ML042930379
| ML042930379 | |
| Person / Time | |
|---|---|
| Site: | Columbia  |
| Issue date: | 09/22/2004 |
| From: | Nowak B, Woosley L Energy Northwest |
| To: | Office of Nuclear Reactor Regulation |
| References | |
| -RFPFR, GO2-04-170 NE-02-04-1, Rev 2 | |
| Download: ML042930379 (25) | |
Text
Page No.
ENERGY 1.000 a,
NORTHWEST CALCULATION COVER SHEET Calculation No. NE-02-04-1 People
- Vision-Solutions Revision No.
2 Equipment Piece No.
Project: Columbia I Quality Class: I Discipline: NUCLEAR Remarks TITLE/SUBJECT/PURPOSE DOSE CALCULATION DATABASE Purpose The purpose of this data base is to provide documentation of all plant-specific information referenced by the Alternative Source Term (AST) dose calculations for application to LOCA and Non-LOCA design basis radiological release accidents, as well as general information for the supporting calculations.
CALCULATION REVISION RECORD REVISION STATUSI REVISION DESCRIPTION INITIATING TRANSMITTAL NO.
F,P, OR S DOCUMENTS NO.
0 F
ORIGINAL ISSUANCE F
- 1. Changed items 3.12, 3.13, and 3.15 from before/after 30 minutes to dual train/single train.
- 2. Added References 16 and 37.
- 3. Revised item 5.1 and Reference 26.
2 F
Replace Appendix C; make editorial changes CR 2-04-05249 PERFORMANCE VERIFICATION RECORD REVISION PERFORMED BYIDATE VERIFIED BYIDATE APPROVED BY/DATE NO.
0 B. Nowack J. Metcalf J. Metcalf I
B. Nowack J. Metcalf J. Metcalf 2
LS Woosley M Abu-Shehadeh "
(4l LC L Study calculations shall be used only for the purpose of evaluating alternate design options or assisting the engineer in performing assessments.
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Page No.
f,,4ENERGY 1.100 NORTHWEST INDEX Calculation No. NE-02-04-1 People. Vision Solutions Revision No.
I ITEM Calculation Cover Sheet Calculation Index Verification Checklist for Calculations and CMR's Calculation Reference List Calculation Output Interface Documents Revision Index Calculation Output Summary Calculation Method Sketches Manual Calculation PAGE NO. SEQUENCE 1.000 -
1.100 -
1.200 -
1.300 -
1.400 -
2.000 -
3.000 -
4.000 -
5.000 -
1.302 5.010 APPENDICES:
Distance between the Blowout Panels and the Control Room Local Intake Appendix A
I Page A-I - Al Internal Volume of a Main Steam Line between Main Steam Line Isolation Valves Appendix B
I Page B-i - B1 Design Basis FHA Fuel Assembly Drop Height Appendix C
2 Pages C-I - C-2 Appendix Pages AppendixPaes Appendix Pages Appendix Pages 25278 R5
Page No.
Cont. on page ENERGY VERIFICATION 1.200 1.300 2N NORTHWEST Calculation No. NE-02-04-1 People -Vision Solutions I
1 Revision No.
2 Calculation/CMR NE-02-04-01 was verified using the following methods:
Z Checklist Below Revision 2 D
Alternate Calculation(s)
Verifier Initials Checklist Item Clear statement of purpose of analysis....................................................................
Methodology is clearly stated, sufficiently detailed, and appropriate for the proposed application..........................................................................................
Does the analysis/calculation methodology (including criteria and assumptions) differ from that described in the Plant or ISFSI FSAR or NRC Safety Evaluation Report, or are the results of the analysis/calculation as described in the Plant or ISFSI FSAR or NRC Safety Evaluation Report affected?
3 Yes El No.................................................................................................
If Yes, ensure that the requirements of 10 CFR 50.59 and/or 10 CFR 72.48 have been processed in accordance with SWP-LIC-02....................................
Does the analysis/calculation result require revising any existing output interface document as identified in DES-4-1, Attachment 7.3?
0 Yes 5 No..................................................................................................
If Yes, ensure that the appropriate actions are taken to revise the output interface documents per DES-4-1, section 3.1.8 (i.e., document change is initiated in accordance with applicable procedures)......................................
Logical consistency of analysis................................................................................
Completeness of documenting references........................................................
Completeness of input.......................................................................................
Accuracy of input data........................................................................................
Consistency of input data with approved criteria...............................................
Completeness in stating assumptions...............................................................
Validity of assumptions......................................................................................
Calculation sufficiently detailed..........................................................................
Arithmetical accuracy.........................................................................................
Physical units specified and correctly used.......................................................
Reasonableness of output conclusion...............................................................
Supervisor independency check (if acting as Verifier).............................................
Did not specify analysis approach Did not rule out specific analysis options Did not establish analysis inputs........................................................................
If a computer program was used:.............................................................................
Is the program appropriate for the proposed application?
Have the program error notices been reviewed to determine if they pose any limitations for this application?
Is the program name, revision number, and date of run inscribed on the output?
Is the program identified on the Calculation Method Form?
If so, is it listed in Chapter 10 of the Engineering Standards Manual?..............
Other elements considered:
-AS of INNA I k AI
'#A A.I h'Nf
- bA6, kon If separate Verifiers were used for validating these functions or a portion of these functions, each sign and initial below.
Based on the foregoing, the Calculation/CMR is adequate for the purpose intended.
Verifier Signature(s)/Date Verifier Initials
_ L___
L 2 25280 R5
Page No.
Cont. on page ENERGY CALCULATION 1301.0 NORTHW EST REFERENCE LIST Calculation No.
NE-02-04-1 People
- Vision Solutions Revision No.
I ISSUE DATE/
NO AUTHOR EDITION OR TITLE DOCUMENT NO.
REV.
Design Specification for Division 60, "Reactor Core I
Energy Northwest Revision 11 and System Analysis for Columbia Generating Station" 2
Energy Northwest October 3, 2000 Technical Memorandum "Core Inventory TM-2128 Comparison"'
3 Energy Northwest Revision I Origen Run Calculation 5.01.59 5.01.59 Calculation of the Mass of Steam Released CMR-2726 to calc 4
Energy Northwest 10/22/2003 Following a MSLB Using AST RG 1.183 NE-02-0 1-11 Methodology, 10/31/01 NE-02-01-11 5
Energy Northwest Tech. Spec. 3.4.8 "RCS Specific Activity" TS Amendment No.
