ML063110181

Attachment 4 - Calculation No. H-1-BG-MDC-1859, Revision 1, Instrument Line Pipe Break Accident.

ML063110181
Person / Time
Site:	Hope Creek
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-BG-MDC-1859, Rev 1
Download: ML063110181 (28)

Text

Attachment 4 LR-N06-0418 LCR H05-O1, Rev. I Calculation No. H-1-BG-MDC-1859, Revision 1 Instrument Line Pipe Break Accident

NC.DE-AP.ZZ-0002(Q)

MCDE-A,,P.Z002(o Rev. It orm 1' CALCULATION COVER SHEET Page 1 of 27 CALCULATION NUMBERh IH-1-BG-MDC-1859 REVISION: 1 rmL: IInstrument Line Pipe Break Acciddent WSUTS (CL : 127 1 ATrNOsHTh: wIA #IDv/so.s9f2.4 SUTN: 17/4/I0 4TOTAL SITS: 39 CHECK ONE:

IR FINAL 0 INTERIM (Proposed Plant Change) [I VOID

[I FINAL (Future Confirmation Req'd, enter tracking Notification number:)

SALEM OR HOPE CREEK: Q LrST I IeORTANT TO SAFETY E NON-SAFETY RELATED HOPE CREEK OqLY: Q OQS Qsh [F [OR SF: MORTANT W TO SAFETY 0 NOT MPeORTANT TO SAFETY O ARE STATION PROCEDURES IMPACTED? YES 0 NO N F *YES, INTERFACE WITH THESYSTEM ENGINEER &PROCEDURE SPONSORP ALL4MPACoTED PROCEDURES SOULD BE IDENTIFIED INA SECTION INTHE CALCUATION BODY 1CRC 70381,-d280q. INCLUDE AN SAP OPERATION FOR UPDATE AND LST THE SAP ORDERS HERE AND WITHIN THWBODY OF THIS CALCULATION.

, CP and ADs INCORPORATED (IF ANY): ,.._

DoESCRITON OF Q CLTION REVWI-ON OF APQ.):

See Revision I Ihstory on.wx1pge.

X=lBSE:

The purpose ofthis calculation Is to determine the EMAB LPZ. and CR doses dueto an ILPB accident using the TEDE dose critema and ractor coolant activity concentrtions corresponding to the core thernal power level of4,031 MWI, incuding the lnstrument uncertainty.

CONCLUSIONS:

The maly* results presented in Section 7,1 inicate thit the MAB, 17, and CR doses due to anU; accident are within their allowable TEDE limts. The rms indicae that CREFiyAem, Initiation is Aot required durng an ILPB accident.

The comparisons In Secton 7.2 = that the proposed Increases in the EAR, LPZ, & CR doses ar less than fte ininmal dose increase regulatory limits, and that the proposed total doses are less than the allowable regulatory limits. -Therefom pumruant to 10 CFR 5059 guidance as deEned in References923 and 9.24, the proposed ncr=ase in the core thermal power level and resulting post-ILPBA dose consequences can be adopted as current licensing and design bases for the HCGS.

Printed Name I Sigit // -- Date ORIGINATORICOMPANY NAME: Gopal 3. Patel/NUCORE *6/16/2006 REVWEWERKOMPANY NAME:V NA VERIFIER/COM: .AME Mark D01ck=r/NUCORE_06/1,/2006 CONTRACTOR SUPERVISOR (If N/A 2pplicable)

PSEG SUERVISOR APPROVAL: (Alwaysi M) Emin B. Ortaan/PSEGZ6A ;6ý_ LVA Nuclear Common Revision 12

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REFERENCE:

G. Patel/NUCORE ORIGINATOR, DATE REV: 06/16/2006 1 Mark Drucker/NUCORE REVIEWER/VERIFIER, DATE 06/19/2006 REVISION HISTORY Revision Revision Description 0 Initial Issue.

1 Revised to include the extended power uprate reactor coolant activity concentrations and TEDE dose criteria. Updated control room volume is incorporated IAW CD D 508, Package No. 80027981. Increased the reactor coolant released from the break as flashed steam to 40 percent, as calculated using a constant enthalpy process.

As of 12/07/2005, 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.

