Text

Request for Changes to Technical Specifications Refueling Operations - Decay Time, License Amendment Request (LAR) S08-01
	ML080930080
Person / Time
Site:	Salem
Issue date:	03/11/2008
From:	Braun R; Public Service Enterprise Group
To:	; Document Control Desk, Office of Nuclear Reactor Regulation
References
	LAR S08-01, LR-N08-0046
	Download: ML080930080 (155)
	v • d • e

PSEG Nuclear LLC P.O. Box 236, Hancocks Bridge, New Jersey 08038-0236

©Psr-Nuclea)r LL C MAR 11. 2008 10 CFR 50.90 LR-N08-0046 United States Nuclear Regulatory Commission Document Control Desk Washington, DC 20555-0001 SALEM GENERATING STATION - UNIT 1 and UNIT 2 FACILITY OPERATING LICENSENOS. DPR 70 and DPR-75 NRC DOCKET NOS. 50-272 and 50-311

Subject:

REQUEST FOR CHANGES TO TECHNICAL SPECIFICATIONS REFUELING OPERATIONS - DECAY TIME LICENSE AMENDMENT REQUEST (LAR) S08-01

References:

(1) Letter from PSEG to NRC: "Request for One-Time Change to Technical Specifications, Refueling Operations -Decay Time, LAR S07-06,,Salem Nuclear Generating Station, Unit 2, Facility Operating License DPR-75, Docket No. 50-311 ", dated October 17, 2007 (2) Letter from NRC to PSEG: "Salem Nuclear Generating Station, Unit No. 2, Issuance of Amendment Re: Refueling Operations - Decay Time (TAC No. MD7027)," dated March 5, 2008 In accordance with the provisions of IOCFR50.90, PSEG Nuclear LLC (PSEG) hereby requests an amendment of the Technical Specifications (TS) for the Salem Nuclear Generating Station, Units 1 and 2. In accordance with 10CFR50.91(b)(1), a copy of this submittal has been sent to the State of New Jersey.

PSEG proposes to revise the requirements for fuel decay time prior to commencing movement of irradiated fuel. TS 3/4.9.3 "Decay Time" is (1) revised to allow fuel movement to commence at 80 hours9.259259e-4 days 0.0222 hours 1.322751e-4 weeks 3.044e-5 months after the reactor is subcritical between October 15th and May 15 th, and,(2) relocated to the Salem UFSAR, or Technical Requirements Manual (TRM) 1.

TS 3/4.9.3 is configured on a calendar approach; the required fuel decay time, based on heatsink temperature, is longer between May 1 6 th and October 1 4 th (168 hours0.00194 days 0.0467 hours 2.777778e-4 weeks 6.3924e-5 months ) decay 1

A TRM is currently under development at Salem 4/o1/O[

,4ooI

/

MAR 11 2008 Document Control Desk Page 2 LR-N08-0046 time limit versus the time between October 15 th and May 15 th (100 hours0.00116 days 0.0278 hours 1.653439e-4 weeks 3.805e-5 months ) based on average river water temperature which is significantly cooler in the fall through spring months. This calendar approach was previously approved for Salem Units 1 and 2 via Amendments 251 and 232, respectively.

Reference 1 submitted a one-time request to change the 100 hour0.00116 days 0.0278 hours 1.653439e-4 weeks 3.805e-5 months , limit to 86 hours9.953704e-4 days 0.0239 hours 1.421958e-4 weeks 3.2723e-5 months for Salem Unit 2 refuel outage 2R16 (scheduled for Spring 2008). Reference 2 provided Amendment 271 for 2R16. The analysis provided in Reference 1 establishes the framework and basis for the enclosed evaluation requesting the permanent change to 80 hours9.259259e-4 days 0.0222 hours 1.322751e-4 weeks 3.044e-5 months for the October 1 5 th to May 1 5 th period.

The proposed change to relocate TS 3/4.9.3, and associated Bases, to the UFSAR (or TRM) is consistent with NUREG-1431, "Standard Technical Specifications, Westinghouse Plants," Revision 3, dated June 2004, with the NRC's Final Policy Statement on Technical Specifications Improvements for Nuclear Power Reactors (58 FR 39132), dated July 22, 1993, and with 10CFR 50.36.

The safety aspects of this proposed permanent change to 80 hours9.259259e-4 days 0.0222 hours 1.322751e-4 weeks 3.044e-5 months (from October 15 th through May 15th) have been assessed and are discussed in detail in Attachment 1.

Acceptable safety impact is based on the following:

1. The impact of the increase in heat load in the SFP (due to the reduced decay time) on the SFP cooling requirements has been evaluated. The evaluation demonstrates that the SFP temperature will remain within the licensing limits for both normal and abnormal operations. Appropriate procedural and program controls are in place to validate the analysis for each core offload.

2. An analysis for the Fuel Handling Accident has been performed demonstrating an acceptable minimum decay time of 24 hours2.777778e-4 days 0.00667 hours 3.968254e-5 weeks 9.132e-6 months prior to movement of irradiated fuel assemblies within the reactor vessel, which is well below the proposed change to 80 hours9.259259e-4 days 0.0222 hours 1.322751e-4 weeks 3.044e-5 months .

