ML063110194

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Attachment 8 - Calculation No. H-1-AE-MDC-1868, Revision 1, Feedwater Line Break Accident Outside Primary Containment.
ML063110194
Person / Time
Site: Hope Creek PSEG icon.png
Issue date: 08/08/2006
From: Drucker M, Ortalan E, 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-AE-MDC-1868, Rev 1
Download: ML063110194 (22)
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Attachment 8 LR-N06-0418 LCR H05-01, Rev. I Calculation No. H-1-AE-MDC-1868, Revision 1 Feedwater Line Break Accident Outside Primary Containment

NC.DE-APRZZ0002(Q)

WNC.DE-APRx *o400 . Rev. 12, EDrMm11 CALCULATION COVER SHEET Page 1 of 21 CALCULATION NUMBER. IT-1-AE-MDC-1868 EMION: I uTLE: [Feedwater Line Break Accident Outside PrmaryContainment

.#SHTS (CALC;: 2 n#A Tr -SITS I WV/so.,s59 STS.: 71 TO I*ToTALSMs: 33 CHECK ONE:

10 FINAL 'INTERIM (Proposed Plant Change) 0 VOID 0 FINAL (Future Counfirmation Reqld, enter tracking Notification number)

SALEM OR HOPE CREEK: Q-UST IRMPORTANT TO SAFETY 0 1ON-SAFM MATE1T HOPE CREEK ONLY: go {ýs 0Qb or OR ISF5i: O1 R!TANT TOSAFETY 0 NOT DeORTANT TO SAMY 0D ARE STATION PROCEDURES IMPACTED? YES t) NO [

IFYEs, INTERFACeWiTHrwE SYSTEM ENGINEER &PROCEDURE BPONSOR. ALL IMPACITD PROCEDURES SIOULD BE IDENFIIVE INASECTION INTHECALCULATION BODY [PRCAT, 3i8I*2O. INCLUDE AN sAp cOERATION FORUPDATEAD LIST THE SAP ORDERS HERE AND WITHIN 1hE BODY OF THS CALCULATION.

I] CP and ADs INCORPORITED f(IF ANY): ________"_....... ______

PDESCI*TION OF CAL UVIsV f, AOPL SeeRevision I fistory on mnd page.

The purpose of this calaftion isto dermine the Exdusion Area Boundary (EAB) Low Population Zone "n?) and Coutrol Rooui (CR) doses due to a Feedwater e Break Accident (MF A) ocurrwn outside conWtmnt uin the T* EE dose crteria and normal Coolant vonoermtion correspondingto the core thernal power level of4,031 MW, lncluding kixsrwent. ozetainty.

-CONCLUSIO]*;

The anasis results presented in Section 7.1 indicate that theEAB, MZ and CR doses due to aFWLB acddem in within their allowable TDE limis The resuls indicate that CREF system initiation is not required during aFWB accident.

The comparisout in Section 72 confirm that the proposed increases in the EAB, LPZ, & CR doses nrcless than thezia dose increase rclatoy, limits, nd hat the proposed total doses are less than the allowable regulatory limits. Therifore, pursuantto 10 CFR 50,59 guidance as defined n'References 9.17 and 9.19, the proposed increase in the corethermal power level and resulting post-FWLIBA dose consequences can be adopted as current lficcndng and design bases fir the HCOS.

Printed Same/I Sip"jý/-.~ Dat ORIGINATO OMPANY~NAME: Gopal J. Patel/NUCORE 6120 REWRC ICOMPY NAME:NI N/A VERWIER/COMPANY NAME: Mark Drucker/NCORE 0=0 CONTR ACTOR SUPERVISOR (Ifapplicable)'

-PSEG SUPERVISOR AftROVA3U (Always~m Emn]lB. Ortaan/PS ý

-NuclearCommon Revis ion 12

/1

CALCUATION CONTINUATION SHEET SHEET 2 of 21 CALC. NO.: H-1-AE-MDC-1868

REFERENCE:

G. Patel/NUCORE, ORIGINATOR, DATE REV: 06/21/2006 1 M. Drucker/NUCORE REVIEWER/VERIFIER, DATE 06/22/2006 REVISION HISTORY Revision Revision Description 0 Original Issue.

1 Revised to include the TEDE dose criteria and uprated reactor coolant concentration.

As of 12/07W2005, the EPU project decided to adopt the AST analysis performed for the increased core thermal power level for the current design and licensing bases because it conservatively bounds the EPU project design. Section 7.2 indicates that the proposed increase in the EAB, LPZ, and CR doses and total doses are less than the corresponding minimal dose increases and applicable regulatory allowable limits as defined in the 10 CFR 50.59 rule. The implementation or cancellation of the proposed core thermal power related DCP would not have any adverse impact on this analysis. Some of design inputs are taken from the documents that support higher core thermal power operation. If the HCGS license is not amended for the proposed increased power level, these design inputs would become conservative assumptions without having any adverse impact on the validity of this analysis.

