ML063110168
| ML063110168 | |
| Person / Time | |
|---|---|
| Site: | Hope Creek  |
| Issue date: | 01/03/2003 |
| From: | Drucker M, Morrison G, Gita Patel NUCORE, Public Service Enterprise Group |
| To: | Document Control Desk, Office of Nuclear Reactor Regulation |
| References | |
| LCR H05-01, Rev. 1, LR-N06-0418 H-1-AB-MDC-1854, Rev 1IR0 | |
| Download: ML063110168 (31) | |
Text
Attachment 3 LR-N06-0418 LCR H05-01, Rev. I Calculation No. H-1-AB-MDC-1 854, Revision 1 IRO Main Steam Line Break Accident
CALC NO.: H-1-AB-MDC-1854 CALCULATION COVER SHEET Page 1 of 30 REVISION: I1R0 CALC. TITLE: I Main Steam Line Break Accident
- SHTS (CALC): 30 # ATT I # SHTS: 1111 # IDVI50.59 SHTS: 14/#'3 # TOTAL SHTS: 3 CHECK ONE:
El FINAL Z INTERIM (Proposed Plant Change) [ FINAL (Future Confirmation Req'd) El VOID SALEM OR HOPE CREEK El Q - LIST 0 IMPORTANT TO SAFETY El NON-SAFETY RELATED HOPE CREEK ONLY: ZQ E]Qs E]Qsh OF -IR 0 STATION PROCEDURES IMPACTED, IF SO CONTACT RELIABILITY ENGINEER CDs INCORPORATED (IF ANY): co Q 13 SO ra -o.ZooWRItE; c0 D--ol_ P A Q. ,
DESCRIPTION OF CALCULATION REVISION (IF APPL.):
Revised to include the EPU reactor coolant activity concentrations and TEDE dose criteria.
PURPOSE:
The purpose of this calculation is to determine the Exclusion Area Boundary (EAB), Low Population Zone (LPZ), and Control Room (CR) doses due to a-Main Stear Line. Break Accident (MS A) occurring outside containment using the Extended Power Uprate (EPU) reactor coolant activity concentrations and TEDE dose criteria. The thermal power level is expected to iicrea'-e to 4,03 i.MW%.
CONCLUSIONS:
The results of analysis in Section 8 indicate that the EAB, LPZ, and CR doses due to a MSLB accident are within their allowable TEDE dose limits. The results of a MSLBA indicate that CREF system initiation is not required during a MSLB accident.
ed vun9I I NzaA4~cr r~nnininn lQpvicinin 9 1
CALCULATION CONTINUATION SHEET SHEET 2 of 30 CALC. NO.: H-1-AB-MDC-1854
REFERENCE:
DCP 80048085 G. Patel, ORIGINATOR, DATE REV: 11t27/02 1 Mark Drucker, REVIEWER/VERIFIER, DATE 11/29/02 REVISION HISTORY Revision Revision Description 0 Initial Issue.
1 Revised to include the EPU reactor coolant activity concentrations and TEDE dose criteria. Updated control room volume is used IAW CD D506, Package No. 80027981.
The CREF is not credited in the analysis and uprated coolant activity is used. Therefore, the discrepancies addressing the CREF operation and source term are considered resolved IAW CD D501, Package Numbers 80032110 and 80033412.
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CALCULATION CONTINUATION SHEET SHEET 3 of 30 CALC. NO.: H-1-AB-MDC-1854
REFERENCE:
DCP 80048085 G. Patel, ORIGINATOR, DATE REV: 11/27/02 Mark Drucker, REVIEWER/VERIFIER, DATE I 1/29/02 PAGE REVISION INDEX PAGE REV PAGE REV 1 1 18 1 2 1 19 1 3 1 20 1 4 I 21 1 5 1 22 1 6 1 23 1 7 1 24 1 8 1 25 1 9 1 26 1 10 1 27 1 11 1 28 1 12 1 29 1 13 1 30 1 14 1 Attachment A 1 15 1 16 1
. 17 I_.... _ _.... . . . .. . . .... .. .. .
I Nuclear Common Revision 9 I
CALCULATION CONTINUATION SHEET SHEET 4 of 30 CALC. NO.: H-1-AB-MDC-1854
REFERENCE:
DCP 80048085 G. Patel, ORIGINATOR, DATE REV: 11/27/02 1 Mark Drucker, REVIEWERVERIFIER, DATE 1M29/02 TABLE OF CONTENTS Section Sheet No.
Cover Sheet 1 Revision History 2 Page Revision Index 3 Table of Contents 4 1.0 Purpose 5 2.0 Scope 5 3.0 Analytical Approach 5 4.0 Assumptions 8 5.0 Design Inputs 12 6..0 Calculaions 16 7.0 Results Summary 17 8.0 Conclusions 19 9.0 References 20 10.0 Tables 21 11.0 Figures 29 12.0 Affected Documents 30 13.0 Attachments 30 I Nuclear Common Revision 9 1I I ula omnRvso
CALCULATION CONTINUATION SHEET SHEET S of 30 CALC. NO.: H-1-AB-MDC-1854
REFERENCE:
DCP 80048085 G. Patel, ORIGINATOR, DATE REV: 11/27/02 Mark Drucker, REVIEWER/ERIFIER, DATE 1W129/02
1.0 PURPOSE
The purpose of this calculation is to determine the Exclusion Area Boundary (EAB), Low Population Zone (LPZ), and Control Room (CR) doses due to a Main Steam Line Break Accident (MSLBA) occurring outside containment using the TEDE dose criteria and Extended Power Uprate (EPU) reactor coolant activity concentrations. The thermal power level is expected to increase to 4,031 MWt.
2.0 BACKGROUND
The consequences of a MSLBA are analyzed using the plant specific design and licensing bases inputs, which are compatible to the TEDE dose criteria. The MSLBA analysis is performed using the guidance in Regulatory Guide 1.183, Appendix D (Ref. 9.1) and Standard Review Plan 15.6.4 (Ref. 9.6). There are no specific ESF functions credited in the analysis, including initiation of the CR emergency filtration (CREF) system to mitigate the CR dose.
3.0 ANALYTICAL APPROACH:
This analysis uses Version 3.02 of the RADTRAD computer code to calculate the potential radiological consequences of the MSLBA. The RADTRAD code is documented in NUREG/CR-6604 (Ref. 9.2).
The RADTRAD code is maintained as Software ID Number A-0-ZZ-MCS-0225, (Ref. 9.12).