U.S. Nuclear Regulatory Alternative Radiological Source Terms for 6
Commission July, 2000 Evaluating Design Basis Accidents at Nuclear Regulatory Guide Power Reactors 7
Energy Northwest Amendment 55 Columbia Generating Station Final Safety Analysis WNP-2 FSAR Report, Amendment 55 8
General Electric December 1994 GE calculation GE-NE-208-17-0993, page 38-12 GE-NE-208-17-0993 9
Combustion Engineering Toping report CENPD-284 (generic analysis CENPD-284 Combstio Enineeingshowing no pin failures for CRDA) 10 Energy Northwest 11/1/01 Calculate the Maximum Volume of the Control NE-02-00-06 Room Envelope I_
Energy Northwest 1/18/2000 Secondary Containment Bypass Leakage Limit NE-02-85-12 Rev. 1 12 Energy Northwest Tech Spec 5.5.12 "Primary Containment Leakage Amendment 169 Rate:
13 Energy Northwest SER 15.3.1 14 Energy Northwest Revision 0 E/I 02-91-1066, " Setting Range Determination for E/I- 02-91-1066 SGT-LMTR-IlA I1" pg. 2.001 15 Energy Northwest 11/8/00 To Calculate the free Air Volume of Secondary NE-02-99-14 Containment 16 NUCON International, Inc 5/24/04 Control Room Habitability Tracer Gas Testing of NUCON the CRE at Columbia Generating Station 12COLUM1 130/02 1 7 Energy Northwest Technical Specifications Surveillance Requirement SR 3.6.1.3.11 SR 3.6.1.3.11 2?521 R2
Page No.
Cont. on page (EN ENERGY CALCULATION 1.301 1.302 NORTHWEST REFERENCE LIST Calculation No.
NE-02-04-1 Pueple -Vision -Solutions Revision No. 1 Control Room HVAC System Characterization Work Order # 18 Energy Northwest Test" 01062329 9
E g Design Specification for Division 300, Section 311, Residual Heat Removal System "Preliminary Test Results From Leakage Test per P.O. No. 49CN-220.
20 Buffalo Forge Co.
June 26, 1979 Research Report 78-99 (SGT-FN-IA-I, IA-2, lB-i, Contact No. 28. Spec.
IB-2)"
No. 2808-28.
21 Burns & Roe, Inc.
7-1-74, Rev.0 Building Volume and Air Change Calc No. 9.49.33 22 Burns & Roe, Inc.
2-13-74 Drawing CVI 08-00,6 CVI 08-00,6 23 Burns & Roe, Inc.
4-23-71 Condenser Volume Calc No. 4.09.01 24 Energy Northwest 3-14-01 Technical Memorandum "Guidance for Controlling TM-2130 Rev.0 Hydrocarbon Loading of Charcoal Filters" 25 Energy Northwest April 22, 2003 "Certificate of conformance", Flanders Filters, Inc.,
Cat ID: 00584011821 Energy Northwest 26 Energy Northwest Revision 0 Control room X/Q using ARCON96 with the NE-02-03.14 1996-1999 Meteorological Data" 27 Energy Northwest Revision 0 Calculation of EAB and LPZ X/Q using PAVAN NE-02-03.16 with the 1996-1999 Meteorological Data" U.S. Nuclear Regulatory Atmospheric Relative Concentrations for Control Regulatory Guide 28 USNuclear July, 2003 Room Radiological Habitability Assessments at 1.194 Commission Nuclear Power Plants 29 Energy Northwest Rev. 8 10/9/03 Columbia Generating Station Site Wide Procedure, SWP-CHE..02 "Chemical Process Management and Control" 30 Energy Northwest April 29, 2004 "Containment Material" (QlD W01239, Task # 4)
WR 1239-00Q 31 Energy Northwest Revision 0 April EOP/SAG APPENDIX C CALCULATIONS NE-02-97-16 14, 1998 32 General Electric Revision 0 May "Heat Balance, Reactor System 104.8% Uprated 25A5145 15, 1992 Steam Flow' CMR..2726 Calculation of The Mass of Steam Released NE-02-01-11 33 Energy Northwest 1/20/2004 Following a MSLB Using AST (Regulatory Guide CMR-2726 1.183) Methodology 34 Energy Northwest 10/30/03 General Operating Procedure " Reactor Plant PPM 3.1.2 Enery Nrthest 0/3/03 Startup" HVAC 35 Burns & Roe, Inc.
5-18-74 Plans & Sections At EL. 525'-0" Drawing M826 R 33 Radwaste and Control Building 36 Burns & Roe, Inc.
5-7-74 Radwaste and Control Building DWG. A521-1 3 s R IIControl Room Plans and Elevations 25281 R2
Page No.
Cont. on page ENERGY CALCULATION 1.302 NORTHWEST REFERENCE LIST Calculation No.
NE-02-04-1 People
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- Solutions Revision No.
1 37 EnegyNorhwstevsio Secondary Containment Drawdown with 7 WS-129-CALC-001 37 Energy Northwest Revision I AppendicesC-00 Appendices 38 Energy Northwest Revision F Drawings SM136 and SM191 39 Energy Northwest Revision 0 TSG-Operations Manager, TM-2118, TM-2118 40 Energy Northwest Revision 0, July WNP-2 EOP/SAG Technical Document, TM-2120, TM2120 2000 Chart E Technical Specification "Surveillance S 3..
41 Energy Northwest Requirements" SR 3.6.1.3.6 42 Farr Company Revision M Control Room Emergency Filtration Unit CVI 02-18-00,15 8-6-73 General Arrangements Operating Floor Plan Rev. 14 "Turbine Generator Building" EPN MS-RIS-610A 02/20/02 Revision Instrumentation Master Data Sheets EPN MS-RIS-61 OB 44 Energy Northwest 29 Loops MS-RIS-610A Loops MS-RIS-61 OB 9/11/98 WNP-2 SPECIFICATIONS DIV. 100 45 Energy Northwest Re91 6 Section I "Process Piping and Pipe Support" ASME Div. 100 Rev. 6 Section III Design Spec.