Nuclear Common Revision 12

ICALCULATION CONTINUATION SHEET SHEET 3 of 21 CALC. NO.: H-1-AE-MDC-1868

REFERENCE:

G. Patel/NUCORE, ORIGINATOR, DATE REV: 06/21/2006 1 M. Drucker/NUCORE REVIEWERVERIFIER, DATE 06/22/2006 PAGE REVISION INDEX PAGE REV PAGE REV 1 1 15 1___i___

2 ___i16 1_______

3 1___ 17 1 ____

4 8____i____

is__ 1 5 ___ 1 ___19 1 ____

6 ____1 20 1 7 1 ____21 1 8 1 Attachment A 1 9 1_ _ _ _

10 _ _ _1 _ _ _ _ _ _ _

1II I________ ________

12 1 _______ ______ _

13 1 ___ ________ ________

1 14 1 1_ 1___1___ _ _ __ _ _ 1 I Nuclear Common Revision 12 1

CALCULATION CONTINUATION SHEET SHEET 4 of 21 CALC. NO.: H-1-AE-MDC-1868

REFERENCE:

G. PateM/NCORE, ORIGINATOR, DATE REV: 06/21/2006 1 M. Drucker/NUCORE REVIEWER/VERIFIER, DATE 06/22/2006 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 Background 5 3.0 Analytical Approach 6 4.0 Assumptions 8 5.0 Design Inputs 11 6.0 Calculations 14 7.0 Results Summary 15 8.0 Conclusions 16 9.0 References 17 10.0 Tables 18 11.0 Figures 20 12.0 Affected Documents 21 13.0 Attachments 21 INuclear Common Revision 12 I Nula omn eiin1

CALCULATION CONTINUATION SHEET SHEET 5 of 21 CALC. NO.: H-1-AE-MDC-1868

REFERENCE:

G. Patel/NUCORE, ORIGINATOR, DATE REV: 06/21/2006 1 M. Drucker/NUCORE REVIEWER/ERIFIER, DATE 06/22/2006

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 Feedwater Line Break Accident (FWLBA) occurring outside containment using the TEDE dose criteria and normal coolant concentration corresponding to the core thermal power level of 4,031 MWt, including instrument uncertainty.

2.0 BACKGROUND

The consequences of a realistic FWLBA were analyzed in Revision 0 of this calculation using the plant specific design and licensing bases design inputs to comply with the 10 CFR 50, Appendix 50, GDC 19 and 10 CFR 100 dose criteria.

Licensing Change Request LCR H01-002 (Ref. 9.7) was submitted to the NRC staffto amend the Hope Creek plant operating license to implement the full scope Alternative Source Term (AST) and TEDE dose criteria. The staff issued a SER (Ref. 9.8) to approve Amendment No. 134 on 10/03/2001, which revises the plant licensing basis to specify the AST in place of the previous source term and establishes the TEDE acceptance criteria in Table 6 of RG 1.183 (Ref. 9.1) in lieu of the whole body and thyroid dose guidelines provided in 10 CFR 100.11. The AST and TEDE dose acceptance criteria are to be used for the design basis accidents (DBAs).

The HCGS licensed reactor thermal power level is proposed to increase by 20%, and the radiological impact of the extended power uprate needs to be evaluated. The FWLBA is considered less severethan the main steam linelbreak (MSLB) accident, which does not result in any fuel damage (Ref. 9.19, Section 2.3.4). Therefore, its is assumed that no fuel damage occurs during the FWLBA. The AST addresses the time dependent radioactive source spectrums resulting from damaged core and fuel. There is no fuel damage postulated during the FWLB accident, therefore, the AST is not applicable to the FWLBA scenario. However, this calculation analyzes the FWLBA using the TEDE dose criteria and uprated coolant activity concentrations. The RG 1.183 neither provides any guidance for radioactive release from the FWLB accident nor specifies the TEDE dose criteria, therefore, the acceptable site boundary dose criterion is assumed to be the lowest limit allowed in Table 6 of RG 1.183 (Ref. 9.1),

which is a small fraction (10%) of the exposure guideline values. This criteria are incorporated in the assumptions 4.4 and 4.6. There are no specific ESF functions credited in this analysis. The CR is assumed to be in a normal mode of operation and the CR emergency filtration (CREF) system is not credited for mitigation of the CR dose.

The design basis FWLBA is analyzed using design basis reactor coolant iodine concentrations and 95%

atmospheric dispersion factors at the HCGS CR air intake due to the post-accident releases from the Turbine Building (TB) louvers. The design basis FWLBA will supersede the realistic FWLBA analyzed in the Revision 0 of this calculation.