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REFERENCE:

G. Patel/NUCORE ORIGINATOR, DATE REV: 06/1612006 1 Mark Drucker/NUCORE REVIEWER/VERIFIER, DATE 06/19/2006 PAGE REVISION INDEX PAGE REV PAGE REV 1 1 15 1 2 1 16 1 3 1 17 1 4 1 18 1 5 1 19 1 6 1 20 1 7 1 21 1 8 1 22 1 9 1 23 1 10 1 24 1 11 1 25 1 12 1 26 1 13 1 27 1 14 1 Attachment 13.1 1 I Nuclear Common Revision 12 1 I

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CALCULATION CONTINUATION SHEET SHEET 4 of 27 CALC. NO.: H-1-BG-MDC-1859

REFERENCE:

G. Patel/NUCORE ORIGINATOR, DATE REV: 06/16/2006 Mark Drucker/NUCORE REVIEWERNERIFIER, DATE 06/19/2006 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 5 4.0 Assumptions 9 5.0 Design Inputs 12 6.0 Calculations 15 7.0 Results Summary 16 8.0 Conclusions 17 9.0 References 17 10.0 Tables 19 11.0 Figures 25 12.0 Affected Documents 27 13.0 Attachments 27 Revision 12 I II Nuclear Common Revision 12 i

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REFERENCE:

G. Patel/NUCORE ORIGINATOR, DATE REV: 06/16/2006 1 Mark Drucker/NUCORE REVIEWER/VERFIER, DATE 06/19/2006

1.0 PURPOSE

The purpose of this calculation is to determine the EAB, LPZ, and CR doses due to an Instrument Line Pipe Break (ILPB) accident using the TEDE dose criteria and reactor coolant activity concentrations corresponding to the core thermal power level of 4,031 MW, including the instrument uncertainty.

2.0 BACKGROUND

The realistic ILPB accident was reconstituted in Revision 0 of this calculation using the accident scenario described in HCGS-UFSAR Section 15.6.2.5 with the design input parameters listed in UFSAR Table 15.6-2. The ILPB accident was postulated to closely reproduce the EAB/LPZ doses and activity releases at the break location and into the environment, which are comparable to those given in the UFSAR. In Revision 1 of this calculation the reactor coolant released from the break as flashed steam is increased to 40 percent, as calculated using a constant enthalpy process.

Licensing Change Request LCRH01-002 (Ref. 9.7) was submitted to the NRC staff to 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 licensingbasis to specify the AST in place of the previous source term and establishes the TEDE acceptance criteria in 10 CFR 50.67 (Ref. 9.4) and Table 6 of Regulatory Guide 1.183 (Ref.

9.1) in lieu of the whole body and thyroid dose guidelines provided in 10 CFR 100.11. Therefore, the AST and TEDE dose acceptance criteria should 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. Regulatory Guide 1.183 neither provides any guidance for radioactive release from the ILPB accident nor specifies the TEDE dose criteria; therefore, the guidance from the Standard Review Plan 15.6.2 (Ref 9.6) is adopted for this analysis. The acceptable site boundary dose criterion is a small fraction of the exposure guideline values, which means 10% of the exposure guideline values (Ref. 9.6, Section 11.2). This criterion is incorporated in assumptions 4.5 and 4.7. The ILPB accident is analyzed using the TEDE dose criteria and uprated reactor coolant activity concentrations. No specific ESF functions - initiations of the CR Emergency Filtration (CREF) System and the Filtration, Recirculation, and Ventilating System (FRVS) - are credited in this analysis. The reactor building and CR are assumed to be in a normal mode of operation.

3.0 ANALYTICAL APPROACH This analysis uses Version 3.02 of the RADTRAD computer code to calculate the potential radiological consequences of the ILPBA. 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. 9.20).

3.1 Primary Coolant Activity Concentrations The methodology and assumptions used for evaluating the potential consequences of an ILPB accident outside containment are identified in Standard Review Plan Section 15.6.2 (Ref. 9.6) and Safety I

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REFERENCE:

G. Patel/NUCORE I ORIGINATOR, DATE REV: 06/16/2006 1 Mark Drucker/NUCORE REVIEWER/VERIFIER, DATE 06/19/2006 Guide 11 (Ref. 9.13). Since no fuel damage occurs during the ILPBA, the initial fission product concentrations in the primary coolant correspond to the maximum equilibrium values permitted by the technical specifications (Ref 9.6, Section m.3.d). The iodine concentration in the primary coolant is assumed to be 4 J4Ci/g Dose Equivalent (DE) 1-131. The assumptions and design input parameters used in this analysis are described in Sections 4.0 and 5.0. The iodine scaling factor for 4 tCi/g DE 1-131 is calculated in Table 2 using the thyroid dose conversion factors developed in Table 1 utilizing the following definition of DE 1-131 and information in Federal Guidance Report 11 (Ref. 9.18).

DOSE EQUIVALENT 1-131 shall be that concentration of 1-131, gtCi/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 isotopic iodine activity concentrations are calculated in Table 3 based on 4 j+/-Ci/g DE 1-131.