The relocation of TS 3/4.9.3 requirements to the UFSAR will continue to provide an appropriate level of control under the requirements of 10CFR50.59. The proposed change is similar to changes previously approved in Amendment 137 for Hope Creek Generating Station, dated January 17, 2002, and Amendment 240 for Millstone Station Unit 2, dated February 10, 2000. provides the existing TS pages marked-up to show the proposed changes. provides the existing TS Bases pages marked-up to reflect the associated changes to the TS (for information only). provides the Decay Heat-up Rates and Curves for the next scheduled outage (1.R1 9) and the sensitivity analysis for worst case SFP Heat Loads (Calculation S-C-SF-MDC-1810, Revision 8). Attachment 5 provides the Fuel Handling Accident (FHA)

Radiological Consequences Evaluation (Calculation S-C-ZZ-MDC-1920, Revision 41R0).

MAR 1, 2008 Document Control Desk Page 3 LR-N08-0046 PSEG has evaluated the proposed changes in accordance with 10CFR50.91 (a)(1),

using the criteria in 10CFR50.92(c), and has determined this request involves no significant hazards considerations. This amendment to the Salem TS meets the criteria of 10CFR51.22(c)(9) for categorical exclusion from an environmental impact statement.

PSEG requests approval of the proposed License Amendment by September 30, 2008 to be implemented within 30 days, to support Salem Unit 1 refueling outage 1R19.

If you have any questions or require additional information, please do not hesitate to contact Mr. Jeff Keenan at (856) 339-5429.

I declare under penalty of perjury that the foregoing is true and correct.

Executed on __

_/° (Date)

Sincerely, Robert C. Braun Site Vice President Salem Generating Station Attachments: 5 C

Mr. S. Collins, Administrator - Region I U. S. Nuclear Regulatory Commission 475 Allendale Road King of Prussia, PA 19406 Mr. R. Ennis, Project Manager - Salem Unit 1 and Unit 2 U. S. Nuclear Regulatory Commission Mail Stop 08B1 Washington, DC 20555-0001 USNRC Senior Resident Inspector - Salem Unit 1 and Unit 2 Mr. P. Mulligan Bureau of Nuclear Engineering PO Box 415 Trenton, New Jersey 08625 LAR S08-01 LR-N08-0046 REQUEST FOR CHANGE TO TECHNICAL SPECIFICATIONS REFUELING OPERATIONS - FUEL DECAY TIME Table of Contents

1.

DESCRIPTION...........................................................................

I

2.

PROPOSED CHANGE.................................................................

1

3.

BACKGROUND.........................................................................

2

4.

TECHNICAL ANALYSIS................................................................

5

5.

REGULATORY SAFETY ANALYSIS................................................

26

6.

ENVIRONMENTAL CONSIDERATION..............................................

31

7.

REFERENCES.........................................................................

32 LAR S08-01 LR-N08-0046

1.

DESCRIPTION PSEG proposes to revise the requirements for fuel decay time prior to commencing movement of irradiated fuel. TS 3/4.9.3 "Decay Time" is (1) revised to allow fuel movement to commence at 80 hours9.259259e-4 days 0.0222 hours 1.322751e-4 weeks 3.044e-5 months after the reactor is subcritical between October 15th and May 1 5 th, and (2) relocated to the Salem UFSAR, or Technical Requirements Manual (TRM) 2.

Currently, TS 3/4.9.3 is configured on a calendar approach; the required fuel decay time, based on heat sink temperature, is longer between May 16" and October 14" (168 hours0.00194 days 0.0467 hours 2.777778e-4 weeks 6.3924e-5 months decay time limit) versus the time between October 15t and May 1 5 th.( 1 0 0 hours0 days 0 hours 0 weeks 0 months ) based on average river water temperature which is significantly cooler in the fall through spring months. This calendar approach was previously approved for Salem Units 1 and 2 via Amendments 251 and 232, respectively.

In October 20073, PSEG submitted a one-time request to change the 100 hour0.00116 days 0.0278 hours 1.653439e-4 weeks 3.805e-5 months limit to 86 hours9.953704e-4 days 0.0239 hours 1.421958e-4 weeks 3.2723e-5 months for Salem Unit 2 refuel outage 2R16 (scheduled for Spring 2008). Amendment 271 was issued on March 5, 2008 for 2R16. The analysis provided in the October 2007 submittal establishes the framework and basis for this evaluation requesting the permanent change to 80 hours9.259259e-4 days 0.0222 hours 1.322751e-4 weeks 3.044e-5 months for the October 1 5 th to May 15 th period. The safety aspects of this proposed permanent change to 80 hours9.259259e-4 days 0.0222 hours 1.322751e-4 weeks 3.044e-5 months have been assessed and are discussed in Section 4.

The proposed change to relocate TS 3/4.9.3, and associated Bases, to the UFSAR (or TRM) is consistent with NUREG-1431, "Standard Technical Specifications, Westinghouse Plants," Revision 3, dated June 2004, with the NRC's Final Policy Statement on Technical Specifications Improvements for Nuclear Power Reactors (58 FR 39132), dated July 22, 1993, and with 10CFR 50.36. The relocation of TS 3/4.9.3 requirements to the UFSAR will continue to provide an appropriate level of control under the requirements of 1 OCFR50.59. The proposed change is similar to changes previously approved in Amendment 137 for Hope Creek Generating Station, dated January 17, 2002, and Amendment 240 for Millstone Station Unit 2, dated February 10, 2000.

2.