I Nuclear Common Revision 1212 I1 Revision I Nuclear Common

CALCULATION CONTINUATION SHEET SHEET 6 of 21 CALC. NO.: H-1-AE-MDC-1868

REFERENCE:

G. Patel/NUCORE, ORIGINATOR, DATE REV: 06/21/2006 1 M. Drucker/NUCORE REVIEWER/VERIFIER, DATE 06/22/2006 3.0 ANALYTICAL APPROACH:

This analysis uses Version 3.02 of the RADTRAD computer code to calculate the potential radiological consequences of the FWLBA. 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).

The radiological consequences of the FWIB accident were evaluated in Revision 0 of this calculation using the accident scenario described in UFSAR Section 15.6.6.5 (Ref. 9.6). The design basis reactor coolant activity concentrations and the activity released to the environment are calculated in Section 6.2 using the mass of condensate released from the FWLB. Since the condensate demineralizers remove the dissolved fission products in the condensate, the FWLB is postulated to occur between the condenser and condensate demineralizer in order to maximize the dose consequences. The iodine activity from the condensate released from the FWLB is postulated to release to the environment via turbine building.

The potential post-FWLBA release paths are the TB vent, south plant vent, and TB louvers, which are shown in References 9.11 and 9.13 with respect to the CR air intake with its tornado missile barrier. The X/Qs for these release paths are obtained from Reference 9.5 Section 8.0, and listed in the following table:

HCGS Control Room Time 95% Atmospheric Dispersion Factors (X/Qs) (s/m3 )

Interval South Plant TB Louvers TB Vent Vent 3 (s/m ) (s/m)

(110 0-2 5.75E-04 6.17E-04 3.48E-04 2-8 3.84E-04 4.OOE-04 2.55E-04 8-24 1.40E-04 1.44E-04 9.11E-05 24-96 9.08E-05 1.00E-04 5.37E-05 96-720 7.01E-05 7.49E-05 3.82E-05 Comparison of X/Qs in the above table indicates that the TB louvers release path is the most limiting release path for the post-FWLBA release. Therefore, the CR dose is calculated using the post-FWLBA release through the TB louvers. Since the post-FWLBA activity is postulated to release instantaneously as a single puff, the CR Emergency Filtration (CREF) is not credited. The CR is assumed to be in the normal mode of operation for the entire duration of the accident.

DOSE EQUIVALENT 1-131 is the concentration of 1-131 in gCi/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 used for this calculation are listed in Table 1.

Revision 12 I II Nuclear Common Revision 12 1

CALCULATION CONTINUATION SHEET SHEET 7 of 21 CALC. NO.: H-1-AE-MDC-1868

REFERENCE:

G. Patel/NUCORE, ORIGINATOR, DATE REV: 06/21/2006 1 M. Drucker/NUCORE REVIEWERNERIFIER, DATE 06/22=2006 In Table 2, a scaling factor is developed based on the maximum equilibrium iodine concentration of 0.2

.tCi/g DE 1-131. The isotopic iodine isotopic activities released to environment are calculated in Table 3 using the amount of coolant mass that becomes airborne (Section 6.2).

The RADTRAD V3.02 (Ref. 9.2) default nuclide inventory file (NIF) Bwrdef is modified based on the post-FWLB iodine activity release to the environment as shown in Table 3. The modified NIF HEPUFWLBdef.txt is used to calculate the EAB, LPZ, and CR doses, which are shown in Section 7.0 and compared with their allowable dose limits.

Determine Compliance of Increased Dose Consequences With 10CFR50.59 Guidance Consistent with the RG 1.183, Section 1.1.1, once the initial AST implementation has been approved by the staff and has become part of the facility design basis, the licensee may use 10 CFR 50.59 and its supporting guidance in assessing safety margins related to subsequent facility modifications and changes to procedures.

The NRC Safety Evaluation Report for Amendment 134 (Ref. 9.8) approved the AST for the HCGS:licensing basis analyses.

An increase in control room, EAB or LPZ dose consequence is considered acceptable under the 10 CFR 50.59 rule if the magnitude of the increase is minimal (as defined by the guidance in Refs. 9.17 and 9.18), and'if the total calculated dose is less than the allowable regulatory guide 1.183 dose limit. The current licensing basis analysis is documented in the calculation H-1-AE-MDC-1868, Rev 0. The increases in the proposed BAB, LPZ,

& CR doses are compared with the 10 CFR 50.59 allowable minimal dose increases in Section 7.2. Similarly, the proposed calculated total doses are compared with the allowable regulatory guide dose limits. The comparison in Section 7.2 confirms that the increase in the EAB, LPZ, & CR doses and the total calculated doses are less than the corresponding minimal dose increases and allowable regulatory guide limits. Therefore, pursuant to 10 CFR 50.59 guidance as defined in References 9.17 and 9.18, the proposed increase in the core thermal power level and resulting post-FHA doses can be adopted as current design and licensing bases for the HCGS.