Since no fuel damage occurs during the MSLBA at the Hope creek plant, the released activity is the maximum coolant activity allowed by technical specifications. The iodine concentration in the primary coolant is assumed corresponding to the following two cases in the standard technical specifications:
3.1. Pre-accident Iodine Spike The reactor coolant activity concentration for this case is assumed to be at the maximum value of 4.0 pCi/gm Dose Equivalent (DE) 1-131 permitted for a condition of a pre-accident spike (Ref. 9.11). The assumptions and design input parameters used for this release path are described in Sections 4.0 and 5.0.
The iodine scaling factors for the pre-accident iodine spike and equilibrium iodine concentration cases are calculated in Table 2 based on the maximum iodine concentrations of 4.0 p.Ci/g and 0.2 j+/-Ci/g using the following definition of 1-131 DE:
DOSE EQUIVALENT 1-131 shall be that concentration of 1-131, piCi/g, which alone would produce the same thyroid dose as the quantity and isotopic mixture of 1-131, 1-132,1-133,1-134, and 1-135 actually present.
The thyroid dose conversion factors are calculated in Table 1 using Federal Guidance Report 12 (Ref.
9.8) and corresponding isotopic iodine concentrations are calculated in Tables 3 & 4.
I Nuclear Common Revision 9
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CAMRTION CONTINUATION SHEET SHEET 6 of 30 CALC. NO.: H-1-AB-MDC-1854
REFERENCE:
DCP 80048085 G. Patel, ORIGINATOR, DATE REV: I1M27/02 I Mark Drucker, REVIEWERIVERIFIER, DATE 11129/02 The isotopic noble gas concentrations are calculated in Table 5 using the noble gas release rate at time t
= 0 sec (Ref. 9.15, Table V) and the uprated steam mass flow rate (Ref. 9.3, Section 3.2.1). The isotopic noble gas concentrations based on 100/E-BAR are calculated in Table 6 and listed in Table 7 using the following 100/E-BAR definition:
E-BAR shall be the average, weighted in proportion to the concentration of each radionuclide in the reactor coolant at the time of sampling, of the sum of the average beta and gamma energies per disintegration, in MeV, for isotopes, with half lives greater than 15 minutes, making up at least 95% of the total non-iodine activity in the coolant The use of RADTRAD code requires a volume node for the source activity released from a MSLB accident. Therefore, a source volume of 100 ft is introduced for a MSLBA release in a way that all activities released to environment in a single puff with a release rate of 2.OE+05 volume/day (see Figure 1). The reactor coolant mass of 140,000 pounds is assumed to release from the-MSLB (Ref 9.6, Section IMl.2.a). Although this release consists of two phase flow of water and steam mixture with different iodine concentrations in each phase, it is conservatively assumed that the reactor coolant iodine concentration are appropriate ftr .bthphase. Simila the noble gas concentrations are assumed equa for both phases. The isotopic activities available for release to the environment are calculated in
-Table 8 for the pre-accident iodine spike case.
3.2. Maximum Equilibrium Iodine Activity The reactor coolant concentration for this case is assumed to be at a value of 0.2 jLCi/gm DE 1-131 permitted for an equilibrium iodine activity for continued full power operation (Ref. 9.11). The specific release model, assumptions and design input parameters used in the analysis are same as those for the pre-accident iodine case (Sections 4.0 & 5.0) except the isotopic iodine concentrations are calculated based on 0.2 j+/-Ci/gm DE 1-131 in Table 4 and listed in Table 9 with the noble gas 100/E-BAR isotopic concentrations.
The potential post-MSLBA release paths are the blow out panels, south plant vent, and turbine building louvers, which are shown in Reference 9.13 with respect to the CR air intake with its tornado missile barrier. Since the MSLBA is a high energy line break accident, the pressure sensitive blow out panels would break open immediately to relieve the high pressure steam release. The X/Qs for these release paths are obtained from Reference 9.5 Section 8.0, and listed in the following table:
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SCALCULATION CONTINUATION SHEET SHEET 7 of 30 CALC. NO.: H-1-AB-MDC-1854
REFERENCE:
DCP 80048085 G. Patel, ORIGINATOR, DATE REV: 11/27/02 1 Mark Drucker, REVIEWER/VERIFIER, DATE 11/29/02 HCGS Control Room Time 95% Atmospheric Dispersion Factors (X/Qs) (s/m 3)
Interval South Plant TBL Blow Out Panel (hr) Vent (s/m3) (s/m3) 0-2 5.75E-04 6.17E-04 1.20E-03 2-8 3.84E-04 4.OOE-04 8.16E-04 8-24 1.40E-04 1.44E-04 3.08E-04 24-96 9.08E-05 1.OOE-04 2.14E-4)96-720 7.01E-05 7.49E-05 1.63E-04 Comparison of X/Qs in the above table indicates that the blow out panel release path is the most limiting
- l~tl*for the release path post-MSLBAerelease.
B*'*'Ut using calculatedto-*
-_ Therefore, the CR dose isposi the pst-MSLBA_.ly-r--df'*-fdfedkt-we--n-tan~tan as a single puff, the CREF is not credited. The CR is assumed to be in the normal mode of operation for the entire duration of the accident.
The RADTRAD V3.02-(Ref:9:2)-default-ntfid-iwentoiyflle(NIF)BW,_ef. NIF i-iii6difi-d bifd..
on the activity releases to the environment from the MSLBA as shown in Tables 8 & 9. The.plant-,.
specific NIFs HEPU4MSLB_deftxt and HEPU2MSLB..deftxt are further modified to include Kr-83m, Xe-131m, Xe-133m, Xe-135m, and Xe-138 isotopes, which are critical for a puff release. The modified RADTRAD3.02 dose conversion factor (DCF) and Release Fraction and Timing (RFI) Files HEPUMSLBFG1 1&12.txt and HEPUMSLBRFT.txt are used for the MSLBA analysis.
The EAB, LPZ, and CR doses are shown for both cases in Section 7.0 and compared with the allowable dose limits.
I Nuclear Common Revision 9 1 Nu.ler Comon.Reisio 9 ....
SCALCULATION CONTINUATION SHEET SHEETS of 30 CALC. NO.: H-1-AB-MDC-1854
REFERENCE:
DCP 80048085 G. Patel, ORIGINATOR, DATE REV: 11/27/02 1 Mark Drucker, REVIEWER/VERIFIER, DATE 11/29/02
4.0 ASSUMPTIONS
Assumptions for Evaluating the Radiological Consequences of a MSLBA The assumptions in these sections are acceptable to the NRC staff for evaluating the radiological consequences of a MSLBA. These assumptions supplement the guidance provided in Regulatory Guide 1.183, Appendix D (Ref. 9.1). These assumptions are incorporated as design inputs in Sections 5.3.1 through 5.3.4 for the MSLBA analysis.
