 60B 5 26E+05 1.39E-04 6 15E+07 8.57E+ 03 8 0-3280 5 716E-07 6.08 5 26E+05 1.21 E-03 6.15E+07 7.4 7E + 04 9 0.1700 1.036E-07 6 08 5.26E + 05 8.05E-04 6.15E+07 4.95E + 04 10 00865 2.959E-08 6.08 5.26E +05 4 26E-04 6.15E+07 2.62E +04 11 0.1140 7.585E-10 6.08 5.26E+05 5.70E-04 6 15E+07 3.51E+04 3.15E-03 6.15E+07 1.94E+05 41 7.96E+06 n An an Op.Time to + is P/Po Po P 1I3Core-3 Cycles Fit Coeff Fit Coeff. (days) (seconds Power Fr. Full Power (Btu/hr) 1 05980 1.772E+00 1557 1.35E+08 O.OOE + 00 6.15E+07 O.OOE + 00 2 1.6500 5.774E-01 1557 1.35E+08 0 COE+00 6.15 E+ 07 O.OOE+00 3 3.1000 6 743E-02 1557 1.35E+08 O.OOE+00 6.15E+07 O.OOE + 00 4 3.8700 6.214E-03 1557 1.35E+08 O.OOE + 00 6.15E+07 0 COE + 00 5 2.3300 4 739E-04 1557 1.35E+08 0 OOE+00 6.15E+07 O.OOE+00 6 1.2900 4 81OE-O5 1557 1.35E+08 0.OOE +00 6.15E+07 0 COE + 00 7 0.4620 5.344E-06 1557 1.35E+08 O.OOE + 00 6 15E+07 000E +00 8 0.3280 5.716E-07 1557 1.35E+08 6.50E-37 6.15E+07 4.00E-29 9 01700 1.036E-07 1557 1.35E +08 7.51 E-1 0 6.15E+07 4.62E-02 10 00865 2.959E-08 1557 1.35E+08 8 07E-06 6 15E+07 4.97E +02 11 0.1140 7.585E-10 1557 1.35E+08 5.15E-04 6 15E+07 3 17E+04 5 23E.04 6.15E+07 3.22E+04 41 1.32E+06 D.H. Rate 2.95E-03 6.15E+07 1.81E+005 41 7.43E+06 S-C-SF-MEE-1679 Rev 0 Attachment A Page A3

OUTSTANDING CHANGES MUST BE ATTACHED FOR WORKING COPY 20021001 SFP Decay Heat Contribution of Heavy Elements U-239 and Np-239 to ts 1-EXP EXP P/Po PO Elem P U-239 2.28E-03 0.7 4.47E+07 5.26E+05 1.OOE.00 8.4E-1 13 1.33E-1 15 61526310 76 6.23E-106 N-239 2.17E-03 0.7 4.47E+07 5.26E+05 1.00E+00 0.166577 2.55E-04 61526310 76 1.19E+06 U-239 2.28E-03 0.7 8.94E+07 5.26E+05 1.OOE+00 8.4E-113 1.33E-115 61526310 76 6.23E-106 N-239 2.17E-03 0.7 8.94E+07 5.26E+05 1.OOE+00 0.166577 2.55E-04 61526310 76 1.19E+06 U-239 2.28E-03 0.7 1.34E+08 5.26E+05 1.00E+00 8.4E-113 1.33E-115 61526310 41 3.36E-106 N-239 2.17E-03 0.7 1.34E+08 5.26E+05 1.00E+00 0.166577 2.55E-04 61526310 41 6.43E+05 3.03E +06 S-C-SF-MEE-1 679 Rev. 0 Attachment A Page A4

r.oo 0c-"

0-4

• .-I 0o>

z SFP Background Heat C)

T z

IR17 1RI8 1R19 1R20 1R21 1R22 1R23 CO Unit 1 IR13 1R14 1R15 1R16 Oct-I1 Apr-13 Oct-14 Oct-02 Apr-04 Oct-05 Apr-07 Oct-08 Apr-10 Unit 1 Oct-99 Apr-01 2R19 2R20 Frl 2R14 2R15 2R16 2R117 2R18 Unit 2 2R10 2R1 1 2R12 2R13 Apr-08 Oct-10 Apr-I1 Oct-12 Apr-14 0 Oct-00 Apr-02 Oct-03 Apr-05 Oct-06 Unit 2 Apr-99 (..)

"1 rFl Btu/hr Year C:)

2.31 E+00 1999 8.46E+00 2014 Year "1 00 0

0 0

2016 C) 2014 C')

0 2012 "-a Y X 2010 2010 6 819981 I ,2008 - Year:

>- 2006 2004

, 2002

' 2000 1998 -I, 0 2 4 6 8 10 Btu/hr x E6 1 F..MCee-1675, , EP. C Attachment A Page A5

OUTSTANDING CHANGES MUST BE ATTACHED FOR WORKING COPY 20021001 Deleware River Temperature Data Weekly Averages c6mpiled into Specified Era Averages 1967 - 1997 Jan I Feb I March AprilI May JuneI July Aug I Sept Oct I Nov Dec I1967-1969 40.0 36.7 37.6 45.8 53.7 62.41 70.4 74.11 72A4 65.61 56.6 47.4 1970-1979 39.2 360 41.0 47.9 57.3 64.8 72.2 750 731 6451 562 47.0 1980-1989 33.6 34.6 41.4 52.8 63.1 72.1 77.6 76.9 72.51 60.6 497 39.9 1990-1997 350 35.4 41.2 51.6 622 709 76.8 753 70.51 6031 484 41.0 AVERAGE 37.0 35.7 40.3 49.5 59.1 67,5 74.2 75.3 72.1, 630, 527 43.8

4 ' Jan Feb I March April May June July Aug Sept Oct Nov Dec i4.. 42.3 39.51 39.1 44.8 4 1 58.71 68.4 72.6 70.7 656 567 47.9 382 33.1 361 47.3 533 61.2 71.3 746 72.8 67.3 57.7 48 V. 37.2 35.7 35.8 46.5 55.5 64.5 72 75.1 75.2 68.6 59 49.4 441 38.8 38.5 41.7 51.2 61.9 65.4 72.8 702 68.1 53.5 46.7 qý!,t -,, S Z 42.7 35 36.1 43.3 55 63.1 67.81 75.3 71.9 70.3 55 45.1

- 40.3 35.4 37.6 41.1 17.9 66.2 69.81 752 73.9 71 56.6 47.1

""43.2 40.7 37.7 43 48.6 55.1 67 72.5 71.5 66.8 59.6 49 395 37.7 35.1 45.3 51.4 58.2 692 74.6 73.6 67.9 61.3 51.7

- 38.2 36.5 358 44.1 53.6 62.5 70.4 74.5 76 70.1 61 51

-; 42.31 37.2 394 461 52.4 64.11 70.2 72.31 69 63.7 51.3 45.2 M. V3;137.7 34.7 40.3 46.8 5 1 65.5 72.9 74.71 7 65.9 53.- 43.1 37.5 35.71 39.3 48.6 602*7 68 72.5. 75' 4 ' 72.2 656 54.5 44.3 g* J. * * .fN~ 47.5 0,... 7141. 61.9

+",'*", -

37.35 63 1967-1969 371i1 736 63.3 .. 'I AVERAGE 9 3667l 37.57 4.81 53.74 62.42 70.37 74.131 72.40 66.61 56.62 47.38 S-C-SF-MEE-1679 Rev. 0 Attachment B Page B1

OUTSTANDING CHANGES MUST BE ATTACHED FOR WORKING COPY 20021001 Deleware River Temperature Data Weekly Averages compiled into Specified Era Averages 1967 - 1997 I

S-C-SF-MEE-1679 Rev. 0 Attachment B Page B2

OUTSTANDING CHANGES MUST BE ATTACHED FOR WORKING COPY 20021001 Deleware River Temperature Data Weekly Averages compiled into Specified Era Averages 1967 - 1997 Jan Feb March April May June July Aug Sept Oct Nov Dec 33

__-_37.1 35 52.3 2 77 763 60.3 46.6 40.4 S25.6 32.4 40.6 54.7 64.5 68.3 77.9 75 72.9 55.6 43.2 37

, 26.7 33.5 41 50.6 61.2 71.7 80.7 76.7 71.5 60.5 48637.3

, TI 33.3 39.9 39.7 49.8 62.8 74.1 77.8 768 69.2 60.2 46.3 34.8 31.8 406 6.3 8 68.4 78.1 8 72.5 63.5 51.1 44.8 32A4 3461 43.1 52.3 65.4 71 77.1 76370.9 60.7 49.441.6

.',. 37.7 359 36 542 583 74.2 79 76.7 72.4 59.8 52.6 41.5

. 30.7 40.5 391 51.1 5187.5 79.5 78.7 70.2 59 46.5 42.7 36 32.2 46.3 53.4 66.7 70.8 77.6 83.9 73.8 61.9 544 42.3 37.7 36 45.3 46 67.4 71.6 73.8 76.7 70.3 67.2 53 37.7 303 31.4 40.8 51.8 63.274 7. 74.3 74.7 72.7 749 69.4 6.9 61.6 66 5 56.6424 47.1 4.

