Text

Attachment 6, Calculation S-C-ZZ-MDC-1920, Rev. 41R0, 'Fuel Handling Accident Radiological Consequences Evaluation. '
	ML062230324
Person / Time
Site:	Salem
Issue date:	05/18/2006
From:	Drucker M, Andrea Johnson, Gita Patel; NUCORE, Public Service Enterprise Group
To:	; Office of Nuclear Material Safety and Safeguards
References
	LCR S06-07, LR-N05-0309, S-C-ZZ-MDC-1920, Rev. 4IR0 S-C-ZZ-MDC-1920, Rev 4IRO
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Attachment 6 LCR S06-07 LR-N06-0309 Attachment 6 Calculation S-C-ZZ-MDC-1920, Rev. 41R0 Fuel Handling Accident Radiological Consequences Evaluation

(KC.l)F-AP.ZZ,20WZO), Rev. 12. Form 11 CALCULATION COVER SHEET Page I of 45 CALCULATION NUMBER: S-C-ZZ-MD)C-1920 REVISION: 41RO TITLE: Fuel Handling Accidents Radiological Consequences NSHTS (CAW: 145 1#Afl/#SHTS: 2/3 1 fIDV/50.59fl2.48 SHTS: 16 1410 1#TOTAL SHTS: 58 CHECK ONE: 4at 4I1ig/t'&,

FINAL 0IN RIM (Proposed Plant Change) OVOID

[] FINAL (Future Confirmation Req'd. enter tracking Notification number.)_

SALEM OR HOPE CREEK: Li Q - LIST 0 IMPORTANT TO SAFETY L] NON-SAFETY RELATED HOPE CREEK ONLY: OQ O1Q9 E-Qsh []F -]R ISFSI: [] IMPORTANT TO SAFETY El NOT IMPORTANT TO SAFETY 0 ARE STATION PROCEDURES IMPACTED? YES "] NO [D IF "YES, INTERFACE WITH THE SYSTEM ENGINEER &PROCEDURE SPONSOR. ALL IMPACTED PROCEDURES SHOULD BE IDENTIFIED INA SECTION INTHE CALCULATION BODY,[CRCA 70D38194O28a]. INCLUDE AN SAP OPERATION FOR UPDATE AND UST THE SAP ORDERS HERE AND WITHIN THE BODY OF THIS CALCULATION.

0I CP and ADs INCORPORATED (IF ANY): A DESCRIPTION OF CALCULATION REVISION (IF APPL.):

The analysis is revised to calculate doses at various decay times in support of an anticipated submittal for a Technical Specification change. The nature of revision is such that the entire calculation is revised.

PURPOSE:

The purpose of this analysis is to determine the Exclusion Area Boundary (EAB), Low Population Zone (LPZ) and Control Room (CR) doses due to a fuel handling accident (FHA) occurring in the containment building with the containment equipment hatch (CEH) open and in the fuel handling building. The FHA analyses are performed using the Alternative Source Term (AST), guidance in the Regulatory Guide 1.183, Appendix B, TEDE dose criteria, and various fuel decay times.

CONCLUSIONS:

The Sections 8.1 and 8.2 results indicate that the EAB, LPZ, and CR doses are within their respective allowable limits for the FHAs occurring in the containment building and fuel handling building. The FHA occurring in the containment provides basis for changing the following SNGS Technical Specification requirements:

1. The irradiated fuel assemblies can be handled in the reactor pressure vessel (RPV) after the reactor has been sub-critical for at least 24 hours2.777778e-4 days 0.00667 hours 3.968254e-5 weeks 9.132e-6 months . This provides a basis for changing the reactor minimum sub-critical time from 168 hours0.00194 days 0.0467 hours 2.777778e-4 weeks 6.3924e-5 months to 24 hours2.777778e-4 days 0.00667 hours 3.968254e-5 weeks 9.132e-6 months (Technical Specification Limiting Condition for Operation (LCO) 3.9.3)

2. The irradiated fuel assemblies can be moved with the containment equipment hatch and personnel locks opened, and all containment penetrations opened in the piping penetration areas without containment integrity (operability) (Technical Specification LCO 3-9.4)

3. The core alterations can be performed without containment integrity (Technical Specification LCO 3.9.4).

The FRA occurring in the FHB provides basis for relaxing the SNGS Technical Specification Surveillance requirements 4.9.12.b and 4.9.12.c.

Printed Name / Signattp, Date ORIGINATOR/COMPANY NAME: Gopal J. Patel/NUCORE 05/17/2006 REVIEWER/COMPANY NAME: N/A N/A VERIFIER/COMPANY NAME: Mark Drucker/NUCORE! !it , 05/18/2006 CONTRACTOR SUPERVISOR (If applicable) N/A k PSEG SUPERVISOR APPROVAL: (Alwaysereioired) Alan A- Joson/PSEG Nuclear Common 4Revision 1

CALCULATION CONTINUATION SHEET SHEET 2 of 45 CALC. NO.: S-C-ZZ-MDC-1920

REFERENCE:

G. Patel/NUCORE, ORIGINATOR, DATE REV: 05/17/2006 4 M. Drucker/NUCORE, REVIEWERNERIFIER, DATE 05/18/2006 REVISION HISTORY Revision Description 0 Original Issue 1 Editorial changes to various sections 2 Revised EAB X/Qs and changes to various sections 3 Revised to simplify the calculation title, correct a typographical error in Section 4.8a identified in Notification 20104610 IAW NUTS Order 80048072 and correct a typographic error in the heading for Section 6.0. Additionally, revised Section 9.0 to limit the discussion to conclusion and added Section 12.0, identifying affected documents (there are none relating to the revision).

4 The analysis is revised to calculate doses at various decay times in support of an anticipated submittal for a Technical Specification change. The nature of revision is such that the entire calculation is revised.

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I CALCULATION CONTINUATION SHEET SHEET 3 of 45 CALC. NO.: S-C-ZZ-MDC-1920

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G. Patel/NUCORE, ORIGINATOR, DATE REV: 05/17/2006 4 M. Drucker/NUCORE, REVIEWER/VERIFIER, DATE 05/18/2006 PAGE REVISION INDEX PAGE REV PAGE REV PAGE REV 1 4 18 4 35 4 2 4 19 4 36 4 3 4 20 4 37 4 4 4 21 4 38 4 5 4 22 4 39 4 6 4 23 4 40 4 7 4 24 4 41 4 8 4 25 4 42 4 9 4 26 4 43 4 10 4 27 4 44 4 11 4 28 4 45 4 12 4 29 4 13 4 30 4 Attachment 13.1 4 14 4 31 4 Attachment 13.2 4 15 4 32 4 16 4 33 4 17 4 34 4 1 I Nudear Common Revision eiin1 12 I Nula omn

CALCULATION CONTINUATION SHEET SHEET 4 of 45 CALC. NO.: S-C-ZZ-MDC-1920 G. Pate 1/NUCORE,

REFERENCE:

[

ORIGINATOR, DATE REV: 05/17/2006 4 M. Drucker/NUCORE, REVIEWERNVERIFIER, DATE 05/18/2006 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 11 5.0 Design Inputs 16 6.0 Methodology 22 7.0 Calculations 22 8.0 Results Summary 29 9.0 Conclusions/Recommendations 32 10.0 References 33 11.0 Tables 36 12.0 Figures 39 13.0 Attachments 45 14.0 Affected Documents 45 1

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CALCULATION CONTINUATION SHEET SHEET 5 of 45 CALC. NO.: S-C-ZZ-MDC-1920

REFERENCE:

G. Patel/NUCORE, ORIGINATOR, DATE REV: 05/17/2006 4 M. Drucker/NUCORE, REVIEWER/VERIFIER, DATE 05/18/2006 1.0 PURPOSE The purpose of this analysis is to determine the Exclusion Area Boundary (EAB), Low Population Zone (LPZ) and Control Room (CR) doses due to a fuel handling accident (FHA) occurring with the reactor being subcritical for various times in:

1. The containment building (CB) with the containment equipment hatch (CEH), personnel air locks, and other containment penetrations open or

2. The fuel handling building (FHB)

The analyses are performed using the Alternative Source Term (AST), guidance in Regulatory Guide 1.183, Appendix B, and TEDE dose criteria with the different fuel decay times.

2.0 BACKGROUND

PSEG Nuclear is expected to change the minimum fuel decay time requirement for the reactor to be subcritical prior to the movement of irradiated fuel assemblies (Ref. 10.6.2). Fuel handling accidents are postulated in the RB and FHB with the reactor being subcritical for various times. Activity is released to the environment through the opened CEH or the plant vent (PV). The releases are modeled as ground-level releases.

The following technical specification requirements are addressed in the FHA analysis:

3.9.3 DECAY TIME The reactor shall be subcritical for at least 100 hours0.00116 days 0.0278 hours 1.653439e-4 weeks 3.805e-5 months prior to movement of irradiated fuel in the reactor pressure vessel (Ref. 10.6.2). This requirement for the subcritical time is expected to change.

3.9.4 CONTAINMENT BUILDING PENETRATION The containment building penetrations shall be operable during CORE ALTERATIONS or movement of irradiated fuel within containment (Ref. 10.6.1).

3.9.10 WATER LEVEL - REACTOR VESSEL At least 23 feet of water shall be maintained over the top of the reactor pressure vessel (RPV) flange (Ref. 10.6.3).

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CALCULATION CONTINUATION SHEET SHEET 6 of 45 CALC. NO.: S-C-ZZ-MDC-1920

REFERENCE:

G. Patel/NUCORE, ORIGINATOR, DATE REV: 05/17/2006 4 M. Drucker/NUCORE, REVIEWRNERIFIER, DATE 05/18/2006

3.9.11 STORAGE POOL WATER LEVEL At least 23 feet of water shall be maintained over the top of the irradiated fuel assembly seated in the storage racks (Ref. 10.6.9).

1.25 RATED THERMAL POWER (RTP)

RATED THERMAL POWER shall be a total reactor core heat transfer rate to the reactor coolant of 3459 MWt (Ref. 10.6.4).

5.3.1 FUEL ASSEMBLIES The reactor shall contain 193 fuel assemblies (Ref. 10.6.5).

3.3.3.1 RADIATION MONITORING INSTRUMENTATION The radiation monitoring instrumentation channels shown in Technical Specification Table 3.3-6 shall be operable with their alarm/trip setpoints with the specified limits (Ref. 10.6.6).

TABLE 3.3-6 RADIATION MONITOR INSTRUMENTATION The control room normal intake radiation monitors must be operable during fuel movement (Ref 10.6.7).

3.9.12 Fuel Handling Area Ventilation System The fuel handling area ventilation system shall be operable (Ref. 10.6.10).

3.0 ANALYTICAL APPROACH This analysis uses Version 3.02 of the RADTRAD computer code (Ref 10.2) to calculate the potential radiological consequences of an FHA. The RADTRAD code is documented in NUREG/CR-6604 (Ref 10.2).

The RADTRAD code is maintained as Software ID Number A-0-ZZ-MCS-0225 (Ref. 10.33).

The FHA is analyzed using the plant specific design inputs. The design inputs are compatible to the AST and TEDE dose criteria.

The scrubbing of the iodine activity in the reactor cavity and spent fuel storage pool are credited in the analyses.

The scrubbing effects are limited by 23 feet height of water over the top of the RPV flange (Ref. 10.6.3) and over the top of the irradiated fuel assemblies in the spent fuel pool storage racks (Ref. 10.6.9).

The core inventory is obtained from Reference 10.3 (page 33, Table 2), which is calculated based on a thermal power level of 3,600 MWt. The radial peaking factor of 1.7 is conservatively used instead of the 1.65 value i Nuclear Common Revision 12 1

CALCULATION CONTINUATION SHEET SHEET 7 of 45 CALC. NO.: S-C-ZZ-MDC-1920

REFERENCE:

G. Patel/NUCORE, ORIGINATOR, DATE REV: 05/17/2006 4 M. Drucker/NUCORE, REVIEWER/VERIFIER, DATE 05/18/2006 recommended in Reference 10.19. The thermal power level of 3,632 MWt, which is 105% of the rated thermal power level of 3,459 MWt (Ref. 10.6.4), is used in the analysis to provide a margin for future power uprate. The core activity obtained from Reference 10.3 is listed in Table 1 and normalized in Tables 1, 2 & 3 based on the core thermal power level, the gap fission product release fractions in Design Input 5.3.1.3, peaking factor, and one fuel assembly failed during the FHA (Ref. 10.19, page 5). The maximum linear heat generation rate is limited to less than 6.3 kw/ft peak rod average power (Ref. 10.1, Table 3, Note 11). The high power density of cores in Pressurized Water Reactors (PWRs), increased fuel burnup, and extended fuel cycle potentially may increase the maximum heat generation rate to a value exceeding the limit of 6.3 kw/hr peak rod average power for bumups exceeding 54 GWD/MTU at the end of the fuel cycle. Many PWR core design loading analyses have reported fuel assemblies that have exceeded the maximum heat generation rate of 6.3 kw/ft. Therefore, to establish a conservative basis for those fuel assemblies that may in future cycle operations exceed the maximum heat generation rate of 6.3 kw/hr, the gap fission product fractions in Table 3 of RG 1.183 are doubled to the values shown in Section 5.3.1.3 for use in this FHA dose analysis (Table 2). The RADTRAD V3.02 code default nuclide inventory file (NIF) Bwrdef. NIF is modified based on the normalized CitMWt in Table 3. The plant-specific NIF SNGSFHAdef is further modified to include Kr-83m, Xe-131m, Xe-133m, Xe-135m, and Xe-138 isotopes. The RADTRAD V3.02 dose conversion factor (DCF) File Fgrl 1&12 (based on Refs. 10.7 and 10.8) is modified to include the DCFs for the added noble gas isotopes. The modified DCF file SALEMFHA_FGl 1&12 is used in the FHA analyses.