I Nuclear Common Revision 12 I

CALCULATION CONTINUATION SHEET SHEET 8 of 21 CALC. NO.: H-1-AE-MDC-1868

REFERENCE:

G. Patel/NUCORE, ORIGINATOR, DATE REV: 06/21/2006 1 M. Dmcker/NUCORE REVIEWERNERIFIER, DATE 06/22/2006

4.0 ASSUMPTIONS

Assumptions for Evaluating the Radiological Consequences of a FWLBA The assumptions for evaluating the radiological consequences of a FWLBA are listed in the following section.

These assumptions are incorporated as Design Inputs in Sections 5.3.1 through 5.3.4 for the FWLBA analysis.

SOURCE TERM 4.1 No fuel damage is postulated for the FWLB accident because it is less severe than the MSLB accident (Ref. 9.19, Section 2.3.4). The iodine activity concentrations in the condensate are assumed to correspond to 2% of the normal reactor coolant iodine concentrations based on the maximum equilibrium iodine concentration of 0.2 .LCi/g Dose Equivalent 1-131 allowed by the technical specification (Ref. 9.4) for normal operation of the plant. The chemical forms of iodine release from feedwater to the environment are assumed to be 97% elemental and 3% organic. Noble gas activity in the condensate is assumed to be negligible.

TRANSPORT 4.2 Since the condensate dernineralizers remove the dissolved fission products in the condensate, the FWLB is postulated to occur between the condenser and condensate demineralizer in order to maximize the dose consequences. The entire content of the condenser hotwells and piping upstream of the condensate demineralizer is assumed to be released through the feedwater line break. The condensate mass of 4,309,629 lbs (Section 6.2) is assumed to be released from the FWLB, and is incorporated in Design Input 5.3.1.9.

4.3 10% of the iodine activity in the condensate mass isz 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.

Offsite Dose Consequences:

The following guidance is used in determining the TEDE for a maximum exposed individual at EAB and LPZ locations:

4.4 The maximum EAB TEDE for any two-hour period following the start of the radioactivity release is determined and used in determining compliance with the following dose acceptance criterion:

EAB Dose Acceptance Criterion: 2.5 Rem TEDE (Ref. 9.1, Table 6) 4.5 Thebreathing rates for persons at offsite locations are given in Reference 9.1, Section 4.1.3, and are incorporated in Design Inputs 5.3.4.3 and 5.3.4.4.

Revision I Nuclear Common Revision 12 12 1I I Nuclear Common 1

CALCULATION CONTINUATION SKEET SHEET 9 of 21 CALC. NO.: H-1-AE-MDC-1868

REFERENCE:

G. Patel/NUCORE, ORIGINATOR, DATE REV: 06/21/2006 1 M. Dmcker/NUCORE REVIEWER/VERIFER, DATE 06/22/2006 4.6 The maximum Low Population Zone TEDE is deternined 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 following dose acceptance criterion:

LPZ Dose Acceptance Criterion: 2.5 Rem TEDE (Ref 9.1, Table 6) 4.7 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 The following guidance is used in determining the TEDE for maximum exposed individuals located in the control room:

4.8 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,
  • 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 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),
  • Radiation shine from radioactive material in the reactor containment (i.e., containment shine dose; not applicable to a FWLB occurring outside containment),
  • Radiation shine from radioactive material in systems and components inside or external to the cbontrol 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 FWLBA 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 FWLBA.

4.9 The radioactive 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 exclusion area boundary (EAB) and the low population zone (LPZ) TEDE values (Ref. 9.1, Section 4.2.2).

4.10 The occupancy and breathing rate of themaximum exposed individual present in the control room are incorporated in Design Input 5.3 (Ref 9.1, Section 4.2.6).

4.11 10 CFR 50.67 (Ref 9.16) establishes the following radiological criterion for the control room.

Nuclear Common Revision 12

CALCULATION CONTINUATION SHEET SHEET 10 of 21 CALC. NO.: H-1-AE-MDC-1868

REFERENCE:

G. Patel/NUCORE, ORIGINATOR, DATE REV: 06/21/2006 1 M. Drucker/NUCORE REVIEWERVERIFIER, DATE 06/22/2006 CR Dose Acceptance Criterion: 5 Rem TEDE 4.12 Although allowed by Reference 9.1, Section 4.2.4, credit is not taken for the engineered safety features of the CR emergency filtration (CREF) system that mitigate airborne activity within the control room.