345.8 333, 363. 44.96 75 42.4 32.5 39 43.4 62.6 72.9 76 77 62.1 49.5 43.2 lI* " I 43.5 32.5 41.6 50.5 64 758 73.1 75.9 72.8 67.2 47.1 41.4 33.9 339 42.9 52.7 60.9 67.8 73.7 74.5 73.1 70 50.6 41

' 38.2 35.1 41.B 49.1 63.9 73.3 76.6 777 70.9 66.3 49.4 37.3 36.9 37 37.9 51.2 59.573.1 73 802 76 67.5 54.735.8

"", - 367 37.3 42.4 50.9 64.5 74.9 78 76.4 73.9 661 49.5 31

- 35.3 33 50.6 60 686 74.7 80.1 776 635 50.2 L 26.7 305 41.2 54.6 588 698 77.7 76.6 71.8 57.7 52.3 41.5

- 32.6 30.5 36.2 43.9 605 68.8 77.5 79 71.9 65.8 53.7 49.3

.. 38.1 31.5 37.7 46.7 57.3 67.3 75.9 7851 74.6 60.2 56.3 43.8 S38.6 29.8 43.9 51.4 63.9 706 75.7 77.4 761 64.3 54.3 44.1

_ t',

i 1 32.7 33 345 51.7 60.6 72.3 75.9 78.7 69.8 65.1 54.3 44.1 37.4 33.2 367 51.6 59 71.8 78 9 79.3 72.4 62 53.3 43.5 32.1 34.2 40 51.9 56.5 67.1 76.3 83 76.3 62.7 51.4 45

, '353 381 36 51 57.679.1 5 61.9 56.2 39.5 35.8 32.5 43 55.3 661 71 78.5 76.4 71.1 57.7 42.2 32.4

, -,, 28.5 42.6 44 54.9 67. 74.31 76.9 72.5 64.7 54.6 42.9 34.9 24.9 s..

35 .. 44.4 54.4780.8 6. 7. 72.5 97.3 549 47.1 43.4

, , 304 40.1 41 51 65.7 74.6 77.3 76.1 68.9 62.4 45 42.7 289 34.g5 41.8 47.4 61.3 69.3 71.9 78.678 74.9 71 69.3 6 A 60.9 58.7 46 492 34.2 3 .

3.9 33.41 48.7 535 70.2 73.4 7. 1 6. .. 6 3.

.2 35.1 47.4 564 68.5 75.8 81.6 74.3 70.1 57.5 45.5 41.2 32.1 37.1 47.1 51.7 74.6 798 765 69.5 56.5 47.9 38.1 364 36.3 47.1 55 6 79.9 77 69.9 58.3 44.3 26.8 32 58.4 "T 78.7

• 52.2 1 all. 7

---

57.5 76.1 *t 54.2 54.5 lj 27.1 57.6 *, 80.1 31.1 54._ 76.4 77.7 63.2 56.___66i.

29.6 62 32.3i 31.6-- 57.4 5--- ý, 1.'1 r -Ilk 80.--

79.2 ' 55.1 57.2  :

33.5 2 8173 52.7 1980-1989 36.9 5794 6 . 4 77 0 5 7 78 77 .56 , 76829 7 6 8 1 2 9 6058.5 55 4 9 68 1 99 A VE RA G E 3 3 .5 8 1 3 . 6 4 3 1 5 2 .7 9 1 6 S-C-SF-MEE-1679 Rev. 0 Attachment B Page 83

OUTSTANDING CHANGES MUST BE ATTACHED FOR WORKING COPY 20021001 Deleware River Temperature Data Weekly Averages compiled into Specified Era Averages 1967 - 1997 L

S-C-SF-MEE-1679 Rev. 0 Attachment B Page B4

N)O o c 0-I I 09 2z z

C)

C)

Delaware River Temperature Average 1967 to 1997 0 z

MU) 80 C:

a)

--4

--4 70 "t

0

,-1 60 0

M

.-_ 50 o

I

,U Ma 0

L. 40

-- 1967-1997 Q 30 20 10 0

e Ip Ip" 0

'0

'0 S-C-SF-MEE-1679 Rev. 0 Attachment B Page B5

N 0 N,)

0> U),

K)U C?0 0

Delaware River Average Temperatures for Past 20 years (1977 - 1997) z C)

U) 90

-. i CU 80 -I 0

-r

.-r 70 0

60 0

.CV r 50 0 I.. 4 40 71 --

1977 1979 1980 - 1989 0

0-

-- 1990- 1997 01 30 20 10.

0.

(C' ,4p

\S.O 0 S-C-SF-MEE-1679 Rev. 0 Attachment B Page B6

Ko oc 0-4 POW9 Cýz z

C?

G) 0 Delaware Water Temps by Decade 1967 - 1997

--4 90 0

!~ ~-

-'V!V 0Z M

80 w 4

- 4r~M4,,4~ip!

--I

-0 71 70 "'1 60 0

0t

'4. -o L, 50 -- 1967-1969 C3 -U-1970 -1979 (A --- 1980 -1989 40

-X-1990-1997 30 20 10 0

0 40 0

S-C-SF-MEE-1679 Rev. 0 Attachment B Page 87

K)O 0 C O-4 0--i oz K)W Oz 2z z

c) 0 z

Comparison of Reedy Island and PSEG Maximum (-)

C:

-4 rri w

Annual Temperatures --4

.-4 rrl 0

Maximum Annual Actual Maximum Annual Temperature Difference

-1 Temperature at Reedy Temperature from Between HC.A2438 and Year Island (7F) HC.A2438 Reedy Island 0 M

70 78.7 81.8 3.1 1992 zC) 1993 85.5 2.8 0

19937 -U 81.7 86.4 4.7 1994 84.6 85.4 0.8 1995 80.4 82.0 1.6 1996 79.7 84.6 4.9 1997 81.3 84.3 3 Average "5-4-- i F-,

Attachment B 79 e- /6PG6V. 0 Page BS

OUTSTANDING CHANGES MUST BE ATTACHED FOR WORKING COPY 20021001 HTC-STX Version 3 6 Time. 9.01:16 AM Date: 4/30/02 File Spfchx

"- Main. English units I Job No Item No. EVALUATION Case 2 Case Desciption SFPHX 3 TEMA Type BEU - HORZ SheIA/Unit 1 Conn In I Series I Parallel 4 Size: 33 500 in Dia 146.3 in Tube Length in Kettle Dia 5 Surface/Shell ft2 2,353.2 Gross 2.319 3 Eff 151 U-Bend Area 6 Surface/Anit ft1 2,353.2 Gross 2,319 3 Eff 151 U-Bend Area Performance of One Unit SHELLSIDE TUBESIDE 7 Fluid Circulated SFPHX" 8 Total Fluid In lb/hr 1,490.000 0 1,140,000.0 9 Vapor lb/hr 0.0 0 0 10 Liquid lb/hr 1.490.000 0 1,140,0000 11 Fluid Vap'z/Cond lb/hr 00 00 12 Density In/Out Ib/ft. 62 050161 946 61 729161.841 13 Spec. Heat Vap/Lig Btu/ib-F 0.00010 997 0 00010.997 14 Viscosity Vap/Lig cP 000010692 000010.588 15 Therm Cond Vap/Lig Btu/hr-ft-F 0 000 / 0 364 0 00010.370 16 Temperature In/Out 'F 95.01103.0 120.0/109 54 17 OperatLng Pressure (Abs) psi 75 000 50.000 18 Press. Drop Allow/Calc psi 9 000110.981 15.000/18.933 19 Number of Passes/Shell 1 4 20 Velocty, Average f/sec 407 961 21 Film Coef Btu/hr-ft;-F 191281 225637 22 Fouling Resist hr-ft2-F/Btu 0 000500 0 000575 23 Heat Duty 11,883.525 Btu/hr MTD/Wtd/CoI" 14.81 *F F-CORR 0 941 24 Transfer Rate 345.99 Serv 37211 Calc 655.32 Clean 000136 Foul Constructionof One Shell 25 TEMA Shell Type E Rear End Type U.T.