46 Energy Northwest 1-23-75 Steam Line B From Isolation Valve to T.G. Stop MS-529-1.3 Revision I1 Valve IOM "Radial Peaking Factor for FHA and CRDA EN2-RXFE-04-0012 47 Energy Northwest March 29, 2004 Analyses" 48 Energy Northwest Rev. 90, 3/4/02 Flow Diagram "HVAC for Control and Switchgear DWG: M548-1 Rooms, Radwaste Building" 25281 R2
ENERGY CALCULATION OUTPUT Page No.
NORTHWEST INTERFACE DOCUMENT 1.400 People Vision*Solutio REVISION INDEX Calculation No. NE-02-04-1 Prepared by
/Date:
K Verified by/Dat 7 qlO4 Revision No.
1 The below listed output interface calculations and/or documents are impacted by the current revision of the subject calculation. The listed output interfaces require revision as a result of this calculation. The documents have been revised, or the revision deferred with Manager approval, as indicated below.
CHANGED BY CHANGED DEFERRED DEPT.
AFFECTED DOCUMENT NO.
(e.g., RFTS, LETTER NO.)
MANAGER*
- Required for deferred changes only.
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(
ENERGY CALCULATION OUTPUT 2.000
\\N NORTHWEST
SUMMARY
Calculation No. NE-02-04-1 People-Vision Solutions Revision No.
I This data base presented in Section 5 includes all plant-specific information referenced by the alternative source term dose calculations, as well as general information for the supporting calculations prepared for the project. The data base is divided into logical sections, and each entry or group of entries is identified by a section and item number to facilitate reference to that entry. References for each entry or group of entries are shown in parenthesis in a way that clearly relates to the entry.
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FENERGY Page NNORTHWEST CALCULATION METHOD 3.000 People-Vision-Solutions Calculation No. NE-02-04-1 Prepared by:
S Verified by 7191d§j Revision 1
REV.
BAR Analysis Method (Check Appropriate Boxes)
Li Manual (As required, document source of equations in Reference List) a Computer LI Main Frame LI Personal El In-House Program aI Computer Service Bureau Program
[]
Other El Verified Program Code name/Revision El Unverified Program:
Approach I Methodology This data base includes all plant-specific information referenced by the alternative source term dose calculations, as well as general information for the supporting calculations prepared for the project.
The data base presents secondary references, as applicable. These secondary references may include:
- 1. Existing Safety-Related General Electric reports
- 2. Existing Safety-Related Energy Northwest calculations or other technical documents supporting the current dose calculations
- 3. Responses to inquiries made by the Polestar Project Manager to the Energy Northwest Technical Contact
- 4. Current Technical Specifications and their associated Bases furnished by Energy Northwest
- 5. Results of Safety-Related tests, surveillances, and inspections furnished by Energy Northwest
- 6. Safety-Related Polestar calculations or studies prepared in accordance for this project or for the support or generic Polestar Safety-Related activities The data base is divided into logical sections, and each entry or group of entries is identified by a section and item number to facilitate reference to that entry. References for each entry or group of entries are shown in parenthesis in a way that clearly relates to the entry.
EN ENERGY Ppl NORTHWEST People
- Vision
- Solutions SKETCHES AND DIAGRAMS Page 4.000 Calculation No. NE-02-04-1 Prepared by:
Revision 1
C Sketches and Diagrams See Appendix A for a sketch of the MSLB release transport distances to the Control Room air intake.
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Ej ENERGY Page No.
Cont. on page N NORTHWEST MANUAL CALCULATION 5.000 5.001 People
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- Solutions Calculation No. NE-02-04-f Prepared by / Date:,'Verified by/Date:
44 Revision No.
1 C
REV.
- 1.
Radionuclide Data 1.1.
Core Power - 3556 MWt (Reference 1)
The Tech Spec value was increased by 2% to account for power measurement uncertainties in accordance with SRP 15.6.5.
1.2.
Core Inventory* @ t=O (Reference 2)
Nuclide Ci/MWt INcide CiIMWt I Nuclide Ci/MWt 1 Kr83m Kr85m Kr85 Kr87 Kr88 Kr89 Xe131m Xel33m Xe133 Xel35m Xe135 Xe137 Xel38 Xe139 Xel4O 1131Org 11 320rg 11 330rg 11340rg 11350rg 1131 Elem 1132Elem 11 33Elem 1134Elem 1135Elem 1131 Part 3.57E+03 7.35E+03 4.11E+02 1.34E+04 1.90E+04 2.20E+04 2.79E+02 1.66E+03 5.43E+04 1.11E+04 1.31 E+04 4.65E+04 3.59E+04 N/A**
N/A**
2.79E+04 3.94E+04 5.44E+04 6.03E+04 5.03E+04 2.79E+04 3.94E+04 5.44E+04 6.03E+04 5.03E+04 2.79E+04 1132Part 11 33Part 1134Part 11 35Part Rb86 Cs134 Csl36 Cs137 Sb127 Sbl129 Tel27m Tel27 Tel29m Te129 Tel31m Tel32 Bal37m Bal 39 Ba140 Mo99 Tc99m RulO3 Ru105 RulO6 Rh1O5 Y90 3.94E+04 5.44E+04 6.03E+04 5.03E+04 4.47E+01 6.27E+03 1.39E+03 5.05E+03 3.31 E+03 9.48E+03 4.66E+02 3.31 E+03 1.39E+03 8.90E+03 4.20E+03 3.99E+04 3.01 E+03 4.72E+04 4.58E+04 4.90E+04 4.34E+04 4.70E+04 3.46E+04 2.04E+04 3.27E+04 2.04E+03 Y91 Y92 Y93 Zr95 Zr97 Nb95 La140 Lal41 La142 Pr143 Nd147 Am241 Cm242 Cm244 Cel4l Ce143 Ce144 Np239 Pu238 Pu239 Pu240 Pu241 Sr89 Sr9O Sr9l Sr92 2.73E+04 2.90E+04 3.56E+04 4.27E+04 4.33E+04 4.27E+04 4.71 E+04 4.36E+04 4.17E+04 3.78E+04 1.71 E+04 7.67E+00 1.74E+03 1.41 E+02 4.43E+04 4.01 E+04 3.25E+04 7.01 E+05 9.56E+01 1.89E+01 3.11E+01 8.85E+03 2.02E+04 3.34E+03 2.59E+04 3.01 E+04
- Radionuclides omitted - do not appear in TID-14844 or in RADTRAD default 18694 R3
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- Solutions Calculation No. NE-02-04-1 Prepared by t Date:
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7 1 04 Revision No.