4.13 No credits for KI pills or respirators are taken (Ref. 9.1, Section 4.2.5).

eiin1 II Nuclear Nula omn Common Revision 12

CALCULATION CONTINUATION SHEET SHEET 11 of 21 CALC. NO.: H-1-AE-MDC-1868

REFERENCE:

G. Patel/NUCORE, ORIGINATOR, DATE REV: 06/21/2006 1 M. Drucker/NUCORE REVIEWER/VERIFIER, DATE 06/22/2006 5.0 DESIGN INPUTS:

5.1 General Considerations 5.1.1 Applicability of Prior Licensing Basis The characteristics of the 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 TID-14844 analyses were assessed for their validity to represent the as-built condition of the plant and evaluated for their compatibility to meet the TEDE methodology. The analysis in this calculation ensures that analysis assumptions, design inputs, and methods are compatible with 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. None of-the ESF functions are credited in this FWLBA analysis. The dose mitigation function of the CREF system is -notcredited in this 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. 9.16) are compatible with the TEDE dose criteria and selected with the objective of producing conservative radiological consequences. For conservatism, the designbasis reactor coolant iodine concentrations are considered and the CREF is not credited for the CR dose mitigation.

5.1.4 Meteorology Considerations The control room atmospheric dispersion factors (%/Qs) for the Turbine Building louvers release point are developed (Ref. 9.5) using the NRC sponsored computer code ARCON96. The EAB and LPZ X/Qs were reconstituted using the HCGS plant specific meteorology and appropriate regulatory guidance (Ref. 9.9). The off-site X/Qs reconstituted in Reference 9.9 were accepted by the staff in previous licensing proceedings.

5.2 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 TEDE dose criteria and assumptions are consistent with those identified in the applicable UFSAR sections. The design inputs and assumptions in the following sections represent the as-built design of the plant.

Revision 12 1 I Nuclear Common Revision 12 I I Nuclear Common

CALCULATION CONTINUATION SHEET SHEET 12 of 21 CALC. NO.: H-1-AE-MDC-1868

REFERENCE:

G. Patel/NUCORE, ORIGINATOR, DATE REV: 06/21/2006 1 M. Drucker/NUCORE REVIEWERNVERIFIER, DATE 06/22=2006 Design Input Parameter Value Assigned Reference 5.3 FWLBA Parameters 5.3.1 Source Term 5.3.1.1 Proposed core power level 4,031 MWt 9.3, Section 3.2.1 5.3.1.2 Uprated Isotopic Reactor Coolant Concentration (IiCi/gm) 9.3, Appendix A Isotope Activity Isotope Activity Isotope Activity 1-131 1.30E-02 1-133 8.90E-02 1-135 1.30E-01 1-132 1.20E-01 1-134 2.40E-01 5.3.1.3 Total hotwell capacity 430,560 gallons 9.21, Section 3.2.1, Item 9 5.3.1.4 Fuel damage None Section 2.0 and Assumption 4.1 5.3.1.5 Iodine carry-over from < 1% 9.19, Section 2.2.1.2 reactor coolant 2% Assumed in the analysis 5.3.1.6 Iodine flasbing from 10% 9.1, Appendix A, Section 5.5 condensate to TB atmosphere 5.3.1.7 Maximum equilibrium 0.2 pCi/g DE 1-131 9.4 iodine concentration 5.3.1.8 Condensate maximum 140"F 9.20, Section 3.3.1, Item 10 temperature 1_1 5.3.1.9 Mass of condensate 4,309,629 lb Section 6.2 released from FWLBA 1_1 5.3.2 Activity Transport (see Figure 1) 5.3.2.1 Activity release rate 2.OE+05 source volumel/o/day Assumed as a single puff 5.3.2.2 Duration of release Instantaneously in a single puff Conservatively Assumed 5.3.2.3 Type of release Ground level release Conservatively Assumed 5.3.2.4 Chemical form of Iodine in condensate Elemental 97.0% Conservatively Assumed Organic 3.0%

5.3.2.5 Dilution or holdup within Not credited Conservatively Assumed the facility building 5.3.2.6 Source volume 100 ft Assumed to facilitate RADTRAD I nodalization 5.3.3 Control Room Parameters (see Figure 1) 5.3.3.1 CR volume 85,000 ft 9.10, page 10 5.3.3.2 CR normal air inflow rate 3,000 +/- 10% cfm (conservatively 9.14 and Assumption 4.12 (0-720 hrs) 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 Nuclear Common Revision 12

T CALCULATION CONTINUATION SHEET SHEET 13 of 21 CALC. NO.: H-1-AE-MDC-1868

REFERENCE:

G. Patel/NUCORE, ORIGINATOR, DATE REV: 06/21/2006 1 M. Dmcker/NUCORE REVIEWER/VERIFIER, DATE 06/22/2006 Design Input Parameter Value Assigned Reference 5.3.3.4 CR breathing rate (m3 /sec) I 3.5E-04 9.1, Section 4.2.6 3

5.3.3.5 CR atmospheric dispersion factors for Turbine Building louvers release (y/Q) (sec/m )

Time (Hr) X/Q (sec/r_)

0-2 6.17E-04 9.5, Section 8.3 2-8 4.OOE-04 8-24 1.44E-04 24-96 1.00E-04 96-720 7.49E-05 5.3.4 Site Boundary Release Model Parameters 5.3.4.1 EAB atmospheric I 1.9E-04 9.9, Pages 5 & 9 dispersion factor (X/Q) (sec/m 3 )

5.3.4.2 LPZ atmospheric dispersion factors sX/Qs)

Time (Hr) X/Q (sec/rn) 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 3

(m /sec) 5.3.4.4 LPZ breathing rates (m__3/sec ____________

Time (Hr) (m3 /sec) 0-8 3.5E-04 9.1, Section 4.1.3 8-24 1.8E-04 24-720 2.3E-04 I Nuclear Common Revision 12 2I 1

I Nula omnRvso

CALCULATION CONTINUATION SHEET SHEET 14 of 21 CALC. NO.: H-1-AE-MDC-1868

REFERENCE:

G. PateL/NUCORE, ORIGINATOR, DATE REV: 06/21/2006 1 M. Drucker/NUCORE REVIEWER/VERIFIER, DATE 06/22/2006

6.0 CALCULATIONS

6.1 Dose Conversion Factors and Iodine Spike Coolant Activity Iodine isotopic dose conversion factors (DCFs) are obtained from Reference 15, page 136. These DCFs are provided in Sv/Bq for Committed Dose Equivalent (CDE) per Unit Intake, which are converted into rem/Ci in Table 1 using the following conversion factor.

Sv/Bq = 100 rem/Sv x 3.7 x 1010 Bq/Ci = 3.7 x 1012 rem/Ci/Sv/Bq Example: 1-131 DCF 2.92E-07 Sv/Bq x 3.7E+12 rem/Ci/SvIBq = 1.08E+06 rem/Ci 6.2 Airborne Iodine From Condensate Release There is no fuel damage as a consequence of this accident. The activity in the main condenser hotwell prior to occurrence of the break is released to the environment The iodine concentration in the main condenser hotwell is 0.02 times the iodine concentration in the reactor coolant. The reactor coolant iodine activity based on the above off-gas release rate is calculated in Table 1. None of the condensate released from the break flashes to steam because the maximum temperature of condensate is 1400F, which is less than 2120F, therefore, it is assumed that 10% of the iodine in the condensate is assumed to become airborne (Ref. 9.1, Appendix A, Section 5.5). The iodine activity released from the break is calculated in Table 2 based on the condensate mass released from the break, the iodine concentration in the condensate, and the condensate flashing factor:

Total condenserhotwell volume = 430,560 gallons (Ref. 9.21, Section 3.2..1, Item 9)

Piping volume is assumed tobe 20% of total hotwell volume = 0.20 x 430,560 gallons = 86,112 gallons Total condensate volume release from break

= 430,560 gallons + 86,112 gallons = 516,672 gallons x (1/7.481) ft3/gallon x 62.4 lb/ft = 4,309,629 lb However, the carry-over of iodine from reactor coolant to main steam, which condenses into feedwater < 1%

(Ref. 9.19, Section 2.2.1.1), it is conservatively assumed to be 2% to compensate any uncertainty associated with the estimated condensate mass release of 4,309,629 lb from the break.

Flashing fraction = 0.10 Total amount of reactor coolant iodine that becomes airborne =

= 4,309,629 lb x 0.02 x 0.10 = 8,619.3 lb

= 8,619.3 lb x 453.6 g/lb = 3.91E+06 g INuclear Common Revision 12 1 Revision 12 I I Nuclear Common

CALCULATION CONTINUATION SHEET SHEET 15 of 21 CALC. NO.: H-1-AE-MDC-1868

REFERENCE:

G. Patel/NUCORE, ORIGINATOR, DATE REV: 06/21/2006 1 M. Drucker/NUCORE REVIEWER/VERIFIER, DATE 06/22/2006 7.0 RESULTS

SUMMARY

7.1 The results of the FWLBA analysis are summarized in the following table:

Feedwater Line Break Accident TEDE Dose (rem)

Receptor Location Control Room EAB LPZ Calculated Dose 5.69E-03 2.87E-03 2.88E-04 (0.0 hr) I Allowable TEDE Limit 5.OE+00 2.5E+00 j 2.5E+00