PROPOSED CHANGE TS LCO 3.9.3 would be revised as follows, and then relocated to the Salem UFSAR:

2 A TRM is currently under development at Salem 3

Letter from PSEG to NRC: "Request for One-Time Change to Technical Specifications, Refueling Operations -Decay Time, LAR S07-06, Salem Nuclear Generating Station, Unit 2, Facility Operating License DPR-75, Docket No. 50-311", dated October 17, 2007 1

Attachment I LAR S08-01 LR-N08-0046 The reactor shall be subcritical for at least:

a.

400 80 hours9.259259e-4 days 0.0222 hours 1.322751e-4 weeks 3.044e-5 months Applicable through year 2010.

b.

168 hours0.00194 days 0.0467 hours 2.777778e-4 weeks 6.3924e-5 months APPLICABILITY:

Specification 3.9.3.a - From October 1 5 th through May 1 5 th, during movement of irradiated fuel in the reactor pressure vessel.

Specification 3.9.3.b -

From May 1 6 th through October 1 4 tth during movement of irradiated fuel in the reactor pressure vessel.

3.

BACKGROUND 3.1 Minimum Decay Time Reduction On October 17, 2007, PSEG requested a one-time change to TS LCO 3.9.3.a for Salem Unit 2 refueling outage 2R16, scheduled for Spring 2008. Amendment 271 was issued on March 5, 2008 for 2R16. This change lowered the 100 hour0.00116 days 0.0278 hours 1.653439e-4 weeks 3.805e-5 months decay time requirement to 86 hours9.953704e-4 days 0.0239 hours 1.421958e-4 weeks 3.2723e-5 months for refuel outage 2R16.

The 100-hour decay time requirement between October 1 5 th through May 1 5 th was initially included in the Salem TS via Amendments 251 and 232 to DPR-70 and DPR-75, respectively, on October 10, 2002. This change was requested because the 168-hour requirement conservatively covered the entire year; it imposed an unnecessary penalty on plant operators in the cooler months, when refuelings are typically scheduled. The NRC staff concluded that this change was acceptable based (1) on the analysis provided for the SFP cooling capability, (2) the radiological consequences of a Fuel Handling Accident (FHA), and (3) the use of the Integrated Decay Heat Management (IDHM) Program.

(1) For SFP cooling capability, the NRC staff concluded that the proposed revisions to TS 3/4.9.3, in conjunction with the specified operational controls, ensure that the available decay heat removal capability will be maintained consistent with its importance to safety and that the SFP cooling system provides the capability to prevent a significant reduction in coolant inventory under accident conditions. Specifically, the decay heat removal capability is acceptable because: (a) the SFP cooling system will be capable of maintaining an appropriate pool temperature consistent with the current design basis during planned refueling evolutions; and, (b) with the unavailability of one spent fuel pool heat exchanger (SFHX), the cooling system will maintain SFP temperature within analyzed limits for SFP structural integrity with the remaining cooling system in operation to cool both trains.

2

Attachment I LAR S08-01 LR-N08-0046 (2) For the FHA analysis, the NRC staff found that PSEG used analysis methods and assumptions consistent with the conservative guidance of RG 1.183. The

.staff compared the radiation doses estimated by the licensee to the applicable acceptance criteria and to the results estimated by the staff in its confirmatory calculations. The staff concluded that the estimates of the TEDE due to FHA accidents will comply with the requirements of 10CFR50.67 and the guidance of RG 1.183.

(3) For the IDHM Program, the NRC staff reviewed the critical software document (CROSSTIE) and audited the decay heat management program evaluation of the most recent Salem refueling outage. The staff found that the software was calibrated against actual plant data, and that validation against SFP transients, including loss and restoration of cooling, produced acceptable agreement. The staff also found that, with representative input data, the predicted SFP temperature profile was in close agreement with that of the previous refueling outage. The SER for Amendments 251 and 232 states: "Accordingly, the staff concluded that the CROSSTIE software, as implemented in the decay heat management program, will provide an accurate representation of peak SFP temperature."

The proposed change to 80 hours9.259259e-4 days 0.0222 hours 1.322751e-4 weeks 3.044e-5 months (from October 1 5 th through May 15th) will continue to comply with the above three elements, and with the licensing basis described in the Salem UFSAR.

The Salem UFSAR, Section 9.1.3.1 states:

"The Spent Fuel Pool Cooling System maintains pool temperature at or below 1490F, provided both SFP heat exchangers are available. If only one heat exchanger is available, pool temperature is limited to 1800 F."

Salem UFSAR, 9.1.3.2, states:

"In 1998, additional spent fuel pool heat removal analyses were performed. The analyses addressed potential full-core off-loads during upcoming refueling outages as well as end of plant life. These analyses concluded one pump and one heat exchanger can maintain pool temperature below 149°F under all combinations of decay time and Component Cooling Water (CCW) temperature except minimum decay times and very high cooling water temperatures. Under these later conditions, in vessel decay-time would be extended or parallel heat exchanger operation would be used to maintain pool temperature below 149°F."

Inaddition to the above, Section 9.1.3.2 describes the SFP IDHM program under which pre-outage assessments of SFP heat loads are performed prior to core offload as follows:

Calculations to assure SFP temperature does not exceed 149°F following a full-core offload with one heat exchanger per pool.

3

Attachment I LAR S08-01 LR-N08-0046

" Calculations to assure SFP temperature does not exceed 180'F following a full-core offload with one heat exchanger for both pools.