I
,i~
(
REV.
BAR 1.3. Core Inventory by Mass*
(Reference 3)
Element g/Group g/MWt Cs 2.93E+05 3.26E+05 9.33E+01 Rb 3.24E+04 i
2.75E+04 2.97E+04 8.52E+00 Br 2.19E+03 Te 5.33E+04 Sb 3.66E+03 6.26E+04 1.79E+01 Se 5.65E+03 Ba 1.44E+05 2.30E+05 6.59E+01 Sr 8.57E+04 Mo 3.37E+05 Tc 8.23E+04 Ru 2.69E+05 8.92E+05 2.56E+02 Rh 5.00E+04 Pd 1.54E+05 La 1.27E+05 Y
4.44E+04 Zr 3.51 E+05 Nb 3.80E+03 Pr 1.13E+05 Nd 3.76E+05 1.13E+06 3.23E+02 Pm 1.55E+04 Sm 7.78E+04 Eu 1.84E+04 Am N/A Cm N/A Ce N/A Np N/A 1.22E+06 3.50E+02 Pu 1.22E+06 ffi Low mass is conservative for activity transport - therefore, Reference 3 is used for these data instead of Reference 2. When used for pH calculation (high mass conservative), these values multiplied by a factor of 1.1 for conservatism.
1.4. Peaking Factor for FHA - 1.7 (Reference 1.5. Peaking Factor for CRDA - 1.7 (Reference 1.6. Coolant Liquid Mass Release for MSLB - 105,000 Ibm (Referenc 1.7. Coolant DE 1-131 Activity per Unit Mass - 0.2 microCurie/gram (Referenc 1.8. Spiking Multiplier for Coolant DE 1-131 Activity for MSLB - 20 (Referenc e47) e47) ce 4) ce 5) ce 5) 18694 R3
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ENERGY Page No.
cont. on page N NORTHWEST MANUAL CALCULATION 5.002 15.003 People
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- 2.
Source Terms (Reference 6)
Item Release Time Frame Fraction of Core Inventory Released (total* and per hour**)
2.1.
0 - 0.033 hours3.819444e-4 days <br />0.00917 hours <br />5.456349e-5 weeks <br />1.25565e-5 months <br /> No Release (LOCA only) 2.2.
Instantaneous at t =
Gases***
Xe, Kr (general), I (general) 0.05 total 24 hours2.777778e-4 days <br />0.00667 hours <br />3.968254e-5 weeks <br />9.132e-6 months <br /> after Kr-85 0.10 total shutdown (FHA only) 1-131 0.08 total 2.3.
0.033 - 0.533 hours0.00617 days <br />0.148 hours <br />8.812831e-4 weeks <br />2.028065e-4 months <br /> Gases Xe, Kr - 0.1/hr 0.05 total (LOCA only)
Elemental I - 4.9E-3/hr 2.4E-3 total Organic I - 1.5E4/hr 7.5E-5 total Aerosols I, Br - 0.095/hr 0.0475 total Cs, Rb - 0.1/hr 0.05 total 2.4.
0.533 -2.033 hours3.819444e-4 days <br />0.00917 hours <br />5.456349e-5 weeks <br />1.25565e-5 months <br /> Gases Xe, Kr - 0.63/hr 0.95 total (LOCA only)
Elemental I - 8.1 E-3/hr 1.2E-2 total Organic I - 2.5E-4/hr 3.8E-4 total Aerosols I, Br - 0.158/hr 0.2375 total Cs, Rb - 0.133/hr 0.2 total Te Group - 0.033/hr 0.05 total Ba, Sr-0.013/hr 0.02 total Noble Metals - 1.7E-3/hr 2.5E-3 total La Group - 1.3E4/hr 2E-4 total Ce Group - 3.3E-4/hr 5E4 total
- Total releases from tables in Section 3.2 of Reference 5
- ltem 2.3: per hour release = total/0.5 hours5.787037e-5 days <br />0.00139 hours <br />8.267196e-6 weeks <br />1.9025e-6 months <br />. Item 2.4: per hour release = total/1.5 hours5.787037e-5 days <br />0.00139 hours <br />8.267196e-6 weeks <br />1.9025e-6 months <br />.
- ltem 2.2: alkali metals ignored (alkali metals are particulate with infinite water pool DF) 2.5.
Percentage of Fuel Pin Failures for FHA - 0.528%
(Ref. 7, sec 15.7.4, and Ref. 8, P. 38-12)
(This fraction = 250 pins out of 764 assemblies x 62 pins/assembly for 8x8 fuel - conservative for fuel with increased array.)
2.6. Fraction of Fuel Damaged for CRDA - 0.0179446 (Ref. 7, sec 15.4.9, and Ref. 9)
(This fraction = 850 pins out of 764 assemblies x 62 pins/assembly for 8x8 fuel - conservative for fuel with increased array.)
2.7.
Fraction of Fuel Melted for CRDA - 0.0001382 (Ref. 7, sec 15.4.9, and 9)
(This fraction = 0.77% of the 850 damaged pins out of 764 assemblies x 62 pins/assembly for 8x8 fuel -
conservative for fuel with increased array. 0.77% is the mass fraction of the damaged fuel rods assumed to reach or to exceed the initiation temperature of fuel melting.)
2.8.
CRDA Release Fractions (Instantaneous)
(Reference 6)
Radionuclide Release Fraction from Release Fraction from Group Gap to Coolant Melted Fuel to Coolant Noble Gas 10%
100%
Iodine 10%
50%
Br 5%
30%
Cs, Rb 12%
25%
Te Group 0%
5%
Ba, Sr 0%
2%
Noble Mtls 0%
0.25%
Ce Group 0%
0.05%
La Group 0%
0.02%
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- 3.