SOURCE TERM 4.1 Per Reference 9.1, Appendix D, Section 2, since no or minimal fuel damage is postulated for the iting event, the released activity is the maximum coolant activity allowed by technical specification.
The iodine concentration in the primary coolant is assumed to correspond to the following two cases in
.. -the nuclear-steam-supply-system -vendofs standard-technical-specifications.;----- -
4.1.1 The maximum value of reactor coolant concentration typically permitted for an assumed pre-accident spike (Ref. 9.1, Appendix D, Section 2.1), which corresponds to 4.0 pCi/gmr DE 1-131 for the Hope Creek plant (Ref. 9.11), and 4.1.2 The maximum equilibrium value of reactor coolant concentration typically permitted for continued full power operation (Ref. 9.1, Appendix D, Section 2.2), which corresponds to 0.2 p+/-Ci/gm DE 1-131 for the Hope Creek plant (Ref. 9.11).
4.1.3 Per Reference 9.1, Appendix D, Section 3, the activity released from the fuel is assumed to mix instantaneously and homogenously in the reactor coolant. Noble gases are assumed to enter the steam phase instantaneously.
TRANSPORT 4.2 The total mass of coolant released is assumed to be that amount in the steam line and connecting lines at the time of the break plus the amount that passes through the valves prior to closure (Ref. 9.1, Appendix D, Section 4.2). The reactor coolant mass of 140,000 lbs is assumed to be released to the environment (Ref. 9.6, Section HI.2.a).
I Nuclear Common Revision 9 I Nuclear.Common.R.v.sion..
T CALCULATION CONTINUATION SHEET SHEET 9 of 30 CALC. NO.: H-1-AB-MDC-1854
REFERENCE:
DCP 80048085 G. Patel, ORIGINATOR, DATE REV: 11/27/02 1 Mark Drucker, REVIEWERIVERIFIER, DATE 11/29/02 4.3 All the radioactivity in the released coolant is assumed to be released to the atmosphere instantaneously as a ground-level release. No credit is assumed for plateout, holdup, or dilution within facility buildings (Ref.9.1, Appendix D, Section 4.3).
4.4 The iodine species released from the main steam line is assumed to be 95% CsI as an aerosol, 4.85%
elemental, and 0.15% organic (Ref. 9.1, Appendix D, Section 4.4).
Offsite Dose Consequences:
The following guidance is used in determining the TEDE for a maximum exposed individual at EAB and LPZ locations:
45 Ti or any to--hou theri ai lease is determined (Ref. 9.1, Section 4.1.5), and used in determining compliance with the dose acceptance criteria in Reference 9.1, Section 4.4, Table 6:
EAB Dose Acceptance Criterion (pre-accident spike case): 25 Rem TEDE EAB Dose Acceptance Criterion (equilibrium iodine activity case): 2.5 Rem TEDE 4.6 The breathing rates for persons at offsite locations are given in Reference 9.1, Section 4.1.3, and are incorporated in Design Input 5.3.4.
4.7 The maximum Low Population Zone (LPZ) TEDE is determined for the most limiting receptor at the outer boundary of the LPZ (Ref. 9.1, Section 4.1.6), and used in determining compliance with the dose criteria in Reference 9.1, Section 4.4 Table 6" LPZ Dose Acceptance Criterion (pre-accident spike case): 25 Rem TEDE LPZ Dose Acceptance Criterion (equilibrium iodine activity case): 2.5 Rem TEDE 4.8 No correction is made for depletion of the effluent plume by deposition on the ground (Ref 9.1, Section 4.1.7).
Control Room Dose Consequences I Nuclear Common Revision 9 1 NucearComon.eviion9 .
CALCULATION CONTINUATION SHEET SHEET 10 of 30 CALC. NO.: H-1-AB-MDC-1854
REFERENCE:
DCP 80048085 G. Patel, ORIGINATOR, DATE REV: 11/27/02 1 Mark Dnzcker, REVIEWER/VERIFIER, DATE 11/29/02 The following guidance is used in determining the TEDE for maximum exposed individuals located in the control room:
4.9 The CR TEDE analysis considers the following sources of radiation that will cause exposure to control room personnel (Ref 9.1, Section 4.2.1):
- Contamination of the control room atmosphere by the intake or infiltration (i.e., filtered CR ventilation inflow via the CR air intake, and unfiltered inleakage) of the radioactive material contained in the post-accident radioactive plume released from the facility,
. Contarination-f Qntrotro mnatosp1here.bytheintake orýnfiltrtion-(i.e`r-fltered CR *
- ventilation inflow via the CR air intake, and unfiltered inleakage) of airborne radioactive material from areas and structures adjacent to the control room envelope, Radiation shine from the external radioactive plume released from the facility.(i.e., external airborne cloud shine dose),
Radiation shine from radioactive material in the reactor containment (i.e., containment shine dose; not applicable to a MSLB occurring outside containment),
Radiation shine from radioactive material in systems and components inside or external to the control room envelope, e.g., radioactive material buildup in recirculation filters (i.e., CR filter shine dose).
Note: The external airborne cloud shine dose and the CR filter shine dose due to a MSLBA are insignificant compared to those due to a LOCA (see the core release fractions for LOCA and non-LOCA design basis accidents in Tables 1 and 3 of Reference 9.1). Therefore, these direct dose contributions are considered to be insignificant and are not evaluated for a MSLBA.
4.10 The radioactivity material releases and radiation levels used in the control room dose analysis are determined using the same source term, transport, and release assumptions used for determining the I Nuclear Common Revision evso.
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CALCULATION CONTINUATION SHEET SHEET 11 of 30 CALC. NO.: H-1-AB-MDC-1854
REFERENCE:
DCP 80048085 G. Patel, ORIGINATOR, DATE REV: 11M27/02 1 Mark Drucker, REVIEWER/ERIFIER, DATE 11W29/02 exclusion area boundary (EAB) and the low population zone (LPZ) TEDE values (Ref 9.1, Section 4.2.2).
4.11 The occupancy and breathing rate of the maximum exposed individual present in the control room are incorporated in Design Input 5.3 (Ref. 9.1, Section 4.2.6).
4.12 10 CFR 50.67 (Ref 9.4) establishes the following radiological criterion for the control room:
CR Dose Acceptance Criterion: 5 Rem TEDE 4.13 Although allowed by Reference 9.1, Section 4.2.4, credit is not taken for the engineered safety features 44of R iipilsor (Ciro sye thatnf.i9igat1 airborne activity within the control room.