Validation of assumptions in the Integrated Decay Heat Management program including o Availability of both heat exchangers, each with an available pump and o Actual CCW system temperatures consistent with calculated values.

The Salem UFSAR, Section 9.1.3.3 states:

"As a result of self-assessments performed on the fuel pool and associated structures, systems and components (SSCs) in 1995, concerns were identified that called into question the ability of these SSCs to perform their design basis functions under loss of normal fuel pool cooling conditions as a result of a design basis earthquake where the heat load in the pools could cause the pool temperature to exceed 1800F. These concerns were resolved by a seismic upgrade of the Spent Fuel Pool Cooling System to render temperatures above 180 OF non-credible. The upgrade not only evaluated the capability of the system to remain functional following a seismic event, but also evaluated potential single active failures and various external hazards (such as flooding, missiles, seismic-non-seismic interactions, etc.) that could result in interruption of forced cooling. The evaluation concluded pool temperature.would-be maintained 180 0F and below under normal, abnormal, and accident conditions."

3.2 Relocation of Decay Time TS Section 182a of the Atomic Energy Act of 1954, as amended (the Act) requires applicants for nuclear power plant operating licenses to include the TS as part of the license.

The NRC described the purpose of Technical Specifications in its Final Policy Statement on Technical Specifications Improvements for Nuclear Power Reactors (58 FR 39132), dated July 22, 1993. 10CFR50.36 was amended July 19, 1995 (60 FR 36953) to reflect the Final Policy Statement.

10CFR50.36 requires that the TS include items in specific categories, including:

(1) safety limits, limiting safety system settings, and limiting control settings; (2) limiting conditions for operation (LCOs); (3) surveillance requirements; (4) design features; and (5) administrative controls. The regulation does not specify the particular requirements to be included in the TSs.

The four criteria defined by 10 CFR 50.36(c)(2)(ii) for determining whether particular items are required to be included in the TS LCOs, are as follows:

4

Attachment I LAR S08-01 LR-N08-0046 (1) installed instrumentation that is used to detect, and indicate in the control room, a significant abnormal degradation of the reactor coolant pressure boundary; (2) a process variable, design feature, or operating restriction that is an initial condition of a design basis accident or transient analysis that either assumes the failure of or presents a challenge to the integrity of a fission product barrier; (3) a structure, system, or component that is part of the primary success path and which functions or actuates to mitigate a design basis accident or transient that either assumes the failure of or presents a challenge to the integrity of a fission product barrier; (4) a structure, system, or component which operating experience or probabilistic risk assessment has shown to be significant to public health and safety.

Existing TS LCOs which fall within or satisfy any of the above criteria must be retained in the TSs; those which do not fall within or satisfy these criteria may be relocated to other licensee controlled documents (e.g., the UFSAR or TRM).

The-proposed change in-this Amendment request to relocate-the Decay.Time TS is similar to changes previously approved in Amendment 137 for Hope Creek Generating Station, dated January 17, 2002, and Amendment 240 for Millstone Station Unit 2, dated February 10, 2000. In the SER for Amendment 137 for Hope Creek the NRC stated:

"The NRC staff has reviewed the licensee's submittal and agrees with the licensee's conclusion that the Decay Time LCO does not meet the criteria in 10 CFR 50.36(c)(2)(ii) requiring inclusion of this item as a TS LCO.

Based on this review and the preceding evaluation, the staff finds it acceptable to relocate TS 3/4.9.4 to the HCGS UFSAR. Any changes to these requirements after relocation to the UFSAR will require a 10 CFR 50.59 evaluation. Under 10 CFR 50.59, sufficient regulatory controls exist to ensure continued protection of public health and safety."

4.

TECHNICAL ANALYSIS 4.1 Minimum Decay Time Reduction In the effort to continuously improve outage operation and efficiency, PSEG has determined that the required activities and plant configuration needed to support commencing movement of irradiated fuel from the reactor vessel to the Spent Fuel Pool (SFP) will be completed prior to the current TS 3/4.9.3 fuel movement constraint of 100 hours0.00116 days 0.0278 hours 1.653439e-4 weeks 3.805e-5 months after subcriticality (October 1 5 th to May 1 5 th). Based on 5

Attachment I LAR S08-01 LR-N08-0046 current optimization of refueling preparatory activities, the minimum time required to establish the needed plant configuration is 80 hours9.259259e-4 days 0.0222 hours 1.322751e-4 weeks 3.044e-5 months .

PSEG previously evaluated the acceptability of 86 hours9.953704e-4 days 0.0239 hours 1.421958e-4 weeks 3.2723e-5 months for Salem Unit 2 refueling outage 2R16 (scheduled for Spring 2008). This analysis was submitted to the NRC on October 17, 2007. Amendment 271 was issued on March 5, 2008 for 2R16. The 2R16 analysis provides the framework and basis for the permanent change to 80 hours9.259259e-4 days 0.0222 hours 1.322751e-4 weeks 3.044e-5 months (from October 15 th through May 15th), as discussed below.