3.1 FIIA Occurring In Containment Building There are one CEH, two personnel air locks, and containment piping penetrations in the containment boundary (Ref. 10.17). The CEH provides a direct release path to the environment (Refs. 10.17.a, 10.17.b, 10.17.g). The personnel air locks and penetrations provide release paths to the environment through the plant vent via piping penetration areas (Refs. 10.17 & 10.18). The most limiting atmospheric dispersion factors for these release paths are obtained from Reference 10.5 and compared in the following table.

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I CALCULATION CONTINUATION SHEET SHEET 8 of 45 CALC. NO.: S-C-ZZ-MDC-1920

REFERENCE:

G. Patel/NUCORE, ORIGINATOR, DATE REV: 05/17/2006 4 M. Drucker/NUCORE, REVIEWER/VERIFIER, DATE 05/18/2006 Salem 1 CR Intake X/Qs (s/m3 )

Time Unit 1 Unit 1 Interval Equip Hatch Plant Vent (hr) Unit 1 Unit 1 CR Intake CR Intake 0-2 2.86E-03 1.78E-03 2-8 2.22E-03 1.31E-03 8-24 9.155E-04 5.22E-04 24-96 6.60E-04 3.77E-04 96-720 5.62E-04 3.17E-04 The comparison of x/Qs in the above table indicates that the CEH provides a conservative release path for the FHA occurring in the containment. Therefore, the EAB, LPZ, and CR doses are calculated using the post-FHA release through the CEH. The activity release rate from the CEH is calculated in Section 7.2 based on the removal of 99% of radioactive material released from the damaged fuel to the environment over a 2-hour period. (Ref. 10.1, Appendix B, Regulatory Position B.5.3). The resulting doses at the EAB, LPZ, and CR locations are compared with the regulatory allowable limits in Section 8.1.

3.2 FHA Occurring In Fuel Handling Building A parametric study is performed to determine a conservative release model using either a post-FHA release rate based on a 0-2 hour release, or a rapid release rate based on one FIIB volume per minute. The results of the parametric study shown in Sections 8.2 & 8.3 indicate that a release based on the rapid release rate of one FHB volume per minute yields a higher CR dose. The puff release yields a higher CR dose because it results in a larger amount of unfiltered iodine activity entering the CR volume prior to the one minute start of the Control Room Emergency Air Conditioning System (CREACS) outside air inflow filtration.

Should a FIIA occur in the FHB, the activity can be either released through the plant vent (Ref. 10.18) or the FHB rollup door at ground level (Ref.10.23). However, the following post-FHA release paths are identified in Reference 10.21 during the FHB pressurization due to a single failure of one FHB exhaust fan:

1. Release through the plant vent, via one operational FHB exhaust fan, at a rate of 15,300 cfin

2. Leakage through truck bay roll-up door at a rate of 3,883 cfm

3. Leakage through gravity damper (that replaced the truck bay exhaust fan) at a rate of 256 cfm Nuclear Common Revision 12 ]

CALCULATION CONTINUATION SHEET SHEET 9 of 45 CALC. NO.: S-C-ZZ-MDC-1920

REFERENCE:

G. Pate /NUCORE, ORIGINATOR, DATE REV: 05/17/2006 4 M. Drucker/NUCORE, REVIEWER/VERIFIER, DATE 05/18/2006 The atmospheric dispersion factors (X/Qs) for the plant vent and FHB rollup doors are calculated in Reference 10.5, Sections 8.2 & 8.3, respectively, using the ARCON96 computer code. The X/Qs for the gravity damper release are conservatively assumed to be same as those for a smoke hatch. The smoke hatch X/Qs are developed in Reference 10.9, Section 8.4 using the ARCON96 computer code. Since the FHA in the FHB release duration is two hours (Ref. 10.1, Appendix B, RGP B.4.1), the plant vent, FHB rollup doors and smoke hatch 0-2 X/Q values are used to calculate the equivalent 0 to 2 hr X/Q in Section 7.5 for a combined post-FHA release path.

The equivalent X/Q is used with the post-FHA unfiltered release from the FIHB to calculate the EAB, LPZ, and CR doses. Activity from the FHB is assumed to be released to the environment at a rate of 21,439 cfin (design flow rate + 10%). The resulting doses at the EAB, LPZ, and CR locations are compared with the regulatory allowable limits in Section 8.2.

3.3 Post-FHA Technical Support Center (TSC) Habitability The TSC habitability is additionally evaluated to fulfill the PSEG Licensing request to evaluate the post-FHA TSC dose. The TSC is located in the Clean Facilities Building (CFB) at the second and third floors (Refs.

10.27.b & 10.27.c). The CFB is located southeast of the Unit 1 containment building (Ref. 10.28). As discussed in Section 3.1 above, the CEH and PV are the release points for the FHA occurring in the containment. As discussed in Section 3.2 above, the plant vent, FHB rollup doors and gravity damper (modeled as the smoke hatch) are the release points for the FH-A occurring in the FHB. The TSC emergency air intake is in the Mechanical Equipment Room located on the roof of CFB (Refs. 10.26, 10.27, & 10.29). The TSC is located closer to Unit 1 containment compared to Unit 2 containment, therefore, the distances between the Unit 1 CEH

& PV and TSC intake are calculated in Section 7.6. These distances are compared with the corresponding distances to the Unit 1 CR intake in Section 7.6. The CR doses are considered bounding for TSC for the F-HA occurring in the containment and FHB because:

1. The TSC intake is located farther from the subject release points in comparison to the CR intakes.

Therefore, the values of corresponding TSC intake X/Qs will be lower than CR intake X/Qs and the resulting post-FHA TSC doses will be lower in the same proportion of X/Qs values.

2. The comparison of CR X/Qs in Reference 10.5, Section 8.1, indicates that the variation of x/Qs due to change in wind direction is insignificant. Therefore, the TSC X/Qs will not be impacted by the differences in wind direction for 0-2 hr period.

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CALCULATION CONTINUATION SHEET SHEET 10 of 45 CALC. NO.: S-C-ZZ-MDC-1920

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G. Patel/NUCORE, ORIGINATOR, DATE REV: 05/17/2006 4 M. Drucker/NUCORE, REVIEWER/VERIFIER, DATE 05/18/2006

3. Manning the TSC occurs some time after initiation of the postulated accident. Therefore, at first there will be a period with no occupancy during the initial phase of the accident.

3.4 CR Intake Monitor Response 3

There are two radiation monitors in each normal CR air intake duct having the alarm/trip set point of 2.48 x 10 cpm (Refs. 10.6.7 & 10.13). These monitors are classified as safety related (Ref. 10.13), are required to be operable in all modes and during movement of irradiated fuel assemblies and during CORE ALTERATION (Refs. 10.6.6 & 10.6.7), are powered by emergency power sources (Ref. 10.22), and are instantaneously actuated by exceeding a predetermined setpoint (Ref. 10.6.7 & Section 7.4). The post-FHA activity at the CR air intake will instantaneously reach the Alert/Trip setpoint (Section 7.4) and actuate the monitors. Therefore, these monitors are credited for automatic initiation CR Emergency Air Conditioning System (CREACS). The CR intake monitor preferential alignment of less contaminated air intake is conservatively not credited. The delay associated with the CR intake damper closure time (20 seconds) (Ref. 10.14, page 8), diesel generator speedup time (13 seconds) (Ref. 10.6.8) if the loss of offsite power is assumed to occur at the time of damper closure, and over-all monitor response time (4 seconds) (Ref. 10.14, Appendix A). The total delay time is less than 1.0 minute. A delay of 1 minute is assumed in the analysis for the initiation of the Control Room Emergency Air Conditioning System (CREACS) and the control room envelope isolation.

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CALCULATION CONTINUATION SHEET SHEET 11 of 45 CALC. NO.: S-C-ZZ-MDC-1920

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G. Pate /NUCORE, ORIGINATOR, DATE REV: 05/17/2006 4 M. Drucker/NUCORE, REVIEWER/VERIFIER, DATE 05/18/2006 4.0 ASSUMPTIONS The regulatory requirements in the Regulatory Guide 1.183, Appendix B (Ref. 10.1) are adopted as assumptions in the following section, which are incorporated as design inputs in Section 5.3 along with other plant-specific as-built design parameters. The assumptions in this section are acceptable by the Staff for evaluating the radiological consequences of FHA occurring in the containment building.

Source Term Assumptions 4.1 Per Reference 10.1, Regulatory Position 3.2, for non-LOCA events, the fractions of the core inventory assumed to be in the gap for the various radionuclides are given in Table 3 of RG 1.183. The release fractions from Table 3 are incorporated in the Design Input 5.3.1.3 in conjunction with the core fission product inventory in Design Input 5.3.1.2 with the maximum core radial peaking factor of 1.70 (Ref.

10.19) and the core inventory at 3,632 MWt power level. The bromines are neglected from thyroid dose consideration due to their low thyroid dose conversion factors, relatively short half-lives, and decay into insignificant daughters.

4.2 Per Reference 10.1, Appendix B, Regulatory Position B. 1.1, the number of fuel rods damaged during the accident should be based on a conservative analysis that considers the most limiting case. One spent fuel assembly is assumed to be damaged (see Design Input 5.3.1.5). Reference 10.31, Section 3.1.3, Risk Significance, indicates that there have been several occasions when fuel bundles have been dropped during fuel handling. In each case, the actual releases from fuel have been minimal or nonexistent. This evidence shows that the assumption of damage of one fuel assembly in the radiological analysis for a FHA is conservative.

4.3 Per Reference 10.1, Appendix B, Regulatory Position B.1.2, the fission product release from the breached fuel is based on fraction of fission product inventory in gap (RGP 3.2) and the estimate of the number of fuel rods breached (See Table 3).

Core Inventory The inventory of fission products in the reactor core and available for gap release from damaged fuel is based on the maximum power level of 3,632 MWt corresponding to current fuel enrichment and fuel burnup. All the gap activity in the damaged rods is assumed to be instantaneously released. The radionuclides included are xenons, kryptons, and iodines. The fraction of fission product in gap activity i Nuclear Common Revision 12

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G. Patel/NUCORE, I ORIGINATOR, DATE REV: 05/17/2006 4 M. Drucker/NUCORE, REVIEWERNERIFIER, DATE 05/18/2006 is shown in Design Input 5.3.1.3. It is further assumed that irradiated fuel shall not be removed from the reactor until the unit has been sub-critical for at least 24 hours2.777778e-4 days 0.00667 hours 3.968254e-5 weeks 9.132e-6 months (Design Input 5.3.1.7).

4.4 Timing of Release Phase Per Reference 10.1, Regulatory Position 3.3, for non-LOCA DBAs in which fuel damage is projected, the release from the fuel gap and the fuel pellet is assumed to occur instantaneously with the onset of the projected damage.

4.5 Chemical Form Per Reference 10.1, Appendix B, Regulatory Position B.1.3, The chemical form of radioiodine released from the fuel to the surrounding water should be assumed to be 95% cesium iodide (CsI), 4.85 percent elemental iodine, and 0.15 percent organic iodine. The CsI released from the fuel is assumed to completely dissociate in the pool water. Because of the low pH of the pool water, the iodine re-evolves as elemental iodine. This is assumed to occur instantaneously.

4.6 Water Depth If the depth of water above the damaged fuel is 23 feet or greater, the decontamination factors for the elemental and organic species are 500 and 1, respectively, giving an overall effective decontamination factor of 200 (i.e., 99.5% of the total iodine released from the damaged rods is retained by the water).

This difference in decontamination factors for elemental (99.85%) and organic iodine (0.15%) species results in the iodine above the water being composed of 57% elemental and 43% organic species (Ref.

10.1, Appendix B, RGP B.2).

4.7 Noble Gases The retention of noble gases in the water in the fuel pool or reactor cavity is negligible (i.e.,

decontamination factor of 1). Particulate radionuclides are assumed to be retained by the water in the fuel pool or reactor cavity (i.e., infinite decontamination factor) (Ref. 10.1, Appendix B, RGP B.3).

Fuel Handling Accidents Within Containment For fuel handling accidents postulated to occur within the containment, the following assumptions are acceptable to the NRC staff (Ref. 10.1, Appendix B, RGP B.5).

4.8a If the containment is open during fuel handling operations (e.g., personnel air lock or equipment hatch is open) the radioactive material that escapes from the reactor cavity pool to the containment is released to I Nuclear Common Revision 12 I ula omo eiin1 I

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G. Patel/NUCORE, ORIGINATOR, DATE REV: 05/17/2006 4 M. Drucker/NUCORE, REVIEWERNERIFIER, DATE 05/18/2006 the environment over a 2-hour time period (Ref. 10.1, Appendix B, RGP B.5.3). The activity release from the damaged fuel is postulated to mix in the RB volume and release to the environment at a rate such that 99% of post-FHA activity is removed from the RB volume (Section 7.2) (Figure 1).

Fuel Handling Accidents Within The Fuel Building For fuel handling accidents postulated to occur within the fuel building, the following assumptions are acceptable to the NRC staff.

4.8b The radioactive material that escapes from the fuel pool to the fuel building is assumed to be released to the environment over a 2-hour time period (Ref 10.1, Appendix B, RGP B.4.1). The activity released from the damaged fuel is postulated to mix in the FHB volume and be released to the environment over a two hour period at a rate of 21,439 cfmn per Design Input 5.3.3.3 (See Figure 2).