The isotopic noble gas activity concentrations are calculated in Table 4 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 activity concentrations based on 100/E-BAR are calculated in Table 5 and listed in Table 6 using the following 100/E-BAR definition:

E-BAR shall be the average, weighted in proportion to the activity 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, -withhalf lives greater than 15 minutes, making up at least 95% of the total non-iodine activity in the coolant.

The coolant mass release from the break in Revision 0 of this analysis reconstituted site boundary doses identical to those presented in HCGS UFSAR Table 15.6.5 (Ref. 9.21). Therefore, the existing coolant mass release information is used with the uprated coolant activity concentration, and the newly developed 95 percentile X/Q spectrums for the potential release point to calculate the radiological consequences at the EAB, LPZ, and CR receptor locations. In Revision I of this calculation the reactor coolant released from the break as flashed steam is increased to 40 percent, as calculated using a constant enthalpy process.

3.2 Fission Product Release to the Environment Of the 25,000 pounds of coolant released from the instrument line break, 10,000 pounds (i.e., 40 percent) flashes to steam. It is assumed that all the iodine in the coolant that flashes to steam enters the steam phase with the coolant, and that 10 percent of the iodine in the initially unflashed coolant becomes airborne. The activity released from the break is assumed to mix with 50% of the reactor building (RB) volume prior to being released to the environment. To simulate 50% mixing in the RB volume, the normal RB venting system flow rate is doubled. No credit for plateout is taken. The iodine and noble gas activities airborne at the break location in the reactor building are calculated in Table 7 using the coolant mass release.

The potential post-ILPBA release paths from the reactor building and to the environment are the south plant vent and the FRVS vent, which are shown in Reference 9.17 with respect to the CR air intake with I Nuclear Common Revision 12 1 Revision 12 INuclear Common

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REFERENCE:

G. Patel/NUCORE ORIGINATOR, DATE REV: 06/16/2006 1 Mark Drucker/NUCORE REVIEWER/VERIFIER, DATE 06/19/2006 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 95% Atmospheric Dispersion Factors Time (X/Qs) (s/im)

Interval South Plant FRVS Vent (110 Vent (s/m3) 0-2 5.75E-04 1.25E-03 2-8 3.84E-04 8.09E-04 8-24 1.40E-04 3.04E-04 24-96 9.08E-05 2.10E-04 96-720 7.01E-05 1.59E-04 Comparison of X/Qs in the above table indicates that the FRVS vent release path is the most limiting release path for the post-ILPB release. Therefore, the CR dose is calculated using the post-ILPB release through the FRVS vent.

Since the post-ILPB 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.

The RADTRAD V3.02 (Ref 9.2) default nuclide inventory file (NIF) Bwr_def. is modified based on the post-ILPBA activity releases to the environment Table 7. The newly developed NIF HEPUTLPBdef.txt is 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 (RFT) Files HEPUILPBFG1l&12.txt and HEPUT.PB_RFT.txt are used for the ILPBA analysis.

The EAB, LPZ, and CR doses are shown in Section 7.0 and compared with the 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. 9.23 and 9.24), and if the total calculated dose is less than the allowable regulatory guide 1.183 dose limit.

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CALCULATION CONTINUATION SHEET SHEET 8 of 27 CALC. NO.: H-1-BG-MDC-1859

REFERENCE:

G. Patel/NUCORE ORIGINATOR, DATE REV: 06/16/2006 1 Mark Drucker/NUCORE REVIEWER/VERIFIER, DATE 06/19/2006 The current licensing basis analysis is documented in the calculation H-1-BG-MDC-1 859, Rev 0. The increases in the proposed EAB, 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 limits. The comparisons in Section 7.2 confirms that the proposed increase in the EAB, LPZ, & CR doses and the total calculated doses are less than the corresponding minimal dose increases and allowable regulatory guide dose limits. Therefore, pursuant to 10 CFR 50.59 guidance as defined in References 9.23 and 9.24, the proposed increase in the core thermal power level and resulting post-ILPBA doses can be adopted as current design and licensing bases for the HCGS.

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REFERENCE:

G. Patel/NUCORE ORIGINATOR, DATE REV: 06/16/2006 1 Mark DruckerN'UCOE REVIEWER/ERIFIER, DATE 06/19t2006 4.0 ASSUMPTIONS Assumptions for Evaluating the Radiological Consequences of an Instrument Line Pipe Break Accident The assumptions for evaluating the radiological consequences of an ILPBA are listed in the following section.

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

SOURCE TERM 4.1 No fuel damage is postulated for the ILPBA because the Safety Guide 11 (Ref. 9.13) Regulatory Position (RGP) C.1.b.(2) requires that the release rate and extent of coolant loss are within the capability of the reactor coolant makeup system, which assures that the core is covered during this event. The released activity is assumed to be the maximum reactor coolant activity concentration permitted, which corresponds to 4.0 tiCi/gm DE 1-131 for the Hope Creek plant (Ref.9.14).