In order to evaluate the acceptability of permanently reducing the 100 hour0.00116 days 0.0278 hours 1.653439e-4 weeks 3.805e-5 months limit to 80 hours9.259259e-4 days 0.0222 hours 1.322751e-4 weeks 3.044e-5 months , the following evaluations were performed:

1. SFP Heat-up Rates and Temperature Curves for the next scheduled outage (1R19) and Sensitivity Analysis for Worst Case SFP Heat Loads (IDHM Program Calculation)

2. Fuel Handling Accident (FHA) Radiological Consequences

3. Validation of the Integrated Decay Heat Management Program Calculation

4. Robust Elements of the IDHM Program and Salem procedures

5. Fuel Handling Building Temperature Sensitivity 4.1.1 SFP Heat-up Rates and Temperature-Curves for. next scheduled outage (1 R1 9) and Sensitivity Analysis for Worst Case SFP Heat Loads (IDHM Program Calculation)

An evaluation, S-C-SF-MDC-1 810, Revision 8, "Decay Heat-up Rates and Curves" (Attachment 4), was performed to determine the SFP heat-up rates and temperatures based on the maximum potential heat load in the Spent Fuel Pool during 1 R1 9, with start of core off-load at 80 hours9.259259e-4 days 0.0222 hours 1.322751e-4 weeks 3.044e-5 months . A Sensitivity Analysis was also performed (Appendix A to the calculation) to address subsequent SFP Heat Loads for future refuel outages for both Unit 1 and Unit 2, demonstrating the acceptability of core offload at 80 hours9.259259e-4 days 0.0222 hours 1.322751e-4 weeks 3.044e-5 months using the IDHM Program. This analysis also resolves the current year 2010 restriction in TS 3/4.9.3.

The computer program CROSSTIE is used in this analysis to predict the SFP temperatures with both SFHXs available, and when the Unit 1 or Unit 2 heat exchanger becomes unavailable. The CROSSTIE program evaluates the SFP heat-up rates and equilibrium temperatures. The CROSSTIE program is critical software, with quality assurance requirements specified by applicable PSEG procedures.

Key elements of the 1 R1 9 analysis:

A. Decay heat (background heat in SFP plus offloaded core) is calculated based on the methodology given in NRC Branch Technical Position ASB 6

LAR S08-01 LR-N08-0046 9-2. The calculated decay heat is based on actual fuel assembly burnups and is conservatively based on 100% capacity factor.

B. Credit is taken for heat loss to the Fuel Handling Building (FHB) atmosphere. The FHB ambient temperature is conservatively assumed to be 110°F. The design value is 105 0F, which bounds the maximum calculated temperature based on the design SFP temperature of 180 0F.

The FHB relative humidity is conservatively assumed to be 100%.

C. The net SFP water volume is based on the minimum Technical Specification SFP level of 23 feet above the fuel assemblies (Technical Specification 3.9.11) and a fully offloaded core.

The program performs analyses for the following cases:

(1) Normal Cooling where each unit's SFP is aligned to its SFHX. This case determines the peak SFP temperature as a function of CCW temperature, and the CCW temperature that results in a peak SFP temperature of 1490F. A corresponding maximum SW temperature is then determined, and is compared to the historical maximum for the associated time of year.

For this case it is assumed that the outage SFP is aligned to a single SFHX. It also assumes that only one Component Cooling Heat Exchanger is aligned, which lowers the maximum allowable SW temperature.

--(2)-C.ross-co n.n.ect.

Ope.ratio.n-.whe.re.the-S F.Ps for each-unit-are -swapped-between a single SFHX, with the other Unit's SFHX unavailable. Both SFHX's are required to be available and aligned to their associated SFP prior to the start of core offload. In the unlikely event that a SFHX becomes unavailable post core off-load, this case determines the time available to swap cooling between the SFPs prior to the uncooled SFP reaching the design limit of 1800F. For this case it is conservatively assumed that cross-connect operation begins right after core offload is complete, with maximum decay heat load in the SFP.

The cross-connect re-alignment of the SFP cooling system requires less than one hour to complete; this has been validated by system walkdown.

The IDHM Program calculation provides not only the normal cooling temperature profile for the SFP; it also provides the heat-up rates of the SFP if cooling becomes temporarily unavailable (and thus requiring the need for cross-connecting). Consequently, Operations has the expected pool heat-up rates prior to fuel off-load, to support potential abnormal operations.

The following is an outline of the cross-connect process:

a. The outage (Unit 1) SFP is initially aligned to the available SFHX. If the Unit 1 SFHX is unavailable, the Unit 1 SFP would be cross-connected to the Unit 2 SFHX.

7

Attachment I LAR S08-01 LR-N08-0046

b. When the non-outage (Unit 2) SFP reaches the swapover temperature limit, cooling is swapped to the Unit 2 SFP (initial swapover) via cross-connect valve manipulations.

c. Swapping of cooling between the two SFPs via cross-connect valve manipulations continues as needed until the unavailable SFHX is returned to service.

Note that the CROSSTIE computer program has been validated against plant data as discussed in Section 4.1.3. For normal cooling following core offload, comparison of the program results to plant data demonstrates the calculated SFP temperatures are conservative. For a loss of cooling, the program results accurately predict SFP temperatures.

Results Based on the maximum heat loads for an offload time of 80 hours9.259259e-4 days 0.0222 hours 1.322751e-4 weeks 3.044e-5 months , the CROSSTIE program determines the maximum allowable CCW temperature (and corresponding SW temperature) to ensure the SFP licensing temperature limits will not be exceeded. The results for 1 R1 9 demonstrate that:

(1) For normal SFP cooling (both SFHXs available, one per unit), a peak SFP temperature of 148.3°F is-reached with a__ CCW temperature of 860 °F, correlating to a maximum SW temperature of 770F. The historical maximum SW temperature for mid-October is approximately 700F.