Volumes and Volumetric Flowrates 3.1. Volume of Drywell - 200,540 ft3 3.2. Max Volume of Wetwell - Vapor: 144,184 ft3 (Ref. 7, Table 6.2-1 and Ref. 8, Table 29-1)
(Ref. 7, Table 6.2-1 and Ref. 8, Table 29-1) 3.3. Min Volume of Wetwell - Water: 137,262 ft3*
(Ref. 7, Table 6.2-1 and Ref. 8, Table 29-1)
- Includes water in pedestal and water lower than 12' below vent exit (Sum of SP-water + pedestal + water < 12' = 112,197 + 10,065 + 15,000) 3.4. Volume of Wetwell - Total: 281,446 ft3 (Item 3.2 + Item 3.3) 3.5. Volume of Control Room (free volume) - 214,000 ft3 (Reference 10) 3.6. Volume of One Main Steam Line between MSIVs - 64.8 ft3 (See Appendix B) 3.7. Volumetric Flowrate, Drywell to Environment, Non-MSIV Secondary Containment (SC) Bypass = 0.04 % DW Volume per Day. This is the drywell leakage directly to the atmosphere bypassing the secondary containment. This leakage occurs before and after drawdown.
(Reference 11)
Tech Spec DW Leakage = 0.5 % DW Volume per Day. This is the drywell leakage which is assumed to go to the atmosphere bypassing the SC before drawdown, but goes to the SC after drawdown, then to the atmosphere via SGT.
(Reference 12)
Therefore:
The Total SC bypass leakage before drawdown = 0.04% + 0.5 % = 0.54 % DW Volume per Day The Total SC bypass leakage after drawdown = 0.04 % DW Volume per Day (Reference 11) 3.8. Volumetric Flowrate, Wetwell (WW) to Environment, Non-MSIV Secondary Containment (SC) Bypass = 0.04 % WW Volume per Day. This is the drywell leakage directly to the atmosphere bypassing the secondary containment. This leakage occurs before and after drawdown.
(Reference 11)
Tech Spec DW Leakage = 0.5 % WW Volume per Day. This is the wetwell leakage which is assumed to go to the atmosphere bypassing the SC before drawdown, but goes to the SC after drawdown, then to the atmosphere via SGT.
(Reference 12)
Therefore:
The Total SC bypass leakage before drawdown = 0.04% + 0.5 % = 0,54 % WW Volume per Day The Total SC bypass leakage after drawdown = 0.04 % WW Volume per Day (Reference 11)
Note: The wetwell volume is the space above the suppression pool water which has the same characteristics as the drywell.
3.9. Volumetric Flowrate, ESF Leakage - 1 gpm (0.134 cfm)
(Reference 13) 3.10. Volumetric Flowrate, Secondary Containment to Env - Through SGTS = 5000 cfm**
(Reference 14)
- Actual value = 5378 +/- 433.5 cfm. Nominal 5000 cfm used since no credit taken for Secondary Containment holdup.
3.11. Approximate Secondary Containment Volume - 3.48E6 ft3 (Reference 15)
(Note: no Secondary Containment holdup credited in dose analysis. Value needed only for RB shine calc) 18694 R3
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Revision No.
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kJ 3.12. Volumetric Flowrate, Environment to CR (Unfiltered) a 75 cfm (dual train) or 50 cfm (single train) (including 10 cfm for ingress/egress)
(Reference 16)
REV.
BAR 3.13. Min. Vol. Flowrate, Environment to CR (Charcoal Filter):
1300 cfm (dual train) 800 cfm (single train)
(Reference 16) 3.14. Vol. Flowrate, DW to All Main Steam Lines (Total Leakage): 64.4 SCFH*
(Reference 17)
- Current Tech Spec value = 46 SCFH (11.5 SCFH per line). Analysis to be done with 40% greater leakage.
As with current Tech Spec, total leakage to be evenly distributed over four lines.
3.15. Max. Vol. Flowrate, Environment to CR (Charcoal Filter):
1600 cfm (dual train) 900 cfm (single train)
(Reference 18) 3.16. DW Spray flow rate - 7450 gpmltrain (2 trains available)
(Reference 19) 3.17. Volumetric Flowrate, Secondary Containment to Env (Bypassing SGTS) - 50 cfm (Reference 20)
Per reference 20, the original SGT leakage bypass is 14 cfm. In order to relax the surveillance requirement it has been increased to 50 cfm in this AST analysis.
3.18. Volumetric Flowrate, Drywell to Secondary Containment:
0 % DW Volume per Day (before drawdown) 0.5 % DW Volume per Day (after drawdown)
See explanation under item 3.7.
(Reference 12) 3.19. Volumetric Flowrate, Wetwell to Secondary Containment:
0 % WW Volume per Day (before drawdown) 0.5 % WW Volume per Day (after drawdown)
See explanation under item 3.8.
(Reference 12) 3.20. Normal Makeup Volumetric Flowrate to Control Room - 1100 cfm (Unfiltered)**
- Nominal value actually 1000 cfm. 1100 cfm used for conservatism.
(Reference 48) 3.21. Approximate Turbine Generator Building Volume - 5.71 E6ft3 (Reference 21)
Per reference 21, this is the free Turbine Building volume after subtracting 20% occupied by equipment.
3.22. Main Condenser Mechanical Vacuum Pump Volumetric Flowrate - 3000 cfm (Reference 22)
Per reference 22, the test results range from 2218 - 2870 cfm. The MVP model is U352 03002 4, which could mean that the flow rate can reach 3000 cfm.
3.23. Approximate Main Condenser Volume Above Hotwell - 121,400 ft3 (Reference 23)
Per reference 23, this is the total volume of the steam inside the condenser which is the sum of:
1-Vol. around tube bundles (49,000 ft3) 2-Vol. between bundles and shell (10,000 ft3) 3-Vol. above hotwell (3390 ft3) 4-Net vol. within hoods ( 59,000 ft3)
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Filter Efficiencies, Removal Lambdas, and Decontamination Factors 4.1.
Filter Efficiency - Standby Gas Treatment System 99% efficiency for all gaseous species except noble gases (Reference 24) 99% HEPA efficiency for all particulate species (Reference 25)
Reference 25 and FSAR Section 6.5.1.2 state 99.97% efficiency, but only 99% is credited.
4.2. Filter Efficiency - CR Intake Filter 95% efficiency for all gaseous species except noble gases 99% efficiency for all particulate species (Reference 24)
(Reference 25) 4.3. Release Fraction of Radioiodine in ESF Leakage: 10%
(Reference 6)
- 5.
X/Q Values, Wind Speed for MSLB Puff, Breathing Rates and Occupancy Factors (Times Relative to Start of Release to Environment) 5.1. XIQ (sec/m 3 )
The effective X/Q was calculated for two control room flow rates: the 1300/800 cfm and the 1600/900 cfm given in items 3.13 and 3.15.