TRANSPORT 4.2 Of the 25,000 pounds of coolant released from the instrument line break, 10,000 pounds (i.e.,

40 percent) flashes to steam. It is assumed that all the iodine in the coolant that flashes to steam enters the steam phase with the coolant.

4.3 10 percent of the iodine in the initially unflashed coolant becomes airborne. All the radioactivity is assumed to be -eleased into the reactor building and then to the atmosphere instantaneously as a ground-level release. No credit is taken for plateout, holdup, or dilution within facility buildings.

4.4 The iodine species released from the instrument line break are assumed to be 97% elemental and 3%

organic.

OFFSITE DOSE CONSEQUENCES:

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

4.5 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 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 Inputs 5.3.4.3 and 5.3.4.4.

4.7 The maximum Low Population Zone 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 following dose acceptance criterion:

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REFERENCE:

G. PateL/NUCORE ORIGINATOR, DATE REV: 06/16/2006 1 Mark Drucker/NUCORE REVIEWER/VERIFIER, DATE 06/19/2006 LPZ Dose Acceptance Criterion: 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 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, 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, 4 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),

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, containment shine dose, and the CR filter shine dose due to an ILPBA 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 an ILPBA.

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 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 5.3.3.4 (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.

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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 of the CR emergency filtration (CREF) system that mitigate airborne activity within the control room.

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

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REFERENCE:

G. Patel/NUCORE ORIGINATOR, DATE REV: 06/16/2006 1 Mark Dnicker/NUCORE REVIEWER/VERIFIER, DATE 06/19/2006 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 ILPBA. The dose mitigation functions of the FRVS and CREF systems are not credited 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.9.4) are compatible to TEDE dose criteria and selected with the objective of producing conservative radiological consequences. For conservatism, the maximum isotopic iodine coolant concentrations based on 4.0 j+/-Ci/g are used.

5.1.4 Meteorology Considerations The control room atmospheric dispersion factors (X/Qs) for the FRVS vent release path 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.

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REFERENCE:

G. Patel/NUCORE ORIGINATOR, DATE REV: 06/16/2006 1 Mark Drucker/NUCORE REVIEWER/VERIEIER, DATE 06/19/2006 Design Input Parameter Value Assigned Reference 5.3 Instrument Line Break Accident Input Parameters 5.3.1 Source Term 5.3.1.1 Proposed extended power 4,031 MWt 9.3, Section 3.2.1 uprate level I 5.3.1.2.a Uprated Iodine Coolant Activity Concentration ( tCi/g) 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 2.00E+04 XE-135M 2.60E+04 KR-85M 6.10E+03 XE-131M 1.50E+01 XE-135 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 Maximum reactor coolant 4.0 j+/-Ci/gm 9.14 iodine activity concentration 5.3.1.4 Mass of total coolant 25,000 lb 9.21 released from break 5.3.1.5 Fuel damage No Fuel Damage 9.13, RGP C.l.b.(2) & Section 4.1.

5.3.2 Activity Transport in Reactor Building (See Figure 1) 5.3.2.1 Reactor Building Volume 4.00E+06 W 9.16 5.3.2.2 Normal RBVS flow rate 99,100 +/- 10% cfm 9.11 218,020 cfm Used in the analysis to simulate 50%

(= [99,100 x 1.10-+0.5) mixing in the RB 5.3.2.3 Chemical Form of iodine released from ILPB Elemental 97.0% Assumed Organic 3.0%

5.3.2.4 Type of release to the Ground level release from FRVS 9.5 atmosphere vent 5.3.3 Control Room Model Parameters (See Figure 2) 5.3.3.1 CR Volume 85,000 fWJ 9.10, Page 10 5.3.3.2 CR Normal Air Inflow 3,000 +/- 10% cfm for 0-720 hrs 9.12 and Assumption 4.13 Rate (conservatively modeled as 13,300 efmn) 5.3.3.3 CR Occupancy Factors Time (fir)  % 9.1, Section 4.2.6 0-24 100 24-96 60 96-720 40 3

3.5E-04 m /sec 9.1, Section 4.2.6 "5.3.3.4 CR Breathing Rate Nuclear Common Revision 12 1

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REFERENCE:

I G. PateL/NUCORE ORIGINATOR, DATE REV: 06/16/2006 1 Mark Drucker/NUCORE REVIEWER/VERIFIER, DATE 06/19/2006 Design Input Parameter Value Assigned Reference 5.3.3.5 CR Atmospheric Dispersion Factors for FRVS Release (YfQ) (secm 3 )

