ML063110190

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Attachment 7 - Calculation No. H-1-CG-MDC-1795, Revision 4, Control Rod Drop Accident Radiological Consequences.
ML063110190
Person / Time
Site: Hope Creek PSEG icon.png
Issue date: 08/07/2006
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-CG-MDC-1795, Rev 4
Download: ML063110190 (25)
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Attachment 7 LR-NO6-0418 LCR H05-01, Rev. I Calculation No. H-1-CG-MDC-1795, Revision 4, Control Rod Drop Accident Radiological Consequences

NC.DE-AP.ZZ-01302(Q) ffC.DE4-PY. 2(O). Rev. 12. Form 11 CALCULATION COVER SHEET Page I of 24 CALCULATION NUMBER: H-1-CG-MDC-1795 REVISION: 4 Tri'L: IControl Rod Drop Accident Radiological Consequences NSJWS ICALQ 1 24 1#AITI#SHT: I1n IIDWVlSO.59t7Z48 SMT: 104161 #TOTAL SERTS: 3 CHECK ONE:

0 FINAL ] INTERIM (Proposed Plant Change) OVOID o FINAL (Future Confirmation Req'd, enter tracking Notification number:.)___

sAL*E*M

  • oPE CREEK: 0 Q- uST 0 IMPORTANTT* SAFETY 0 NON-SAFETY RELATED HOPE CREEK ONLY: NQ 0Qs OQs [IF OR ISFSl: 0 IMPORTANT TO SAFETY 0 NOT IMPOkTANT TO SAFETY O ARE STATION PROCEDURES IMPACTED? YES [ NO ID IF YES'.INTERFACE WITH THE SYSTEM ENGINEER &PROCEDURE SPONSOR. ALL IMPACTED PROCEDURES SHOULD BE IDENTIFE%).IN ASECTION INTHE cALC.mJATION BODY cRCA70038194o02801. INCLUDE AN SP OPERATION FOR UPDATE AND

.ISTTHE3SAP ORDERS HERE AND WtMIfN THE BODY-OFTUS CALCULATION.

[0 CP and ADs INCORPORATED (IF ANY); ..... _ __'

JDESQRn'TON,O* &*AA*-JTMREMv ON AFMAr*

See Revision 4 Ilt"r olineic page.

PRPMOSE:

Thepurpose ofthis calculation is to determine the Exclusion AraR Boundary (EAB), LowPopulation Zone "LZ),and Control Room.

(CR) doses due to a Control Rod Drop Accident (CRDA) us*ng the Altrnative Source Term (AST) and core thermal power level of 4.031 MW(, mddg1crany

-CONCLUSIONS:

The analysis results presented in Section 7.1 indicatetha the EAB,LPZ,ý ad CR doses due to aCDA are within their allowable TEDE Emits. The tesults indicate that CREF system Initiation is not required during a CRDA-.

The comparisons in Section 72 document adecrea. i the proposed EAB dose; the EXB dose decrease is dueto the lower proposed Iodine activity release. The comparisons in Section72 confirmthat the proposed increase in the CR dose is less than the minimal dose increase regulatory limit, and that the total calculated EA and CR doses are less than the allowable regulatory guide imits. Therefore, pursuant to 10 CFR 50.50 guidance ps defined in 1Aeferen*es 923 and 9.24, the proposed increase in the core thermal power level and resulting post-CRDA doses can be adopted as Current design and licensrng bases for the HCGS.

I Nuclear Common Revision eiin1 12 1 I Nula omn

CALCULATION CONTINUATION SHEET SHEET 2 of 24 CALC.NO.: H-1-CG-MDC-1795

REFERENCE:

G. Patel/NUCORE, ORIGINATOR, DATE REV: 05/10/2006 4 M. Drucker/NUCORE, REVIEWER/VERIFIER, DATE 05/15/2006 REVISION HISTORY Revision Description 0 Original Issue 1 Revised (see Order 70009023, Activity 0020) to provide information relative to:

" Specific assumptions made (that is, the mechanical vacuum pumps are assumed to be tripped)

" Evaluation against regulatory limits (that is, 10CFR100 and SRP Section 6.4 guidelines)

" Explanation of any qualitative relationships to any other accidents described in the HCGS-UFSAR (that is, LOCA)

Moreover, the analysis is revised to correct the TACT5 input error identified in Notification 20035343.

Revision bars are not used due to the extent of the revision.

2 Revised (see Order 70020574, Activity 0010) to incorporate a revised 10CFR50.59 Screening relating to Revision 1 of this calculation.

3 Revised (see Order 70022227, Activity 0010) to incorporate a revised 10CFR50.59 Screening relating to Revision 1 of this calculation.