A reduction in the amount of radioactive material released from the fuel pool by engineered safety feature (ESF) filter systems is not accounted for in the radioactivity release analyses.

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 and used in determining compliance with the dose acceptance criterion in Reference 10.1, Appendix B, RGP 4.4 and RGP Table 6.

EAB Dose Acceptance Criteria: 6.3 Rein TEDE 4.10 The breathing rates for persons at offsite locations are given in Reference 10.1, RGP 4.1.3, which are incorporated in Design Input 5.3.5.4.

4.11 TEDE is determined for the most limiting receptor at the outer boundary of the low population zone (LPZ) and is used in determining compliance with the dose acceptance criterion in Reference 10.1, RGP 4.4 and RGP Table 6.

LPZ Dose Acceptance Criteria: 6.3 Rein TEDE I Nuclear Common Revision 12 I I ulerCm oReiin1

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G. Patel/NUCORE, ORIGINATOR, DATE REV: 05/17/2006 4 M. Drucker/NUCORE, REVIEWER/VERIFIER, DATE 05/18/2006 4.12 No correction is made for depletion of the effluent plume by deposition on the ground (Ref 10.1, RGP 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 10.1, RGP 4.2.1):

0 Contamination of the control room atmosphere by the intake or infiltration of the radioactive material contained in the post-accident radioactive plume released from the facility (via CR air intake),

0 Contamination of the control room atmosphere by the intake or infiltration of airborne radioactive material from areas and structures adjacent to the control room envelope (via CR unfiltered inleakage),

Radiation shine from the external radioactive plume released from the facility (external airborne cloud),

Radiation shine from radioactive material in the reactor containment (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 (CR filter shine dose).

Note: The external airborne cloud dose, containment shine dose, and CR filter shine dose due to FRA 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 10.1), therefore, these direct dose contributions are considered to be insignificant and are not evaluated for a FHA.

4.14 The radioactivity releases and radiation levels used in the control room dose is 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 10.1, RGP 4.2.2).

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CALCULATION CONTINUATION SHEET SHEET 15 of 45 CALC. NO.: S-C-ZZ-MDC-1920

REFERENCE:

IG. Patel/NUCORE, ORIGINATOR, DATE REV: 05/17/2006 4__

M. Drucker/NUCORE, REVIEWER/VERIFIER, DATE 05/18/2006 4.15 The occupancy and breathing rate of the maximum exposed individuals present in the control room are incorporated in design inputs 5.3.4.8 & 5.3.5.3 (Ref. 10.1, RGP 4.2.6).

4.16 10 CFR 50.67 (Ref 10.4) establishes the following radiological criterion for the control room.

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

4.17 Credit for engineered safety features that mitigate airborne activity within the control room may be assumed including control room isolation or pressurization, intake or recirculation filtration (Ref. 10.1, RGP 4.2.4). The control room pressurization as a result of CREACS actuation following CR intake monitor response to a FHA (Ref. 10.6.6 & Sections 3.4 & 7.4) is assumed. No credit is taken for the preferential alignment of the outside air emergency intake dampers.

4.18 No credit is taken for KI pills or respirators (Ref. 10.1, RGP 4.2.5).

Revision 12 I Ii Nuclear Nuclear Common Common Revision 12

CALCULATION CONTINUATION SHEET SHEET 16 of 45 CALC. NO.: S-C-ZZ-MDC-1920

REFERENCE:

G. Patel/NUCORE, I ORIGINATOR, DATE REV: 05/17/2006 4 M. Drucker/NUCORE, REVIEWER/VERIFIER, DATE 05/18/2006 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 SNGS plant specific design inputs and assumptions used in the current facility's design basis FHA analysis 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 comply with RG 1.183, Appendix B requirements.

5.1.2 Credit for Engineered Safeguard 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 normal CR air intake monitors are required to be operable by TS 3.3.3.1 in ALL MODES and during movement of irradiated fuel assemblies and during CORE ALTERATIONS. The normal CR air intake monitor's function of preferential alignment of the less contaminated outside air emergency intake is conservatively not credited (Ref. 10.10, page 49). The CREACS charcoal filtration operation is credited (Ref.

10.6.15) with a 1-minute system response delay. The FHB safety related charcoal filtration system is conservatively not credited in the analysis.

5.1.3 Meteorology Considerations The control room atmospheric dispersion factors (X/Qs) for the CEH, PV, and FHB rollup door release point are developed (Ref. 10.5) using the NRC sponsored computer code ARCON96 and guidance provided for the use of ARCON96 in the Regulatory Guide 1.194. The EAB and LPZ X/Qs are calculated using the SNGS plant specific meteorology and appropriate regulatory guidance (Ref. 10.16). The site boundary X/Qs in Reference 10.16 were accepted by the staff in the previous licensing proceedings.

Nuclear Common Revision 12

CALCULATION CONTINUATION SHEET SHEET 17 of 45 CALC. NO.: S-C-ZZ-MDC-1920

REFERENCE:

G. Patel/NUCORE, I ORIGINATOR, DATE REV: 05/17/2006 4 M. Drucker/NUCORE, REVIEWER/VERIFIER, DATE 05/18/2006 5.2 Accident-Specific Design Inputs/Assumptions The design inputs/assumptions utilized in the post-FHA 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 Regulatory Position 3 and Appendix B of RG 1.183 (Ref. 10.1). The design inputs and assumptions in the following sections represent the as-built design of the plant.

I Nuclear Common Revision 122I 1 I Nula omnRvso

CALCULATION CONTINUATION SHEET SHEET 18 of 45 CALC. NO.: S-C-ZZ-MDC-1920

REFERENCE:

G. Patel/NUCORE, 1 ORIGINATOR, DATE REV: 05/17/2006 4 M. Drucker/NUCORE, REVIEWERNERIFIER, DATE 05/18/2006 Value Assigned Reference Design Input Parameter 5.3 Source Term and Transport Parameters 5.3.1 Source Term 5.3.1.1 Core Power Level 3,459 MWt 10.6.4

[3,632 MWt (3,459 MWt x 1.05) Used in the analysis 5.3.1.2 Isotopic Core Inventory @ 3,600 MWt 10.3, Table 2 Core Inventory (Ci)

Isotope Activity Isotope Activity Isotope Activity KR-83M 1.20E+07 1-132 1.40E+08 XE-133 2.OOE+08 KR-85M 2.60E+07 1-133 2.OOE+08 XE-135 5.OOE+07 KR-85 1.10E+06 1-134 2.20E+08 XE-135M 4.OOE+07 KR-87 4.70E+07 1-135 1.90E+08 XE-138 1.60E+08 K.R-88 6.70E+07 XE-131M 7.OOE+05 1-131 9.90E+07 XE-133M 2.90E+07 5.3.1.3 Radionuclide Release Fractions (10.1, RGP 3.2, Table 3)

Group Fraction Fraction Used in Analysis 1-131 0.08 0.16 Kr-85 0.10 0.20 Other Noble Gases 0.05 0.10 Other Halogens 0.05 0.10 Alkali Metals 0.12 0.24 5.3.1.4 Radionuclide Composition Group Elements 10.1, RGP 3.4, Table 5 Noble Gases Xe, Kr Halogens I, Br Alkali Metals Cs, Rb 5.3.1.5 Number of Damaged Fuel 1 Assumed per Assumption 4.2 Assembly 5.3.1.6 Number of Fuel 193 10.6.5 Assemblies In Core 5.3.1.7 Irradiated Fuel Decay 96 Hrs used in the analysis Assumed Time 72 Hrs 60 Hrs 48 Hrs 24 Hrs 5.3.1.8 Radial Peaking Factor 1.65 (1.70 used in the analysis) 10.19 Nuclear Common Revision 12

CALCULATION CONTINUATION SHEET SHEET 19 of 45 CALC. NO.: S-C-ZZ-MDC-1920

REFERENCE:

G. Patel/NUCORE, ORIGINATOR, DATE REV: 05/17/2006 4 M. Drucker/NUCORE, REVIEWER/VERIFIER, DATE 05/18/2006 Design Input Parameter Value Assigned Reference 5.3.2 Activity Transport in Containment Building 5.3.2.1 Refueling Cavity Water 23 feet 10.6.3 Depth 5.3.2.2 Containment Building 2.62E+06 fW 3 10.11 Free Air Volume 5.3.2.3 Iodine Decontamination Factors (DFs)

Elemental 500 10.1, Appendix B, Section 2 Organic 1 5.3.2.4 Overall Effective Decontamination Factor (DFs) for Iodine Total Iodine [ 200 110.1, Appendix B, Section 2 5.3.2.5 Chemical Form of Iodine Released From Pool Water Elemental 57% 10.1, Appendix B, Section 2 Organic 43%

5.3.2.6 DF of Noble Gas 1 10.1, Appendix B, Section 3 5.3.2.7 Duration of Release (hr) 2 10.1, Appendix B, Section 5.3 5.3.2.8 Containment Exhaust 35,000 cfrn 10.18.g & 10.18.h From Ring Header 1 5.3.2.9 Activity release rate 100,600 cfm See Section 7.2.1 5.3.3 Activity Transport in Fuel Handling Building 5.3.3.1 Spent Fuel Pool Storage 23 feet 10.6.9 Water Depth 5.3.3.2 Fuel Handling Building 558,550 ft3 Section 7.2.2 Volume 5.3.3.3 Activity release rate 21,439 cfm 10.18.a, 10.18.d, & 10.21 (19,490 x 1.1 = 21,439 cfin) 5.3.3.4 FHB Charcoal Filter Not credited in the analysis 10.6.10 Efficiencies I The remaining FHA occurring in the FHB source term and activity transport design input parameters are the same as those for a FHA occurring in the containment (see design inputs 5.3.1 and 5.3.2) 5.3.4 Control Room Model Parameters 5.3.4.1 CR Volume 81,420 ft3 10.12, page 33 5.3.4.2 CR Normal Flow Rate 1,320 cfm Section 7.3 5.3.4.3 CREACS Design Makeup 2,200 cfm 10.6.13 Flow Rate 5.3.4.4 CREACS Ventilation 8,000 cfm +/- 10% cfm 10.6.12 Flow Rate 5,000 cfin (used in analysis) Section 7.3 5.3.4.5 CREACS Charcoal Filter 95% Section 7.7 Efficiency Nuclear Common Revision 12

CALCULATION CONTINUATION SHEET SHEET 20 of 45 CALC. NO.: S-C-ZZ-MDC-1920

REFERENCE:

G. Patel/NUCORE, ORIGINATOR, DATE REV: 05/17/2006 4 M. Drucker/NUCORE, REVIEWER/VERIFIER, DATE 05/18/2006 Design Input Parameter Value Assigned Reference 5.3.4.6 CREACS HEPA Filter 99% Section 7.7 Efficiency 95% Used in Analysis 5.3.4.7 CR Unfiltered Inleakage 150 cfm (nominal value measured 10.32, Table 1 is less than 100 cfm) 5.3.4.8 CR Occupancy Factors Time (Hr) % 10.1, RGP 4.2.6 0-24 100 24-96 60 96-720 40 3

3.5E-04 m /sec 10.1, RGP 4.2.6 5.3.4.9 CR Breathing Rate) 5.3.4.10 Unit 1 CR x/Qs - Post-FHA Release From Unit 1 CEH Time (Hr) X/Q (sec/m 3) 0-2 2.86E-03 10.5, page 33 2-8 2.22E-03 8-24 9.15E-04 24-96 6.60E-04 96-720 5.62E-04 5.3.4.11 F-B 0-2 hr Equivalent X/Q 1.85E-03 s/m 3 Section 7.5 5.3.4.12 Unit 1 CR x/Qs - Post-FHA Release From Unit 1 Plant Vent Time (Hr) x/Q (sec/m3) 0-2 1.78E-03 10.5, page 34 2-8 1.31E-03 8-24 5.22E-04 24-96 3.77E-04 96-720 3.17E-04 _

5.3.4.13 Unit 1 CR X/Qs - Post-FHA Release From FHB Rollup Door Time (Hr) X/Q (sec/m3) 0-2 1.50E-03 10.5, page 35 2-8 1.20E-03 8-24 4.48E-04 24-96 3.22E-04 96-720 2.50E-04 Nuclear Common Revision 12

CALCULATION CONTINUATION SHEET SHEET 21 of 45 CALC. NO.: S-C-ZZ-MDC-1920 G.___________

G. Patel/NUCORE,[

REFERENCE:

ORIGINATOR, DATE REV: 05/17/2006 4 M. Drucker/NUCORE, REVIEWER/VERIFIER, DATE 05/18/2006 Design Input Parameter Value Assigned Reference 5.3.4.14 Unit 1 CR X/Qs - Post-FHA Release From Smoke Hatch Time (Hr) z/Q (sec/m 3) 0-2 1.15E-02 10.9, Section 8.4 2-8 9.28E-03 8-24 3.50E-03 24-96 2.49E-03 96-720 2.02E-03 5.3.5 Site Boundary Release Model Parameters 5.3.5.1 EAB Atmospheric Dispersion Factor 5.3.5.2 LPZ Atmospheric Dispersion Factors (X/Qs) 1.30E-04 JX1)0(sec/m3)

10. 16, Table 5 Time (Hr) X/Q (sec/m 3) 10.16 Table 5 0-2 1.86E-05 2-8 7.76E-06 8-24 5.01E-06 24-96 1.94E-06 96-720 4.96E-07 5.3.5.3 CR Breathing Rate 3.5E-04 10.1, RGP 4.2.6 (m3/sec) 1 1 5.3.5.4 Offsite Breathing Rate (m3/sec)

Time (Hr) (m3/sec) 10.1, RGP 4.1.3 0-8 3.5E-04 8-24 1.8E-04 24-720 2.3E-04 5.3.5.5 CR Intake Monitor Xe-133 6.2 x 107 cpm/j+/-Ci/cc 10.13, page 12 Sensitivity 5.3.5.6 CR Intake Monitor 2.48 x 103 cpm 10.6.7 Alert/Trip Setpoint II I Nuclear Common Revision eiin1 12 1 I Nula omn

I CALCULATION CONTINUATION SHEET SHEET 22 of 45 CALC. NO.: S-C-ZZ-MDC-1920

REFERENCE:

G. Patel/NUCORE, ORIGINATOR, DATE REV: 05/17/2006 4 M. Drucker/NUCORE, REVIEWERN/ERFIER, DATE 05/18/2006 6.0 METHODOLOGY 6.1 Post-FHA Activity Release Rates Activity released from the reactor cavity is uniformly distributed in the entire volume of containment building and released to environment over a two hour time period such that 99% of the activity released from the damaged spent fuel assembly is released to the environment. The post-FHA activity release rate from the containment is calculated in Section 7.2.1.