Time (Hr) X/Q (sec/mr) 0-2 1.25E-03 9.5, Section 8.1 2-8 8.09E-04 8-24 3.04E-04 24-96 2.10E-04 96-720 1.59E-04 5.3.4 Site Boundary Release Model Parameters J 1.9E-04 5.3.4.1 EAB Atmospheric(sec/m 3) I 1.9E_04 (X/Q)

J 9.9, Pages 5 & 9 Dispersion Factor 5.3.4.2 LPZ Atmospheric Dispersion Factors (X/Qs)

Time (Hr) X/Q (sec/mr) 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/seq)

(m 5.3.4.4 LPZ Breathing Rates (m/s(ec)

Time (ir) (m&/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 12I

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REFERENCE:

G. Patel/NUCORE ORIGINATOR, DATE REV: 06/16/2006 1 Mark Drucker/NUCORE REVIEWER/VERIFIER, DATE 06/19/2006

6.0 CALCULATIONS

6.1 Post-ILPBA Airborne Mass of Coolant The portion of reactor coolant release from the break that flashes to steam is calculated using the constant enthalpy method (Ref. 9.6, Section I1.3.d). The flashing fraction (FF) is derived as follows:

FF x (steam enthalpy at 212 F) + (1-FF) x (liquid enthalpy at 212 F) =

(saturated liquid enthalpy at temperature of steam at reactor vessel outlet)

Vessel dome temperature = 5470 F (Ref. 9.25, Section 3.3.1, Item 14)

Liquid enthalpy of@ 5470 F = 545.65 BTU/lb (Ref. 9.26, page 182)

Steam enthalpy @ 2120 F = 1150.5 BTU/Ib (Ref. 9.26, page 184)

Liquid enthalpy = 180.17 BTU/lb (Ref. 9.26, page 184)

Substituting, FF = (545.65 - 180.17) / (1150.5 - 180.17) = 0.377 For conservatism, a value of 0.40 or 40% is used below.

Mass of water carrying activity into the flashed steam is calculated as follows:

Reactor coolant directly flashes into steam = 25,000 lb x 0.40 = 10,000 lb Reactor coolant flashed from un-flashed coolant = 0.10 x (25,000 lb- 10,000 lb) =1,500 lb Total amount of coolant flashed into steam and became airborne

= 10,000 lb + 1,500 lb = 11,500 lb x 453.6 g/lb = 5.216E+06 g used in Table 7 6.2 Dose Conversion Factors and Iodine Spike Coolant Activity Iodine isotopic dose conversion factors (DCFs) are obtained from Reference 9.18, 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 renm/Ci/Sv/Bq = 1.08E+06 rem/Ci 6.3 Steam Mass Flow Rate:

Uprated Steam Flow Rate

= 17,774,000.0 lb/br (Ref. 9.3, Section 3.2.1) = 17,774,000.0 lb/hr x 453.6 g/lb x 1/3600 hr/sec

= 2,239,524.0 g/sec =-2.24E+06 g/sec This conversion factor is used to convert noble gas release in pCi/sec to pCi/g in Table 4.

I Nuclear Common Revision eiin1 12 I Nula!omn

CALCULATION CONTINUATION SHEET SHEET 16 of 27 CALC. NO.: H-1-BG-MDC-1859

REFERENCE:

G. Patel/NUCORE ORIGINATOR, DATE REV: 06/16/2006 1 Mark Drucker/NUCORE REVIEWER/VERIFIER, DATE 06/19/2006 7.0 RESULTS

SUMMARY

7.1 The results of ILPBA analysis with the maximum iodine concentration are summarized in the following table:

Instrument Line Pipe Break Accident TEDE Dose (rem)

Receptor Location Control Room EAB LPZ Calculated Dose 3.24E-01 7.40E-02 7.41E-03 (0.0br)

Allowable TEDE Limit 5.0OE+00 2.50E+00 2.50E+00

,,_ RADTRAD Computer Run No.

HEPUILPBOO BHEPUILPBOO HEPUILPBOO Significant assumptions used in this analysis:

• Maximum iodine activity concentration is 4.0 ýtCi/gm DE 1-131 v FRVS is not credited.