26 TubeType PLAIN Bundle Dia in 32.50 27 Tube 0 D in 0.750 No. Holes/TubeSheet 920 28 Tube I D in 0 652 No. Holes Counted 29 Area Ratio 1.150 Tube Pitch In 0.9375 30 Tube Length Total ft 12.19 Tube Layout Angle 30 31 Tube Length Effective ft 1200 Impingement Plate - NO 32 Baffle Type VERT-DBL.SEG Crosspasses/S hell 9 33 Baffle Cut Frac Dia/NFA 0 160/0 200 Central Spacing in 16558 34 Window Area in, 94 9941 In/Out Spacing in 23 914.2 35 Seal Strips YES Drop Under Noz In/Out in 1.711.7 Shell Nozzles Inlet Outlet Tube Nozzles Inlet Outlet 36 Inside Dia, in 1000 1000 Inside Dia, in 10.00 1000 37 Velocity Wsec 1223 12,25 Velocity Wsec 9.41 9.39 38 Rho-V-Sqr lb/ft-secO 9280 9296 Rho-V-Sqr Ib/ft-sec' 5461 5451 39 Nozzles]Shell (OPP. SIDE) 1 1 Shellside Performance Pressure Drop 40 Bundle Flow Fraction 0.761 Shell Cross/Wind 4 37814 408 41 Mass Vel Cross/Vind 252.11627.4 Tubes 17.750 42 Mass Vel LonglMean 125 5/397.7 Nozzles Shell/Tube 2 195/1 183 Bundle Diameter Clearances Tube Metal Temperatures 43 Bundle-Shell in 1 00000 Avo Tube Metal Temo 1068 44 Baffle-Shell in 0 18750 Shellside Avg Surf Temp "F 102 0 45 Tube-Baffle In 003625 Tubeside Av* Surf Tern *F 111 6 46 Baffle Thk. in 0.313 1

-rýF-m

-- e-. 16 ?7 1~ gcv Attachment C Page C I

OUTSTANDING CHANGES MUST BE ATTACHED FOR WORKING COPY 20021001 HTC-STX Version 3 6 Time 90116 AM Date 4030/02 File Spfchx Summary English units Item No Service SFPHX Calculation Mode Evaluation Case Size 34 x 146 Type BEU - HORZ Connectons 1 Senes 1 Parallel SurfacerUnid 2,319 Shellstunit 1 SurflShelt 2,31935 CostiUrmt 42,668 Cost/Surf 184 Weight!Shell 9,531 Heat Duty 11,883,525 MTD 14 81 F-corr 0 9409 Rate-Service 345 99 Calculated 372 11 Calc Fouling 0 00136 Shell Tubes Tubes 0 750 x 0 049 on 0 9375 30 deg FRow Rate 1490000 1140000 Tube No 920 Type PLAIN Temperature In 950 120 0 Baffies. VERT DBL-SEG 16 6 space 20 0 cut Temperature Out 1030 1095 Pressure Drop 10981 18933 Surface Area OK Overdesilgnby 7.55%

Velocity 4.066 9611 Shell pressure Drop -Allowable exceeded Passes 1 4 Tube Pressure Drop Allowable exceeded Film Coef 19128 22564 Vibrabon Tube vibration [ikely.

Nozzle In 1 x 10 0 100 Shell Nozzles Rho-V-Sgr exceeds 4000 Nozzle Out 1 x 100 100 Chan Nozzles OK. Rho-V-Sqr witNn 6000 Attachment C Page C2

OUTSTANDING CHANGES MUST BE ATTACHED FOR WORKING COPY 20021001r: -_

?17 HK ITEM NO. A~-:~

i 7v -.

m "A ly. I L PER SUN1 z -zo 5 PERFORMA4NCE (PZER UNI-1 SKM:.L StOC Tust slot

• t rA A s-u I I iI'.

LIOU;C L.5M CSNO,~,-NocmsIqL--3 jJMR STEAN LZ/KN I 1 i = S -Z-vTY- LU IO C e v r r tV

$ WC.WVApc"t wEJL1R pVgSLut!, VT /EC. P~

P ~ ~~~it~AmS'rrR ftN XSERVICErt/ ~ ~ FJIUP~

Sosr-~uN "-*S . P v

-9r k o. -V o s.:i- I IN24 N Y nI m aZI t.- fo I

- ~ ~oýrSkL6. /F e2-- o r- ( a VN.ýu.

2 .C~cu: Z3 V~S v5 ~ ~T '14 o"7C CL.L= / =ST. SPE.C.

I PCfl? :t6;: tl'E =1ERA OOM..LL pUL PANairm I M ACc L!ASU IN. '7 l.z. LANGjm SAFPI.E I C

~ANOVR1~A~J~~3tI 7-

• A- =V- R 1%.3LC ju Attachmnent C 1age C3 P

- - t

OUTSTANDING CHANGES MUST BE ATTACHED FOR WORKING COPY 20021001 HTC-STX Version 3 6 Time. 2 12 11 PM Date 611212002 File: sfphx-71ccw$%-158

"* Summary English units Item No Service SFPHX-71CCnfeat Calculation Mode Ratng Case Size 34 x 146 Type BEU - HORZ Connechons I Series 1 Parallel Surface/Unit 2,179 Shellsunit I Surf/Shell 2,179 24 Cost/Unit 41,206 Cost/Surf 1891 Weight/Shell 9,238 Heat Duty 43.968.300 MTD 64 41 F-corr 09456 Rate-Servce 370.84 Calculated 35637 Calec Foulia 000105 Shell Tubes Tubes 0 750 x 0 049 on 0 9375 30 dog Raw Rate 1501800 1222716 Tube No 866 Type PLAIN Temperature In 71.0 1609 Baffies" VERT DBL-SEG 16 5 space 22 0 cut Temperature Out 100 1 1254 Pressure Drop 10413 4458 Surface Area "Under design by -3 91%

Velocity 4215 5642 Shell pressure Drop Allowable exceeded Passes 1 2 Tube Pressure Drop OK Within allowable Film Coef 2047,0 17305 Vibration -Tube vibration likely.

Nozzle In 1 x 10 0 100 Shell Nozzles - Rho-V-Sqr exc:eeds 4000 Nozzle Out xI00, 100 Chan Nozzles - Rho-V.Sr e0<eeds 6000 Attachment C Page C4

OUTSTANDING CHANGES MUST BE ATTACHED FOR WORKING COPY 20021001 HTC-STX Version 3.6 Time 244"21 PM Date 6/122002 Fie

' Summary** English units Item No Service SFPHX-71CCinlet Calculation Mode Rating Case Size 34 x 146 Type BEU - HORZ Connections I Series I Parallel Surface/Unit 2,179 Shet/unit 1 Surf/Shell 2,17924 CostAJnIt 41,206 Cost/Surf 18 91 Weight/Shell 9,236 Heat Duty 21,997,584 MTD 3263 F-or 0,9371 Rate-Service 30938 CalcuL0ted,0003 Shell Tub.. Tubes 0 750x0049on 0 9375 30 deg Flow Rate 1501800 742640 Tube No 888 Type, PLAIN Temperatureln 710 1284 Baffles. VERTDBL-SEG 165space 220cut Temperature Out 856 98.9 Pressure Drop 10 441 1 783 Surface Area "Under design by -3 97%

Velocity 4.185 3349 Shell pressure Drop "Allowable ecceeaded Passes 1 2 Tube Pressure Drop OK Within allowable Film Ccef 19442 9634 Vlbrabon - Tube vlbration likely Nozzle In 1x100 100 She, Nozzles - Rho-V-Sr exceeds 4000 Nozzle Out 1x100 100 ChanNozzles OK. Rho-V-Sqr within 6000 Attachment C Page CS

OUTSTANDING CHANGES MUST BE ATTACHED FOR WORKING COPY 20021001 HIC-STX Version 3 6 Time: 1 55 50 PM Date 6112/2002 File:

"-" Summary English units Item No Service SFPHX-71CCInlet Calculaton Mode Rating Case Size 34 x 146 Type BEU - HORZ Connections I Series I Parallel Surface/Unit 2,179 Sheas/unit 1 Surf/Shell 2,17924 Cost/Unil 41.206 Cost/Surf 1891 Weight/Shell 9,236 Heat Duty 42,971,104 MTD 54 37 F-corr 0 9478 Rate-Service 36265 Calculated 35601 C F(Wllno imc 000111 Shell Tub", Tubes 0 750 x 0 049 on 093 75 30 dog Flow Rate 1501800 1222716 Tube No 868 Type, PLAIN Temperature In 71 0 1600 Baffles: VERT DBL-SEG 165 space 220 cul Temperature Out 995 1253 Pressure Drop 10412 4.459 Surface Area OK Over design by -1 84%

Velocity 4214 5640 Shell pressure Drop "* Allowable exceeded Passes 1 2 Tube Pressure Drop OK Within allowable Film Coef 20435 17257 Vibration - Tube vibrabon Ukely Nozzle In 1x100 100 Shell Nozzles - Rho-V-Sqr exceds 4000 Nozzle Out 1 x 100 100 Chan Nozzles Rho-V-Sar exceeds 6000 Attachment C Page C6

OUTSTANDING CHANGES MUST BE ATTACHED FOR WORKING COPY 20021001 HTC-STX Version 3 6 Time* 1,49 54 PM Date 611212002 File: sfphx-7lccwS%-170

"- Summary Engllsh units Item No Se'vIce SFPHX-71CCIn!et Calculation Mode Rabna Case Size 34 x 146 Type BEU - HORZ Connections I Series I Parallel SurfaceAUnit 2,179 Shells/unit I Surf/Shell 2,17924 Cost/Unit 41,206 Cost/Surf 18 91 Weight/Shell 9,236 Heat Duty 47,964,360 MTO 6030 F-corr 09471 Rate-Servce 36501 380.3

_asculawd Galc Fv flng 0,0011.1 Shell Tub". Tubes 0.750 x 0.049 on 0-9375 30 deg Flow Rate 1501800 1217710 Tube No 866 Type" PLAIN Temperature In 710 1700 !affles- VERT DBL-SEG 165 swace 220 cut Temperature Out 1027 1312 Pressure Drop 10407 4410 Surface Area OK Over design by -1 3%

Velocity 4221 5652 Shell pressure Drop - Allowable exceeded Passes 1 2 Tube Pressure Drop OK Within alTowable.

Film Coef 20659 17963 Vlbraton -Tube vibration likely Nozzle In I x 10 0 100 Shell Nozzles 'Rho-V-Spr exceeds 4000 Nozzle Out 1x100 100 Chan Nozzles *Rho-V-Sqr exceeds 6000

' " - f F-A,'fe -4[67f ,E---49' I'

Attachment C Page C7

OUTSTANDING CHANGES MUST BE ATTACHED FOR WORKING COPY 20021001 HTC-STX Version 3 6 Time 1 43 40 PM Date. 611212002 File sfphx-71ccw6%

Summary English units Item No Service SFPHX-71CCinlet Calculation Mode Rating Case Size 34 *A146 Type BEU - HORZ Connections 1 Series 1 Paraliel Surface/Unit 2,179 Shelslunlit 1 Surf/Shell 2,17924 Cost/Unit 41206 Cost/Surf 1891 Welght/Shell 9,236 Heat Duty 53.118176 MTD 66 10 F-con' 09461 Rate-Service 368.76 Calculad 384 51 cac F3Mlin 000113 Shell Tube" Tubes 0 750 x 0 049 on 0 9375 30 de Flow Rate 1501800 1213955 Tube No 866 Tpeo PLAIN Temperature In 710 1800 Baffles, VERT DBL-SEG 16 5 space 22 0 cut Temperature Out 106 1 137.0 Pressure Drop 10402 4372 Surface Area OK Over design by-1 16%

Velocity 4 228 5671 Shelf pressure Drop " Allowable exceeded Passes 1 2 Tube Pressure Drop OK Within allowable Rim Coef 2088 7 1873 2 Vibration

• Tube vibration likely Nozzle In I x 10 0 100 Shell Nozzles - Rho-V-Scr etcceeds 4000 Nozzle Out I x 100 100 Chan Nozzles - Rho-V-Sqr eseeds W000 Attachment C Page C8

OUTSTANDING CHANGES MUST BE ATTACHED FOR WORKING COPY 20021001 HTC-STX Version 3 6 Time. 4.01.40 PM Date 6112/2002 File. sfphx-71ccw6%-135

- Summary " Engltsh units Item No Service SFPHX-71 CCinlet Calculaton Mode Rafina Case Size 34 x 146 Type BEU - HORZ Connectons 1 Series I Parallel Surface/Unit 2,179 Shelslunit I Surf/Shell 2,179,24 Cost/Unit 41,206 Cost/Surf 1891 Wesght/Shell 9,236 Heat Duty 30,407,950 MTD 3957 F-corr 09505 Rate-Service 35263 Celculated 34449 CaIc Foulino 000109 Shell Tubes Tubes 0750x0 049on0 9375 30deg Flow Rate 1501800 1232728 Tube No 868 Type PLAIN Temperature In 710 1350 Baffles- VERT DBL-SEG 16 5 space 22 0 cut Temperature Out 91 2 1105 Pressure Drop 10428 4576 Surface Area Under design by-2 31%

Velocity 4 197 5598 Shell pressure Drop "Allowable exceeded Passes 1 2 Tube Pressure Drop OK. Within allowable Film Coef 19869 1547.3 VIbrato *"Tube vibraorn likely Nozzle In Ix100 100 Shell Nozzles - Rho-V-Sqr exceeds 4000 Nozzle Out 1 x 10,0 10.0 Chan Nozzles -- Rho-V-Sqr exceeds 6000 1w-c - 5F- 1EC~-h57? jE' Attachment C Page C9

OUTSTANDING CHANGES MUST BE ATTACHED FOR WORKING COPY 20021001 HTC-STX Version 3.6 Time: 2.1958 PM Date 6112/2002 File. sfphx-71ccw6%-149 ISummary ' English units Item No Service SFPHX-71CCinlet Calculaton Mode Rating Case SIze 34 x 146 TYPe BEU - HORZ Connsctons 1 Series I Parallel SwufacelUnit 2,179 Shells/unit 1 Surf/Shell 2,17924 Cost/Unit 41206 Cost/Surf 1891 Walght/Shell 91236 Heat Duty 29,943,250 MTD 44.18 F-corr 0 933 Rale-Serwce 31102 Calculated 30652 Calc Foulina 000111 Shell Tubes Tubes 0 750 x 0 049 on 0 9375 30 dec Flow Rate 1501800 736633 Tube No 866 Type PLAIN Tamperature In 710 1490 Baffles" VERT DBL-SEG 165 SPace 22 0 cut Temperature Out 90.9 1087 Pressure Drop 10427 1 738 Surface Area OK. Over design by -1.45%

Velocity 4.196 3382 Shell pressure Drop " Allowable exceeded Passes 1 2 Tube Pressure Drop OK Within allowable Film Coea 19795 1043 9 Vibrabon -"Tube vtbration rikely Nozzle In I x 100 100 Shell Nozzles " Rho-V-Sqr exceeds 4000 Nozzle Out I x 10 0 100 Charn Nozzles OK. Rho-V-Sqr wlthIn 6000 Attachment C Page C 10

o) 0 C 0 0-1 NW)A oz z

(")

"0 z

Btu/hr E6 Temp SFPC HX Capability vs. Pool Temp (CCW at 70F) C:

V) 220 125 26.1 135 Cu 32.1 149 200 --I

-.4 22.0 115 0 32.0 135 "C-44.0 158 180 P12 50.1 170 54.7 180 -rl 0

160 LL 0 P 140 I-'-TemPl 00-CL E 120 I

100 80 60 20.0 25.0 30.0 35.0 40.0 45.0 50.0 55.0 60.0 Btulhr x E6 t-C-f'P..A7- 1 6 79f ,4e&,.