Consequently the SW temperature for 1 R1 9 will be less than 770F, therefore the SFP temperature limit of 149°F will not be exceeded. Note that a SW temperature of 70°F can result in a SFP temperature as low as 1420F.

(2) An initial run was performed for cross-connect operation at 180°F. With the Unit 1 SFP aligned to the Unit 2 SFHX, the isolated Unit 2 peak SFP temperature is shown to reach the licensing basis limit of 180°F with one SFHX isolated, and thus swapping of SFPs between the available SFHX is required. A second run was performed for cross-connect operation at 1700F.

This swapwover temperature limit is based on a Unit 1 SFP heatup rate of 9.1 OF/hour without cooling and a one-hour duration to complete cross-connect manipulations to ensure the uncooled SFP does not reach 180OF prior to cooling being restored. The worst-case scenario would be a loss of one SFHX just upon completing a full core offload. The results show that, after the initial swap of cooling to the non-outage (Unit 2) SFP, Operators would have a minimum of 3.1 hours1.157407e-5 days 2.777778e-4 hours 1.653439e-6 weeks 3.805e-7 months at the maximum allowable CC supply temperature of 86 0F before cross-connect valve manipulations would be required to swap cooling back to the recently offloaded SFP, and 32.2 hours2.314815e-5 days 5.555556e-4 hours 3.306878e-6 weeks 7.61e-7 months at a CC supply temperature of 860F to swap cooling back to the non-outage SFP. Additional cross-connect manipulations (if required) are bounded by 8

Attachment I LAR S08-01 LR-NO8-0046 these initial times. Cross-connect manipulations at lower CC temperatures are also bounded by these times.

The results demonstrate a decay time of 80 hours9.259259e-4 days 0.0222 hours 1.322751e-4 weeks 3.044e-5 months is acceptable for 1 Ri19. A maximum CCW temperature of 86°F is required to be procedurally verified prior to the start of core offload. This conclusion is based on the capability of the SFP cooling system to (1) maintain both Salem pools below 1490F with two SFHX available and (2) maintain both pools below 180°F with only one heat exchanger available. This capability meets the requirements of UFSAR Chapter 9.1.3.1.

Additional assurances, conservatisms and rigor have been built into the program, as discussed in Section 4.1.3 and 4.1.4.

Key elements of the Worst Case Heat Load Sensitivity Analysis:

In addition to the specific 1R19 analyses, a sensitivity analysis was also performed to address bounding SFP Heat Loads for future refuel outages, for both Unit 1 and Unit 2, using a decay time of 80 hours9.259259e-4 days 0.0222 hours 1.322751e-4 weeks 3.044e-5 months (from October 1 5 th through May 1 5th).

The worst-case decay heat load will occur when the SFP reaches full capacity following a full core offload. During a refueling outage, the SFP decay heat load

-is dominated by the " fresh".full. coreoffload. The.. background -decay heat-load.

from the pre-existing assemblies is a small percentage of the total heat load. For a given offload start time and discharge rate from the core into the SFP, the core decay heat will be relatively constant. Since the background decay heat is small in comparison, the increase in background decay heat from future outage discharges will have a small impact on the total SFP decay heat load during an outage. Thus it is expected that the impact on SFP heatup rates and peak temperatures for future outages, due to the increase in background decay heat, will be small as compared to those calculated for 1R19.

Each SFP has a total storage capacity of 1632 cells. This capacity will not be expanded; PSEG is planning to implement Dry Cask Storage at Salem. Each SFP currently has approximately 100 unusable cells due to inaccessibility, damage or non-fuel items. For the sensitivity analysis, it is conservatively assumed that all the cells are usable, to maximize the SFP heat load. Thus the outage in which each SFP will reach its net maximum capacity (accounting for unusable cells) will be earlier than that predicted to reach full capacity (assuming all cells are usable). Based on this assumption and the number of open cells following 1R19 (339), the Unit 1 SFP will reach full capacity, following a full core offload, in 1R22. Based on the number of open cells following 2R16 (511), the Unit 2 SFP will reach full capacity, following a full core offload, in 2R21.

For 1 R22, with the Unit 1 SFP at full capacity, the calculated peak SFP temperature is 143.4°F (for normal cooling), based on a CC temperature of 800F.

This is only 0.7°F above the peak SFP temperature determined for 1 R1 9 of 9

Attachment I LAR S08-01 LR-NO8-0046 142.7°F at the same CC temperature. For 2R21, with the Unit 2 SFP at full capacity, the calculated peak SFP temperature is 143.7'F (for normal cooling),

based on a CC temperature of 80'F. This is only 0.3°F above the peak Unit 1 SFP temperature at full capacity. Thus the difference between the two units with the SFP at full capacity is insignificant. Based on these results, the other cases were not performed as it is evident the additional background decay heat up through 1R22 and 2R21 has minimal impact.

A CC temperature of 80'F CC corresponds to a SW temperature of 71 OF, which bounds the historical maximum SW temperature for the period of October 15 through May 15. Therefore, with core offload starting within the period of October 15 through May 15, a full core offload start time after shutdown of 80 hours9.259259e-4 days 0.0222 hours 1.322751e-4 weeks 3.044e-5 months is acceptable for the worst case decay heat load, and thus is acceptable for all future outages on both units. This analysis also resolves the previous

'Year 2010' restriction in TS LCO 3.9.3 established by Amendments 251 and 232.