4 Complete revision to perform AST analysis for the EPU 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 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 INuclear Common Revision 12 1 INuclear Common Revision 12 I

CALCULATION CONTINUATION SHEET SHEET 3 of 24 CALC. NO.: H-1-CG-MDC-1795

REFERENCE:

G. PateIINUCORE, ORIGINATOR, DATE REV: 05/10/2006 4 M. Drucker/NUCORE, REVIEWERNVERIFIER, DATE 05/15/2006 PAGE REVISION INDEX PAGE REV PAGE REV 1 4 15 4 2 4 16 4 3 4 17 4 4 4 18 4 5 4 19 4 6 4 20 4 7 4 21 4 8 4 22 4 9 4 23 4 10 4 24 4 11 4 Attachment 13.1 4 12 4 13 4 14 4 eiin1 II Nuclear Nula omn Common Revision 12

CALCULATION CONTINUATION SHEET SHEET 4 of 24 CALC. NO.: H-1-CG-MDC-1795

REFERENCE:

G. Patel/NUCORE, ORIGINATOR, DATE REV: 05/1042006 4 M. Drucker/NUCORE, REVIEWER/VERIFIER, DATE 05/1512006 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 8 5.0 Design Inputs 11 6.0 Calculations 16 7.0 Results Summary 17 8.0 Conclusions 18 9.0 References 18 10.0 Tables 20 11.0 Figures 22 12.0 Affected Documents 24 13.0 Attachments 24 I Nuclear Common Revision 12 1 I Nula omnRvso 2I

CALCULATION CONTINUATION SHEET SHEET S of 24 CALC. NO.: H-I-CG-MDC-1795

REFERENCE:

G. Patel/NUCORE, ORIGINATOR, DATE REV: 05/10/2006 4 M. Dnicker/NUCORE, REVIEWER/ERIFIER, DATE 05/15/2006

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 Control Rod Drop Accident (CRDA) using the Alternative Source Term (AST) and core thermal power level of 4,031 MWt, including the instrument uncertainty.

2.0 BACKGROUND

Hope Creek Technical Specification (TS) LIMITING CONDITION FOR OPERATION (LCO) 3/4.3.10 requires that two channels of the main steam line radiation - high, high function for the mechanical vacuum pump (MVP) trip shall be operable. This LCO 3/4.3.10 assures that the post-CRDA fission product release path to the environment would be through the main condenser.

The MVP trip is required to be OPERABLE in OPERATIONAL CONDITIONS 1 and 2 when any mechanical vacuum pump is in service (i.e., taldng a suction on the main condenser) and any main steam line is not isolated, to mitigate the consequences of a postulated CRDA. In this condition fission products released during a CRDA could be discharged directly to the environment. Therefore, the MVP trip is necessary to assure conformance with this calculation's assumption that the post-CRDA radiological release path is via the condenser. In OPERATIONAL CONDITION 3, 4 or 5, the consequences of a CRDA are insignificant, and are not expected to result in any fuel damage or fission product releases. When the MVP is not in service or the main steam lines are isolated, fission product releases via the MVP pathway would not occur.

The function of MVP is to evacuate the condenser during startup. Operating Procedure HC.OP-SO.CG-0001(Q) (Ref. 9.11) includes Precaution 3.1.2, which identifies that operation of the mechanical vacuum pumps while radioactive steam is being admitted to the main condenser will result in high radiation levels at the south plant vent. The procedure also includes Limitation 3.2.4, which calls for securing the mechanical vacuum pumps from service and placing the steam jet air ejectors (SJAE) in service prior to reactor power exceeding 5%. The expected MVP response following a CRDA is to be automatically tripped due to either loss of offsite power or a main steam radiation monitor signal (Ref 9.13). The post-CRDA activity release through the MVP during startup will be insignificant due to the MVP operation limited to 5% core power. For the post-CRDA release through the Gaseous Waste Management System (GWMS) including the SJAE, all of the iodine that enters the off-gas treatment system is retained indefinitely and does not contribute to the CR and off-site dose (Ref. 9.12, page 3). Therefore, the post-CRDA dose impact for the releases through the MVP during the startup at a low power level and GWMS during normal operation at a rated power level will be bounded by the post-CRDA release through the isolated condenser, which is analyzed in the following section.

3.0 ANALYTICAL APPROACH:

This analysis uses Version 3.02 of the RADTRAD computer code to calculate the potential radiological consequences of the CRDA. The RADTRAD code was developed by Sandia National Laboratories, the NRC's technical contractor, for the staff to use in establishing fission product transport and removal models and in estimating radiological doses at selected receptors at nuclear power plants. The RADTRAD code is documented I Nuclear Common Revision 12 1I Revision 12 I Nuclear I

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CALCULATION CONTINUATION SHEET SHEET 6 of 24 CALC. NO.: H-1-CG-MDC-1795

REFERENCE:

G. Patel/NUCORE, ORIGINATOR, DATE REV: 05/10/2006 4 M. Drucker/NUCORE, REVIEWER/VERIFIER, DATE 05/15/2006 in NUREG/CR-6604 (Ref. 9.2). The RADTRAD code is maintained as Software ID Number A-0-ZZ-MCS-0225 (Ref. 9.15).

The consequences of a CRDA are analyzed using the as-built plant specific as-built design and licensingbases inputs, which are compatible to the AST and TEDE dose criteria. There is no specific ESF function credited in the analysis.

For the CRD accident, the release from the breached fuel is based on an NRC approved fuel vendor methodology for the number of fuel rods breached and the assumption that 10% of the core inventory of noble gases and iodine, and 12% of the core inventory of alkali metals are in the fuel gaps. The release attributed to fuel melting is based on the fraction of the fuel that reaches or exceeds the initiation temperature for fuel melting and on the assumption that 100% of the noble gases and 50% of the iodines contained in that fraction are released to the reactor coolant. The activities released from the fuel gaps and melted fuel are assumed to be instantaneously mixed in the reactor coolant within the pressure vessel. Of the activity released to the reactor coolant, 100% of the noble gases, 10% of the iodine, and 1% of the remaining radionuclides are assumed to reach the turbine and condenser. Of the activity that reaches the turbine and condenser, 100% of the noble gases, 10% of the iodine, and 1% of the remaining radionuclides are available for release to the environment.