Attachment C Page CIl

OUTSTANDING CHANGES MUST BE ATTACHED FOR WORKING COPY 20021001 Temp Btu/hr(E6) Evaporative Heat Losses 149 0 86 165 1.579 179 3.64 10 180 3.87 9 190 6.3 200 9.41 8 7

w 6 x

L5 I -~-Btulhr(E6);

S4

• 3 2

1 0

0 50 100 150 200 250 Degrees F S-C-SF-MEE-1679 Rev. 0 Attachment D Page Dl

OUTSTANDING CHANGES MUST BE ATTACHED FOR WORKING COPY 20021001 DOCUMENTED TELEPHONE CONVERSATION Reference 5.2 Date: 5/2/02 From: Glen Schwartz, PSEG Fuels To: Ted DelGaizo, MLEA Inc.

Subject:

Future Refueling Plans I. Based on current projections, Salem Station will replace 76 spent fuel assemblies during upcoming refueling outages. Consequently, at the end of each cycle, the core would contain the following types of assemblies:

76 assemblies with 1 operating cycle 76 assemblies with 2 operating cycles 41 assemblies with 3 operating cycles 193 total assemblies S-C-SF-MEE-1679 Rev. 0 ATTACHMENT E Page E1

OUTSTANDING CHANGES MUST BE ATTACHED FOR WORKING COPY 20021001 DOCUMENTED TELEPHONE CONVERSATION Reference 5.4 Date: 5/3/02 From: Glenn Schwartz, PSEG Fuels To: Ted DelGaizo, MLEA Inc.

Subject:

Spent Fuel Pool Information

1. There are currently 920 fuel assemblies in the Unit I pool as of 1R14 (April 2001) and 812 elements in the Unit 2 pool as of2Rl2 (April 2002).

2. Refueling was performed during the recent past Salem outages as shown below:

Off-Load Started Off-Load Complete Re-Load Started 1R13 9/28/99 at 1855 10/1/99 at 0607 10/8/99 at 0411 IR14 4/14/01 at 1508 4/16/01 at 2044 4/26/01 at 1811 2R10 4/14/99 at 0527 4/16/99 at 1549 4/28/99 at 1930 2R11 10/16/00 at 0104 10/18/00 at 0616 10/24/00 at 0807 S-C-SF-MEE-1679 Rev. 0 ATTACHMENT E Page E2

OUTSTANDING CHANGES MUST BE ATTACHED FOR WORKING COPY 20021001 DOCUMENTED TELEPHONE CONVERSATION Reference 5.8 Date: 5/6/02 From: Kevin King, PSEG Engineering To: Ted DelGaizo, MLEA Inc.

Subject:

CCW Temperatures with Shutdown Conditions Question: Based upon shutdown conditions with Service Water inlet temperature at 66 0 F and approximately 4 x 107 Btu/hr of heat duty, what is the CCW outlet temperature according to the ProtoFlo model of the CCW system.

Answer: With on SW/CCW heat exchanger in operation, the CCW outlet temperature is approximately 7°F higher than the inlet SW temperature. If both CCW heat exchangers are operating and sharing the heat duty, the CCW temperature is approximately 30 F higher than SW temperature.

S-C-SF-MEE-l1679 Rev. 0 ATTACHMENT E Page E3

OUTSTANDING CHANGES MUST BE ATTACHED FOR WORKING COPY 20021001 Ted DelGaizo From: King, Kevin C. [Kevin.King@pseg.com]

Sent: Wednesday, May 15, 2002 5:32 PM To: Ted DelGaizo (E-mail)

Subject:

CC temperature confirmation Ted I ran my P-Flo model, and got the following results with I and 2 SFHXs. For both cases, SW temp = 660F, SW flow 10000 gpm, CC flow to SFHX = 3000 gpm.

1 1 SFHX (Q - 44 MBtu/hr):

SFP flow S2500 gpm 0

SFP temp -61.8 0 F CC temp - 69.3 F 2 SFHXs (0 = 22 MBtu/hr per hx):

SFP flow = 1740 gpm 0

SFP temp = 121.0 F CC temp 67.79F Thus your assumption for 70OF CC temp is valid (and slightly conservative).

Kevin

!r-Z-F-Mef-A~ 16 7 q XF(/. 0 1

OUTSTANDING CHANGES MUST BE ATTACHED FOR WORKING COPY 20021001 S-C-SF-MEE-1679 Rev. 0 Attachment F CC Temperature Assumption Validation Preparer: Kevin King Date: 5/16/02 Reviewer: Ted Delgaizo Date: 5/16/02

1.0 PURPOSE

To determine the CC inlet temperature to the SFHX (CC supply temperature) based on the SFP heat load and SW temperature requirements specified in Section 2.

2.0 INPUTS/ASSUMPTIONS:

2.1 SW temperature = 66'F [= 630 (Reedy Island historical data) + 30 (Reedy Island to plant intake) - Calc, Section 3.6]

2.2 SW flow to CCHXs = 10000 gpm (max allowable flow). For the plate CCHX (#12),

this is 5000 gpm per each half.

2.3 CC flow to SFHX = 3000 gpm (Calc, Section 3.2.3) 2.4 The tube and shell CCHX (#11) is assumed to be 2% plugged (Reference 3.2, Section 3.3.6). No. tubes = 3400*0.98 = 3332; Surface area = 16954

• 0.98 = 16615 ft2 .

2.5 The SFHX is modeled as a fixed heat load. The required SFHX heat load from Calculation Section 3.6 is 44 MBtu/hr (1 SFHX aligned) and 22 MBtu/hr (2 SFHXs aligned). For the two SFHX condition, it is assumed that the total SFP heat load is split equally between the two SFHXs.

3.0 REFERENCES

3.1 S-I -CC-MDC-1 788, Rev. 0, Component Cooling System Thermal-Hydraulic Model (Unit 1) 3.2 S-1-CC-MDC-1817, Rev. 2, Component Cooling System Thermal-Hydraulic Analysis - Unit 1 3.3 S-C-CC-MDC-1798, Rev. 2, Component Cooling System Heat Exchangers 3.4 Procedure S1.OP-SO.RHR-0001, Rev. 14. Initiating RHR Page 1 of 6

OUTSTANDING CHANGES MUST BE ATTACHED FOR WORKING COPY 20021001 S-C-SF-MEE-1679 Rev. 0 Attachment F CC Temperature Assumption Validation Preparer: Kevin King Date: 5/16/02 Reviewer: Ted Delgaizo Date: 5/16/02

4.0 METHODOLOGY

The Unit I CC Thermal-Hydraulic Model developed per Reference 3.1 will be used for this analysis. The default model database "S 1CCRO.dbd" from Reference 3.1 will be the baseline database. A new working database "SI CCRO - Refueling.pdb" will be created for this analysis, and will be saved as default database "S ICCRO - Refueling.dbd".

Approach:

1. Set the CC model alignment to match actual field conditions.

2. Input the known parameters from Section 2.0 into the model.

3. Run model, and determine the CC System supply temperature (SFHX inlet)*.

"*12 CCHX modeling:

The 12 CCHX is a plate type heat exchanger. It is modeled in Proto-Flo as a UA-counter flow type heat exchanger since the current version of Proto-Flo cannot plate type heat exchangers. That is, a fixed U value is inputted into the model. This requires a trial and error solution within Step 3 above to determine U, using the plate CCHX model developed per Reference 3.3, as follows:

1. Perform an initial run of the system model to determine the CC flows to each half of the plate CCHX.

2. Input the CC flows determined from above, SW flow (5000 gpm per half), SW inlet temperature (66'F) and an initial estimate of the CC inlet temperature into the plate CCHX model.

3. Run the plate CCHX model to determine the U values.

4. Input the U values into the system model.

5. Run system model.

6. Repeat until U values and CC inlet temperatures agree.

Page 2 of 6

OUTSTANDING CHANGES MUST BE ATTACHED FOR WORKING COPY 20021001 S-C-SF-MEE-1679 Rev. 0 __I Attachment F CC Temperature Assumption Validation Preparer: Kevin King Date: 5/16/02 Reviewer: Ted Delgaizo Date: 5/16/02

5.0 ANALYSIS

Discussion This analysis will use the CC System Thermal-Hydraulic Model, which will perform a thermal balance between the CCHXs and the SFHX. The CC system temperatures are determined by Proto-Flo as a result of this thermal balancing. By setting the SW flow to the CCHXs to the maximum value of 10000 gpm, the resultant CC supply temperature (CCHX CC outlet temperature) represents the minimum temperature for a given heat load and SW temperature. Thus if the CC supply temperature is set in the field at a value less than this, the setpoint value could not be maintained as the flow controls would limit SW flow to 10000 gpm.