The FHB volume is back calculated in Section 7.2.2 knowing the FHB exhaust rate of 21,439 cfiri and the requirement to remove 99% of the activity in a two hour period.

6.2 Fuel Handling Accident in the FEB with a Failure of an Exhaust Fan The post-FHA activity releases through three different release paths due to pressurization of the FHB are discussed in Section 3.2 and a composite 0-2 hr x/Q is calculated for a combined release path is calculated in Section 7.5.

7.0 CALCULATIONS 7.1 SNGS Plant Specific Nuclide Inventory File (NIF) For RADTRAD V%3.02 Input The parameter Ci/MWt in the RADTRAD V3.02 default nuclide inventory file Bwr def NIF is dependent on the plant-specific core thermal power level, reload design, fuel burnup, and fuel cycle, therefore, the NIF is modified based on the plant-specific isotopic Ci/fMWt information developed in Table 3. The RADTRAD nuclide inventory file SNGSFHA_def.txt is used in the analysis.

7.2 Release Rates 7.2.1 Containment Building The release rate from the source node - reactor cavity to containment - is calculated such that 99% of the activity released into the containment is released to the environment in two hours. The 1% of the activity remaining in the containment is insignificant.

A= A0 e-1t I Nuclear Common Revision 12 1

CALCULATION CONTINUATION SHEET SHEET 23 of 45 CALC. NO.: S-C-ZZ-MDC-1920

REFERENCE:

G. Patel/NUCORE, ORIGINATOR, DATE REV: 05/17/2006 4 M. Drucker/NUCORE, REVIEWERVERIFIER, DATE 05/18/2006 Where; Ao = Initial Activity in Source Node A Final Activity in Source Node

= Removal Rate (vol/hr) t = Removal Time (hr) = 2.0 hr Assuming that 99% of activity is released into the environment, A/A 0 = 0.01 Therefore, A/Ao =e"t 0.01 = e"21 In (0.01)= - 2* ln(e)

- 4.605 = -2 .

X= - 4.605/-2 = 2.303 volume/hr Containment Building Release Rate = 2.303 1/hr x 2,620,000 ft3 x 1 hr/60 min 100,600 ft3/min 7.2.2 Fuel Building Fuel building exhaust flow rate = 19,439 cfin (Ref. 10.18.a & 10.18.d) x 1.10 = 21,439 cfin A removal rate of 2.303 volume/hr (calculated in the above section) corresponds to removal of 99% of activity from the FEB volume over a two-hour period.

Therefore, the FHB volume can be arbitrarily calculated as follows:

2.303 vol/hr x 1 hr/60 min = 3.8388E-02 vol/min 3

FHB Volume = 21,439 t 3 /min =558,550 ft 3.838E-02 vol/min This volume is used in the RADTRAD model.

For the scenario of a FHA occurring in the fuel handling building with a rapid release of one volume per minute, a FHB release rate is 558,550 ft3/min is used in the RADTRAD model.

I Nuclear Common Revision eiin1 12 1 I Nula!omn I

CALCULATION CONTINUATION SHEET SHEET 24 of 45 CALC. NO.: S-C-ZZ-MDC-1920

REFERENCE:

G. PateIINUCORE, ORIGINATOR, DATE REV: 05/17/2006 4 M. Drucker/NUCORE, REVIEWER/VERIFIER, DATE 05/18/2006 7.3 Control Room Flow Rates Normal Flow Rate Reference 10.20, Note "S" provides the outside air flow rates to Zone 1 from the Unit 1 and Unit 2 air intakes.

Zone 1 is the combined control room envelop. The control area air conditioning system (CAACS) normal airflow rate is calculated as follows:

Total CAACS Air Flow Rate = 32,600 cfmn (2,200 cfm outside air + 30,400 cfm recirc air)

Zone 1 (control room pressure boundary) Supply Air Flow Rate = 8,000 cfm Amount of Outside Air To Zone 1

= (Fraction of Total CAACS Air Flow Rate to Zone 1) x (2,200 cfm outside air inflow rate)

= (8,000 cfm / 32,600 cfmn) x 2,200 cfrn = 0.2454 x 2,200 cfin = 540 cfin Use 600 cfm for Zone 1 During Normal Plant Operation Total Amount of Outside Air Flow Rate From Both Intakes = 2 x 600 cfin = 1,200 cfm Maximum Amount of Outside Air Flow Rate = 1.1 x 1,200 cfm = 1,320 cfm CREACS Recirculation Flow Rate CREACS ventilation flow rate = 8,000 cfin +/- 10% cfmn (Ref. 10.6.12)

Minimum CREACS flow rate = 8,000 cfm - 0.10 x 8,000 cfm = 8,000 cfm - 800 cfrn = 7,200 cfmn Net CREACS recirculation flow rate = Minimum CREACS flow rate - CREACS makeup flow rate 7,200 cfm--2,200 cfm (Ref. 10.6.13) = 5,000 cfin 7.4 CR Intake Monitor Setpoint FHA In Containment Building Minimum Xe-133 Concentration at CR Intake

= total Xe-133 release (Table 3) divided between two CR intakes = 3.555E+06 Ci / 2 = 1.778E+06 Ci

= 1.778E+06 Ci x 1 x 35,000 ft3/min (Ref. 10.18.c & e) x 1 x 2.86E-03 sec/m3 2.62E+06 ft 3 60 sec/min

= 1.132 Ci/m 3 = 1.132 pCi/cc CR Intake Monitor Xe-133 Sensitivity = 6.2 x 107 cpm/pCi/cc (Ref. 10.13, page 12)

CR Intake Monitor Alarm/Trip Setpoint = 2.48 x 103 cpm (Ref 10.6.7)

CR Monitor Count Rate Due Post-FHA Activity Concentration At CR Intake Nuclear Common Revision 12

CALCULATION CONTINUATION SHEET SHEET 25 of 45 CALC. NO.: S-C-ZZ-MDC-1920

REFERENCE:

G. Patel/NUCORE, ORIGINATOR, DATE REV: 05/17/2006 4 M. Drucker/NUCORE, REVIEWER/VERIFIER, DATE 05/18/2006

= 1.132 ptCi/cc x 6.2 x W07 cprn/mCi/cc = 7.02 x 10 7 cpm >> 2.48 x 103 cpm FHA In Fuel Handling Building Minimum Xe-133 Concentration at CR Intake

= total Xe-133 release (Table 3) divided between two CR intakes = 3.555E+06 Ci / 2 = 1.778E+06 Ci

= 1.778E+06 Ci x 1 x 21,439 ft3/min (Section 7.2.2) x 1 x 1.85E-03 sec/m 3 558,550 ft3 60 sec/min

= 2.104 Ci/m3 = 2.104 pCi/cc CR Intake Monitor Xe-133 Sensitivity = 6.0 x W cpm/p.Ci/cc (Ref. 10.13, page 12)

CR Intake Monitor Alarm/Trip Setpoint = 2.48 x 103 cpm (Ref. 10.6.7)

CR Monitor Count Rate Due Post-FHA Activity Concentration At CR Intake

= 2.104 jiCi/cc x 6.2 x 107 cpm/ýtCi/cC = 1.30 x 107 cpm >> 2.48 x 103 Cpm It is clear that the CR intake monitor will instantaneously reach its Alarm/Trip setpoint following a FHA occurring in the containment or fuel handling building.

7.5 Equivalent 0-2 hr y/Q For FMB Release Path Plant Vent 0-2 X/Q = 1.78E-03 s/m 3 (Ref. 10.5, page 34)

F-B Rollup Door 0-2 X/Q = 1.50E-03 s/m 3 (Ref. 10.5, page 35)

Smoke Hatch 0-2 X/Q = 1.15E-02 s/r 3 (Ref. 10.9, Section 8.4)

1. Release through the plant vent at a rate of 15,300 cfm (Ref. 10.21)

2. Leakage through truck bay roll-up door at a rate of 3,883 cfm (Ref. 10.21)

3. Leakage through gravity damper 256 cfin (Ref. 10.21) 0-2 hr FHB X/Q 3 3

= 15,300 cfm x 1.78E-03 s/m 3 + 3,883 cfm x 1.50E-03 s/m + 256 cfin x 1.15E-02 s/m (15,300 cfm + 3,883 cfmn + 256 cfin) 3 3

= 36.00 cfin.s/rm = 1.85E-03 s/m 19,439 cfin Nuclear Common Revision 12

CALCULATION CONTINUATION SHEET SHEET 26 of 45 CALC. NO.: S-C-ZZ-MDC-1920

REFERENCE:

G. Patel/NUCORE, ORIGINATOR, DATE REV: 05/17/2006 4 M. Drucker/NUCORE, REVIEWER/VERIFIER, DATE 05/18/2006 7.6 Distance of TSC Air Intake

-E2.27' Plant North S480' Unit 1 181.64' Containment South Coordinate of Unit 1 Containment = South Coordinate of Plant + Distance between Centerlines of Plant and Unit 1 Containment

= S320.0' (Ref. 10.23.a) + 160'-0" (Ref.10.23.b) = S480.0' South Coordinate of Column 1B of Clean Facility Building (CFB)

= South Coordinate of CFB + Distance between South Coordinate and Column lB

= S715.88' (Ref. 10.28) +1'-6" (Ref. 10.28) = S717.38' Distance between Column lB and TSC Air Intake Distance between Columns lB and 2B - Distance between 2B and TSC Air Intake

= 22'-3-1/2" (Ref. 10.27.a) - (4'-8-3/4" + 6-1/8") (Ref. 10.29) = 22'-3-1/2" - 5'-2-7/8" = 17.05' South Distance between Centerline of Unit 1 Containment and TSC Air Intake

= S717.38 - S480.0' + 17.05' = 237.38' + 17.05' = 254.43' Distance between Centerlines Unit 1 Containment and CEH = 49.82 (Ref. 10.5, page 26)

I Nuclear Common Revision 122I 1 I Nula omnRvso

CALCULATION CONTINUATION SHEET SHEET 27 of 45 CALC. NO.: S-C-ZZ-MDC-1920

REFERENCE:

G. Patel/NUCORE, ORIGINATOR, DATE REV: 05/17/2006 4 M. Drucker/NUCORE, REVIEWER/VERIFIER, DATE 05/18/2006 South Distance between Centerline Unit 1 CEH and TSC Air Intake

= 254.43'- 49.82' = 204.61' West Coordinate of Containment Centerline = W120.0' (Ref. 10.23.a)

East Coordinate of Centerline of TSC Air Intake

= East Coordinate of East Wall of CFB - (Distance between East Wall of CFB and Row AB + Distance between Centerlines of Rows AB and BB) + Distance between Row BB and Centerline of TSC Air Intake

= (E30.79' (Ref. 10.28) 6" (Ref. 10.28) - 28'-10-1/4" (Ref 10.28)) + (1'-0" + (1'-8")/2) (Ref. 10.29.J)

= E2.27' Distance between Centerlines of Unit 1 Containment and CEH = 59.37' (Ref. 10.5, page 26)

East-west Distance between Centerlines of Unit 1 Containment and TSC Air Intake

= E2.27' + W1 20.0' = 122.27' East-west Distance between Unit 1 CEH and TSC Air Intake

= E2.27' + W120.0' + = 122.27' East-west Distance between Centerlines of Unit 1 CEH and TSC Air Intake East-west Distance between Centerlines of Unit 1 Containment and TSC Air Intake + Distance between Centerlines of Unit 1 Containment and CEH

= 122.27' + 59.37' = 181.64' Slant Distance between Centerlines of Unit 1 Containment (Plant Vent) and TSC Air Intake

= [(254.43)2+ (122.27)2]112 = 282.28' = 86.06 m Slant Distance between Centerlines of Unit 1 CEH and TSC Air Intake

= [(204.61)2+ (1 8 1 .6 4 )2] " = 273.60' = 83.42 m The distance between the source locations (Plant Vent and CEI-) and receptor locations (Unit 1 CR and Unit 1

& 2 TSC) are compared in the following table:

I Nuclear Common Revision 122I 1

[NcerCmonRvso

CALCULATION CONTINUATION SHEET SHEET 28 of 45 CALC. NO.: S-C-ZZ-MDC-1920

REFERENCE:

G. Patel/NUCORE, ORIGINATOR, DATE REV: 05/17/2006 4 M. Drucker/NUCORE, REVIEWERIVERIFIER, DATE 05/18/2006 Comparison of Distance Between Source & Receptor Control Room Intake Vs Technical Support Center Intake Slant Distance Between Source and Receptor Unit 1 Unit 1 Unit 1 Unit 1 Plant Vent Plant Vent CEH CEH and and and and Unit I TSC Intake Unit 1 TSC Intake CR Intake (Meter) CR Intake (Meter)

(Meter) (Meter) 30.25 86.06 46.62 83.42 7.7 CREACS Charcoal/HEPA Filter Efficiencies Charcoal Filter In-place penetration testing acceptance criteria for the safety related Charcoal filters are as follows:

CREACS Charcoal Filter - in-laboratory testing methyl iodide penetration < 2.5% (Ref. 10.6.11)

GL 99-02 (Ref 10.30) requires a safety factor of at least 2 should be used to determine the filter efficiencies to be credited in the design basis accident.