The turbine and condenser leak to the atmosphere as a ground-level release at a rate of 1% per day for a period of 24 hours2.777778e-4 days <br />0.00667 hours <br />3.968254e-5 weeks <br />9.132e-6 months <br />, at which time the condenser leakage is assumed to terminate. No credit is taken for dilution or holdup within the turbine building. The post-CRDA activity from the turbine and condenser can be released to the turbine building (TB) and to the environment at ground level through the south plant vent when offsite power is available; and through the TB louvers/TB vent during a loss of offsite power (Refs. 9.16 & 9.17). 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/mr)

Interval South Plant TB Louvers TB Vent (hr) Vent (s/m3) (s/m3) 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 9.1 lE-05 24-96 9.08E-04 1.00E-04 5.37E-05 96-720 7.01E-04 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 0 to 24 hour2.777778e-4 days <br />0.00667 hours <br />3.968254e-5 weeks <br />9.132e-6 months <br /> post-CRDA release prior to the condenser leakage being terminated. Therefore, the CR dose is calculated using the post-CRDA release through the TB louvers. The Control Room Emergency Filtration (CREF) system is not credited in the analysis. The CR is assumed to operate in a normal mode of operation with a normal HVAC inflow rate of 3,300 cfin (3,000 cfm + 10 % uncertainty) for the entire duration of the accident. The resulting doses at the EAB, LPZ, and CR locations are compared with the dose acceptance criteria in Section 7.0.

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CALCULATION CONTINUATION SHEET SHEET 7 of 24 CALC. NO.: H-1-CG-MDC-1795

REFERENCE:

G. Patel/NUCORE, ORIGINATOR, DATE REV: 05/10/2006 4 M. Drucker/NUCORE, REVIEWERNVERIFIER, DATE 05/15/2006 The core activity inventory is obtained from Reference 9.3, which is calculated based on a thermal power level of 4,031 MWt. A radial peaking factor of 1.75 is conservatively used instead of the 1.5 value recommended in Reference 9.6. The isotopic activity available for release from the condenser are calculated in Tables 1 & 2 based on the core activity inventory obtained from Reference 9.3 and the CRDA failed and melted fuel fractions from Reference 9.12 (Section 6.2.2).

The RADTRAD V3.02 (Ref 9.2 & 9.15) default nuclide inventory file (NIF) Bwr def. NIP is modified based on the isotopic activities calculated in Table 2. The newly developed plant-specific nuclide inventory file (HEPUCRDA_def.txt) is further modified to include Kr-83m, Xe-13lm, Xe-133m, Xe-135m, Xe-138, Rb-88, and Cs-138 isotopes. The RADTRAD3.02 dose conversion factor (DCF) File (Fgrl 1&12) is modified to include the DCFs obtained from References 9.7 & 9.8 for the added noble gas isotopes. The modified DCF file HCRDAFG1I&12.txt is used in the CRDA analysis. The newly developed release fraction and timing file (HCRDARFT.txt) is used to postulate an instantaneous post-CRDA release. The NIP is developed based on the actual activity in curies released to the environment from the condenser; therefore, the thermal power level is set to unity in the RADTRAD input.

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. 9.26) 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. The current licensing basis analysis is documented in the calculation H-1-CG-MDC-1975, Rev 3. The increases in the proposed EAB and 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 comparisons in Sections 7.2 confirm that the proposed increases in the EAB & 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 9.23 and 9.24, the proposed increase in the core thermal power level and resulting post-CRDA doses can be adopted as current design and licensing bases for the HCGS.

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CALCULATION CONTINUATION SHEET SHEET 8 of 24 CALC. NO.: H-1-CG-MDC-1795

REFERENCE:

G. Patel/NUCORE, ORIGINATOR, DATE REV: 05/10/2006 4 M. Drucker/NUCORE, REVIEWERJVERIFIER, DATE 05/15/2006

4.0 ASSUMPTIONS

Assumptions for Evaluating the Radiological Consequences of a Control Rod Drop Accident (CRDA)

The assumptions in these sections are acceptable for evaluating the radiological consequences of a CRDA.

These assumptions supplement the guidance provided in Regulatory Guide 1.183, Appendix C (Ref. 9.1). These assumptions are incorporated as design inputs in Sections 5.3 through 5.5 for the CRDA analysis.

Source Term Assumptions 4.1 Per Reference 9.12 (Section 6.2.1), in the event of a CRDA 850 fuel rods are breached, and 0.77 percent of these breached rods experience fuel melt. Per Reference 9.14 there are 764 fuel assemblies contained in the reactor core, and per Reference 9.20 there are 62 fuel rods in each reactor assembly.