System Alignment The Normal Operations alignment from the default model database "S lccr0.dbd", which has two pumps aligned to the entire system, except the RHRHXs, is modified as follows:

1. The BAE Package is isolated by closing valve 1CC48. This is in accordance with Reference 3.4, which isolates the BAE Package prior to initiating RHR.

2. Letdown HX (LDHX) temperature control valve ICC71 is closed, as letdown is isolated during shutdown modes.

3. The containment isolation valves are closed, as the containment loads are isolated during shutdown modes. This includes: 1CC 113 & 1CC215 (Excess LDHX); ICC 117, 1CC 118, 1CC]31, 1CC136, 1CC187 & ICC190 (RCPs)

4. The RHRHX isolation valves (1 1&12 CC16) remain closed as RHR is not required after a full core offload.

5. Flow to the SFHX is set to 3000 gpm by establishing throttle valve 1CC37 as the flow balancing parameter.

6. With the above valve alignments, only one CC pump is required - 13 CC Pump is selected. Since flow to the SFHX is being set to a specific value, the pump curve to be used is not critical - the "benchmark" curve is selected.

7. All heat exchanger heat loads are set to 0, except the CCHXs and SFHX. The parameters for these HXs (flows, temperatures, 12 CCHX Us) are inputted.

Page 3 of 6

OUTSTANDING CHANGES MUST BE ATTACHED FOR WORKING COPY 20021001 S-C-SF-MEE-1679 Rev. 0 Attachnient F CC Temperature Assumption Validation Preparer: Kevin King Date: 5/16/02 Reviewer: Ted Delgaizo Date: 5/16/02 Results Cases were run with both one and two SFHXs and with both one and two CCHXs. Since Unit 1 has one tube and shell CCHX and one plate type CCHX, separate cases were run with each individual CCHX. A summary of the pertinent results are included below. The complete Proto-Flo reports are saved as report files, and are included on the disk included with this evaluation. The 12 CCHX spreadsheet model results are included on pages 5 and 6 of this attachment.

Case CCHXs # SFHXs QSFHX CC supply (MBtu/hr) temperature ('F) 1 11 & 12 1 44 69.3 2 11 1 44 75.0 3 12 1 44 74.5 4 11 & 12 2 22 67.7 5 11 2 22 70.7 6 12 2 22 70.6

6.0 CONCLUSION

The minimum CC supply temperature with a SFP heat load of 44 MBtu/hr and a SW temperature of 66°F is as follows:

1. CCHXs # SFHXs CC supply temperature (IF) 2 1 69.3 1 1 75.0 2 2 67.7 1 2 70.7 Page 4 of 6

"I~0 0

0-4 S-C-SF-MEE-1679, Rev. 0 Attachment F z 0

z EXCEL Spreadsheet for 12 CCHX Evaluation - EOL Full Core SFP Discharge "r

z 11 & 12 CCHXs; 1 SFHX 11 & 12 CCHXs: 2 SFHXs o u}

A half B half Total A half B half Total C:

SW CC SW CC SW CC SW CC

• rr Inlet temp (°F) 66 00 91.37 66.00 91.37 66.00 78.80 66.00 78.80 Outlet temp (*F) 70 61 69.44 70.64 69.47 68.32 67.82 68.33 67.84 Mass Flow (lb,/hr) 2,521,178 530,990 2,521,178 535,008 2,521,178 533,600 2,521,178 537,124 -)

M Volumetric Flow (gpm) 5000 1057 5000 1065 5000 1060 5000 1067 M

-4 Fouling (hr-f--FIBtu) 0.001000 0.001000 0.001000 0.001000 Properties:

0 Tavg (°F) 68.31 80.40 68.32 80.42 67.16 73.31 67.17 73.32 ";U C')

Density @ TI (Ibd,1i9) 62.86 62.63 62.86 62.63 62.86 62.76 62.86 62.76 0

Density @ Tav (lb/ft3 ) 62.84 62.21 62.84 62.21 62.85 62.27 62.85 62.27 Cp (Btu/Ib,*-°F) 1.0008 0.9998 1.0008 0.9998 1.0006 1.0003 1.0006 1.0003  ;)

k (Btulhr-ft-°F) 0.3483 0.3547 0.3484 0.3547 0.3478 0.3515 0.3478 0.3515 C)

Dynamic visc (lbJft-hr) 2.473 2.064 2.473 2.064 2.512 2.260 2.512 2.259 0

-u Kinematic visc (ft2/s) 1.093E-05 9.217E-06 1.093E-05 9.215E-06 1.110E-05 1.008E-05 1.110E-05 1.008E-05 Pr 7.106 5.820 7.105 5.818 7.227 6431 7.227 6.430 Film Resistance:

Velocity (ftls) 1.631 0.345 1.631 0.347 1.631 0346 1.631 0.348 Re 4476 1122 4477 1131 4408 1029 4408 1036 Nu 128.22 40.35 128.23 40.59 127.56 39.26 127.56 39.45 h (Btulhr-tm-F) 1488.9 477.0 1489.0 479.9 1478 8 460.0 1478.9 462.4 C (Btu/hr-*F) 2,523,314 530,907 2,523,320 534,924 2,522,809 533,771 2,522,812 537,296 Cd1n (Btu/hr-'F) 530,907 534,924 533,771 537,296 C,,L, (Btu/hr-°F) 2,523.314 2,523,320 2,522.809 2.522,812 r (CWC.) 0.2104 0.2120 0.2116 0.2130 R (hr-flt-F/Btu) 0.0039719 0.0039594 0.0040542 0.0040430 U (BtuThr-ft 2-'F) 251.8 252.6 246.7 247.3 NTU 2.2767 2.2668 2.2186 2.2101 Effectiveness 0 8645 0.8631 0.8576 0.8564 LMTD ('F) 9.63 9.66 4.95 4.96 Q (MBtu/hr) 11.64 11.71 23.36 5.86 5.89 11.75 Page 5 of 6

0 S-C-SF-MEE-1679, Rev. 0 *-

0o 0

zý Attachment F 2z 0

EXCEL Spreadsheet for 12 CCHX Evaluation - EOL Full Core SFP Discharae 0

z0 12 CCHX only! I FHX 12 CCHX Only" 2 SFHX*

A half B half Total A half B half Total C)

En SW Cc SW Cc SW CC SW CC U)

-4 Inlet temp ('F) 66.00 96.91 66.00 96.91 66.00 82.02 66.00 82.02 o

Outlet temp (OF) 74.73 74.60 74.76 74.66 70.49 70.60 70.50 70.62 ri, Mass Flow (lb,/hr) 2,521,178 988,110 2,521,178 994,132 2,521,178 991,204 2,521,178 997,242 -4 Volumetric Flow (gpm) 5000 1969 5000 1981 5000 1970 5000 1982 Fouling (hr-itt-°FBtu) 0.001000 0.001000 0.001000 0.001000 Properties:

0 Tavg (OF) 70.36 85.76 70.38 85,78 68.24 76.31 68.25 76.32 0

Density @ TI (1bJt9) 62.86 62.56 62.86 62.56 62.86 62.73 62.86 62.73 Density @ Tav (Ibe/ft 3) 62.83 62.16 62.83 62.16 62.85 62.25 62.85 62.25 Cp (BtuIbm-°F) 1.0012 0.9995 1.0012 0.9995 1.0008 1.0001 1.0008 1.0001 z k (Btu/hr-ft-"F) 0.3493 0.3570 0.3493 0.3570 0.3483 0.3528 0.3483 0.3529 0

Dynamic vise (lb,/ft-hr) 2.406 1.934 2.406 1.934 2.475 2.174 2.475 2.173 o

Kinematic visc (fels) 1.064E-05 8.644E-06 1.064E-05 8.642E-06 1.094E-05 9.700E-06 1.094E-05 9.698E-06 Pr 6.896 5.416 6.895 5.414 7.113 6.161 7.112 6.160 Film Resistance:

Velocity (flls) 1.631 0.642 1.631 0.646 1.631 0.643 1.631 0.647 Re 4600 2229 4601 2244 4472 1988 4473 2000 Nu 129.41 66.87 129.42 67.20 128.19 64.40 128.19 64.71 h (Btu/hr-ft2 -°F) 1506.9 795.8 1507.1 799.6 1488.3 757.4 1488.4 761.1 C (Btu/hr-°F) 2,524,232 987,660 2,524,239 993,678 2,523,286 991,311 2,523,290 997,348 Cr,, (Btu/hr-°F) 987,660 993,678 991,311 997,348 C,., (Btuthr-°F) 2,524,232 2,524,239 2,523,286 2,523,290 r (C,,/IC ) 0.3913 0.3937 0.3929 0.3953 R (hr-fe-°F/Btu) 0.0031243 0.0031181 0.0031961 0.0031898 U (Btu/hr-ft2-F) 320.1 320.7 312.9 313.5 NTU 1.5559 1.5495 1.5153 1.5091 Effectiveness 0.7217 0.7200 0.7131 0.7114 LMTD (°F) 14.34 14.36 7.54 7.55 0 (MBtulhr) 22.03 22.11 44.14 11.33 . ,~~11.37, i ,.,6 ~

___ 1,, 22.69 ,

Page 6 of 6

OUTSTANDING CHANGES MUST BE ATTACHED FOR WORKING COPY 20021001 FORM NC.DE-AP.ZZ-0010-1 CERTIFICATION FOR DESIGN VERIFICATION Reference S-C-SF-MEE-1679, Rev.0 No. "SFP Cooling System Capability With Core Offload Starting 100-hours After Shutdown"

SUMMARY

STATEMENT Refueling operations are currently restricted to fuel movement no sooner than 168 hours0.00194 days <br />0.0467 hours <br />2.777778e-4 weeks <br />6.3924e-5 months <br /> after subcriticality. This Is primarily based on evaluation of cooling capability provided by spent fuel pool cooling (SFPC). Decay heat from a full core off-load combined with existing background heat in the spent fuel pool can be removed without exceeding design basis temperature limits provided in vessel decay has occurred for 168 hours0.00194 days <br />0.0467 hours <br />2.777778e-4 weeks <br />6.3924e-5 months <br />. Additional heat removal is required by SFPC ifin-vessel decay is reduced and the core is off-loaded sooner. The capability of SFPC to removal increased decay heat from a full core off-load 100 hours0.00116 days <br />0.0278 hours <br />1.653439e-4 weeks <br />3.805e-5 months <br /> after shutdown has been reviewed in this evaluation.

Additional SFPC margin for 100-hour fuel movement is available using lower component cooling water (CCW) supplied to SFPC and total background heat in the pool is less than 6.8E6 Btufhr. Lower CCW temperature is 0

achieved provided Delaware River temperature Is maintained at or below 63 F. Maximum background heat re sulting Is restricted to that produced from stored spent fuel accumulated through year 2010. This results in the bounding heat load for this evaluation. Maximum pool temperature will increase above UFSAR described tem perature of 149°F if a single SFP heat exchanger is dedicated to the refueling unit. Tandem operation using the non-outage unit's SFPC is required to maintain limits for normal operation. Pool temperatures under postulated accident conditions (i.e. loss of a single heat exchanger and SBO) are calculated to be within current design basis.

In addition to increasing SFP temperature, 100 hour0.00116 days <br />0.0278 hours <br />1.653439e-4 weeks <br />3.805e-5 months <br /> core movement will exacerbate humidity/condensation is sues that result from surface evaporation as pool temperature increase above 135 OF. This can be mitigated by maintaining reduced pool temperatures using tandem operation with reduced swap-over times.

The methodology for this evaluation Is based on benchmarking the SFPC heat exchanger using TEMA design criteria and iterating heat duty, shell (CCW) or tube (SFP) temperatures. Adequate conservatism is used in the evaluation and results are consistent with previous reviews of SFPC performance.

The undersigned hereby certifies (in the right column) that the design verification for the subject document has been completed, the questions from the generic checklist have been reviewed and addressed as appropriate, and all comments have been adequately Incorporated.

Design riffidr Assigned by G. Morrison 4Johnwieaemann (MLEA,Inc.) IdAte '

(Signature of Manager/Supervisor)* "(Signature of Design Verifier)

Design Verifier Assigned By Signature of Design Verifier / Date (signature of Manager/Supervisor)"

Design Verifier Assigned By Signature of Design Verifier I Date (signature of Manager/Supervisor)'

Design Verifier Assigned By Signature of Design Verifier / Date (signature of Manager/Supervisor)'

"Ifthe Manager/Supervisor acts as the Design Verifier, the signature of the next higher level of technical man agement is required.

Page 1 of 1 Nuclear Common Rev. 3

OUTSTANDING CHANGES MUST BE ATTACHED FOR WORKING COPY 20021001 NC.DE-AP.ZZ-0010(Q)

FORM-2 COMMENT i RESOLUTION FORM FOR DESIGN DOCUMENT REVIEW/CHECKING OR DESIGN VERIFICATION REFERENCE DOCUMENT NO. /REV.

CALC NO: S-C-SF-MEE-1679/Rev 0 COMMENTS RESOLUTION

1. GENERAL:

Section 3.0 - Questions resolved by this MEE Section 3.0 already implies the time period limits jw include: of October I to May 15 in question 1. However,

1. Determination when is it acceptable to the questions in this section were revised to include perform a full core off-load if fuel the aspect of time limits based upon background movement begins 100 hours0.00116 days <br />0.0278 hours <br />1.653439e-4 weeks <br />3.805e-5 months <br /> after heat in the pool (e.g. through 2010) and when shutdown (based on differential parallel heat exchangers are needed.

between Delaware River and CCW temperature).

2. Determined maximum SFP inventory (i e last acceptable refueling) that can accommodate a full core off-load at t=100 hrs.

3. Determination of when parallel SFP heat exchanger operation is required.

2.

MEE is based on max. decay heat added to SFP A 113 core offload, if it could be instantaneously jw cooling at time = 146 hrs. Credit is taken for 46 discharged to the SFP at time 10-hours (with no hrs of decay time to determine maximum heat further delay) would have a heat value slightly less added to pool from completely off loaded core. than half the heat produced by a full core offloaded Is the heat added from a hotter, partial core at 146 hours0.00169 days <br />0.0406 hours <br />2.414021e-4 weeks <br />5.5553e-5 months <br />. Consequently, the peak heat will offload less than total core at 146 hrs? always be associated with the full core offload.

3.

Typo last line p. 6: "77" Corrected jw 4.

If parallel SFP heat exchanger operation is An additional table is added to page 1I- e jw required prior to fuel movement and hot pool is HX cycle needed to kee 5 s below 135F.

maintain at 1 15'F, can parallel operation maintain both pools below a temperature th*a--

• _

evaporative cooling does not Acceptance J. Wiedemann 5/13/02 elGaizo 5/13/02 of SUBM~ITTED BY DATE I RESOLVED BY DATE Resolution Page 1 of 1 Nuclear Common Rev. 4

OUTSTANDING CHANGES MUST BE ATTACHED FOR WORKING COPY 20021001 NC.NA-AS.ZZ-0059(Q)

FORM-1 REGULATORY CHANGE PROCESS DETERMINATION Document ID I S-C-SF-MEE-1 679 I Revision 0 Title SFP Cooling with Core Offload Starting 100-hours After Shutdown PAGE 1 OF 4 Activity

Description:

The activity determines that during the months from October through May through the year 2010, a fully radiated 193 element core can be off-loaded to either Salem Spent Fuel Pool with a 100-hour In-vessel decay, rather than a 168 hour0.00194 days <br />0.0467 hours <br />2.777778e-4 weeks <br />6.3924e-5 months <br /> decay, because the SFPC system is capable of (1) maintaining both Salem pools below 149OF with two 0 only one heat exchanger avail SFPC heat exchangers available and (2) maintaining both pools below 180 F with able. While this capability meets the requirements of UFSAR Chapter 9.1.3.1, a Technical Specification change will be required because a 168-hour delay is currently required regardless of the time of year or cooling water tem peratures.