Testing methyl iodide penetration (%) = (100% - ii)/safety factor = (100% - 11)/2 Where ri = charcoal filter efficiency to be credited in the analysis CREACS Charcoal Filter 2.5% = (100% - ir)/2 5% = (100% -,q) il = 100% - 5% = 95%

HEPA Filter HEPA filter efficiency = 99% (Ref. 10.6.14). HEPA filter efficiency of 95% is used in the analysis Safety Grade Filter Efficiency Credited (%)

Filter Aerosol I Elemental Organic CREACS 95 95 95 Revision 12 I II Nuclear Common Revision 12 1

CALCULATION CONTINUATION SHEET SHEET 29 of 45 CALC. NO.: S-C-ZZ-MDC-1920

REFERENCE:

IG. Patel/NUCORE, ORIGINATOR, DATE REV: 05/17/2006 4 M. Drucker/NUCORE, REVIEWER/VERIFIER, DATE 05/18/2006 8.0 RESULTS

SUMMARY

8.1 The EAB, LPZ, & CR doses due to a FHA occurring in the containment building with the CEH, personnel air locks, and containment penetrations open are summarized in the following table for different fuel decay times:

Fuel Decay Fuel Handling Accident Occurring In Containment Building Time (hr) TEDE Dose (rem)

Computer Run Receptor Location Number Control Room EAB LPZ 24 1.13 1.26 0.18 S24FHA150.o0 0.95 1.05 0.15 S48FHAI50.o0 60 0.89 0.99 0.14 S60FHA150.o0 72 0.84 0.93 0.13 S72FHA150.o0 96 0.76 0.84 0.12 S96FLA150.oO Allowable TEDE 5.0 6.3 6.3 Limits Significant assumptions used in this analysis:

" CEH, personnel air locks, and other containment penetrations remain open for the duration of the accident

" Containment integrity is not credited in the analysis

" Gap fission product fractions doubled

" Activity is released to the environment at a rate of 100,600 cfm

" CR envelope is pressurized with actuation of the CREACS following a FHA

" CR monitors' preferential alignment to less contaminated CR intake is not credited

" Worst X/Qs are used for entire duration of the accident

" CR unfiltered inleakage of 150 cfin is assumed

" All fuel rods in one spent fuel assembly are damaged

" Reactor cavity overall effective DF = 200

" Core thermal power = 3,632 MWt

" Radial Peaking Factor = 1.70 I Nuclear Common Revision 12 1

CALCULATION CONTINUATION SHEET SHEET 30 of 45 CALC. NO.: S-C-ZZ-MDC-1920

REFERENCE:

G. Patel/NUCORE, ORIGINATOR, DATE REV: 05/17/2006 4 M. Drucker/NUCORE, REVIEWERNERIFIER, DATE 05/18/2006 8.2 The EAB, LPZ, & CR doses due to a FHA occurring in the fuel handling building with a failure of an exhaust fan, are summarized in the following table for different fuel decay times:

Fuel Decay Fuel Handling Accident Occurring In Fuel Handling Building Time (hr) TEDE Dose (rem)

Computer Run Receptor Location Number Control Room EAB LPZ 24 0.73 1.26 0.18 FB24FHA150.o0 48 0.62 1.05 0.15 FB48FHA150.o0 60 0.58 0.99 0.14 FB60FIA15O.oO 72 0.55 0.93 0.13 FB72FHA150.o0 96 0.49 0.84 0.12 FB96FHA150.o0 Allowable TEDE 5.00 6.3 6.3 Limits Significant assumptions used in this analysis:

FHB charcoal filtration is not credited

Gap fission product fractions doubled

" Activity is released to the environment at a rate of 21,439 cfm

" CR envelope is pressurized with actuation of the CREACS following a FHA

CR monitors' preferential alignment to less contaminated CR intake is not credited

" Worst x/Qs are used for entire duration of the accident

CR unfiltered inleakage of 150 cfm is assumed

" All fuel rods in one spent fuel assembly are damaged

Spent fuel pool overall effective DF = 200

Core thermal power = 3,632 MWt

Radial Peaking Factor = 1.70 I Nuclear Common Revision 12 ;i I Nula omn eiin1

CALCULATION CONTINUATION SHEET SHEET 31 of 45 CALC. NO.: S-C-ZZ-MDC-1920

REFERENCE:

G. Patel/NUCORE, ORIGINATOR, DATE REV: 05/17/2006 4 M. Drucker/NUCORE, REVIEWER/VERIFIER, DATE 05/18/2006 8.3 The EAB, LPZ, & CR doses due to a FHA occurring in the fuel handling building with a rapid release of one volume per minute are summarized in the following table for different fuel decay times:

Fuel Decay Fuel Handling Accident Occurring In Fuel Handling Building Time (hr) TEDE Dose (rem)

Computer Run Receptor Location Number Control Room EAB LPZ 24 2.06 1.27 0.18 FB24PUFF15O.oO 48 1.78 1.06 0.15 S

FB48PUFF150.o0 60 1.67 1.00 0.14 FB60PUFF150.oO 72 1.58 0.94 0.13 FB72PUFF15O.oO 96 1.43 0.85 0.12 FB96PUFF150.o0 Allowable TEDE 5.00 6.3 6.3 Limits Significant assumptions used in this analysis:

" FHB charcoal filtration is not credited

" Post-FHA activity is released to the environment at a rate of one volume/minute (558,550 efin)

Gap fission product fractions doubled

" CR envelope is pressurized with actuation of the CREACS following a FHA

CR monitors' preferential alignment to less contaminated CR intake is not credited

" Worst X/Qs are used for entire duration of the accident

CR unfiltered inleakage of 150 cfm is assumed

All fuel rods in one spent fuel assembly are damaged

Spent fuel pool overall effective DF = 200

Core thermal power = 3,632 MWt

Radial Peaking Factor = 1.70 I Nuclear Common Revision 12 12I 1

I NulearCommn Reisio

CALCULATION CONTINUATION SHEET SHEET 32 of 45 CALC. NO.: S-C-ZZ-MDC-1920

REFERENCE:

G. Patel/NUCORE, ORIGINATOR, DATE REV: 05/17/2006 4 M. Drucker/NUCORE, REVIEWER/VERIFIER, DATE 05/18/2006

9.0 CONCLUSION

S 9.1 FHA Occurring In Containment The Section 8.1 results indicate that the EAB, LPZ, and CR doses are within allowable limits for a FHA occurring in the Containment building without containment integrity (with the CEH, personnel locks, and containment penetrations in the piping penetration areas opened) with a minimum fuel decay time of 24 hours2.777778e-4 days 0.00667 hours 3.968254e-5 weeks 9.132e-6 months .

The results demonstrate that the following Salem 1 & 2 Technical Specification requirements can be relaxed:

1. The irradiated fuel can be moved in the reactor pressure vessel after the reactor has been sub-critical for at least 24 hours2.777778e-4 days 0.00667 hours 3.968254e-5 weeks 9.132e-6 months (relaxation to Technical Specification LCO 3.9.3)

2. Irradiated fuel assemblies can be moved without containment integrity (relaxation to Technical Specification LCO 3.9.4)

3. Core alterations can be performed without containment integrity (relaxation to Technical Specification LCO 3.9.4) 9.2 FHA Occurring In Fuel Handling Building The Sections 8.2 and 8.3 results indicate that the EAB, LPZ, and CR doses are within allowable limits for a FHA occurring in the fuel handling building without crediting the charcoal filtration in the fuel handling ventilation system with a minimum fuel decay time of 24 hours2.777778e-4 days 0.00667 hours 3.968254e-5 weeks 9.132e-6 months . The results demonstrate that the Salem I & 2 Technical Specification Surveillance requirements 4.9.12.b and 4.9.12.c can be relaxed.

I Nuclear Common Revision eiin1 12 1I I ula omo

CALCULATION CONTINUATION SHEET SHEET 33 of 45 CALC. NO.: S-C-ZZ-MDC-1920

REFERENCE:

G. Patel/NUCORE, ORIGINATOR, DATE REV: 05/17/2006 4 M. Drucker/NUCORE, REVIEWER/VERIFIER, DATE 05/18/2006

10.0 REFERENCES

1. U.S. NRC Regulatory Guide 1.183, Alternative Radiological Source Terms for Evaluating Design Basis Accidents at Nuclear Power Reactors, July 2000

2. S.L. Humphreys et al., "RADTRAD: A Simplified Model for Radionuclide Transport and Removal and Dose Estimation," NUREG/CR-6604, USNRC, April 1998

3. Westinghouse Calculation No. CN-CRA-93-144, Rev 0, Salem LOCA Dose Analysis

4. 10 CFR 50.67, "Accident Source Term."

5. Calculation No. S-C-ZZ-MDC-1912, Rev 0, Control Room X/Qs Using ARCON96 Code - Equipment Hatch & Plant Vent Releases

6. SNGS Technical Specifications:

6.1 Specification 3.9.4, Containment Building Penetrations 6.2 Specification 3.9.3, Decay Time 6.3 Specification 3.9.10, Water Level - Reactor Vessel 6.4 Specification 1.25, Rated Thermal Power 6.5 Specification 5.3.1, Fuel Assemblies 6.6 Specification 3.3.3.1, Radiation Monitoring Instrumentation LCO 6.7 Table 3.3-6, Radiation Monitoring Instrumentation 6.8 Specification Surveillance Requirement 4.8.1.1.2, Each diesel generator shall be demonstrated to be operable 6.9 Specification 3.9.11, Storage Pool Water Level 6.10 Specification 3.9.12, Fuel Handling Area Ventilation System 6.11 Specification Surveillance Requirement 4.7.6.1.b.3 and 4.7.6.1.c, CREACS Methyl Iodide Penetration 6.12 Specification Surveillance Requirement 4.7.6.1.d.1, CREACS Ventilation Flow Rate 6.13 Specification Surveillance Requirement 4.7.6.l.d.3, CREACS Design Makeup Flow Rate 6.14 Specification Surveillance Requirement 4.7.6.1.e, HEPA Filter DOP 6.15 Specification 3.7.6.1, Control Room Emergency Air Conditioning System (CREACS)

7. Federal Guidance Report 11, EPA-5201/1-88-020, Environmental Protection Agency

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

9. Calculation No. S-C-ZZ-MDC-1959, Rev 0, CR x/Qs Using ARCON96 Code - Non-LOCA Releases.

10. Design Change Package (DCP) No. 1EC-3505, CP Rev 2, Package No. 3, Control Area Ventilation -

Radiation Monitoring Mod

11. Specification 5.2.1, Salem Unit 1/Unit 2 Containment Configuration Nuclear Common Revision 12

CALCULATION CONTINUATION SHEET SHEET 34 of 45 CALC. NO.: S-C-ZZ-MDC-1920

REFERENCE:

G. Patel/NUCORE, ORIGINATOR, DATE REV: 05/17/2006 4 M. Drucker/NUCORE, REVIEWERNERIFIER, DATE 05/18/2006

12. CD P534 of Design Change Package (DCP) No. lEC-3505, Rev 7, Package No. 1, Control Area Air Conditioning System Upgrade

13. SNGS Calculation No. SC-RM005-01, Rev 2, RIB Radiation Monitors

14. Vendor Technical Document No. 322265-4, Rev 2, Fuel Handling Accident In Containment (Non-Design Basis)

15. Not Used.

16. Vendor Technical Document No. 321035, Rev 3, Accident X/Q Values At the Salem Generating Station Control Room Fresh Air Intakes, Exclusion Area Boundary And Low Population Zone

17. SNGS Architectural Drawings:

a. 207069, Rev 12, Unit 1 Reactor Containment Floor Plan EL 130'-0"

b. 207070, Rev 14, Unit 2 Reactor Containment Floor Plan EL 130'-0" C. 207080, Rev 23, Unit 1 Auxiliary Building Floor Plan EL 100'-0"

d. 207081, Rev 29, Unit 2 Auxiliary Building Floor Plan EL 100'-0"

e. 207084, Rev 13, Unit 1 Auxiliary Building Roof Plan EL 140'-0" & 141'-0"

f. 207085, Rev 10, Unit 2 Auxiliary Building Roof Plan EL 140'-0" & 141'-0"

g. 204803, Rev 10, Auxiliary Building EL 122', Reactor Cont & Fuel Building Area EL 130'