The results demonstrate a decay time of 80 hours9.259259e-4 days 0.0222 hours 1.322751e-4 weeks 3.044e-5 months (from October 15 th through May 15th) is acceptable for all future Unit 1 and Unit 2 outages. The IDHM Program will calculate a maximum allowable CCW temperature for each outage, which is required to be procedurally verified prior to the start of core offload. The analysis and controls in place will ensure the capability of the SFP cooling system to (1)rmaintain-both.-Salem.pools below.149 F. with itwoSFHX's.-available and (2) maintain both pools below 180°F with only one heat exchanger available.

This capability meets the requirements of UFSAR Chapter 9.1.3.1.

Note: The information provided below in Sections 4.1.2 through 4.1.5 was previously provided in support of LAR S07-06 and corresponding Amendment 271.

4.1.2 Fuel Handling Accident Radiological Consequences 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 and in the Fuel Handling Building (FHB). The FHA analyses are performed using the Alternative Source Term (AST) guidance in the Regulatory Guide 1.183, Appendix B, and TEDE dose criteria. For a FHA in containment, containment closure during fuel movement is not credited in the analyses. The analyses assume that activity is released to the environment through either the opened Containment Equipment Hatch (CEH) or the plant vent (PV).

The regulatory requirements in the Regulatory Guide 1.183, Appendix B are adopted as assumptions, which are incorporated as design inputs along with other plant-specific as-built design parameters.

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 10

Attachment I LAR S08-01 LR-N08-0046 assumed to be retained by the water in the fuel pool or reactor cavity (i.e., infinite decontamination factor).

The results of the evaluation (Attachment 5) demonstrate that the irradiated fuel can be moved after the reactor vessel has been sub-critical for 24 hours2.777778e-4 days 0.00667 hours 3.968254e-5 weeks 9.132e-6 months . The 24 hour2.777778e-4 days 0.00667 hours 3.968254e-5 weeks 9.132e-6 months FHA limit is well below the proposed TS change to 80 hours9.259259e-4 days 0.0222 hours 1.322751e-4 weeks 3.044e-5 months (from October 1 5 th through May 1 5 th). The doses shown in Attachment 5 of this application are less than the TEDE criteria set forth in RG 1.183 and are a small fraction of the dose criteria in 10CFR 50.67.

4.1.3 Validation of the Inteqrated Decay Heat Management Program Calculation The IDHM program calculates peak SFP temperature for each refueling outage using the CROSSTIE computer program. The CROSSTIE model was developed by HOLTEC in 1993. The model was benchmarked against Unit 1 and Unit 2 plant data obtained during Refueling Outage 1 R1 1. The model development and benchmarking is documented in HOLTEC's Verification and Validation Document HI-931099, contained in Appendix 2 of Critical Software Document S-C-SF-MCS-0113. All plant data was measured using calibrated M&TE. For each benchmark case run, the results show that the CROSSTIE curve and field-measured data

......are in,.close agreement,_with excellent comparisons between temperatures and.

heat-up rate.

To support the 2002 amendment that reduced the "time to move fuel after shutdown" from 168 hrs to 100 hrs, the calculated 2R12 SFP heat-up data (CROSSTIE) was validated against actual SFP temperature data obtained during 2R12. Actual recorded values of Component Cooling Water (CCW) supply temperatures were used as input to the IDHM calculation for 2R12, resulting in calculated SFP temperatures that closely correlated with actual SFP temperatures, and provided validation of the ability of the IDHM calculations to accurately predict maximum SFP temperatures. The decay heat loads used for the 2R12 IDHM program predictions are indicative of a typical 18-month production run, and subsequent refueling outage decay heat loads would not vary significantly. Consequently, the IDHM will provide a precise and accurate SFP heat-up, while maintaining the inherent'conservatisms of the SFP Cooling capability. The NRC Staff concurred with this conclusion in the SER for Salem Amendments 251 and 232.

In addition to the original benchmarking and validation against 2R12 plant data, recent validation of the IDHM Program has been performed. Actual data from 2R15 (Fall 2006 Unit 2 outage) and 1R18 (Spring 2007 Unit 1 outage) was compared to the predicted IDHM calculations for these outages. This information has been included in S-C-SF-MDC-1800, Revision 6. The results are depicted in the Charts 1, 2a and 2b.

11

Attachment I LAR S08-01 LR-N08-0046 Chart 1 (2R15) compares IDHM projected temperature versus actual temperature following core offload, with normal cooling available. Comparison of the IDHM calculation results to plant data demonstrates the calculated SFP temperatures conservatively correlate to actual SFP temperatures.

Charts 2a (2R15) and 2b (1R18) compare IDHM projected temperature versus actual temperature before core off-load begins, when SFHX cooling was unavailable to support a planned maintenance evolution. The IDHM Program results accurately predicted actual SFP temperatures. The SFHX was returned to service prior to core off-load, as required by the IDHM Program (both SFHXs available prior to core off-load).

PSEG concludes that the CROSSTIE model is an accurate prediction of SFP temperatures and that the criteria defined by the IDHM calculation will ensure that the maximum SFP temperatures remain within the licensing limits.