4.2 Per Reference 9.1, Appendix C, Section 1, the release from the breached fuel is based on the estimate of the number of fuel rods breached and the assumption that 10% of the core inventory of the noble gases and iodine is in the fuel gap, as incorporated in design input 5.3.1.7. The release attributed toWfuel melting is based on the fraction of the fuel that reaches or exceeds the initiation temperature for fuel melting and on the assumption that 100% of the noble gases and 50% of the iodine contained in that fraction are released to the reactor coolant, as incorporated in design input 5.3.1.11. In additionper Reference 9.1, Section 3.2, for non-LOCA events the release fraction of Alkali Metals from Table 3 is incorporated in Design Input 5.3.1.7 in conjunction with the core fission product inventory in Design Input 5.3.1.2 for the core thermal power level of 4,031 MWt. The-bromines are neglected from thyroid dose consideration due to their low thyroid dose conversion factors, relatively short half lives, and decaying into insignificant daughters.

4.3 Per Reference 9.1, Appendix C, Section 3.1, the activity released from either the gap or from fuel pellets is assumed to be instantaneously mixed in the reactor coolant within the pressure vessel.

4.4 Per Reference 9.1, Appendix C, Section 3.2, credit is not assumed for partitioning in the pressure vessel or for removal by the steam separators.

4.5 Per Reference 9.1, Appendix C, Section 3.3, of the activity released from the reactor coolant within the.

pressure vessel, 100% of the noble gases, 10% of the iodine, and 1% of the remaining radionuclides are assumed to reach the turbine and condensers, which is incorporated in the design input 5.3.1.8.

4.6 Per Reference 9.1, Appendix C, Section 3.4, of the activity that reaches the turbine and condenser, 100%

of the noble gases, 10% of the iodine, and 1% of the particulate radionuclides are available for release to the environment, which is incorporated in design input 5.3.1.9. The turbine and condenser leak to the atmosphere as a ground-level release at a rate of 1% per day for a period of 24 hours2.777778e-4 days <br />0.00667 hours <br />3.968254e-5 weeks <br />9.132e-6 months <br />, at which time the leakage is assumed to terminate (see design inputs 5.3.2.1 through 5.3.2.3). No credit is taken for dilution or holdup within the turbine building, which is incorporated in the design input 5.3.2.6.

Radioactive decay during holdup in the turbine and condenser is assumed.

4.7 Per Reference 9.1, Appendix C, Section 3.6, the iodine species released from the reactor coolant within the pressure vessel is assumed tobe 95% CsI as an aerosol, 4.85% elemental, and 0.15% organic, which is incorporated in the design input 5.3.2.4. The release from the turbine and condenser is assumed to be 97% elemental and 3% organic, which is incorporated in the design input 5.3.2.5.

Nuclear Common Revision 12

CALCULATION CONTINUATION SHEET SHEET 9 of 24 CALC. NO.: H-1-CG-MDC-1795

REFERENCE:

G. Patel/NUCORE, ORIGINATOR, DATE REV: 05/10/2006 4 M. Drucker/NUCORE, REVIEWER/VERIFIER, DATE 05/15/2006 Offsite Dose Consequences:

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

4.9 The maximum EAB TEDE for any two-hour period following the start of the radioactivity release 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: 6.3 Rem TEDE 4.10 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.

4.11 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: 6.3 Rem TEDE 4.12 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:

4.13 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),
  • Radiation shine from radioactive material in the reactor containment (i.e., containment shine dose; not applicable to a CRDA release occurring outside containment),

I Nuclear Common Revision 12

CALCULATION CONTINUATION SHEET SHEET 10 of 24 CALC. NO.: H-1-CG-MDC-1795

REFERENCE:

G. Patel/NUCORE, ORIGINATOR, DATE REV: 05/10/2006 4 M. Drucker/NUCORE, REVIEWER/VERIFIER, DATE 05/15/2006 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 CR filter shine dose due to a CRDA 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 CRDA.

4.14 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).

4.15 The occupancy and breathing rate of the maximum exposed individual presents in the control room are incorporated in design input 5.3.3 (Ref. 9.1, Section 4.2.6).

4.16 10 CFR 50.67 (Ref 9.4) establishes the following radiological criterion for the control room.

CR Dose Acceptance Criterion: 5 Rem TEDE (50.67(b)(2)(iii))

4.17 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.

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

Revision 12 I IINuclear Common Revision 12

CALCULATION CONTINUATION SHEET SHEET 11 of 24 CALC. NO.: H-1-CG-MDC-1795

REFERENCE:

G. Patel/NUCORE, ORIGINATOR, DATE REV: 05/10/2006 4 M. Drucker/NUCORE, REVIEWER/VERIFIER, DATE 05/15/2006 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 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 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 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 safety-related CR emergency filtration system is not credited for dose mitigation.

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. As a conservative alternative, the limiting value applicable to each portion of the analysis is used in the evaluation of that portion.

5.1.5 Meteorology Considerations The control room atmospheric dispersion factors (X/Qs) for the turbine building louver release point are developed (Ref. 9.5) using the NRC sponsored computer code ARCON96. The EAB and LPZ X/Qs were INuclear Common Revision 12 1I Revision 12 I Nuclear Common

CALCULATION CONTINUATION SHEET SHEET 12 of 24 CALC. NO.: H-1-CG-MDC-1795

REFERENCE:

G. Patel/NUCORE, ORIGINATOR, DATE REV: 05/10/2006 4 M. Drucker/NUCORE, REVIEWER/VERIFIER, DATE 05/15/2006 reconstituted using the HCGS plant specific meteorology and appropriate regulatory guidance. 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 AST and TEDE dose criteria and assumptions are consistent with those identified in Appendix C 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 12 I II'Nuclear Nuclear Common Revision 12I