4.14 No credits for KI pills or respirators are taken (Ref. 9. 1, Section 4.2.5).
INuclear Common Revision 9 I Nula omnRvso
CALCULATION CONTINUATION SHEET SHEET 12 of 30 CALC. NO.: H-1-AB-MDC-1854
REFERENCE:
DCP 80048085 G.Patel, ORIGINATOR, DATE REV: 11/27/02 1 Mark Drucker, REVIEWER/VERIFIER, DATE 11/29/02 5.0 DESIGN INPUTS:
5.1 General Considerations 5.1.1 Applicability of Prior Licensing Basis The implementation of an AST is a significant change to the design basis of the facility and to the assumptions and design inputs used in the analyses. The characteristics of the ASTs and the revised TEDE dose calculation methodology may be incompatible with many of the analysis assumptions and methods currently used in the facility's design basis analyses. The HCGS plant specific design inputs and assumptions used in the current TIDA1484-4 analyseswere assessed for theirvalidity-to-represent the-as-built condition of the. plant and evaluated for their compatibility to meet the AST and TEDE methodology. The analysis in this calculation ensures that analysis assumptions, design inputs, and methods are compatible with the ASTs and the TEDE criteria.
5.1.2 Credit for Engineered Safety Features Credit is taken only for accident mitigation features that are classified as safety-related, are required to be operable by technical specifications, are powered by emergency power sources, and are either automatically actuated or, in limited cases, have actuation requirements explicitly addressed in emergency operating procedures. The dose mitigation function of the CREF system is not credited in the analysis.
5.1.3 Assignment of Numeric Input Values The numeric values that are chosen as inputs to the analyses required by 10 CFR 50.67 (Ref. 9.4) are compatible to AST and TEDE dose criteria and selected with the objective of producing conservative radiological consequences. For conservatism, the limiting values of reactor coolant iodine concentrations listed in the HCGS Technical Specification are used in the analysis.
5.1.4 Meteorology Considerations The control room atmospheric dispersion factors (X/Qs) for the blowout panel release point are developed (Ref.
9.5) using the NRC sponsored computer code ARCON96. The EAB and LPZ X/Qs were reconstituted using the Rei io9 Nucea. Como I Nuclear Common Revision 9 1
TCALCULATION CONTINUATION SHEET SHEET 13 of 30 CALC. NO.: H-1-AB-MDC-1854
REFERENCE:
DCP 80048085 ORIGINATOR, DATE REV: I1/27/02 I Mark Drucker, REVIEWERNERIFIER, DATE I M29/02 HCGS plant specific meteorology and appropriate regulatory guidance (Ref. 9.9). The off-site XIQs reconstituted in Reference 9.9 were accepted by the staff in previous licensing proceedings.
52 Accident-Specific Design Inputs/Assumptions The design inputs/assumptions utilized in the EAB, LPZ, and CR habitability analyses are listed in the following sections. The design inputs are compatible with the AST and TEDE dose criteria and assumptions are consistent with those identified in Appendix D of RG 1.183 (Ref.9.1). The design inputs and assumptions in the following sections represent the as-built design of the plant.
Revision 9 Common I Nuclear Common Revision 9 i W-7
CAL4CULATION CONTINUATION SHEET SHEET 14 of 30 CALC. NO.: H-1-AB-MDC-1854
REFERENCE:
DCP 80048085 G. Patel, ORIGINATOR, DATE REV: 11t27/02 Mark Drucker, REVIEWERNERIFIER, DATE 1W129/02 Design Input Parameter Value Assigned Reference 5.3 Main Steam Line Break Accident Parameters 5.3.1 Source Term 5.3.1.1 Proposed extended power 4,031 MWt 9.3, Section 3.2.1 uprate levelI 5.3.1.2.a Uprated Iodine Coolant Concentration (tCi/gm) 9.3, Appendix A Isotope Activity Isotope Activity Isotope Activity 1-131 1.30E-02 1-132 1.20E-01 1-133 8.90E-02 1-134 2.40E-01 1-135 1.30E-01 5.3.1.2.b Uprated Noble Gas Release Rate @ time t = 0 (pCi/sec) 9.15, Table V KR-83M 3.40E+03 KR-88 2.OOE+04 XE-135M 2.60E+04 KR-85M
... 6.1E03 .E31M . 1.50E40-I1 -- XE-1435 2.20E+04 KR-85 2.OOE+01 XE-133M 2.90E+02 XE-138 8.90E+04 Kr-87 2.OOE+04 XE-133 8.20E+03 5.3.1.3 Maximurn reactor coolant 4.0 piCi/gm 9.11 iodine concentration for pre-accident spike 5.3.1.4 Maximum equilibrium 0.2 ptCi/gm 9.11 reactor coolant iodine concentration for continued full 5.3.1.5 Mass of reactor coolant 140,000 lbs 9.6, Section lI.2.a released from MSLBA 5.3.2 Activity Transport (see Figure 1) 5.3.2.1 Activity release rate 2.0E+05 source volumes/day Assumed to postulate a single puff 5.3.2.2 Duration of release Instantaneously in a single puff 9.1, Table 6 and Appendix D, Section 4.3 5.3.2.3 Type of release to the Ground level release 9.1, Appendix D, Section 4.3 atmosphere 1 1 5.3.2.4 Chemical form of Iodine in reactor coolant released from the main steam line Aerosol 95% 9.1, Appendix D, Section 4.4 Elemental 4.85%
Organic 0.15%
5.3.2.5 Dilution or holdup within Not credited 9.1, Appendix D, Section 4.3 the facility building 5.3.2.6 Source volume 100 fi Assumed to facilitate RADTRAD nodalization 5.3.3 Control Room Parameters (see Figure 1) 5.3.3.1 CR volume 85,000 W 9.10, page 10 Nuclear Common Revision 9
CALCULATION CONTINUATION SHEET SHEET 15 of 30 CALC. NO.: H-1-AB-MDC-1854
REFERENCE:
DCP 80048085 G. Patel, ORIGINATOR, DATE REV: 11/27/02 1 Mark Drucker, REVIEWERNERIFIER, DATE 11/29/02 Design Input Parameter Value Assigned Reference 5.3.3.2 CR normal air inflow rate 3,000 +/- 10% cfm for 0-720 hrs 9.14 and Assumption 4.13 during MSLBA (conservatively modeled as 3,300 cfin) 5.3.3.3 CR occupancy factors Time (Hr) % 9.1, Section 4.2.6 0-24 100 24-96 60 96-720 40 5.3.3.4 CR atmospheric dispersion factors for blowout panel release (X/Qs)
Time (Hr) X/Q (sec/ml) 0-2 1.20E-03 9.5, Section 8.8
........ ..2-8 -....... ....... ........... ....... 8.16-6F 04 - -- -- - -- - - .A-...- .