___,, ___________ RADTRAD Computer Run No.

HEPUFWLBO1 HEPUFWLBO1 HEPUFWLBOI1 Significant assumptions used in this analysis:

  • Post-FWLBA activity is released to the environment in a single puff at ground level through TB Louvers
  • CREF system is not credited.
  • Core thermal power = 4,031 MWt 7.2 Compliance of proposed dose increases with the 10 CFR 50.59 rule is shown as follows:

Current Licensing Basis Proposed Regulatory RG Design Basis Accident Dose (rm) Total Dose Proposed Minimal Dose Thyroid Whole Equivalent Dose Limit Increase Increase Limit Body TEDE (rem) (rem) (rem) (rem) (rem)

TEDE TEDE TEDE TEDE TEDE A B CfA*0.03+B D E Ff-D-C 1ff.9-C) H Feedwater Line Break H-1-AE-MDC-1868, RO H1l-AE-MDC-1868, Rev 1 Accident Control Room 4.38E-03 0 0.0001314 0.00569 5.00 0.006 0.50 5.00 Exclusion Area Boundary 1.69E-03 0 0.0000507 0.00287 25.00 0.003 2.50 2.50 Low Population Zone 1.29E-04 0 0.00000387 0.000288 25.00 0.000 2.50 1 2.50 C Equivalent TEDE calculated using equation presented in Regulatory Guide 1.183 (Ref. 9.1, Footnote 7)

E From 10 CFR 50.67 (Ref. 9.16)

H Assumed To Be 10% of Regulatory Dose Limit 1

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CALCULATION CONTINUATION SHEET SHEET 16 of 21 CALC. NO.: H-1-AE-MDC-1868 REFERENCE-G. PatelNUCORE, ORIGINATOR, DATE REV: 06/21/2006 1 M. Drucker/NUCORE REVIEWER/VERIFIER, DATE 06/22/2006

8.0 CONCLUSION

S:

The analysis results presented in Section 7.1 indicate that the EAB, LPZ, and CR doses due to a FWLB accident are within their allowable TEDE limits. The results indicate that CREF system initiation is not required during a FWLB accident.

The comparisons in Section 7.2 confirm that the proposed increases in the EAB, LPZ, & CR doses are less than the minimal dose increase regulatory limits, and that the proposed total doses are less than the allowable regulatory limits. Therefore, pursuant to 10 CFR 50.59 guidance as defined in References 9.17 and 9.18, the proposed increase in the core thermal power level and resulting post-FWLBA dose consequences can be adopted as current licensing and design bases for the HCGS.

Revision I Nuclear Common Revision 12 12 1I I Nuclear Common

CALCULATION CONTINUATION SHEET SHEET 17 of 21 CALC. NO.: H-1-AE-MDC-1868

REFERENCE:

G. Patel/NUCORE, ORIGINATOR, DATE REV: 06/21/2006 1 M. Drucker/NUCORE REVIEWERNERIFIER, DATE 06/22/2006

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 3.02: A Simplified Model for Radionuclide Transport and Removal and Dose Estimation," NUREG/CR-6604, USNRC, April 1998
3. Vendor Technical Document (VTD) No. 430059, Volume 002, Rev 1, EPU TR T0807 Coolant Radiation Sources.
4. HCGS Technical Specification 3/4.4.5, "Specific Activity' Limiting Condition for Operation
5. Calculation No. H-1-ZZ-MDC-1 879, Rev 1, Control Room & Technical Support Center X/Qs Using ARCON96 Code
6. HCGS UFSAR Section 15.6.6.5, ',adiological Consequence."'
7. HCGS Licensing Change Request LCR No. H01-002, Dated 05/17/01, Request For Change to Technical Specifications, Increase in Allowable MSIV Leakage Rate and Elimination of MSIV Sealing System.
8. NRC Safety Evaluation Report, Hope Creek Generating Station - Issuance of Amendment No. 134 for Increase in Allowable MSIV Leakage Rate and Elimination of MSIV Sealing System.
9. Calculation No. H-1-ZZ-MDC-1 820, Rev 0, Offsite Atmospheric Dispersion Factors
10. Calculation No. H-1-ZZ-MDC-1882, Rev 0, Control Room Envelope Volume
11. HCGS Architectural Drawing No. A-0221-0, Sheet 1, Rev 10, General Plant RoofPlan
12. Critical Software Package Identification No. A-0-ZZ-MCS-0225, Rev 2, 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. Federal Guidance Report 11, EPA-520/1-88-020, Environmental Protection Agency
16. 10 CER 50.67, "Accident Source Term."
17. PSEG Procedure No. NC.NA-AS.ZZ-0059(Q), Rev 11, 10CFR50.59 Program Guidance.
18. Nuclear Energy Institute Report No. NEI 96-07, Rev 1, Guidelines for 10 CFR 50.59 Implementation.
19. Vendor Technical Document (VTD) No. PNO-A46-4100-0047, Rev 2, GE Specification Document No.