18. SNGS Mechanical P&IDs:

a. 205321, Rev 21, Sheet 1 of 3, Unit 1 - Auxiliary Building Diesel Generator & Fuel Handling Area Ventilation

b. 205237, Rev 42, Sheet 1 of 3, Unit 1 - Auxiliary Building - Ventilation

c. 205237, Rev 30, Sheet 2, Unit 1 - Auxiliary Building - Ventilation

d. 205322, Rev 23, Sheet 1 of 3, Unit 2 - Auxiliary Building Diesel Generator & Fuel Handling Area Ventilation

e. 205337, Rev 36, Sheet 1 of 3, Unit 2 - Auxiliary Building - Ventilation

f. 205337, Rev 22, Sheet 2, Unit 2 - Auxiliary Building - Ventilation

g. 205238, Rev 33, Sheet 2, Reactor Containment - Ventilation

h. 205338, Rev 27, Sheet 2, Reactor Containment - Ventilation

19. Core Operating Limits Reports for Salem 1 & 2:

a. NFS-0190, Rev 0, Cycle 15, February 20001

b. NFS-0209, Rev 0, Cycle 13, January 2002

20. SNGS Mechanical P&IDs:

a. 205248, Rev 43, Sheet 2, Unit 1 Aux Bldg Control Area Air Conditioning & Ventilation

b. 205348, Rev 34, Sheet 2, Unit 2 Aux Bldg Control Area Air Conditioning & Ventilation I Nuclear Common Revision 122i 1 I Nula omnRvso

CALCULATION CONTINUATION SHEET SHEET 35 of 45 CALC. NO.: S-C-ZZ-MDC-1920 G. Patel/NUCORE,

REFERENCE:

ORIGINATOR, DATE REV: 05/17/2006 4 M. Drucker/NUCORE, REVIEWER/VERIFIER, DATE 05/18/2006

21. Memorandum From Paul Wood To John Duffy, dated 12/15/96,

Subject:

Estimate of Unfiltered Inleakage from FBB with One Exhaust Fan and the Supply Fan Operating (Attached)

22. SNGS Wiring Diagram No. 220813, Rev 22, No. 2 Unit-Control Area No. 2 B 115 V AC Vital Instrument Bus

23. SNGS General Arrangement Drawings:

a. 204805, Rev 5, Aux Bldg El. 84', Reactor Cont. 78' & 81', Fuel Handling Area El. 85'& 89'-6"

b. 204808, Rev 1, Auxiliary Building & Reactor Containment Section A-A

24. Not Used.

25. Not Used.

26. SNGS Mechanical P&IDs:

a. 602513, Sheet 1 of 3, Rev 0, No. 1 & 2 Units Technical Support Center - Ventilation

b. 602513, Sheet 2 of 3, Rev 0, No. 1 & 2 Units Technical Support Center - Ventilation

27. SNGS Mechanical Arrangement Drawings:

a. 602511, Rev 0, Clean Facilities Bldg, - Technical Support Center/Computer Room HVAC Systems El. 132'-6"

b. 602512, Rev 0, Clean Facilities Bldg - Technical Support Center HVAC Equipment Room -

Elevation 147'-4/12"

c. 602514, Rev 0, Clean Facilities Bldg - Technical Support Center Technical Document & Annex Room HVAC Systems EL 119'-0"

28. SNGS Concrete Structural Drawing No. 242914, Rev 3, Clean Facilities Building Foundation Plan

29. SNGS Architectural Drawing No. 245685, Rev 2, Clean Facilities Bldg, Technical Support Center Floor, Roof Plans & Sections

30. USNRC, "Laboratory Testing of Nuclear-Grade Activated Charcoal", NRC Generic Letter 99-02, June 3, 1999

31. NRC Safety Evaluation for Calvert Cliffs Nuclear Power Plant Unit Nos. 1 and 2, Docket Nos. 50-317 and 50-318, License Amendment Nos. 242 and 216, dated March 12, 2001

32. Vendor Technical Document No. 326043, Control Room Envelope Inleakage Testing At Salem Nuclear Generating Station 2003.

33. Critical Software Package Identification No. A-0-ZZ-MCS-0225, Rev.2, RADTRAD Computer Code, Version 3.02 I Nuclear Common Revision 12 1 I Nula omn eiin1

CALCULATION CONTINUATION SHEET SHEET 36 of 45 CALC. NO.: S-C-ZZ-MDC-1920

REFERENCE:

G. Patel/NUCORE, ORIGINATOR, DATE REV: 05/17/2006 4 M. Drucker/NUCORE, REVIEWERNVERIFIER, DATE 05/18/2006 11.0 TABLES Table 1 Salem 1 & 2 Noble Gas & Iodine Normalized Core Inventory Core Core Normalized Inventory Power Core Isotope At 3600 MWt Normalizing Inventory Factor (Ci) (Ci)

A B C=AxB KR-83M 1.200E+07 1.009 1.211E+07 KR-85 1.100E+06 1.009 1.110E+06 KR-85M 2.600E+07 1.009 2.623E+07 KR-87 4.700E+07 1.009 4.742E+07 KR-88 6.700E+07 1.009 6.760E+07 1-131 9.900E+07 1.009 9.988E+07 1-132 1.400E+08 1.009 1.412E+08 1-133 2.000E+08 1.009 2.018E+08 1-134 2.200E+08 1.009 2.220E+08 1-135 1.900E+08 1.009 1.917E+08 XE-131M 7.000E+05 1.009 7.062E+05 XE-133M 2.900E+07 1.009 2.926E+07 XE-133 2.OOOE+08 1.009 2.018E+08 XE-135 5.000E+07 1.009 5.044E+07 XE-135M 4.000E+07 1.009 4.036E+07 XE-138 1.600E+08 1.009 1.614E+-08 A From Reference 10.3, Table 2 B = (3459 MWt x 1.05)/3600 MWt = (3632/3600) = 1.009 I Nuclear Common Revision eiin1 12 1 I Nula omn

CALCULATION CONTINUATION SHEET SHEET 37 of 45 CALC. NO.: S-C-ZZ-MDC-1920

REFERENCE:

G. Patel/NUCORE, ORIGINATOR, DATE REV: 05/17/2006 4 M. Drucker/NUCORE, REVIEWER/VERIFIER, DATE 05/18/2006 Table 2 Normalized Core Inventory Used In FHA Analysis Normalized Gap Gap Normalized Core Release Release Core Isotope Inventory Fraction Fraction Inventory IN Used In Used In (Ci) RFT File Analysis FHA A B C D=(A*C)/B KR-83M 1.211E+07 0.05 0.10 2.421E+07 KR-85 1.110E+06 0.05 0.20 4.439E+06 KR-85M 2.623E+07 0.05 0.10 5.246E+07 KR-87 4.742E+07 0.05 0.10 9.484E+07 KR-88 6.760E+07 0.05 0.10 1.352E+08 1-131 9.988E+07 0.05 0.16 3.196E1+08 1-132 1.412E+08 0.05 0.10 2.825E+08 1-133 2.018E+08 0.05 0.10 4.036E+08 1-134 2.220E+08 0.05 0.10 4.439E+08 1-135 1.917E+08 0.05 0.10 3.834E+08 XE-131M 7.062E+05 0.05 0.10 1.412E+06 XE-133M 2.926E+07 0.05 0.10 5.852E+07 XE-133 2.018E+08 0.05 0.10 4.036E+08 XE-135 5.044E+07 0.05 0.10 1.009E+08 XE-135M 4.036E+07 0.05 0.10 8.071E+07 XE-138 1.614E+08 0.05 0.10 3.228E+08 A From Table 1 C From Design Input 5.3.1.3 NularCmonRvsin1 I Nuclear Common Revision 12

CALCULATION CONTINUATION SHEET SHEET 38 of 45 CALC. NO.: S-C-ZZ-MDC-1920

REFERENCE:

G. Patel/NUCORE, ORIGINATOR, DATE REV: 05/17/2006 4 M. Drucker/NUCORE, REVIEWERNVERIFER, DATE 05/18/2006 Table 3 Post-FHA Activity Released In Containment Building Used In RADTRAD Nuclide Inventory File Core Radial Total Number Activity Post-FHA Activity In RB Bldg Isotope Initial Peaking Number of Fuel In Damaged For RADTRAD Code Inventory Factor of Fuel Assembly Fuel DF Nuclide Inventory File Assembly Damaged Rods RADTRAD (Ci) In Core (Ci) (Ci) (Ci/MWt) (CiIMWt)

A B C D E=A*B*DIC F G=E/F H=G/3632 I=H*1 KR-83M 2.421E+07 1.70 193 1 2.133E+05 1.0 2.133E+05 5.872E+01 .5872E+02 KR-85 4.439E+06 1.70 193 1 3.910E+04 1.0 3.910E+04 1.077E+01 .1077E+02 KR-85M 5.246E+07 1.70 193 1 4.621E+05 1.0 4.621E+05 1.272E+02 .1272E+03 KR-87 9.484E+07 1.70 193 1 8.353E+05 1.0 8.353E+05 2.300E+02 .2300E+03 KR-88 1.352E+08 1.70 193 1 1.191E+06 1.0 1.191E+06 3.279E+02 .3279E+03 1-131 3.196E+08 1.70 193 1 2.815E+06 200.0 1.408E+04 3.876E+00 .3876E+01 1-132 2.825E+08 1.70 193 1 2.488E+06 200.0 1.244E+04 3.425E+00 .3425E+01 1-133 4.036E+08 1.70 193 1 3.555E+06 200.0 1.777E+04 4.893E+00 .4893E+01 1-134 4.439E+08 1.70 193 1 3.910E+06 200.0 1.955E+04 5.383E+00 .5383E+01 1-135 3.834E+08 1.70 193 1 3.377E+06 200.0 1.688E+04 4.649E+00 .4649E+01 XE-131M 1.412E+06 1.70 193 1 1.244E+04 1.0 1.244E+04 3.425E+00 .3425E+01 XE-133M 5.852E+07 1.70 193 1 5.154E+05 1.0 5.154E+05 1.419E+02 .1419E+03 XE-133 4.036E+08 1.70 193 1 3.555E+06 1.0 3.555E+06 9.787E+02 .9787E+03 XE-135 1.009E+08 1.70 193 1 8.887E+05 1.0 8.887E+05 2.447E+02 .2447E+03 XE-135M 8.071E+07 1.70 193 1 7.109E+05 1.0 7.109E+05 1.957E+02 .1957E+03 XE-138 3.228E+08 1.70 193 1 2.844E+06 1.0 2.844E+06 7.830E+02 .7830E+03 A From Table 2 I Nuclear Common Revision 12 1 I.Nula omn eiin1

CALCULATION CONTINUATION SHEET SHEET 39 of 45 CALC. NO.: S-C-ZZ-MDC-1920 IG. Patel/NUCORE,I

REFERENCE:

ORIGINATOR, DATE REV: 05/17/2006 4 M. Drucker/NUCORE, REVIEWER/VERIFIER, DATE 05/18/2006 12.0 FIGURES E

N V

R Containment 100,600 cfmToC Reactor Building 0 Cavity V f 2.62E+06 ft 3 N

M E

N

:.,, ii", * "EI T Figure 1: FHA In Containment Building With Equipment Hatch Open RADTRAD Nodalization I Nuclear Common Revision 12 I Nula omn eiin1

CALCULATION CONTINUATION SHEET SHEET 40 of 45 CALC. NO.: S-C-ZZ-MDC-1920

REFERENCE:

G. Patel/NUCORE, ORIGINATOR, DATE REV: 05/17/2006 4 M. Drucker/NUCORE, REVIEWER/VERIFIER, DATE 05/18/2006 Fuel Spent Handling Fuel Building Pool Storage V = 558,550 ft3 Figure 2: FI[A Occurring In Fuel Handling Building RADTRAD Nodalization Nuclear Common Revision 12

CALCULATION CONTINUATION SHEET SHEET 41 of 45 CALC. NO.: S-C-ZZ-MDC-1920

REFERENCE:

G. Patel/NUCORE, I ORIGINATOR, DATE REV: 05/17/2006 4 M. Drucker/NUCORE, REVIEWER/VERIFIER, DATE 05/18/2006 L

Figure 3: Salem Control Room RADTRAD Nodalization Revision 12 I I

I Nuclear Common Revision 12 1

CALCULATION CONTINUATION SHEET SHEET 42 of 45 CALC. NO.: S-C-ZZ-MDC-1920

REFERENCE:

G. Patel/NUCORE, ORIGINATOR, DATE REV: 05/17/2006 4 M. Drucker/NUCORE, REVIEWER/VERIFIER, DATE 05/18/2006 Figure 4: Post-FHA CR TEDE Dose Vs Fuel Decay Time (hr)

(CB) 1.2 1.0 0.8 Li o0.6 0.4 0.2 0.0 0 20 40 60 80 100 120 Fuel Decay Time (hr)

NulaICmo Rvson1 11 i Nuclear Common Revision 12 1

CALCULATION CONTINUATION SHEET SHEET 43 of 45 CALC. NO.: S-C-ZZ-MDC-1920

REFERENCE:

Gi. Patel/NUCORE,I ORIGINATOR, DATE REV: 05/17/2006 4 M. Drucker/NUCORE, REVIEWERNERIFIER, DATE 05/18/2006 Figure 5: Post-FHA CR TEDE Dose Vs Fuel Decay Time (FHB) 0.8 0.7 0.6 E 0.5 0.4 S0.3 0.2 0.1 0.0 0 20 40 60 80 100 120 Fuel Decay Time (hr)

INuclear Common Revision 12 2I I Nula omnRvso I

CALCULATION CONTINUATION SHEET SHEET 44 of 45 CALC. NO.: S-C-ZZ-MDC-1920

REFERENCE:

G. Patel/IUIORE, ORIGINATOR, DATE REV: 05/17/2006 4 M. Drucker/NUCORE, REVIEWEERNERIFIER, DATE 05/18/2006 Figure 6: Post-FIA CR TEDE Dose Vs Fuel Decay Time (FHB Puff) 2.5 -

2.0" S1.5 : . .:

91.0 -

0.5 .