8-24 3.08E-04 24-96 2.14E-04 96-720 1.63E-04 5.3.3.5 CR breathing rate 3.5E-04 9.1, Section 4.2.6 3
(m /sec) 5.3.4 Site Boundary Release Model Parameters 5.3.4.1 EAB atmospheric - 1.9E-04 3
dispersion factor (X/Q) (sec/m )
-5.3.4.2LPZ-atmospheric-dispersion factors (XIQs)-
Time (Hr) X/Q (se/nmi) 0-2 1.9E-05 9.9, Pages 5 & 9 2-4 1.2E-05 4-8 8.OE-06 8-24 4.OE-06 24-96 1.7E-06 96-720 4.7E-07 5.3.4.3 EAB breathing rate 3.5E-04 9.1, Section 4.1.3 (m3/sec) 5.3.4.4 LPZ breathing rates (m3 /sec)
Time (Hr) (m0/sec) 0-8 3.5E-04 9.1, Section 4.1.3 8-24 1.8E-04 24-720 2.3E-04 Nuclear Common Revision 9
CALCULATION CONTINUATION SHEET SHEET 16 of 30 CALC. NO.: H-1-AB-MDC-1854
REFERENCE:
DCP 80048085 G. Patel, ORIGINATOR,4 DATE REV:- 11/270 Mark D*cker, REVIEWER/VERIFIER, DATE I I 29/02
6.0 CALCULATIONS
Miscellaneous Conversion Factors Steam Mass Flow Rate:
Uprated Steam Flow Rate
= 17,774,000.0 lb/hr (Ref. 9.3, Section 3.2.1) = 17,774,000.0 lb/hr x 453.6 g/ib x 1/3600 hr/sec
= 2,239,524.0 g/sec = 2.24E+06 g/sec This conversion factor is used to convert the noble gas release rates in IiCi/sec to noble gas activity concentrations in pCi/g in Table 5.
Coolant Mass Release:
Coolant Mass Release From MSLB
= 140,000 lb (Ref. 9.6, Section mI.2.a) = 140,000 lb x 453.6 g/lb = 6.35E+07 g This conversion factor is used in Tables 8 & 9.
eiso Nula omo I Nuclear Common Revision 9 1
CALCULATION CONTINUATION SHEET SHEET 17 of 30 CALC. NO.: H-1-AB-MDC-1854
REFERENCE:
CP 80048085 G. Patel,,
ORIGINATOR, DATE REV: 11/27/02 1 Mark Drucker, REVIEWERNERIFIER, DATE 11/29/02 7.0 RESULTS
SUMMARY
7.1 The results of the MSLBA analysis with the pre-accident iodine spike are summarized in the following table:
Main Steam Line Break Accident with Pre-accident Iodine Spike TEDE Dose (rem)
Receptor Location Control Room EAB LPZ Calculated Dose 3.60E+00
_ 9.42E-01 (0.ohr) 9.45E-02 I Allowable TEDE Limit 5.OE+00 .. 2.SE+01 .... ..SE;01.
RADTRAD Computer Run No.
IEPU4MSLBAOO HEPU4MSLB00 HEPU4MSLBOO I Significant assumptions used in this analysis:
- Post-MSLBA activity is released to the environment in a single puff at ground level through blowout panels.
- CREF system is not credited.
- Core thermal power = 4,03 MWt I I Nuclear Common Revision 9 I
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CALCULATION CONTINUATION SHEET SHEET 18 of 30 CALC. NO.: H-1-AB-MDC-1854
REFERENCE:
DCP 80048085 G. Patel, ORIGINATOR, DATE REV: I 1/27/02 1 Mark Drucker, REVIEWERIVERIFIER, DATE 11/29/02 7.2 The results of the MSLBA analysis with the maximum equilibrium iodine concentration permitted for I
continued full power operation are summarized in the following table:
Main Steam Line Break Accident with Maximum Equilibrium Iodine Concentration for Continued Full Power Operation TEDE Dose (rem)
Receptor Location Control Room EAB LPZ Calculated Dose 1.81E-01 5.61E-02 5.63E-03
_ _ _ __ __ _ _ _ _ _ _ _ _ _ _ _ _(0.0 hr) _ _ _ _
Allowable TEDE Limit 5.OE+00 2.5E+00 2.5E+00 RADTRAD Computer Run No.
HEPU2MSLBOO HEPU2MSLBOO HEPU2MSLB00 Significant assumptions used in this analysis:
- Post-MSLBA activity is released to the environment in a single puff at ground level through blowout panels
- CREF system is not credited.
- Core thermal power = 4,031 MWt I r ,....
I Nuclear Common Revision 9 1 eiso I Nula omo
CALCULATION CONTINUATION SHEET SHEET 19 of 30 CALG NO.: H-1-AB-MDC-1854
REFERENCE:
DCP 80048085 G. Patel, ORIGINATOR, DATE REV: 11/27/02 1 Mark Drucker, REVIEWER/VERIFIER, DATE 11/29/02
8.0 CONCLUSION
S:
The results of MSLB accident analyses in Section 7.0 indicate that the EAB, LPZ, and CR doses due to a MSLB accident are within their allowable limits and CREF system actuation is not required during a MSLB accident.
I Nuclear ommon Revision 9 .1 e .so 9 Nuler..mon.
CALCULATION CONTINUATION SHEET SHEET 20 of 30 CALC. NO.: H-1-AB-MDC-1854
REFERENCE:
DCP 80048085 G. Patel, ORIGINATOR, DATE REV: I /27/02 1 Mark Drucker, REVIEWER/VERIFIER, DATE IIt29/02
9.0 REFERENCES
- 1. U.S. NRC Regulatory Guide 1.183, Alternative Radiological Source Terms for Evaluating Design Basis Accidents at Nuclear Power Reactors, July 2000
- 2. S.L. Humphreys et al., "RADTRAD: A Simplified Model for Radionuclide Transport and Removal and Dose Estimation," NUREG/CR-6604, USNRC, April 1998
- 3. GE-NE-0000-0008-3534-02, DRF 0000-0004-6923, Revision 0, Class M, November 2002, Project Task Report T0807 Draft, Coolant Radiation Sources
- 4. 10 CFR 50.67, "Accident Source Term."