Filtered CR Intake Flow (using the 1600/900 cfm flow rate and assuming one train is Secured between 0-2 hrs) and Unfiltered inleakage X/Q (s/m3)
(Reference 26)
Filtered Unfiltered SGT KK RBW SC Turbine SGT KK RBW SC Turbine Roofline doors SC Bypass Building Roofline doors SC Bypass Building Bypass Bypass 0 - t hrs 1.60E-04 4.07E-04 2.22E-04 4.41 E-04 6.95E-04 5.34E-04 8.69E-04 4.70E-03 t -2 hrs 1.47E-04 3.75E-04 2.05E-04 7.83E-04 6.95E-04 5.34E-04 8.69E-04 4.70E-03 2 - 8 hrs 1.08E-04 2.97E-04 1.48E-04 3.33E-04 3.36E-04 1.97E-04 4.40E-04 2.OOE-03 8 - 24 hrs 4.25E-05 1.21 E-04 5.88E-05 1.72E-04 1.28E-04 8.41 E-05 1.75E-04 1.03E-03 1-4 days 3.61 E-05 1.01 E-04 5.13E-05 1.34E-04 9.72E-05 7.26E-05 1.38E404 8.01 E-04 4 - 30 days 3.1OE-05 8.83E-05 4.29E-05 1.28E-04 7.69E-05 7.OOE-05 1.1OE-04 7.69E-04 Filtered CR Intake Flow (using the 1300/800 cfm flow rate and assuming one train is Secured between 0-2 hrs) and Unfiltered inleakage X/Q (s/m3)
(Reference 26' Filtered Unfiltered SGT KK RBW SC Turbine SGT KK RBW SC Turbine Roofline doors SC Bypass Building Roofline doors SC Bypass Building Bypass Bypass 0 - t hrs 1.56E-04 3.98E-04 2.17E-04 5.42E-04 6.95E-04 5.34E-04 8.69E-04 4.70E-03 t - 2 hrs 1.43E-04 3.65E-04 1.99E-04 8.81 E-04 6.95E-04 5.34E-04 8.69E-04 4.70E-03 2 - 8 hrs 1.05E-04 2.89E-04 1.44E-04 3.75E-04 3.36E-04 1.97E-04 4.40E-04 2.OOE-03 8 - 24 hrs 4.14E-05 1.18E-04 5.73E-05 1.93E-04 1.28E-04 8.41E-05 1.75E-04 1.03E-03 1 -4 days 3.52E-05 9.83E-05 5.OOE-05 1.50E-04 9.72E-05 7.26E-05 1.38E-04 8.01 E-04 4 - 30 days 3.03E-05 8.61 E-05 4.18E-05 1.44E-04 7.69E-05 7.OOE-05 1.10E-04 7.69E-04
_i_-
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Filtered CR Intake Flow (using the 1300/800 cfm flow rate and assuming one train is Secured Between 2-8 hrs) and Unfiltered inleakage X/Q (slm3)
(Reference 26)
Filtered Unfiltered SGT KK RBW SC Turbine SGT KK RBW SC Turbine Roofline doors SC Bypass Building Roofline doors SC Bypass Building Bypass Bypass 0 - 2 hrs 1.56E-04 3.98E-04 2.17E-04 5.42E-04 6.95E-04 5.34E-04 8.69E-04 4.70E-03 2 - t hrs 1.15E-04 3.15E-04 1.57E-04 2.31 E-04 3.36E-04 1.97E-04 4.40E-04 2.OOE-03 t - 8 hrs 1.05E-04 2.89E-04 1.44E-04 3.75E-04 3.36E-04 1.97E-04 4.40E-04 2.OOE-03 8 - 24 hrs 4.14E-05 1.18E-04 5.73E-05 1.93E-04 1.28E-04 8.41E-05 1.75E-04 1.03E-03 1 -4 days 3.52E-05 9.83E-05 5.OOE-05 1.50E-04 9.72E-05 7.26E-05 1.38E-04 8.01 E-04 4 - 30 days 3.03E-05 8.61 E-05 4.18E-05 1.44E-04 7.69E-05 7.OOE-05 1.10E-04 7.69E-04 Filtered CR Intake Flow of 1300 cfm (Assuming Both Trains Remain on For 30 Days and Unfiltered inleakage X/Q (s/m3)
(Reference 26)
Filtered l
Unfiltered l
SGT KK RBW SC Turbine SGT KK RBW SC Turbine Roofline doors SC Bypass Building Roofline doors SC Bypass Building Bypass Bypass 0 - 2 hrs 1.56E-04 3.98E-04 2.17E-04 5.42E-04 6.95E-04 5.34E-04 8.69E-04 4.70E-03 2 - 8 hrs 1.15E-04 3.15E-04 1.57E-04 2.31 E-04 3.36E-04 1.97E-04 4.40E-04 2.OOE-03 8 - 24 hrs 4.51 E-05 1.28E-04 6.24E-05 1.19E-04 1.28E-04 8.41 E-05 1.75E-04 1.03E-03 1 - 4 days 3.83E-05 1.07E-04 5.44E-05 9.24E-05 9.72E-05 7.26E-05 1.38E-04 8.01 E-04 4 - 30 days 3.30E-05 9.38E-05 4.56E-05 8.87E-05 7.69E-05 7.OOE-05 1.10E-04 7.69E-04 X/Qs for EAB (Reference 27) 0 -30 days 1.BI E-04 X/Qs for LPZ X/Qs for CST (Reference 27) 0-8 hr 8-24 hr 1 -4 days 4 - 30 days 0-2 hr 2-8 hr 8-24 hr 1 -4 days 4 - 30 days 4.95E-05 3.69E-05 1.95E-05 7.81 E-06 (Reference 26) 4.18E-04 1.59E-04 6.31 E-05 5.78E-05 5.57E-05 5.2. Minimum Windspeed for MSLB Puff Migration - 1 m/s (Reference 28) 18694 R3
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BAR 5.3. Breathing rates CR Breathing Rates (m3/s) 0 - 30 days 3.5E-4 a
EAB, LPZ and Environment Breathing rates (m3/s) 0 - 8 hr 3.5E-4 8 - 24 hr 1.8E-4 1 - 30 days 2.3E-4 5.4. CR Occupancy Factors (Reference 6)
(Reference 6)
Values in fractions:
0-24 hr 1 -4 days 4-30 days 1.0 0.6 0.4
- 6.