0.0 0 20 40 60 80 100 120 Fuel Decay Time (hr)

I Nuclear Common Revision 12 12 1

I NulearCommn Reisio

CALCULATION CONTINUATION SHEET SHEET 45 of 45 CALC. NO.: S-C-ZZ-MDC-1920

REFERENCE:

G. Patel/NUCORE, ORIGINATOR, DATE REV: 05/17/2006 4 M. Drucker/NUCORE, REVIEWER/VERIFIER, DATE 05/18/2006 13.0 ATTACHMENTS 13.1 CD containing the following electronic files Design Calculation S-C-ZZ-MDC-1920, Rev 4 Nuclide Inventory File SNGSFHAdef.txt Nuclide Release Fraction & Timing File SNGSFHArft.txt FGR Dose Conversion File SALEMFHAFG1 1&12.txt RADTRAD Input and Output Files for FHA Inside Containment:

S24FHA150.psf and S24FHA150.oO S48FHA150.psf and S48FHA150.oO S60FHA15O.psf and S60FHA15O.oO S72FHA150.psf and S72FHA15O.oO S96FHA15O.psf and S96FHA15O.oO RADTRAD Input and Output Files for FHA Inside Fuel Handling Building:

FB24FHA150.psf and FB24FHA150.o0 FB48FHA15O.psf and FB48FHA150.oO FB60FHA150.psf and FB60FHA150.o0 FB72FHA150.psf and FB72FHA150.o0 FB96FHA150.psf and FB96FHA150.oO RADTRAD Input and Output Files for FHA Inside Fuel Handling Building (Puff Release):

FB24PUFF150.psf and FB24PUFF150.o0 FB48PUFF150.psf and FB48PUFF150.oO FB60PUFF150.psf and FB60PUFF150.o0 FB72PUFF1 50.psf and FB72PUFF1 50.oO FB96PUFF150.psf and FB96PUFF15O.oO 13.2 Copy of Reference 10.21 (2 pages) 14.0 AFFECTED DOCUMENTS S-C-ZZ-MDC-1920, Revision 3 will be superseded.

I Nuclear Common Revision 122I 1 I Nula omnRvso

Attachment 13.1 S-C-ZZ-MDC-1920, Rev. 4 CD With Various Electronic Files

/I/

Attachment 13.2 S-C-ZZ-MDC- 1920, Rev. 4 1 of2

- _ANDU From the Desk of Paul Woods 7, -.eým.c.`.-

TO: John 0uffy - "- -

DATE: 12/16/5

SUBJECT:

Esftiatt of Unfiltered Lukage from FH8 with One Exhaust Fan 'witne Supply Fan Operiutng.

Initially, ft FHV system will be in OW no*il alignment foe fuel handling. 2.

exhaust fans and *t supply fan o"ptrng. Upon loss of one exhaust fan the bulding pressure contrmller will attempt to riodulsta open to maintain building negative pressure. Eventually the rnmaxum trveol stop wil be rached and no further exhaust flow Incraw is possible. Building pressue will onlinun to kIncease due to tMe inbalanoe botwen Me exhaust flow and the su*py ftOW.

when the buidn pressure rladw Me awm setpont (1pprox. 0.16 Inhes of wvte negave. WRT to outside) th contrl room annundator vwll Warm. Per the Aljam Response procedure Owoptrie is dreeted to shut down the opr*alsf sup* fan when fa building PMessuM lm Is roesived. Durn the podod when the operatot is evaluating the ilarm and taklno action-,tle fuel handling building may go postw.

Potential roeleas points am the tuck bay nll-up door on ti west end of the Fuel Hkndling Building, leakage t rogh the dosed gravity damper that replaced th *ruck bay exhaust fan (ghown on Dwg 207047 and located In the south wall at elevation 124'-."; ' west of Me N-N grid tecalon Ine), end leakge hrmugh the 2FHV8 uipply k handlirg unit damper located in UNe nO well at the 100' elevn &ma tl agl of nrd Ion In R.-R, also Inthuck by are (aso shown an P"g 207647).

During normal operatin ie supply fan is sot to approximately 2000 fm less than the exhaust flow nrte. The nonral exhaust flow rate Is approximately 19.490 dim. Therefore the normal supply Is wound 17.500 cdm.

The FHV exhaust fans am Ident*ial fans operated In parallel, wfM back dmet dampern cn the ofaust to permit the full flow from cne fan to be exhusted if the ther fan is stopped. The pressure vs. f response for tis effangement can be moealed by piotting te aquivalent fan auvs for both fans from the fan curve foreaiNe fan by doubting to flow rate for each cnstant total presswe point Whem oae fan Is los tMe rnsult Is that to sysltm resistance curve Is foaowed down to the Itsrsoctim of ft systm curve whh ft single fan curve.

The i and 2 fan curves can be ploe using fte vendor supplied fan curve. The equatin of i system Is gven by AP-RV". Where AP Is the fan total presse, R Inm constant, ad V isthe volurmeft flow. The operating point on the 2-fan crve i 19,490 cfmn. From 1he combhnd 2 Fan curve this cormspwda to 8.2 InW.G. totsl prmir. SlVi for R R AP/ V2 R- 2,1(19490)' -

Re 2.113(10)4 In W.GJCm Io4 IFi No7ten - R -ý

Attachment 13.2 S-C-ZZ-MDC-1920, Rev. 4 2 of 2 0 £RN-I rM. the O~sk of Paul AWods This consant is te used to plot t oystem curve, which intersocs the 1-fan curve it 14.500 elm (roter to te atscted curves).

Therefore. he flow thrvugh the normal ewtihust path (~1tered) Is14,500 efm. A consloriitive as wnupton Isthat the supply fan. continues to operats at 17.500 dmrn with 14,500 dmel*xusted to th stidc, and the remainier or 3,000 dlm lealung un14w" to the envionmo.r The me y for this type of evaluation Is pmvded in S*ction 19.4 of thirdbooI of Air ConditionAig and Rtfrfeftrweon by Shan K.Wang; McOraw-HIll, 1993.

The gravity dampor Identifled above Is Class a leakeoa pew ASME AG-I (Refer to PSBP 317245). CIass INdarnprs ar raeid forS dml per sq. ft o! face am at 1 i WO. For the 48"x 48" grav relief daW. fteisequal to 128 dm0 I In.

The maxdum pressure from the sup ply fan ISappmxiately 3 in W.G.

Uthretfors to be cormivatr, th leakage from fte dameWr Is eliimated to be 2 tfnes 128 o 256 dlm.

Uing a sirillar miolhod to the aboves the 10% And .10% system curves wW plofttd, ard to imfted leakage for jich cas was evaluated, tie results arm pnosented.bolow.

Design Airflow Deilgn +10% Noiwn -10%

2-Exhaust Fens 19,490 21.439 17,541 Sup*y Fan 17,490 1l9.439 15.541 1-Exhaust Fan 14,500 15,300 13,800 Delta (Supply- 2,990 4.139 1,$41 sing exhaust) -

Leakage t Truck 2.734 3,683 1,88S Bay.

Leakage Through 258 258 258 G-ravity Damper Themi ame see cons:erialsmsbuIlt kft the above estiatle. FMit, The estnmamd systmn cuive ignores me off** of the supply fan and only looks at the ~ft of let bvo exhaust fars Inparalel. The prourttation ff of fe' supply fan wS land to push mre ir t. ugh She exhaust filters Seond calcilah fh. *uslig WI* U. gravity tamer ignores te Manual daMper ricenaly placed in sfes with tlh damper.

The mwau damper Shcid be dosed during fuel haniling a will proveoda*"nl resistance to leokag at tis kloctlon. Th i" leaage through to grvity dame is muined to W et a pressure of 3 in W.G. although this squ "s fa riladomip, and ft leakage rot would actually " Increase by a factor of 1.7. Fourth. fth blmding pressure Is assumed to be 3 in W.G. po*i;iv. i.e. uie rnmdmum attaiable by the supply fan, In fact Widblng pressure will be Ian than w'lstrndmwn. end may be nerly retrL Reviewed By: ........ DA*ate:

NC.CC-AP.ZZ-OO10(Q)

FORM-1 CERTIFICATION FOR DESIGN VERIFICATION Reference No. S-C-ZZ-MDC-1920, Rev. 41R0

SUMMARY

STATEMENT Design verification consisted of a detailed check of the completed engineering evaluation. The method of verification included design review and line-by-line" examination.

Use of a generic design verification checklist is waived. Design input considerations and assumptions are adequately identified in the body of the design calculation.

The design calculation completely revised existing design calculation S-C-ZZ-MDC-1920, Rev 3 to perform a sensitivity study to determine the Exclusion Area Boundary (EAB), Low Population Zone (LPZ), and Control Room (CR) doses for various fuel decay times for the FHA occurring in the reactor building and fuel handling building. The analysis is revised to calculate doses at various decay times in support of an anticipated license change request LCR No. S06-07 for a Technical Specification change.

Each individual named below in the right column hereby certifies that the design verification for the subject document or document portion has been completed, the questions from the generic checklist have been reviewed and addressed as appropriate, I

and all comments have been adequately incorporated. The top right column individual is the Lead Design Verifier. SAP Order/Operation final rmations are the legal equivalent of signatures. - = 4-.

4s.b.A 3\,, V V116 1-, Mark 6tucker ,5,'-- 18 -2006*

Design Verifier Assigned By Name of Lead Design Verifier / Date (print name of Supv/Manager/Director)"

Design Verifier Assigned By Name of Design Verifier Date (print name of Supv/ManagerlDirector)

Design Verifier Assigned By Name of Design Verifier" I Date (print name of Supv/ManagerlDirector)"

Design Verifier Assigned By Name of Design Verifier" I Date (print name of SupvlManager/Director) I

If the MMaIcr/Supe*risor acts as the Design Veifier. the name of the next higher level of technical m-nagement is rcquird in the left column.

Nuclear Common Rev. 3

NC.CC-A.z-O010(Q)

FORM-2 COMMENT / RESOLUTION FORM FOR DESIGN DOCUMENT REVIEW/CHECKING OR DESIGN VERIFICATION (SAP Standard Text Key "NR/CDV2")

REFERENCE DOCUMENT NO. /REV. S-C-ZZ-MDC-1920. Revision 4TRO COMMENTS RESOLUTION ACCEPTANCE OF RESOLUTION General Incorporated.

Editorial Comments are being provided separately in the form of a redline/strikeout mark-up. The Originator may determine which editorial comments should be incorporated.

2 Section 3.0 incorporated.

It is recommended that an introductory paragraph be added to explicitly state that this analysis uses Version 3.02 of the RADTRAD computer code (Ref.

10.2) to calculate the potential radiological consequences of an FHA. The RADTRAD code is documented in NUREG/CR-6604 (Ref. 10.2). The RADTRAD code is maintained as Software MD Number A-0-ZZ-MCS-0225 (Ref. 10.33).

3 Section 3.2 Typo corrected.

The text erroneously states that the results of the tthe parametric study shown in Sections 8.2 & 8.3 indicate that a release over a two-hour period yields a higher CR dose due to a larger amount of activity entering the CR volume. In fact, the results ofthe parametric study indicate that a release based on the rapid release rate of one FHB volume per minute yields a higher CR dose. The puff release yields a higher CR dose because it results in a larger amount of unfiltered iodine activity entering the CR volume prior to the one minute start of the CREACS outside air inflow filtration.

4 Section 3.2 Information deleted.

The text discussion ofthe Reference 10.9 ARCON95 analysis of the smoke hatch is not necessary.

Reference 10.9 used the ARCON96 code. This, should also allow for the deletion of Reference 10.24 from Section 10.

5 Design Input 5.3.4 Incorporated.

Please add a new design input section (5.3.4.14) to document the X/Qs for the smoke hatch release taken S / g- '

from Reference 10.9 page 42.

Nuclear Common Rev. 3

NC.CC-AP.ZZ-0010(Q)

FORM-2 COMMENT/ RESOLUTION FORM FOR DESIGN DOCUMENT REVIEW/CHECKING QR DESIGN VERIFICATION (SAP Standard Text Key "NR/CDV2")

REFERENCE DOCUMENT NO. /REV. S-C-ZZ-MDC-1920, Revision 4110 COMMENTS RESOLUTION ACCEPTANCE OF RESOLUTION 6 Design Inputs 5.3.5.1 & 5.3.5.2 and Ref. 10.9 New Reference 10.16 added.

Design Inputs 5.3.5.1 and 5.3.5.2 present the EAB wt4 -

and LPZ X/Q values. In previous calculation 57-- 1 g-O revisions the data was taken from Reference 10.9 (which was probably Vendor Technical Document No. 321035, Rev. 3). The current analysis has replaced Reference 10.9 with Calculation SC-ZZ-MDC-1959 which provides the smoke hatch X/Q values, but which does not provide the offsite XIQ values. Please add a reference for oflsite X/Q values.

7 Section 7.2.1, Section 5.3.2.9, and FRA Inside New release rate is 100,600 fin used in the Containment RADTRAD runs analysis and the RADTRAD runs for the FH The containment: building release rate of 99,800 cfmn occurring in the containment building are 3-/'6 modeled in the RADTRAD runs is calculated in revised.

Section 7.2.1 for a containment volume of2.6E6 cf Design Input 5.3.2.2 revised this volume to 2.62E6 cf. When recalculated, the release rate will increase to approximately 100,600 cfin.

8 Section 7.4 and Design Input 5.3.5.5 Information is made consistent.

The Section 7.4 calculations model a CR intake monitor Xe-133 sensitivity of 6.0E7 rather than the 6.2E7 value shown in Design Input 5.3.5.5 (which cites Ref. 10.13 page 12). Please revise as necessary to ensure consistency with Reference 10.13.