- 5. Calculation No. H-1-ZZ-MDC-1879, Rev 1, Control Room & Technical Support Center X/Qs Using ARCON96 Code
- 6. NUREG-0800, Standard Review Plan 15.6.4, Revision 2, "Radiological Consequences of Main Steam Line Failure Outside Containment (BWR)," July 1981.
- 7. Federal Guidance Report 11, EPA-520/1-88-020, Environmental Protection Agency
- 8. Federal Guidance Report 12, EPA-402-R-93-08 1, Environmental Protection Agency
- 9. Calculation No. H-1-ZZ-MDC-1820, Rev 0, Offsite Atmospheric Dispersion Factors
- 10. Calculation No. H-1-ZZ-MDC-1882, Rev 0, Control Room Envelope Volume
- 11. HCGS Technical Specification 3/4.4.5, "Specific Activity" Limiting Condition for Operation
- 12. Critical Software Package Identification No. A-0-ZZ-MCS-0225, Rev 0, RADTRAD Computer Code.
- 13. HCGS General Arrangement Drawings:
- a. P-0006-0, Rev 7, Plan EL 153'-0" & EL 162'-0"
- b. P-0007-0, Rev 7, Plan EL 171'-0" & EL 201'-0"
- c. P-0010-0, Rev 6, Sections A-A & B-B
- d. P-001 1-0, Rev 5, Sections C-C & D-D
- 14. HCGS Air Flow Diagram No. M-78-1, Rev 21, "Aux Bldg Control Area Air Flow Diagram."
- 15. GE Specification Document No. 22A2703F, Rev 3, Radiation Sources.
I Nuclear Common Revision 9 1 l
NularCmmnReiio
CALCULATION CONTINUATION SHEET SHEET 21 of 30 CALC. NO.: H-1-AB-MDC-1854
REFERENCE:
DCP 80048085 G. Patel, ORIGINATOR, DATE REV: 11I27/02 1 Mark Drucker, REVIEWER/VERIFIER, DATE 11Y29/02 10.0 TABLES:
I Table 1 Iodine Isotopic Dose Conversion Factors Isotopic Conversion Iodine Dose Factor Dose Isotope Conversion Conversion Factor Factor (Sv/Bq) (rem/CitSv/Bq) (rem/C)
A B C-AxB 1-131 2.920E-07 3.700E+12 1.080E+06 1-132 1.740E-09 3.700E+12 6.438E+03 1-133 4.860E-08 3.700E+12 1.798E+05 1-134 2.8801-10 3.700E+12 1.066E+03 1-135 8.460E-09 3.700E+12 3.130E+04 A From Reference 9.7, Page 136 Table 2 Iodine Scaling Factors Pre-accident Iodine Spike & Equilibrium Iodine Concentration Normal Iodine Iodine Dose Product Isotope Activity Conversion Concentration Factor pCLrem/Ci.g PCI/g (reim/CQ (rem)
A B (AxB) 1-131 1.300E-02 1.080E+06 1.404E+04 1-132 1200E-01 6A38E+03 7.726E+02 1-133 8.900E-02 1.798E+05 1.600E+04 1-134 2.4001-01 1.066E+03 2.558E+02 1-135 1.300E-01 3.130E+04 4.069E+03 Total 3.514E+04 A From Reference 9.3, Appendix A
[ 1-131 DE Based on Normal Iodine Concentration 3.254E-02 [
Iodine Scaling Factor Based on 4 gCi/g DE 1-131 1.229E+02 Iodine Scaling Factor Based on 0.2 pCl/g DE 1-131 6.147E+00 I Nudear Common Revision 9 ]
eiso Nula omo
CALCULATION CONTINUATION SHEET SHEET 22 of 30 CALC. NO.: H-1-AB-MDC-1854
REFERENCE:
DCP 80048085 ORIGINATOR, DATE REV: 11t27/02 1 Mark Drucker, REVIEWER/VERIFIER, DATE 11/29/02 Table 3 I
Iodine Concentrationn Based On Pre-accident Iodine Spike Normal Iodine Iodine Iodine Scaling Activity Isotope Activity Factor Concentration Concentration A B C-AxB 1-131 1.300-02 1.229E+02 1.598E+00 1-132 1200-1 1.229E+02 1.475E+01 1-133 8.90013-02 1.229E+02 1.094E+01 1-134 2.400E-01 1.229E+02 2.951E+01 1-135 1.30013-01 1-229E+02 1.598E+01 A From Reference 9.3, Appendix A B Scaling Factor Based on 4 pCi/g DE 1-131 From Table 2 Table 4 Iodine Concentration Based On Equilibrium Iodine Concentration Normal Iodine Iodine Iodine Scaling Activity Isotope Activity Factor Concentration Concentration A B C=AxB 1-131 1.300E-02 6.147E+00 7.991E-02 1-132 1200E-01 6.147E+00 7.376E-01 1-133 8.900E-02 6.147E+00 5.471E-01
-134 2.4001-01 6.147E+00 1.475E+00 1-135 1.300E-01 6.147E+00 7.991E-01 A From Reference 9.3, Appendix A B Scaling Factor Based on 0.2 gCi/g DE 1-131 From Table 2 eiin9I Nula omn I Nuclear Common Revision 9 I
CALCULATION CONTINUATION SHEET SHEET 23 of 30 CALC. NO.: H-1-AB-MDC-1854
REFERENCE:
DCP 80048085 G. Patel, ORIGINATOR, DATE REV: 11/27/02 Mark DDucker, REVIEWER/VERIFIER, DATE W129/02 Table 5 Normal Noble Gas Concentration Noble Gas Uprated Normal Release Rate Steam Mass Noble Gas Isotope At t -0 Flow Rate Activity (ACi/sec) (g/sec) Concentration (0lCi/g)
A B C-A/B Kr-83m 3.400E+03 2.240E+06 1.518E-03 Kr-85m 6.100E+03 2.240E+06 2.724E-03 Kr-85 2.000E+01 2.240E+06 8.93 1E-06 Kr-87 2.000E-04 2.240E+06 8.93 1E-03 Kr-88 2.OOOE+04 2.240E+06 8.931E-03 Xe-131m 1.500E+01 2.240E+06 6.698E-06 Xe-133m 2.900E+02 2.240E+06 1.295E-04 Xe-133 8.200E+03 2.240E+06 3.662E-03 Xe-135m 2.600E+04 2.240E+06 1.161E-02 Xe-135 2.200E+04 2.240E+06 9.824E-03 Xe-137 1.500E+05 2.240E+06 6.698E-02 Xe-138 S.900E+04 2.240E+06 3.974E-02 A From Reference 9.15, Table V B = 17774000 lb/hr x 453.6 g/lb x 1/3600 hr/sec = 2.240E-06 g/sec I Nuclear Common Revision 9 1 Nula omo eiso !