Chemistry Data 6.1. Initial Pool pH - 5.3 (Reference 29)
Per reference 29, the suppression pool pH is between 5.3 - 8.6; low value was conservatively selected.
6.2. Mass of Chloride-Bearing Cable Insulation in Containment:
(Reference 30)
Hypalon: 1.673E6 grams Neoprene: 0.798E6 grams 6.3. Cable Dimensions (table taken from reference - EPR omitted because of no Cl content):
(Reference 30)
Hypalon Neoprene OD (cm) 2.979774 0.588988 Thickness (cm) 0.106905 0.106086 6.4. Mass of Sodium Pentaborate Available for Injection - 4062.8 Ibm 6.5. Formula of Sodium Pentaborate - Na2O-5B203-10H 20 (using natural boron) 6.6. Volume in RCS - 23,679 ft3 (Reference 31)
(Reference 31))
(Reference 8) 18694 R3
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Thermal-Hydraulic Data 7.1. Normal Operational Steam Line Temperature - 544 F (for MSIV Leakage)
(Reference 32)
Reference 32 provided the normal operational pressure of 1000 psi, from steam tables this corresponds to a saturation temperature of 544 F.
7.2. MSIV Test Conditions: Test Pressure > 25 psig (Reference 17) 7.3. Assumed Drywell Conditions for Converting Bypass SCFH to CFH (including MSIV):
Calc Pressure = 37.4 psig, Calc Temp. = 283 F (Reference 8) 7.4. Steam Release Accompanying MSLB Liquid Release - 25,000 Ibm (Reference 33) 7.5. RCS Pressure/Temperature (for MSLB Liquid Flash) - 1060 psia (552 F)
(Reference 7, sec 15.6.4) 7.6. Max Power/Min RCS Pressure for Operating Mech Vacuum Pump - 5%/400 psia (Reference 34)
- 8.
System-Related Data (Other than Volumetric Flows) 8.1. Elevation of CREF Filter Centerline - El 535' 4 3/8' (Reference 35) 8.2.
Elevation of Equipment Room Floor - El 525' (Reference 35) 8.3. Thickness of Floor (below El 525') - 1.0' (Reference 35) 8.4. Elevation of False Floor in Control Room - El 501' (Reference 36) 8.5. Secondary Containment Drawdown time - 20 Minutes (Reference 37) 8.6. Spray Initiation Time - 15 minutes (Reference 7, sec. 6.3.3.4)
Per FSAR Section 6.3.3.4 "System Performance During the Accident" operator action is not required during the short-term cooling period following the LOCA. During the long-term cooling period (after 10 minutes), the operator may take actions to:
- a. Use ECCS for vessel level control,
- c. Place systems into operation, such as containment cooling, standby liquid control, or drywell spray.
An assumption to credit drywell spray initiation in 15 minutes is a conservative time duration relative to the FSAR analysis for ECCS system performance during a LOCA which would allow operators to spray the drywell after 10 minutes. (Assumed and confirmed in dose analysis) 8.7. Spray System Parameters (Reference 38)
From DW Upper Header (level 562' 7.25') to DW floor (level 501':61' 7.25")
From DW Lower Header (level 518' 6") to DW floor (level 501': 17' 6")
8.8. Drywell Pressure Manual Spray Shutoff Pressure - 1.68 psig (Reference 39, p. 153) 8.9. Drywell Spray Initiation Based on Radiation - 14,000 R/Hr in Drywell (Reference 39, p. 152)
.1.
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BAR 8.10. Drywell Spray Initiation Limit Curve:
(Reference 40)
Columbia Generating Station Drywell Spray Initiation Limit U.
I.
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5 10 15 20 25 30 35 40 45 50 55 60 Drywell pressure, psig 8.11. Assumed Start Time for Filtered Control Room Intake for FHA - None (No Credit) 8.12. Maximum Time for MSIV Closure - 6 seconds Per reference 41, the MSIV closure time is 3-5 seconds.
Conservatively 6 seconds will be used in the analysis.
8.13. Blowout Panel Locations for MSLB Panels A and C (N. end of tunnel) & B and D (E. end of tunnel) 8.14. Distance from Panel A to CR Air Intake (via TGB) - -200' 8.15. Distance from Panels D to CR Air Intake - -240' 8.16. Dimensions of CREF HEPA Filter - 24" x 24" x 11 Y2" deep (in direction of flow)
(Reference 41)
(see Appendix A)
(see Appendix A)
(see Appendix A)
(Reference 42) 18694 R3
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BAR 8.17. Distance between Main Steam Line Centerlines at Rad Monitor Location - 7' 4" (Reference 43) 8.18. Maximum Safety Limit for MSL Rad Monitor - 4320 mR/hr (Reference 44)
"Maximum Analytical Value" 8.19. MSL Rad Monitor Location Relative to Steam Line:
Midpoint between Steam Lines A/B and C/D-Offset - 2' from Centerline (Reference 43) 8.20. Steam Line Dimensions at Rad Monitor Location:
(Reference 45, Table 2, p. 1-120)
- 26" OD x 1.125" wall
- 30" OD x 1.25" wall Material Specifications, Main Steam (MS), Primary Service Rating: ANSI 900#-4, ASME IlIl Class 2 8.21. Approximate Length of Steam Line "Seen" by Rad Monitor - 11 + feet (Reference 46) 8.22. RCS Inventory - 6.59E5 Ibm (Reference 7, Table 6.24) 8.23. Minimum Depth of Water above Damaged Fuel for FHA - 23' (See Appendix C) 18694 R3
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St Distance between the blowout panels and e control room local intake Blowout Panel B (Vertical)
(11917, 1264)
Steam Line Tunnel I
References DWG. S717 DWG. C500 Columbia Generating Station System Description -Secondary Containment, volume 9, chapter 1, sec IV (Fig. 5 & 6)
Distance = 1(1 1917 - 11874)2 + (1460 - 1264)2
= 200.6 ft = 61 m from panel A (release to TGB) to CR Local Intake
= 238 ft = 75 m from panel D to CR Local Intake (B and C release to ventway, then D to env) 18694 R3
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AKg.°f BAR Calculate the internal volume of a Main Steam line between the Main Steam Isolation Valves inside and outside containment when the valves are in the closed position References-Pipe Spool MSIV:
GE Drawing 131C8403, Rev. 1 (CVI 2-02B22-04,46)
Rockwell Drawings SKA-100378. Rev. 1 (CVI 2-02B22-99,1) and PD-4228B5, Sht. 1, Rev T (CVI 2-02-663400-1)
C:.D. = ZC t6 M 1.0I I('1 ZLi
( 4._1 ) -
f'7 q1(,I3 3)1 Q'i 4 4> M#,).