22A2703F, Rev 3, Radiation Sources.

20. Vendor Technical Document (VTD) No. 430054, Volume 2, Rev 1, EPU TR T0702, Condensate Demineralizer
21. Vendor Technical Document (VTD) No. 430055, Volume 2, Rev 1, EPU TR T0703, BOP Power Cycle Mechanical Performance.

I I Nuclear Common Revision 12 1 I Nuclear Common Revision 12 I

CALCULATION CONTINUATION SHEET SHEET 18 of 21 CALC. NO.: H-1-AE-MDC-1868

REFERENCE:

G. Patel/NUCORE, ORIGINATOR, DATE REV: 06/21/2006 1 M. Drucker/NUCORE REVIEWER/VERIFIER, DATE 06/22/2006 10.0 TABLES:

Table I Iodine Isotopic Dose Conversion Factors Isotopic Conversion Iodine Dose Factor Dose Isotope Conversion Conversion Factor Factor (Sv/Bq) (rem/Ci/Sv/Bq) (rem/Cl)

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 2.880E-10 3.700E+12 1.066E+03 1-135 8.460E-09 3.700E+12 3.130E+04 A From Reference 9.15, Page 136 B From Section 6.1 Table 2 Scaling Factor for Maximum Equilibrium Iodine Concentration Normal Iodine Iodine Dose Product Isotope Activity Conversion Concentration Factor (ILCi/g) (rem/Cl) (LCizrem/Ci.g)

A B (AxB) 1-131 1.300E-02 1.080E+06 1.404E+04 1-132 1.200E-01 6.438E+03 7.726E+02 1-133 8.900E-02 1.798E+05 1.600E+04 1-134 2.400E-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 1odine Scaling Factor Based on 0.2 tCi/g DE 1-1311 6.147E+00 I Nuclear Common Revbion 12 2I 1

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CALCULATION CONTINUATION SHEET SHEET 19 of 21 CALC. NO.. H-1-AE-MDC-1868

REFERENCE:

G. PatelI/NUCORE, ORIGINATOR, DATE REV: 06/21/2006 1 M. Drucker/NUCORE REVIEWER/VERIFIER, DATE 06/22/2006 Table 3 Isotopic Maximum Equilibrium Iodine Activity Release Design Iodine Mass of Maximum Basis Scaling Coolant Equilibrium Isotope Iodine Factor Becomes Iodine Activity Concentration Airborne Release (110/g) (g) (C)

A B C AxBxC/1E6 1-131 1-300E-02 6.147E+00 3.910E+06 .3124E+00 1-132 1 .200E-01 6.147E+00 3.910E+06 .2884E+01 1-133 8.900E-02 6.147E+00 3.910E+06 .2139E1+01 1-134 2.400E-01 6.147E+00 3.910E+06 .5768E+01 1-135 1.300E-01 6.147E+00 3.910E+06 .3124E+01 A From Table 2 B Scaling Factor Based on 0.2 pCi/g DE 1-131 From Table 2 C From Section 6.2 eiin1 II Nula omn Nuclear Common Revision 12 1

CALCULATION CONTINUATION SHEET SHEET 20 of 21 CALC. NO.: H-1-AE-MDC-1868

REFERENCE:

G. Patel/NUCORE, ORIGINATOR, DATE REV: 06/21/2006 1 M. Drucker/NUCORE REVIEWER/VERIFIER, DATE 06/22/2006 11.0 FIGURES:

Figure 1: RADTRAD Nodalization For FWLBA Release I

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-CALCULATION CONTINUATION SHEET SHEET 21 of 21 CALC. NO.: H-1-AE-MDC-1868

REFERENCE:

G. Patel/NUCORE, ORIGINATOR, DATE REV: 06/21/2006 1 M. Drucker/NUCORE REVIEWER/VERIFIER, DATE 06/22/2006 12.0 AFFECTED DOCUMENTS:

The following documents will be revised:

UFSAR Section 15.6.6 UFSAR Section 15A.7 Feedwater Line Break Accident (Section 15.6.6)

UFSAR Table 15.6-21 UFSAR Table 15.6-22 UFSAR Table 15.6-24 13.0 ATTACHMENTS:

Attachment A : 1 Diskette with the following electronic files:

Calculation No: H-1-AE-MDC-1868, Rev 1.

Peer Review Comment Resolutions - Mark Drucker Owner's Acceptance Comment Resolution Form 2 - Michael E. Crawford Certification for Design Verification Form-1 RCPD Form-1 I Nuclear Common Revision eiin1 12 I Nula!omn

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