CALCULATION CONTINUATION SHEET SHEET 24 of 30 CALC. NO.: H-1-AB-MDC-1854
REFERENCE:
DCP 80048085 G. Patel, ORIGINATOR, DATE REV: 11/27/02 I Mark Drucker, REVIEWERVERIFIER, DATE 11/29/02 Table 6 HCGS Reactor Coolant Concentration Based on 100E-BAR Normal Average Energy Weighted EPU Mev/Dis Energy Isotope Activity Beta Gamma Total E-Bar Concentration Mev.FtCi/dis.g pCi/g Al. Bi Ci Di-Bi+Ci Ei-Ai*Di Br-83 1.50E-02 0.321 0.008 0.329 0.0049 Br-84 2.70E-02 1.229 1.788 3.017 0.0815 Kr-83m 1.52E-03 0.039 0.003 0.042 0.0001 Kr-85m 2.72E-03 0.255 0.158 0.413 0.0011 Kr-85 8.93E-06 0.251 0.002 0.253 0.0000 KR 87 8.93E-03 1.324 0.793 2.117 0.0189 KR 88 8.93E-03 0.364 1.955 2.319 0.0207 Xe-131m 6.70E-06 0.144 0.020 0.164 0.0000 Xe-133m 1.29E-04 0.192 0.041 0.233 0.0000 Xe-133 3.66E-03 0.136 0.046 0.182 0.0007 Xe-135m 1.16E-02 0&098 0.429 0.527 0.0061 Xe-135 9.82E-03 0.317 0.249 0.566 0.0056 Sr-89 3.1.E-03 0.583 0.000 0.583 0.0018 Sr-90 2.30E-04 0.196 0.000 0.196 0.0000 Sr-91 6.90E-02 0.656 0.697 1.353 0.0934 Sr-92 1.1OE-01 0.196 1.339 1.535 0.1689 Zr-95 4.OOE-05 0.116 0.739 0.855 0.0000 Zr-97 3.20E-05 0.700 0.179 0.879 0.0000 Nb-95 4.20E-05 0.044 0.766 0.810 0.0000 Mo-99 2.20E-02 0.392 0.150 0.542 0.0119 Tc-99m 2.80E-01 0.016 0.126 0.142 0.0396 Ru-103 1.90E-05 0.075 0.469 0.544 0.0000 Ru-106 2.60E-06 0.010 0.000 0.010 0.0000 Te-129m 4.OOE-05 0.260 0.038 0.298 0.0000 Te-132 4.90E-02 0.102 0.234 0.336 0.0165 Cs-134 1.60E-04 0.164 1.555 1.719 0.0003 Cs-136 1.1OE-04 0.139 2.166 2.305 0.0003 Cs-137 2.40E-04 0.187 0.000 0.187 0.0000 Cs-138 1.90E-01 1.207 2.361 3.568 0.6779 Ba-139 1.60E-01 0.898 0.043 0.941 0.1506 Ba-140 9.00E-03 0.313 0.183 0.496 0.0045 Revision 9 1 I Nuclear Common Revision 9 I Nuclear Common
CALCULATION CONTINUATION SHEET SHEET 25 of 30 CALC. NO.: H-1-AB-MDC-1854
REFERENCE:
DCP 80048085 G. atel, ORIGINATOR, DATE REV: 11/27/02 1 Mark Drucker, REVIEWER/VERIFIER, DATE 11/29/02 Table 6 (Con'd)
HCGS Reactor Coolant Concentration Based on 100AE-BAR Normal Average Energy Weighted EPU Mev/Dis Energy Isotope Activity Beta Gamma Total E-Bar Concentration Mev.pCi/dis.g 11Ci/g Al Bi CI DI-Bi+Ci EI-AM*Di Ba-141 1.70E-01 0.901 0.845 1.746 02968 Ce-141 3.90E-05 0.171 0.076 0.247 0.0000 Ce-143 3.501-05 0.433 0282 0.715 0.0000 Ce-144 3.50E-05 0.092 0.021 0.113 0.0000 Pr-143 3.801-05 0.314 0.000 0.314 0.0000 Nd-147 1.4DB-05 0270 0.140 0.410 0.0000 Np-239 2.401-01 0260 0.173 0.433 0.1039 Na-24 2.001-03 0.554 4.121 4.675 0.0094 P-32 2.001-05 0.695 0.000 0.695 0.0000 Cr-51 5.001-04 0.004 0.033 0.036 0.0000 Mn-54 4.001-05 0.004 0.836 0.840 0.0000 Mn-56 5.00E-02 0.830 1.692 2.522 0.1261 Co-58 5.00E-03 0.034 0.976 1.009 0.0050 Co-60 5.00E-04 0.097 2.504 2.601 0.0013 Fe-59 8.00E-05 0.118 1.189 1.307 0.0001 Ni-65 3.OOE-04 0.017 0.000 0.017 0.0000 Zn-65 2.00E-06 0.007 0.584 0.591 0.0000 Zn-69m 3.0DB-O5 0.022 0.417 0.439 0.0000 Ag-i 1Dm 6.001-05 0.072 2.751 2.823 0.0002 W-187 3.00E-03 0.312 0.481 0.793 0.0024 F-I8 4.OOE-03 0.250 1.022 1272 0.0051 Total 1.46E+00 Total 1.86E+00 Ai From Reference 9.15, Tables II & IV Bi & Ci From Reference 9.8, Appendix A for isotope having half life > 15 minutes E-BAR - SUM (Weighted E-Bar)/Sum (Al) 1.273 100/E-BAR Coolant Concentration 78.571 Percent Fuel Defect Based on E-BAR - (100/E-BAR)/Sum (Al) 53.887 omo eiso Nula I Nuclear Common Revision 9 1
CALCULATION CONTINUATION SHEET SHEET 26 of 30 CALC NO.: H-1-AB-MDC-1854
REFERENCE:
DCP 80048085 G. Patel, ORIGINATOR, DATE REV: 1127/02 1 Mark Drucker, REVIEWER/VERIFIER, DATE 1129/02 Table 7 Normal Noble Gas Concentration Based on 100/E-BAR Normal Noble Gas Noble Gas Noble Gas Scaling Concentration Isotope Activity Factor Based On Concentration Based On 100/E-BAR (uCI/g) 1O0/2-BAR (jLCLg)