2s
'I b _
I 10 I
VALVI Jgou,4 ;6)e!t&
t4(t.5,'r-P. [Z+
-L (t 1t It
, 2 (. 4 3 0 0 4 70 RZT.
j Pc: '.s.,44 i 5.SOC&ASL) 4 (-23.5\\' s-s5o) =
zi 5evcJ 3
- 4 e
- 'f4I
~01,w4 !I V
C_
AP i. tCe PoW, I-L
)f wRE04
)t IWtiC s~>nI, 1
^f>V t5/
Z4-, o44 iiJ3 4I
- 2) Dis C.7
((,
.C&~-, 2')
i 6
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ENERGY Page IE-,; R Y Page Cont'd on page NORTHWEST APPENDIX C C-I People Vision Solutions Calculation No. NE-02-04-1 Prepared by / Date: LS Woosley Verified by/Date: M Abu-Shehadeh Revision No.
2 Background
0-eA-,JCA-
/2I iL/OL4 The design basis Fuel Handling Accident (FHA) involves a drop of an irradiated fuel assembly over the reactor vessel (Reference 1). At this location, the dropped assembly is assumed to fall approximately 34 feet, and damage is postulated to occur to both the dropped assembly and to some portion of those assemblies impacted in the reactor core. The extent of damage is directly proportional to the kinetic energy (i.e., the weight of the assembly times the free fall distance less the distance-integrated drag force) of the dropped assembly.
FHAs at other locations during the fuel movement have also been considered. For a drop in the fuel transfer area (between the reactor vessel and the spent fuel pool) or over the spent fuel pool, the kinetic energy would be significantly less. TS 3.9.6 (Reference 2) requires 22 feet of water above the reactor vessel flange. The normal water level is approximately 22 feet 9 inches (273 inches) (Reference 3). Licensee Controlled Specifications (LCS) 1.9.1 (Reference 4) specifies requirements for the Refueling Platform. Surveillance Requirement (SR) 1.9.1.3 requires verification of the uptravel electrical stop on the refueling platform that limits the height of the fuel handling crane to maintain a minimum of 7 feet 6 inches (90 inches) above the top of the active fuel in an assembly. The point of reference for this uptravel electric stop is the TS minimum required water level of 22 ft. This limit is needed to ensure at least 7 ft 6 inches of water is always above the top of active fuel to maintain adequate shielding for the radiological protection of the crane operator. This surveillance requirement is implemented via Procedure OSP-NSSE-C401 (Reference 5). The uptravel stop is set at 1 foot 6 inches (18 inches) below the bottom of the bail handle. This location equates to the top of active fuel. The nominal length of a fuel assembly is approximately 176 inches overall and the bail handle is approximately 4 inches in height (Reference 6). Given these TS/LCS limits and the other applicable dimensions, the nominal drop for a fuel bundle at a location above fuel pool or the transfer area would be: the depth of the water above the reactor vessel flange (same as the depth of the water above the spent fuel pool) minus the uptravel stop minus the length of the fuel assembly below the uptravel stop elevation. Referring to Figure 1 (page C-2) for clarity, the TS required normal water level is 22 feet.The length of the fuel assembly below the uptravel stop elevation would be: 176 inches minus 1 inch (for the bail handle) minus 1 foot 6 inches, which equals 157 inches. Therefore, the maximum credible drop outside of the reactor cavity area is 264 inches - 90 inches - 157 inches or 17 inches. For this 17 inch drop, the water depth available for decontamination would be approximately 22 feet.
Summary The design basis FHA analysis is based on a drop of 34 feet over the reactor vessel (versus a drop elsewhere where the maximum credible drop is 17 inches) which damages the dropped assembly as well as other recently irradiated assemblies in the core (versus a drop elsewhere where the dropped bundle would not likely sustain any damage based on empirical data from actual events in the industry and the drop elsewhere may or may not impact additional bundles which may or may not be recently irradiated). The decontamination factor credited for the design basis analysis is based on 23 feet of water. The credited water depth for the design basis FHA is essentially the same as the water depth for a drop elsewhere (approximately 1% difference under normal water level conditions and a maximum difference of approximately 4% at the TS minimum level).
Conclusion The design basis FHA analysis bounds any drop that could be postulated during the movement of irradiated fuel between the reactor vessel and the spent fuel pool. The current TS and design are adequate to ensure this conclusion.
References
- 1. Columbia FSAR 15.7.4, "Fuel Handling Accident," Amendment 57.
- 2. Columbia Technical Specifications, TS 3.9.6, 'Reactor Pressure Vessel (RPV) Water Level - Irradiated Fuel,"
Amendment 196.
- 3. Energy Northwest Calculation 245-NOME-CALC-0124, Rev. 00, Page A2 of A4.
- 4. Columbia Licensee Controlled Specifications, LCS 1.9.1, 'Refueling Platform," Rev. 27.
- 5. Energy Northwest Procedure OSP-NSSE-C401, "Refuel Platform Crane & Hoist Interlock Surveillance," Rev. 1.
- 6. Framatome ANP, Inc. document EMF-2865, Revision 1, Supplement 2, 'Mechanical and Thermal-Hydraulic Design Report for Columbia Generating Station Atrium-10 Fuel Assemblies," dated January 2004
- 7. 245-NOME-CALC-0124, Revision 00, Sketch of Vertical Elevations c
- 8. FSAR Figure 3.A.2.1-1, Primary and Secondary Containment Structure
- 9. NE-02-94-15 Fuel Handling Accident 18694 R3
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1 Figure 1 (Not to scale) v
- Iv Wat-ReS~dI I
I I
139 i
Ir
+ Reactor ' a Vessel amI r.
I CO Drawing Based on Figure 9.1-15 CGS FSAR Amendment 53 T T T 178"I Fuel Bundle*
'not to scale 1 8694 R3