I CALCULATION CONTINUATION SHEET ISHEET 13 of 24 CALC. NO.: H-1-CG-MDC-1795

REFERENCE:

G. Patel/NUCORE, ORIGINATOR, DATE REV: 05/10/2006 4 M. Drucker/NUCORE, REVIEWER/VERIFIER, DATE 05/15/2006 Design Input Parameter Value Assigned Reference 5.3 CRDA Parameters 5.3.1 Source Term 5.3.1.1 Proposed extended power 4,031 MWt Section 6.1 uprate level 1 5.3.1.2 Isotopic Core Inventory In Ci/MWt 9.3 Isotope Activity Isotope Activity Isotope Activity KR-83M 2.981E+03 1-134 5.937E+04 RB-86 1.300E+02 KR-85 4.711E+02 1-135 5.117E+04 RB-88 1.574E+04 KR-85M 5.908E+03 XE-131M 3.129E+02 CS-134 1.319E+04 KR-87 1.097E+04 XE-133 5.306E+04 CS-136 3.704E+03 KR-88 1.539E+04 XE-133M 1.743E+03 CS-137* 1.096E+04 1-131 2.779E+04 XE-135 1.482E+04 1-132 3.991E+04 XE-135M 1.118E+04

  • CS-137 inventory includes BA-1-133 5.454E+04 XE-138 4.322E+04 137M inventory 5.3.1.3 Radionuclide Composition Group Elements 9.1, Section 3.4, Table 5 Noble gases Xe, Kr Halogens I, Br Alkali metals Cs,Rb 5.3.1.4 Number of fuel rods in 62 9.20 fuel assembly 5.3.1.5 Damaged fuel rods:

Breached Fuel Rods 850 9.12, Section 6.2.2 Melted Fuel Rods 0.77% of the breached fuel rods 5.3.1.6 Number of fuel assemblies 764 9.14 in core 5.3.1.7 Fission products release 10% noble gas in breached rods 9.1, Appendix C, Section 1 from breached fuel rods to reactor 10% iodine in breached rods 9.1, Appendix C, Section 1 coolant 12% Alkali metal inbreached 9.1, Section 3.2, Table 3 rods 5.3.1.8 Fission products transfer 100% noble gas 9.1, Appendix C, Section 3.3 from reactor coolant to turbine/ 10% iodine condenser 1% Alkali metal 5.3.1.9 Fission products available 100% noble gas 9.1, Appendix C, Section 3.4 for release to the environment 10% iodine from turbine/ condenser 1% Alkali metal 5.3.1.10 Radial peaking factor 1.5 9.6, Appendix A, Section 111.7 (1.75 conservatively assumed)

Nuclear Common Revision 12

CALCULATION CONTINUATION SHEET Design Input Parameter Value Assigned Reference 5.3.1.11 Fission products release 100% noble gas in melted fuel 9.1, Appendix C, Section 1 from melted fuel rods to reactor 10% iodine in melted fuel 9.1, Appendix C, Section 1 coolant 25% Alkali metal in melted fuel Assumed based on 9.1, Table 1 5.3.2 Activity Transport in Turbine Building (see Figure 1) 9 .1 , Appendix C, Section 3.4 5.3.2.1 Condenser leak rate 1% per day 5.3.2.2 Duration of 24 hours2.777778e-4 days <br />0.00667 hours <br />3.968254e-5 weeks <br />9.132e-6 months <br /> 9.1, Table 6 and Appendix C, turbine/condenser leak rate Section 3.4 5.3.2.3 Turbine/Condenser leak to Ground level release 9.1, Appendix C, Section 3.4 the atmosphere 1I1 5.3.2.4 Chemical form of Iodine in reactor coolant released within the pressure vessel Aerosol 95% 9.1, Appendix C, Section 3.6 Elemental 4.85%

Organic 0.15%

5.3.2.5 Chemical form of iodine available for release from turbine and main condenser Elemental 97% 9.1, Appendix C, Section 3.6 Organic 3% _

5.3.2.6 Dilution or holdup within Not credited 9.1, Appendix C, Section 3.4 the turbine building 5.3.2.7 Condenser free volume 235,000 t 3 9.13, Page 3 5.3.3 Control Room Parameters (see Figure 2) 85,000 fW3 9.10, page 10 5.3.3.1 CR volume 5.3.3.2 CR normal air inflow rate 3,000 +/- 10% cfru for 0-720 hrs 9.18 and Assumption 4.17 during CRDA (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 breathing rate 3.5E-04 m3 /sec 9.1, Section 4.2.6 5.3.3.5 CR atmospheric dispersion factors for Turbine Building louvers release (X/Qs)

Time (Hr) X/Q (secim) 0-2 6.17E-04 9.5, Section 8.3 2-8 4.OOE-04 8-24 1.44E-04 24-96 1.OOE-04 96-720 7.49E-05 Nuclear Common Revision 12

I CALCULATION CONTINUATION SHEET I SHEET 15 of 24 CALC. NO.: H-1-CG-MDC-1795

REFERENCE:

G. Patel/NUCORE, ORIGINATOR, DATE REV: 05/10/2006 4 M. Drucker/NUCORE, REVIEWER/VERIFIER, DATE 05/15/2006 Design Input Parameter Value Assigned Reference 5.3.4 Site Boundary Release Model Parameters 5.3.4.1 EAB atmospheric J 1.9E-04 (see/n) 9.9, Pages 5 &.9 dispersion factor (X/Q) I 5.3.4.2 LPZ Atmospheric dispersion factors (XIQs)