• 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 mr)n 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 fiC=A*0.03+B D E F=D-C Gff0.1(E-C) H Instrument Line Pipe H-1-BG-MDC-1859, R0 H-1-BG-MDC-1859, Rev 1 Break Accident Control Room 0.196 0 0.00588 0.324 5.00 0.32 0.50 5.00 Exclusion Area Boundary 0.442 0 0.01326 0.074 25.00 0.06 2.50 2.50 Low Population Zone 0.035 0 0.00105 0.0074 25.00 0.01 2.50 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. 9.4)

H Assumed To Be 10% of Regulatory Dose Limit I Nuclear Common Revision 12 1I Revision 12 I Nuclear Common

CALCULATION CONTINUATION SHEET SHEET 17 of 27 CALC. NO.: H-1-BG-MDC-1859

REFERENCE:

I ~ G. Patel/NUCORE ORIGINATOR, DATE REV: 06/16/2006 1 Mark Drucker/NUCORE REVIEWER/VERIFIER, DATE 06/19/2006

8.0 CONCLUSION

S The analysis results presented in Section 7.1 indicate that the EAB, LPZ, and CR doses due to an ILPB accident are within their allowable TEDE limits. The results indicate that CREF system initiation is not required during an ILPB 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 9.23 and 9.24, the proposed increase in the core thermal power level and resulting post-ILPBA dose consequences can be adopted as current licensing and design bases for the HCGS.

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. Humpbreys 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. 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 15.6.2, Revision 2,"Radiological Consequences of The Failure of Small Lines Carrying Primary Coolant Outside Containment," July 1981.

7. 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.

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-1 882, Rev 0, Control Room Envelope Volume.

11. HCGS Mechanical P&ID No. M-76-1, Rev 19, Reactor Building Air Flow Diagram.

12. HCGS Mechanical P&ID No. M-78-1, Rev 21, Auxiliary Building Control Area Air Flow Diagram.

13. Safety Guide 11, 3/10/71, "Instrument Lines Penetrating Primary Reactor Containment."

14. HCGS Technical Specification 3/4.4.5, "Specific Activity" Limiting Condition for Operation I Nuclear Common Revision 12 1

CALCULATION CONTINUATION SHEET SHEET 18 of 27 CALC. NO.: H-1-BG-MDC-1859

REFERENCE:

G. Patel/NUCORE ORIGINATOR, DATE REV: 06/16/2006 1 Mark Drucker/NUCORE REVIEWER/VERIFIER, DATE 06/19/2006

15. Vendor Technical Document (VTD) No. PNO-A46-4100-0047, Rev 2, GE Specification Document No.

22A2703F, Rev 3, Radiation Sources.

16. Specification 5.2.3, Secondary Containment

17. 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

18. Federal Guidance Report 11, EPA-520/1-88-020, Environmental Protection Agency

19. Federal Guidance Report 12, EPA-402- R-93-081, Environmental Protection Agency

20. Critical Software Package Identification No. A-O-ZZ-MCS-0225, Rev 2, RADTRAD Computer Code.

21. HCGS UFSAR Section 15.6.2.5.2.2, Fission Product Release to the Environment.

23. PSEG Procedure No. NC.NA-AS.ZZ-0059(Q), Rev 11, 10CFR50.59 Program Guidance.

24. Nuclear Energy Institute Report No. NEI 96-07, Rev 1, Guidelines for 10 CFR 50.59 Implementation.

25. Vendor Technical Document (VTD) No. 430003, Volume 002, Rev 1, EPU TR T0100, Reactor Heat Balance.

26. ASME Steam Tables, Sixth Edition.

I Nuclear Common Revision 12 1 eiin1 I Nula omn

CALCULATION CONTINUATION SHEET SHEET 19 of 27 CALC.NO.: H-1-BG-MDC-1859

REFERENCE:

G. Patel/NUCORE I ORIGINATOR, DATE REV: 06/16/2006 1 Mark Drucker/NUCORE REVIEWER/VERIFIER, DATE 06/19/2006 10.0 TABLES Table 1 Iodine Isotopic Dose Conversion Factors Isotopic Conversion Iodine Dose Factor Dose Isotope Conversion Conversion Factor Factor (Sv/Bq) (remnCl/Sv/Bq) (rem/Ci)

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 9.18, Page 136 Table 2 Iodine Scaling Factors Normal Iodine Iodine Dose Product Isotope Activity Conversion Concentration Factor pCi.rem ICLg ACi/g (rem/Cl) (rem)

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 Iodine Scaling Factor Based on 4 Cg DE 1-131 1.229E+02 I

I Nuclear Common Revision Revision 12 12 I1 INuclear Common

I CALCULATION CONTINUATION SHEET SHEET 20 of 27 CALC. NO.: H-1-BG-MDC-1859

REFERENCE:

G. Patel/NUCORE ORIGINATOR, DATE REV: 06/16/2006 1 Mark Drucker/NUCORE REVIEWER/VERIFIER, DATE 06/19/2006 Table 3 Iodine Concentration Based On Maximum DE Iodine Concentration Normal Iodine Iodine Iodine Scaling Activity Isotope Activity Factor Concentration Concentration AC /g LCl/g A B C-AxB 1-131 1.30E-02 1.229E+02 1.598E+00 1-132 1.20E-01 1.229E+02 1.475E+01 1-133 8.90E-02 1.229E+02 1.094E+01 1-134 2.40E-01 1.229E+02 2.951E+01 1-135 1.30E-01 1.229E+02 1.598E+01 A From Reference 9.3, Appendix A B Scaling Factor Based on 4 jtCi/g DE 1-131 From Table 2 Table 4 Normal Noble Gas Concentration Noble Gas Uprated Normal Release Rate Steam Mass Noble Gas Isotope At t -0 Flow Rate Activity (pCi/sec) (g/sec) Concentration (PCi/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.OOOE+01 2.240E+06 8.931E-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 2.9001+02 2.240E+06 1.295E-04 Xe-133 8.200E1+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-138 8.900E+04 2.240E+06 3.974E-02 A From Reference 9.15, Table V B From Section 6.3 I Nuclear Common Revision 12 1I eiin1 I ula omo

CALCULATION CONTINUATION SHEET SHEET 21 of 27 CALC. NO.: H-1-BG-MDC-1859

REFERENCE:

G. Patel/NUCORB ORIGINATOR, DATE REV: 06/16/2006 1 Mark Drucker/NUCORE REVIEWER/VERIFIER, DATE 06/19/2006 Table 5 HCGS Reactor Coolant Concentration Based on 100/E-BAR Normal Average Energy Weighted EPU Mev/Dis Energy Isotope Activity Beta Gamma Total E-Bar Concentration Mev.ttClldis.g PCL/g Al Bi Ca 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 0364 1.955 2.319 0.0207 Xe-131m 6.70E-06 0.144 0.020 0.164 0.0000 Xe-133m 1.29B-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 3.10E-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 1.10E-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 4.00E-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 1.10E-04 0.139 2.166 2.305 0.0003 Cs-137 2AOE-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 I Nuclear Common Revision 12 I Revision 12 1 I Nuclear Common

CALCULATION CONTINUATION SHEET SHEET 22 of 27 CALC. NO.: H-1-BG-MDC-1859

REFERENCE:

G. Patel/NUCORE ORIGINATOR, DATE REV: 06/16/2006 1 Mark Drucker/NUCORE REVIEWER/VERIFIER, DATE 06/19/2006 Table 5 (Con'd)

HCGS Reactor Coolant Concentration Based on 100/E-BAR Design Average Energy Weighted Basis Mev/Dis Energy Isotope Activity Beta Gamma Total E-Bar Concentration Mev.+Cl/dis.g 1C0/g Al Bi Ci Di=Bi+Ci EI=AI*Di Ba-141 1.70E-01 0.901 0.845 1.746 0.2968 Ce-141 3.90E-05 0.171 0.076 0.247 0.0000 Ce-143 3.50E-05 0.433 0.282 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 1AOE-05 0.270 0.140 0.410 0.0000 Np-239 2.40E-01 0.260 0.173 0A33 0.1039 Na-24 2.00E-03 0.554 4.121 4.675 0.0094 P-32 2.00E-05 0.695 0.000 0.695 0.0000 Cr-51 5.00E-04 0.004 0.033 0.036 0.0000 Mn-54 4.00E-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.OOE-04 0.097 2.504 2.601 0.0013 Fe-59 8.00E-05 0.118 1.189 1.307 0.0001 Ni-65 3.00E-04 0.632 0.549 0.869 0.0003 Zn-65 2.00E-06 '0.007 0.584 0.591 0.0000 Zn-69m 3.00E-05 0.022 0.417 0.439 0.0000 Ag-110mi 6.00E-05 0.072 2.751 2.823 0.0002 W-187 3.002-03 0.312 0.481 0.793 0.0024 F-18 4.00E-03 0.250 1.022 1.272 0.0051 Total 1A6E+00 Total 1.86E+00 Ai From Reference 9.15, Tables IMI & IV Bi & Ci From Reference 9.19, Appendix A for isotopes 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 - (100IE-BAR)/Sum (Al) 53.887 Revision 12 ]

Nuclear Common 1 Nuclear Common Revision 12 [

CALCULATION CONTINUATION SHEET SHEET 23 of 27 CALC. NO.: H-1-BG-MDC-1859

REFERENCE:

G. Pat__ _NUCORE__ _ _

ORIGINATOR, DATE REV: 06/16/2006 1 Mark Drucker/NUCORE REVIEWERVERIFIER, DATE 06/19/2006 Table 6 Normal Noble Gas Concentration Based on 100AE-BAR Normal Noble Gas Noble Gas Noble Gas Scaling Concentration Isotope Activity Factor Based On Concentration Based On 100/E-BAR (jCi1g) 100IE-BAR (PCig)