A B C-AxB Kr-83m 1.518E-03 5.389E+01 8.181E-02 Kr-85m 2.7242-03 5.389E+01 1.468E-01 Kr-85 8.93 1E-06 5.389E+01 4.8122-04 Kr-87 8.931E-03 5.389E+01 4.8122-01 Kr-88 8.93 1E-03 5.389E+01 4.812E-01 Xe-131m 6.698E-06 5.389E+01 3.609E-04 Xe-133m 1295E-04 5.389E+01 6.978E-03 Xe-133 3.6622-03 5.389E+01 1.973E-01 Xe-135m 1.161E-02 5.389E+01 6.2561-01 Xe-135 9.8242-03 5.389E+01 5.294E-01 Xe-138 3.9742-02 5.389E+01 2.1422+00 A From Table 5 I
INuclear Commnnm
CALCULATION CONTINUATION SHEET SHEET 27 of 30 CALr NO.: H-1-AB-MDC-1854
REFERENCE:
DCP 80048085 G. Patel, ORIGINATOR, DATE REV: I It27/02 I Mark Drucker, REVIEWER/VERIFIER, DATE I It29/02 Table 8 Post-MSLB Activity Release - Pre-accident Iodine Spike Iodine & Post-MSLB Post-MSLB Noble Gas Coolant Activity Isotope Activity Mass Release Concentration Release 11C119 (9) (CiO A B C-AxB/E6 1-131 1.598E+00 6.350E+07 .1015E+03 1-132 1.475E+01 6.350E+07 .9368E+03 1-133 1.094E+01 6.350E+07 .6948E+03 1-134 2.951E+01 6.350E+07 .1874E+04 1-135 1.598E+01 6.350E+07 .1015E+04 Kr-83m 8.181E-02 6.350E+07 .5195E+01 Kr-85m 1.468E-01 6.350E+07 .9320E+01 Kr-85 4.812E-04 6.350E+07 .3056E-01 Kr-87 4.812E-01 6.350E+07 .3056E+02 Kr-88 4.812E-01 6.350E+07 .3056E+02 Xe-131m 3.609E-04 6.350E+07 .2292E-01 Xe-133m 6.978E-03 6.350E+07 .4431E+00 Xe-133 1.973E-01 6.350E+07 .1253E+02 Xe-135m 6.256E-01 6.350E+07 .3973E+02 Xe-135 5.294E-01 6.350E+07 .3361E+02 Xe-138 2.142E+00 6.350E+07 .1360E+03 A - Iodine Activity Concentration From Table 3 A - Noble Gas Activity Concentration From Table 7 I Nuclear Common Revision 9 1
CALCULATION CONTINUATION SHEET SHEET 28 of 30 CALC. NO.: H-1-AB-MDC-1854
REFERENCE:
DCP 80048085
- 0. Patel, ORIGINATOR, DATE REV: 11/27/02 1 Mark Drucker, REVIEWER/VERIFIER, DATE I It29/02 Table 9 Post-MSLB Activity Release - Equilibrium Iodine Concentration Iodine & Post-MSLB Post-MSLB Noble Gas Coolant Activity Isotope Activity Mass Release Concentration Release 110/9 (9) (a)
A B C-AxB/1E6 1-131 7.991E-02 6.350E+07 .5074E+01 1-132 7.376E-01 6.350E+07 .4684E+02 1-133 5.471E-01 6.350E+07 .3474E+02 1-134 1.475E+00 6.350E+07 .9368E+02 1-135 7.991E-01 6.350E+07 .5074E+02 Kr-83m 8.181E-02 6.350E+07 .5195E+01 Kr-85m 1.468E-01 6.350E+07 .9320E+01 Kr-85 4.812E-04 6.350E+07 .3056E-01 Kr-87 4.812E-01 6.350E+07 .3056E+02 Kr-88 4.812E-01 6.350E+07 .3056E+02 Xe-131m 3.609E-04 6.350E+07 .2292E-01 Xe-133m 6.978E-03 6.350E+07 .4431E+00 Xe-133 1.973E-01 6.350E+07 .1253E+02 Xe-135m 6.256E-01 6.350E+07 .3973E+02 Xe-135 5.294E-01 6.350E+07 .3361E+02 Xe-138 2.142E+00 6.350E+07 .1360E+03 A - Iodine Activity Concentration From Table 4 A - Noble Gas Activity Concentration From Table 7 Revision 9 I I Nuclear Common Common Revision 9 1
CALCULATION CONTINUATION SHEET SHEET 29 of 30 CALC. NO.: H-1-AB-MDC-1854
REFERENCE:
DCP 80048085 G. Patel, ORIGINATOR, DATE REV: 11/27/02 1 Mark Drucker, REVIEWERNVERIFIER, DATE 1 1/29/02 11.0 FIGURES:
Figure 1: RADTRAD Nodalization For MSLBA Release I Nuclear Common Revision 9 1 I NcerCm oReiin9I
T CALCULATION CONTINUATION SHEET SHEET 30 of 30 CALC. NO.: H-1-AB-MDC-1854
REFERENCE:
DCP 80048085 G. Patel, ORIGINATOR, DATE REV: 11/27/02 1 Mark Drucker, REVIEWER/VERIFIER, DATE 11/29/02 12.0 AFFECTED DOCUMENTS:
Upon approval of Licensing Change Request LCR H02-01 and implementation of DCP 80048085, the following documents will be either superseded or revised:
Document to be superseded Engineering Evaluation H-1-ZZ-MDC-1854, Rev 0 Documents to be revised:
UFSAR Section 15.6.4 UFSAR Table 15.6.-7 UFSAR Table 15.6.-9 13.0 ATTACHMENTS:
Attachment A : 2 Diskettes with the following electronic files:
Calculation No: H-1-AB-MDC-1854, Rev 1.
Peer Review Comment Resolutions - Mark Drucker Nuclide Inventory File HEPU4MSLBdef Nuclide Inventory File HEPU2MSLBdef Nuclide Release Fraction & Timing File HEPUMSLBrft Dose Conversion File HEPUMSLBFGI l&12 RADTRAD Input File HEPU4MSLBOO.psf RADTRAD Output File HEPU4MSLBOO.oO RADTRAD Input File HEPU2MSLBOO.psf RADTRAD Output File HEPU2MSLB00.o0 I Nuclear Common Revision 9 1 Nula Cmo evso