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 4.0E-06 24-96 1.7E-06 96-720 4.7E-07 5.3.4.3 EAB breathing rate 3.5E-04 m3 /sec 9.1, Section 4.1.3 5.3.4.4 LPZ breathing rates (m 3/sec)

Time (Hr) (m3 lsec) 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 12 1I I Nuclear Common Revision

CALCULATION CONTINUATION SHEET SHEET 16 of 24 CALC. NO.: H-1-CG-MDC-1795

REFERENCE:

G. Patel/NUCORE, ORIGINATOR, DATE REV: 05/10/2006 4 M. Drucker/NUCORE, REVIEWER/VERIFIER, DATE 05/15/2006

6.0 CALCULATIONS

6.1 Extended Uprated Power Level Original Licensed Power Level = 3,293 MWt (Ref. 9.21)

Proposed Power Level Increase = 20%

Instrument Uncertainty = 2% (Ref. 9.22)

Extended Uprated Power Level = 3,293 MWt x 1.20 x 1.02 s 4,031 MWt 6.2 Composite Percentage Release Fractions This calculation uses the gap activity inventory fractions in Table 3 of RG 1.183 and assumes the release of 50% of the iodine and 100% of the noble gases for fuel reaching melted conditions (per RG 1.183, Appendix C, Section 1). Since the fuel gap can also contain the alkali metals (per RG 1.183 Table 1), this calculation applies a gap activity inventory fraction of 12% consistent with RG 1.183 Table 3. Since Appendix C of RG 1.183 does not address the melt release fraction for alkali metals for a CRDA, this calculation will assume 25% of the alkali, metals are released from the melted fuel consistent with RG 1.183 Table 1. Although RG 1.183 Table 1 reports, that a small fraction of other nuclide groups are also released from the melted fuel, these source terms are neglected in this calculation due to 1) a very small fraction of fuel exposed to melt condition (<1%), 2) the small in-vessel release fractions for these nuclide groups, and 3) the low volatility of these aerosols from both reactor coolant and condenser.

Gap Release Melt Release Group Fraction Fraction Noble Gases 10% 100%

Iodine 10% 50%

Alkali Metals 12% 25%

Iodine Release Fraction = (1-0.0077)*10% + 0.0077*50% = 10.308% = 0.10308 NG Release Fraction = (1-0.0077)*10% + 0.0077*100% = 10.693% = 0.10693 Alkali Metals Release Fraction = (1-0.0077)*12% + 0.0077*25% = 12.100% = 0.12100 (These composite rod Iodine and NG release fractions are consistent with Reference 9.12,Section 6.2.2)

Total Number of Rods Per Core = 62 rods/assembly (Ref. 9.20) x 764 assemblies (Ref. 9.14) = 47368 rods/core I Nuclear Common Revision 12 1

CALCULATION CONTINUATION SHEET SHEET 17 of 24 CALC. NO.: H-1-CG-MDC-1795

REFERENCE:

G. Patel/NUCORE, ORIGINATOR, DATE REV: 05/10/2006 4 M. Drucker/NUCORE, REVIEWERJVERIFIER, DATE 05/15/2006 7.0 RESULTS

SUMMARY

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

Control Rod Drop Accident TEDE Dose (rem)

Receptor Location Control Room EAB LPZ Calculated Dose 1.37E-01 2.92E-02 6.23E-03 1_ (0.0 hr)

Allowable TEDE Limit 5.0 E+00 6.3E+00 6.3E+00 RADTRAD Computer Run No.

.... _ HEPU3300CRDAOO 1 HEPU3300CRDAOO HEPU3300CRDAOO Significant assumptions used in this analysis:

a Radial peaking factor = 1.75

  • All activity released to the environment at ground level through TB louvers
  • CREF system is not credited.
  • 850 fuel rods breached
  • 0.77% of the breached fuel rods have fuel melt
  • 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 (rem) Total Dose Proposed Minimal Dose Thyroid Whole Equivalent Dose Limit Increase Increase Lmit Body TEDE (rem) (rem) (rem) (rem) (rem)

TEDE TEDE TEDE TEDE TEDE A B C--A*0.03+B D E F=D-C Gf0.1(E-Q H Control Rod Drop H-1-CG-MDC-1795, Rev 3 H-1-CG-MDC-1795, Rev 4 Accident (CRDA)

Control Room 0.657 10.01231 0.03201 0.137 5.00 0.105 0.50 5.00 Exclusion Area Boundary 0.35 1 0.35 1 0.3605 0.0292 25.00 -0.331 2.46 6.30 Low Population Zone Not Calculated 0.00623 25.00 6.30 E From 10 CFR 50.67 (Ref. 9.25)

H From RG 1.183, Table 6 (Ref. 9.1)

Revision 12 I I

I Nuclear Common Revision 12 1

CALCULATION CONTINUATION SHEET SHEET 18 of 24 CALC. NO.: H-1-CG-MDC-1795

REFERENCE:

G. Patel/NUCORE, ORIGINATOR, DATE REV: 05/10/2006 4 M. Drucker/NUCORE, REVIEWER/VERIFIER, DATE 05/15/2006

8.0 CONCLUSION

S:

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

The comparisons in Section 7.2 document a decrease in the proposed EAB dose; the EAB dose decrease is due to the lower proposed iodine activity release. The comparisons in Section 7.2 confirm that the proposed increase in the CR dose is less than the minimal dose increase regulatory limit, and that the total calculated EAB and CR doses are less than the allowable regulatory guide 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-CRDA doses can be adopted as current design andlicensing 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. 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. 430058, Volume 002, Rev 1, EPU TR T0802, Radioactive Source Term - Core Inventory
4. 10 CFR 50.67, "Accident Source Term."
5. Calculation No. H-1 -ZZ-MDC-1 879, Rev 1, Control Room & Technical Support Center X/Qs Using ARCON96 Code
6. NUREG-0800, Standard Review Plan 15.4.9 Appendix A, Revision 2, "Radiological Consequences of Control Rod Drop Accident (BWR)," July 1981
7. Federal Guidance Report 11, EPA-520/1-88-020, Environmental Protection Agency
8. Federal Guidance Report 12, EPA-402-R-93-081, Environmental Protection Agency
9. Calculation No. H-1-ZZ-MDC-1820, Rev 0, Offsite Atmospheric Dispersion Factors
10. Calculation No. H-4-ZZ-MDC-1 882, Rev 0, Control Room Envelope Volume
11. HCGS Procedure No. HC.OP-SO.CG-0001(R), Rev 32, Condenser Air Removal System Operation
12. GE Report NEDO 31400A, October 1992, "Safety Evaluation for Eliminating The Boiling Water Reactor Main Steam Isolation Valve Closure Function and Scram Function of The Main Steam Line Radiation Monitor."

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

CALCULATION CONTINUATION SHEET SHEET 19 of 24 CALC. NO.: H-1-CG-MDC-1795

REFERENCE:

G. Patel/NUCORE, ORIGINATOR, DATE REV: 05/10/2006 4 M. Drucker/NUCORE, REVIEWER/VERIFIER, DATE 05/15/2006

13. HCGS Procedure No. HC.OP-AB.RPV-0008(Q), Rev 0, Reactor Coolant Activity
14. Hope Creek Technical Specification 5.3, Reactor Core - Fuel Assemblies
15. Critical Software Package Identification No. A-O-ZZ-MCS-0225, Rev 2, RADTRAD Computer Code.
16. HCGS General Arrangement Drawings:
a. P-0007-0, Rev 7, Plan EL 171'-0" & EL 201'-0"
b. P-001 1-0, Rev 5, Sections C-C & D-D
17. HCGS Architectural Drawing No. A-0221-0, Sheet 1, Rev. 10, General Plant Roof Plan
18. HCGS Mechanical P&ID No. M-78-1, Rev 21, Aux Bldg Control Area Air Flow Diagram.
19. HCGS Technical Specification 3/4.3.10, Mechanical Vacuum Pump Trip Instrumentation.
20. Nuclear Fuel Section Design Input File, T03.5-043, Revised Refueling Accident (Bundle Drop)

Analysis

21. NRC Safety Evaluation Report NUREG-1048, October 1984, Operation of Hope Creek Generating Station
22. U.S. NRC Regulatory Guide 1.49, Rev 1, Power Levels of Nuclear Power Plants
23. PSEG Procedure No. NC.NA-AS.ZZ-0059(Q), Rev 10, 10CFR50.59 Program Guidance.
24. Nuclear Energy Institute Report No. NEI 96-07, Rev 1, Guidelines for 10 CFR 50.59 Implementation.
25. 10 CFR 50.67, "Accident Source Term."
26. 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.

INuclear Common RevIsion 12 I1 Revision 12 I Nuclear Common

CALCULATION CONTINUATION SHEET SHEET 20 of 24 CALC. NO.: H-1-CG-MDC-1795

REFERENCE:

G. PatelVNUCORE, ORIGINATOR, DATE REV: 05/10/2006 4 M. Drucker/NUCORE, REVIEWER/VERIFIER, DATE 05/15/2006 10.0 TABLES:

Table 1 CRDA Activity In Peak Failed Fuel Core Uprated Radial Total Number Post-CRDA Isotope Inventory Core Thermal Peaking Number of Fuel Activity In Power Level Factor of Fuel Rod Damaged (Ci/MWt) (MWt) Rod Damaged Fuel In Core (CQ)

A B C D E F=(A*B*C*E)/D 1-131 2.779E+04 4031 1.75 47368 850 3.518E+06 1-132 3.991E+04 4031 1.75 47368 850 5.052E+06 1-133 5.454E+04 4031 1.75 47368 850 6.904E+06 1-134 5.937E+04 4031 1.75 47368 850 7.515E+06 1-135 5.117E+04 4031 1.75 47368 850 6.477E+06 KR-83M 2.981E+03 4031 1.75 47368 850 3.774E+05 KR- 85 4.711E+02 4031 1.75 47368 850 5.963E+04 KR- 85M 5.908E+03 4031 1.75 47368 850 7A79E+05 KR- 87' 1.097E+04 4031 1.75 47368 850 1.389E+06 KR-88 1.539E+04 4031 1.75 47368 850 1.948E+06 XE-131M 3.129E+02 4031 1.75 47368 850 3.961E+04 XE-133 5.306E+04 4031 1.75 47368 850 6.717E1+06 XE-133M 1.743E+03 4031 1.75 47368 850 2.206E+05 XE-135 1.482E+04 4031 1.75 47368 850 1.876E+06 XE-135M 1118E+04 4031 1.75 47368 850 1.415E+06 XE-138 4.322E+04 4031 1.75 47368 850 5.471E+06 RB-86 1.300E+02 4031 1.75 47368 850 1.646E+04 RB-88 1.574E+04 4031 1.75 47368 850 1.992E+06 CS-134 1.319E+04 4031 1.75 47368 850 1.670E+06 CS-136 3.704E+03 4031 1.75 47368 850 4.689E+05 CS-137* 1.096E+04 4031 1.75 47368 850 1.387E+06 CS-138 4.840E+04 4031 1.75 47368 850 6.127E+06 A From Reference 93