9 Section 7.5 and Design Input 5.3.4 Incorporated.

Section 7.5 models a 0-2 hr smoke hatch X/Q of 1.14E-2. Per Reference 10.9 Section 8.4 the C -AT-maximum X/Q is 1.151-2 for the Ul smoke hatch to UI CR intake path. Please revise Section 7.5. In addition, please add a new design input section (perhaps 5.3,4.14) to document the X/Qs for the smoke hatch release taken from Reference 10.9.

10 Sections 8.1 through 83 Incorporated.

The doses are reported to the ten-thousandths rem (i.e., tenths of a nuillirem). Consider rounding the doses reported in the results with the same level of accuracy that they will be reported in the UFSAI _

11 RADTRAD runs for FIRA in containment Incorporated.

In each run, the CREACS recirculation filter is turned on at 2 minutes, instead of the 1minute value specified in Section 3.4 12 RADTRAD runs for FRA in FHB (2 hr release) Incorporated.

1) In each run, the CREACS recirculation filter is turned on at 2 minutes, instead of the I minute 4:"".

O*

value specified in Section 3.4

2) The FB24FHA24 output file name has an extra dot before its extension: "..oW" Nuclear Common Rev.3 S

COMMENT I RESOLUTION FORM FOR DESIGN DOCUMENT OWNER'S REVIEW REFERENCE DOCUMENT NO. /REV. S-C-ZZ-MDC-1920, Rev. 41R0 COMMENTS

1. Cover Sheet: The original plan was to retain our licensing basis analysis and add a sensitivity study of dose vs. decay time. However, the revised calculation eliminates the current analysis-of-record. The revision should be identified as interim rather than final.

2. Cover Sheet: The description of the revision indicates that the only parameter changed is decay time. The analysis is revised to calculate doses at various decay times in support of an anticipated submittal for a Technical Specification change. However, the other parameters that were also changed should be identified as well as why it was necessary to change other parameters.

3. General Comment: Much of the analysis is not changed. Revision bars should be used to identify the changes.

4. Design Input Parameter 5.3.4.7 unfiltered control room inleakage is changed from 4000 cfm to 150 cfm. The original plan was to retain the current analysis-of-record parameter values and add a sensitivity study of dose vs. decay time. The 4000 cfm value should be retained. Additionally, Table 1 in Reference 10.32 does not show a value of 150 cfm. 150 cfm bounds the nominal inleakage results except for the isolation mode. 4000 cfm bounds all the results even if uncertainty is included. The higher value should be retained.

5. Regulatory Change Process Determination: The original plan was to retain our licensing basis analysis and add a sensitivity study of dose vs. decay time. In that case the calculation revision would not be controlled by any of the processes identified. However, with the elimination of the current analysis-of-record, the calculation revision should be associated with a planned change to the Technical Specifications and an LCR should be identified and all aspects of the calculation revision would be controlled by the license change submittal. The RCPD should be revised to include an explanation that the calculation is revised to support the submittal.

6. 10CFR50.59 Screening: A screening is not required if all aspects of the calculation revision support the license change submittal.

J. Duffy 05/17/2006 SUBMITTED BY DATE Nuclear Common Page 25 of 27 Rev. 3

COMMENT / RESOLUTION FORM FOR DESIGN DOCUMENT OWNER'S REVIEW RESOLUTION

1. Incorporated.

2. The statement is revised.

3. Since the various sections are re-organized to standardize the calculation format, the entire calculation is considered revised.

4. The CR unfiltered inleakage licensing basis was established in the LOCA analysis, which is 150 cfm. The use of additional unfiltered inleakage makes the CR dose unnecessary conservative. The use of a very high CR unfiltered inleakage of 4,000 cfm was acceptable in absence of the tracer gas test result.

5. Incorporated.

6. 50.59 Screening is deleted.

Gopal J. Patel RESOLVED BY -4$*

-4ýDATE 05/17/2006 ACCEPTANCE OF RESOLUTION J. Duffy 05/18/2006 SUBMITTED BY DATE Nuclear Common Page 25 of 27 Rev. 3

NC.CC-AP.ZZ-0010(Q)

FORM-2 COMMENT / RESOLUTION FORM FOR DESIGN DOCUMENT REVIEW/CHECKING QR DESIGN VERIFICATION (SAP Standard Text Key "NRICDV2")

REFERENCE DOCUMENT NO. /REV. S-C-ZZ-MDC-1920. Revision 41R0 COMMENTS RESOLUTION ACCEPTANCE OF RESOLUTION 13 RADTRAD runs for FrA In FEB (puff) Incorporated.

1) In each run, the CREACS recirculation filter is turned on at 2 minutes, instead of the 1 minute value specified in Section 3.4

2) In each run, the CREACS filtered and unfiltered inflow rates are initiated at 2 minutes, instead of the I minute value specified in Section 3.4

3) In each run, the CR unfiltered inleakage rate is modeled as 4000 cfm (not 150 cfin); and consequently the CR outflow rate is also high.

4) In each run, the CR X/Q is modeled as 1.78E-3 (not l.&SE-3).

5) The run titles state that a puff release rate of 350,000 cfm is modeled. The actual modeled puff release rate is 558,550 cflr.

END ark Drucker 05114/2006 Gopal J. Patel 06 SUBMITTED BY DATE RESOLVED BY4r-4 DATE Nuclear Common 616 Rev. 3

NC.NA-AS.ZZ-0059(Q)

FORM-I REGULATORY CHANGE PROCESS DETERMINATION Document I.D.: I S-C-ZZ-MDC-1920 I Revision: 141R0

Title:

I Fuel Handling Accidents Radiological Consequences Page 1 of 4 Activity

Description:

Issuing the design calculation, which performs a sensitivity study to determine the Exclusion Area Boundary (EAB), Low Population Zone (LPZ), and Control Room (CR) doses for various fuel decay times for the FHA occurring in the reactor building and fuel handling building. The analysis is revised to calculate doses at various decay times in support of an anticipated license change request LCR No. S06-07 for a Technical Specification change.

Note that more than one process may apply. If unsure of any answer, contact the cognizant departmentfor guidance.

Activities Affected Yes No Action

1. Does the proposed activity involve a change to the Technical 0 El IfYes, contact Licensing. See NOTE In Specifications or the Operating License? Section 4.1.1 LCR No. S06-07

2. Does the proposed activity involve a change to the Quality 0 0 IfYes, contact Quality Assessment.

Assurance Plan? Example:

0 Changes to Chapter 17.2 of UFSAR 3 Does the proposed activity involve a change to the Security El E IfYes, contact Security Department.

Plan? Examples:

Change program in NC.NA-AP.ZZ-0033(Q)

Change indoor/outdoor security lighting

Placement of component or structure (permanent or temporary) within 20 feet of perimeter fence

Obstruct field of view from any manned post

Interfere with security monitoring device capability

Change access to any protected or vital area

Modify safeguards systems or equipment Does the proposed activity involve a change to the El 0 IfYes, contact Emergency Preparedness Emergency Plan? Examples:

Change ODCM/accident source term

Change liquid or gaseous effluent release path

Affect radiation monitoring instrumentation or EOP/AOP setpoints used in classifying accident severity

Affect emergency response facilities or personnel, including control room

Affect communications, computers, information systems or Met tower Nuclear Common Rev. 11

NC.NA-AS.ZZ-0059(Q)

FORM-I REGULATORY CHANGE PROCESS DETERMINATION Document I.D.: I S-C-ZZ-MDC-1920 I Revision: 141R0

Title:

I Fuel Handling Accidents Radiological Consequences Page 2 of 4 Activities Affected Yes No Action

6. Does the proposed activity involve a change to the IST El 0 IfYes, contact Engineering Programs Program Plan? Example: ISI/IST.

Affect the design or operating parameters of a Nuclear Class 1, 2, or 3 Pump or Valve (Guidance in NC.CC-AP.ZZ-0007(Q))

7. Does the proposed activity involve a change to the Fire El Z IfYes, contact Design Engineering.

Protection Program? Examples:

Change program in NC.DE-PS.ZZ-0001(Q)

Change combustible loading of safety related space

Change or affect fire detection system

Change or affect fire suppression system/component

Change fire doors, dampers, penetration seal or barriers

See NC.CC-AP.ZZ-0007 for details

Change or affect FPP compensatory measures B. Does the proposed activity involve Maintenance which El 0 IfYes, process in accordance with restores SSCs to their original design and configuration? NC.WM-AP.ZZ-0001(Q)

Examples:

CM or PM activity

Implements an approved Design Change?

Troubleshooting (which does not require 50.59 screen per SH.MD-AP.ZZ-0002)

9. Is the proposed activity a temporary change (T-Mod) which El 0 IfYes, contact Engineering.

meets all the following conditions?

Directly supports maintenance and is NOT a compensatory measure to ensure SSC operability.

Will be in effect at power operation less than 90 days.

Plant will be restored to design configuration upon completion.

SSCs will NOT be operated in a manner that could impact the function or operability of a safety related or Important-to-Safet system.

10. Does the proposed activity consist of changes to El 0 if Yes, process in accordance with maintenance procedures which do NOT affect SSC design, NC.NA-AP.ZZ-0001(Q) performance, operation or control?

Note: Procedure information affecting SSC design, performance, operation or control, including Tech Spec required surveillance and inspection, requires50.59 screening. Examples include acceptance criteria for valve stroke times or other SSC function, torque values, and types of materials (e.g., gaskets, elastomers, lubricants, etc.)

Nuclear Common Rev. 11

NC.NA-AS.ZZ-0059(Q)

FORM-1 REGULATORY CHANGE PROCESS DETERMINATION Document I.D.: I S-C-ZZ-MDC-1920 I Revision: I 41R0

Title:

I Fuel Handling Accidents Radiological Consequences Page 3 of 4 Activities Affected Yes No Action

11. Does the proposed activity involve a minor UFSAR change [] 0 IfYes, process in accordance with (including documents incorporated by reference)? NC.NA-AP.ZZ-0035(Q)

Examples:

0 Reformatting, simplification or clarifications that do not change the meaning or substance of information a Removes obsolete or redundant information or excessive detail 0 Corrects inconsistencies within the UFSAR 0 Minor correction of drawings (such as mislabeled ID)

12. Does the proposed activity involve a change to an El 0 If Yes, process in accordance with Administrative Procedure (NAP, SAP or DAP) governing the NC.NA-AP.ZZ-0001 (Q) and conduct of station operations? Examples: NC.DM-AP.ZZ-0001(Q)

Organization changes/position titles

Work control/_modification processes

13. Does the proposed activity involve a change to a regulatory E] Z If Yes, contact Licensing.

commitment?

14. Does the activity impact other programs controlled by El 0 If Yes, process in accordance with regulations, operating license or Tech Spec? Examples: applicable procedures such as:

Chemical Controls Program NC.NA-AP.ZZ-0038(Q)

NJ "Right-to-know" regulations NC.LR-AP.ZZ-0037(Q)

OSHA regulations

NJPDES Permit conditions

State and/or local building, electrical, plumbing, storm water management or "other" codes and standards S1 OCFR20 occupational exposure

15. Does the proposed activity affect the Independent Spent El ED If Yes, contact Licensing and initiate the Fuel Storage Installation (ISFSI) or the Dry Cask Storage 10CFR72.48 screening process per System (DCSS) or their analyses? Examples: NC.NA-AS.ZZ-0041 (NAS-41).

Affect the spent fuel canisters or casks

Affect the method of lifting, rigging or transporting DCSS

Challenge Spent Fuel Pool level limits or reactivity limits

Affect fire hazard analyses for the Heavy Haul Path

Affect procedures for DCSS operation or ISFSI activities

16. Has the activity already received a 10CFR50.59 Screen or El Z Take credit for IOCFR50.59 Screen or Evaluation under another process? Examples: Evaluation already performed.

Calculation

Design Change Package or OWD change ID:

Procedure for a Test or Experiment

DR/Nonconformance

Incorporation of previously approved UFSAR change

17. Is the proposed change a change to a Chemistry procedure LI 0 IfYES, no 50.59 Screen is required.

as described in paragraph 4.1.7?

Nuclear Common Rev. 11

NC.NA-AS.ZZ-0059(Q)

FORM-1 REGULATORY CHANGE PROCESS DETERMINATION Document I.D.: J S-C-ZZ-MIDC-1920 Revision-,F41R0 L

Title:

I Fuel Handling Accidents Radiological Consequences Page 4 of 4 Ifany other program or regulation may be affected by the proposed activity, contact the department indicated for further review in accordance with the governing procedure. If responsible department determines their program is not affected, attach a written explanation.

IfALL of the answers on the previous pages are "No," then check A below:

A. [] None of the activity is controlled by any of the processes above, therefore a 10CFR50.59 review IS required. Complete a 10CFR50.59 screen.

Ifone or more of the answers on the previous pages are "Yes," then check either B or C below as appropriate and explain the regulatory processes which govern the change:

B. [X] All aspects of the activity are controlled by one or more of the processes above, therefore a I OCFR50.59 review IS NOT required.

C. [ ] Only part of the activity is controlled by the processes above, therefore a 10CFR50.59 review IS required.

Complete a 50.59 screen.

Explanation:

The analysis is revised to support a planned licensing change request LCR S06-07 to reduce the fuel decay time in the Salem 1 &2 Technical Specification.

05/18/2006 Gopal J. Patel 04/04/2007 PREPAREA IN DATE NAME (PRINT) QUAL EXPIRES John F. Duffy 10/31/2007 DATE NAME (PRINT) QUAL EXPIRES A/REIEWeVNk Nuclear Common Rev. I I

ML062230324

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