A B C=-AxB Kr-83m 1.518E-03 5.389E+01 8.181E-02 Kr-85m 2.724E-03 5.389E+01 1.468E-01 Kr-85 8.931E-06 5.389E+01 4.8121-04 Kr-87 8.931E-03 5.389E+01 4.812E-01 Kr-88 8.931E-03 5.389E+01 4.812E-01 Xe-131m 6.698E-06 5.389E+01 3.609E-04 Xe-133m 1.295E-04 5.389E+01 6.978E-03 Xe-133 3.662E-03 5_389E+01 1.973E-01 Xe-135m 1.161E-02 5.389E+01 62.56E-01 Xe-135 9.824E-03 5.389E+01 52.94E-01 Xe-138 3.974E-02 5.389E+01 2.142E+00 A From Table 4 B From Table 5 I

I Nuclear Common Revision eiin1 12 1 I Nula omn

CALCULATION CONTINUATION SHEET SHEET 24 of 27 CALC. NO.: H-1-BG-MDC-1859

REFERENCE:

G. Patel/NUCORE ORIGINATOR, DATE REV: 06/16/2006 1 Mark Drucker/NUCORE REVIEWER/VERIFIER, DATE 06/19/2006 Table 7 Post-ILPBA Activity Release To Environment Iodine & Post-ILPB Post-ILPB Noble Gas Coolant Activity Isotope Activity Mass Release Concentration Release 11019 (9) (C)

A B C=AxBIlE6 1-131 1.598E+00 52.16E+06 .8336E+01 1-132 1.475E+01 5.216E+06 .7695E+02 1-133 1.094E+01 5.216E+06 .5707E+02 1-134 2.951E+01 5.216E+06 .1539E+03 1-135 1.598E+01 5.216E+06 .8336E+02 Kr-83m 8.181E-02 5.216E+06 .4267E+00 Kr-85m 1A68E-01 5.216E+06 .7656E+00 Kr-85 4.812E-04 5.216E+06 .2510E-02 Kr-87 4.812E-01 5.216E+06 .2510E+01 Kr-88 4.812E-01 5.216E+06 .2510E+01 Xe-131m 3.609E-04 5.216E+06 .1883E-02 Xe-133m 6.978E-03 5.216E+06 .3640E-01 Xe-133 1.973E-01 5.216E+06 .1029E+01 Xe-I35m 6.256E-01 5.216E+06 .3263E+01 Xe-135 5.294E-01 5.216E+06 .2761E+01 Xe-138 2.142E+00 5.216E+06 1117E+02 A - Iodine Activity Concentration From Table 3 A - Noble Gas Activity Concentration From Table 6 B - From Section 6.1 I Nuclear Common Revision 12 2I 1 I Nula omnRvso

CALCULATION CONTINUATION SHEET SHEET 25 of 27 CALC. NO.: H-1-BG-MDC-1859

REFERENCE:

G. Patel/NUCORE ORIGINATOR, DATE REV: 06/16/2006 1 Mark Drucker/NUCORE REVIEWERNERIFIER, DATE 06/19/2006 11.0 FIGURES Figure 1: RADTRAD Nodalization For Instrument Line Pipe Break Accident I Nuclear Common Revision eiin1 12 1I l ula o mo

CALCULATION CONTINUATION SHEET SHEET 26 of 27 CALC. NO.: H-1-BG-MDC-1859

REFERENCE:

G. Patel/NUCORE ORIGINATOR, DATE REV: 06/16/2006 1 Mark Drucker/NUCORE REVIEWERIVERIFIER, DATE 06/19/2006 Figure 2 - HCGS Control Room RADTRAD Nodalization I Nuclear Common Revision 12 1I Revision 12 I Nuclear Common

CALCULATION CONTINUATION SHEET SHEET 27 of 27 CALC. NO.: H-1-BG-MDC-1859

REFERENCE:

G. Patel/NUCORE ORIGINATOR, DATE REV: 06/16/2006 1 Mark Drucker/NUCORE REVIEWER/VERIFIER, DATE 06/19/2006 12.0 AFFECTED DOCUMENTS The following documents will be revised:

1. UFSAR Table 15.6-2, Instrument Line Failure Accident - Parameters Tabulated For Postulated Accident Analyses.

2. UFSAR Table 15.6-5, Instrument Line Failure Radiological Effects (Realistic Analysis).

3. UFSAR Section 15.6.2.5.2.2, Fission Product Release to the Environment

4. UFSAR Section 15.A-4, Instrument Line Failure Accident (Section 15.6.2).

13.0 ATTACHMENTS 13.1 - 1 Diskette with the following electronic files:

Calculation No: H-1-BG-MDC-1859, Rev 1.

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