  • CS-137 inventory includes BA-137M inventory I Nuclear Common Revision 12' I1 RevisIon 12 I Nuclear Common

CALCULATION CONTINUATION SHEET SHEET 21 of 24 CALC. NO.: H-1-CG-MDC-1795

REFERENCE:

G. Patel/NUCORE, ORIGINATOR, DATE REV: 05/10/2006 4 M. Dmcker/NUCORE, REVIEWER/VERIFIER, DATE 05/15/2006 Table 2 Post-CRDA Activity Released From Condenser Post-CRDA Activity Activity Activity Activity Isotope Activity In Release Release Release Available For Damaged Fraction Fraction To Fraction From Release From Fuel From Fuel Condenser Condenser Condenser (CO (Ci)

A* B C D E=A*B*C*D 1-131 3.518E+06 0.10308 0.10 0.10 .3626E+04 1-132 5.052E+I06 0.10308 0.10 0.10 .5208E+04 1-133 6.904E+06 0.10308 0.10 0.10 .7117E+04 1-134 7.515E+06 0.10308 0.10 0.10 .7747E+04 1-135 6.477E+06 0.10308 0.10 0.10 .6677E+04 KR-83M 3.774E+05 0.10693 1.00 1.00 .4035E+05 KR- 85 5.963E+04 0.10693 1.00 1.00 .6377E+04 KR- 85M I7A79E+05 0.10693 1.00 1.00 .7997E+05 KR- 87 1.389E+06 0.10693 1.00 1.00 .1485E+06 KR-88 1.948E+06 0.10693 1.00 1.00 .2083E+06 XE-131M 3.961E+04 0.10693 1.00 1.00 .4235E+04 XE-133 6.717E+06 0.10693 1.00 1.00 .7182E+06 XE-133M 2.206E+05 0.10693 1.00 1.00 .2359E+05 XE-135 1.876E+06 0.10693 1.00 1.00 2.006E+06 XE-135M 1.415E+06 0.10693 1.00 1.00 .1513E+06 XE-138 5.471E+06 0.10693 1.00 1.00 .5850E+06 RB-86 1.64613+04 0.12100 0.01 0.01 .1991E+00 RB-88 1.992E+06 0.12100 0.01 0.01 .2411E+02 CS-134 1.670E+06 0.12100 0.01 0.01 .2020E+02 CS-136 4.689E+05 0.12100 0.01 0.01 .5673E+01 CS-137 1.38711+06 0.12100 0.01 0.01 .1679E+02 CS-138 6.127E+06 0.12100 0.01 0.01 .7413E+02 A From Table 1 B From Section 6.2 I Nuclear Common Revision 12 1I Revision 12 I1~uc1ear Common

CALCULATION CONTINUATION SHEET SHEET 22 of 24 CALC. NO.: H-1-CG-MDC-1795

REFERENCE:

G. Patel/NUCORE, ORIGINATOR, DATE REV: 05/10/2006 4 M. Drucker/NUCORE, REVIEWER/VERIFIER, DATE 05/15/2006 11.0 FIGURES:

Figure 1: RADTRAD Nodalization For CRDA Release

[Nuclear Common Revision 12 1 I Nuclear Common

CALCULATION CONTINUATION SHEET SHEET 23 of 24 CALC. NO.: H-1-CG-MDC-1795

REFERENCE:

G. Patel/NUCORE, ORIGINATOR, DATE REV: 05/10/2006 4 M. Drucker/NUCORE, REVIEWER/VERIFIER, DATE 05/15/2006 Figure 2 - HCGS Control Room RADTRAD Nodalization INuclear Common Revision 12 2I 1

I Nula omnRvso

CALCULATION CONTINUATION SHEET SHEET 24 of 24 CALC. NO.: H-1-CG-MDC-1795

REFERENCE:

G. Patel/NUCORE, ORIGINATOR, DATE REV: 05/10/2006 4 M. Drucker/NUCORE, REVIEWERIVERIFIER, DATE 05/15/2006 12.0 AFFECTED DOCUMENTS:

The following documents will be either superseded or revised:

Document to be superseded Calculation H-1 -CG-MDC-1795, Rev 3 Documents to be revised:

UFSAR Section 15.4.9 UFSAR Table 15.4-6 13.0 ATTACHMENTS:

13.1 - 1 Diskette with the following electronic files:

Calculation No: H-1-CG-MDC-1795, Rev 4.

Comment Resolution Form 2 - 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 12 1I Revision 12 I Nuclear Common

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