Note that more than one process may apply. If unsure of any answer,contact the cognizant depart ment for guidance. __ ,,

1. Does the proposed activity involve a change to the Tech- E] No If Yes, contact Licensing; process In nical Specifications or the Operating License? accordance with El Yes NC.NA-AP.ZZ-0035(Q)

Does the proposed activity involve a change to the Quality El No If Yes, contact Quality Assessment; Assurance Plan? Examples: process in accordance with

• Changes to Chapter 17.2 of UFSAR El Yes ND.QN-AP.ZZ-0003(Q)

3. Does the proposed activity involve a change to the Secu- [ No If Yes, contact Security Department; rity Plan? Examples: process in accordance with

", Change program in NC.NA-AP.ZZ-0033(Q) El Yes NC.NA-AP.ZZ-0033(Q)

"* Change indoor/outdoor security lighting

"* Placement of component or structure (permanent or temporary) within 20 feet of perimeter fence

"* Obstruct field of view from any manned post

"* Interfere with security monitoring device capability

"* Change access to any protected or vital area

"* Modify safeguards systems or equipment

4. Does the proposed activity involve a change to the Emer- Z No If Yes, contact Emergency Prepared gency Plan? Examples: ness

"* Change ODCM El Yes

"* Change liquid or gaseous effluent release path

• Affect radiation monitoring instrumentation used in classifying accident severity

• Affect emergency facilities, including control rm

• Affect communications, telephone, SPDS or Met tower

5. Does the proposed activity Involve a change to the ISI 0 No IfYes, contact ReliabiLity Programs Program Plan? Examples: ISI/IST; process in accordance with

• Affect Nuclear Class 1, 2, or 3 Piping, Vessels, or El Yes NC.NA-AP.ZZ-0027(O)

Supports (Guidance in NC.DE-AP.ZZ-0007(Q)

Form-11)

Nuclear Common Rev. 4

OUTSTANDING CHANGES MUST BE ATTACHED FOR WORKING COPY 20021001 NC.NA-AS.ZZ-0059(Q)

FORM-1 REGULATORY CHANGE PROCESS DETERMINATION Document ID I S-C-SF-MEE-1679 I Revision 0 Title SFP Cooling with Core Offload Starting 100-hours After Shutdown PAGE 2 OF 4

5. Does the proposed activity invole a change to the IST No If Yes, contact Reliability Programs Program Plan? Examples: ISI/IST; process in accordance with Affect the design or operating parameters of a Nuclear El Yes NC.NA-AP.ZZ-O070(Q)

Class 1, 2, or 3 Pump or Valve (Guidance in NC.DE-AP.ZZ-0007(Q) Form-15)

7. Does the proposed activity involve a change to the Fire [ No if Yes, contact Design Engineering; Protection Program? Examples: process in accordance with

"* Change program In NC.DE-PS.ZZ-0001(Q) F] Yes NC.DE-PS.ZZ-0001(Q)

• Change combustible loading of safety related space

"* Change or affect fire detection system

"* Change or affect fire suppression system/component

"* Change fire door, damper, penetration seal or barrier

"* See NC.DE-AP.ZZ-00D7, Forms 3,4 and 14 for de tails

8. Does the proposed activity involve Maintenance, which 0 No If Yes, contact Maintenance; process restores SSCs to their design conditlon? Examples: in accordance with "a CM or PM activity El Yes NC.WM-AP.ZZ-00OI(Q)

"* Design and configuration remain unchanged

"* Implements an approved Design Change?

9. Is the proposed activity a temporary change (T-Mod) [ No IfYes, contact Engineering: process which meets all the following conditions? n accordance with

"* Directly supports maintenance and is NOT a compen- El Yes NC.NA-AP.ZZ-0013(Q) satory measure to ensure SSC operability.

"* Will be In effect during power operation for less than 90 days.

"* Activity will NOT change the normal system lineup.

"o SSCs will NOT be operated in a manner that could impact the function or operability of a safety related or Important-to-Safety system.

10. Does the proposed activity consist of changes to mainte- 0 No If Yes, contact Maintenance; process nance procedures which do NOT affect SSC destgn, per- n accordance with formance, operation or control? El Yes C.NA-AP.ZZ-0001(Q)

Note: Procedures containing important information con cerning SSC design, performance, operation or control, including those for Tech Spec required surveillance and inspection, require50.59 screening.Examples Include acceptance criteria for valve stroke times or other SSC function, torque values, and types of materials (e.g., gas kets, elastomers, lubricants, etc ) .........

Nuclear Common Rev. 4

OUTSTANDING CHANGES MUST BE ATTACHED FOR WORKING COPY 20021001 NC.NA-AS.ZZ-0059(Q)

FORM-I REGULATORY CHANGE PROCESS DETERMINATION I Document ID S-C-SF-MEE-1679 Revision 0 Title SFP Cooling with Core Offload Starting 100-hours After Shutdown PAGE 3 OF 4

11. Does the proposed activity involve a minor UFSAR [ No IfYes, process in accordance with change (including documents incorporated by reference)? NC.NA-AP.ZZ-0035(Q)

Examples: j] Yes

"* Reformatting, simplification or clarifications that do not change the meaning or substance of information

"* Removes obsolete or redundant information or ex cessive detail

"* Corrects inconsistencies within the UFSAR

"* Minor correction of drawings (such as mislabeled ID)

12. Does the proposed activity involve a change to a Q-listed [ No If Yes, process in accordance with Administrative Procedure (NAP, SAP or DAP) governing NC.NA-AP.ZZ-000 1(Q) and the conduct of station operations? Examples: El Yes NC.DM-AP.ZZ-0001(Q) a Organization changes

13. Does the proposed activity Involve a change to a regula- 0 No If Yes, contact Licensing and proc tory commitment not covered by another regulation-based ess in accordance with change process? El Yes NC.NA-AP.ZZ-0030(Q)

14. Does the activity impact other programs controlled by 0 No If Yes, process in accordance with regulations, operating license or Tech Spec? Examples: applicable procedures such as:

"* Chemical Controls Program [] Yes NC.NA-AP.ZZ-0038(Q)

"* NJ "Right-to-know" regulations NC.LR-AP.ZZ-0037(Q)

"* OSHA regulations

"* NJPDES Permit conditions

"* State and/or local building, electrical, plumbing, storm water management or "other" codes and standards

15. Has the activity already received a 10CFR50.59 review or ED No Take credit for 10CFR50.59 review evaluation under another process? Examples: or evaluation already performed.

"* Calculation El Yes

"* Design Change Package or OWD change

"* Procedure for a Test or Experiment

"* DR/Nonconformance

"* Incorporation of previously approved UFSAR change If any other program or regulation may be affected by the proposed activity, contact the department indicated for further review in accordance with the governing procedure. If responsible department determines their program Is not affected, attach a written explanation.

Nuclear Common Rev. 4

OUTSTANDING CHANGES MUST BE ATTACHED FOR WORKING COPY 20021001 NC.NA-AS.ZZ-0059(Q)

FORM-1 REGULATORY CHANGE PROCESS DETERMINATION I Document ID I S-C-SF-MEE-1679 I Revision 0 Title SFP Cooling with Core Offload Starting 100-hours After Shutdown PAGE 4 OF 4 If all of the answers on the previous pages are "No," then check A below:

A. Q None of the activity is controlled by any of the processes above, therefore a I0CFR50.59 review LS required. Complete a IOCFR50.59 screen.

If one or more of the answers on the previous pages are "Yes," then check either B or C below as appro priate:

B. [ All aspects of the activity are controlled by one or more of the processes above, therefore a IOCFR50.59 review IS NOT required.

C. Q Only part of the activity is controlled by the processes above, therefore a 10CFR50.59 review IS required. Complete a 50.59 screen.

Preparer: T. J. DelGaizo 5/5/02 Printed Name gnature Date Reviewer:

Printed Name Signature PDate Nuclear Common Rev. 4

APPENDIX 1 Error Notice Log Software Number: A-O-ZZZ-MCS-0113 Revision: 00 Software Name: CPO)RTTF*

Application: Mainframe X Personal Computer Error Date Source/Description CR Date Notice Received Number Closed /

Number Distributed 4 1- I I + 4 1 I t 4 4 4 1- 4 1 I 1- 1 1 4 4- 1 1 I I I 1 1- I I 4 4- 4 £ i $- i I I t- I I t 4- 4 £ I I I I

• __________ a A A Form NMEDED-001 -Al-1

APPENDIX 1 Error Notice Disposition Form Error Identification Software Number: A-O-ZZ-MCS -0113 Revision: 00 Software Name: CROSSTIE Application: Mainframe X Personal Computer Error Notice Number: Source:

Error

Description:

Error Evaluation Applicable to Platform: Yes No Preliminary Disposition:

Affected Calculations (list):

DEF Number: Date:

Close-out Action Items:

Software Sponsor: Date:

Form NMEDED-001 -A1-2

APPENDIX 2 INITIAL BENCHMARK VERIFICATION ATTACHMENT 1: Holtec document ID HI-931099, "Verification and Validation Documentation for Computer Program CROSSTIE"