12

Attachment I LR-N08-0046 LAR S08-01 Chart I - 2R15 Offload SFP Heatup (with Cooling) 130 120 110 0100 CL 90 E

80 70 LOG SFP TEMP Crosstie unit 2 (F)

C 9o CO CD 092 N

Co 0

CD 0

092 0

CD 0

N,-

0 92 CO 0

P 0

00 0

92 CA 00 C) 0q 0

CD to C

0 10 0) 0 0

0 000 Dateltime 13

Attachment I LR-N08-0046 LAR S08-01 Chart2a - 2R15 Pre-offload SFP Heatup (without Cooling) 115 110 105 100 LL E

I_

95 90 85 80 S0 0

" 0 c

0 C

9RI 9

P 9

1?90 Co co V

.o

.o W

Co Co o

Co o

c oo Co Qo 10 (D

Co

~~~~L L.-O**-t.

to LO 0

0 0

Date/time 14 LR-N08-0046 LAR S08-01 Chart 2b - 1R18 Pre-OffLoad SFP Heatup (without Cooling) 100 95 90 85 80 E

1-75 70 65 60 C

0 C

CD C) c C)

0) 0 0

C) co 0)LOc C) c~)

(D 0)C Date/time ccz:C 115

Attachment I LAR S08-01 LR-N08-0046 4.1.4 Robust Elements of the IDHM Proqram and Salem Procedures The IDHM program controls have been further strengthened, providing more rigor into the pre-requisite requirements ensuring that adequate controls are in place for normal and abnormal operations (and that the 149 0F and 180°F temperature limits will not be exceeded for i R19 and subsequent refueling outages).

The IDHM Program, described in PSEG Procedure SC.OM-AP.ZZ-0001, Shutdown Safety Management Program - Salem Annex, implements the IDHM process for each refueling outage. PSEG procedure SI (2).OP-IO.ZZ-0007(Q),

Cold Shutdown to Refueling, establishes SFP cooling requirements. This procedure verifies the IDHM requirements (actual CCW temperature, decay time, equipment availability, familiarization of cross-connect procedures). The availability of the spent fuel cooling system is then monitored using the Outage Risk Assessment Model (ORAM) logic and Outage Risk Assessment procedure OU-AA-1 03(Q).

SC.OM-AP.ZZ-0001 has been strengthened to provide more specific controls on implementation of the IDHM Program, as discussed in the procedure steps below.

.........Excerpt from SC.OM-AP.ZZ-0001:........

5.7 Salem Integrated Decay Heat Management Program 5.7.1 The Salem Integrated Decay Heat Management Program is a two phase approach to ensuring that spent fuel cooling capability is adequate prior to the start of core offload. Phase one occurs in the pre-outage period and will be accomplished by performing reactor vessel and spent fuel pool decay heat load and heat-up calculations for use during schedule development and outage risk assessment. These calculations will confirm that:

Spent fuel pool temperatures will not exceed 149 degrees following full core offload using only one heat exchanger for each spent fuel pool

Spent fuel pool temperatures will not exceed 180 degrees following full core offload with only one heat exchanger available for both spent fuel pools Phase two validates the assumptions made in phase one and ensures that (a) the required pre-requisite parameters are met and (b) the appropriate equipment is available prior to removing the first fuel assembly from the core.

5.7.1.1 Prior to each refueling outage, a calculation shall'be performed to ensure that the predicted component cooling water (CCW) temperature is adequate to meet the required heat removal 16

Attachment I LAR S08-01 LR-N08-0046 capability for core offload after the reactor has been subcritical for the time identified by the proposed outage schedule, supported by the calculation, AND within the required Tech Spec 3.9.3 limit. This calculation will validate the time proposed by the outage schedule, and provide the maximum allowable CCW temperature. The results of this calculation shall be documented on Attachment 5 and provided to Operations Management as documentation that Tech Spec hold requirements (TS 3.9.3) will be met. [Ref. LCR S02-03]

5.7.1.2 The IDHM Program will also require the following:

Ensuring the availability of both SFP heat exchangers, each with an available spent fuel pit pump, to support spent fuel cooling for core offload; and

Verifying that actual CCW supply temperatures are consistent with the IDHM calculation input requirements, and Appropriate on shift personnel as determined by the Operations Manager have reviewed SI(2)OP-AB.SF-0001 prior to commencing core offload.

In the unlikely event that a SFP heat exchanger becomes unavailable following the start of core offload, the IDHM program calculation and procedures S1(2)OP-AB.SF-0001 provide the heat-up rates and controls to ensure both SFPs can be maintained below 180 degrees F by cross-connecting SFP cooling between units. These controls ensure there is sufficient time for (a) the initial cross-connect to be made prior to exceeding 180 degrees F in the outage unit's SFP, and if required, (b) the subsequent swapping of the available SFP heat exchanger between the outage unit and the non-outage unit SFP to maintain both SFPs below 180 degrees F.

5.7.2 provides the timeline for milestones which must be met in developing decay heat loads and heat-up curves/tables which will support schedule development, validation, outage risk assessment, and contingency planning. The risk assessment also provides the means to identify the proper time frames for taking major systems out of service for maintenance (i.e., CCW, SW, Electric Power, Other Unit's SFP Heat Exchanger).

Finalized heat-up curves shall be provided to Operations management personnel prior to the start of a refueling outage by the Outage Manager.

17

Attachment I LAR S08-01 LR-N08-0046 PSEG abnormal operating procedure for loss of spent fuel cooling, S1(2).OP-AB.SF-0001, has also been enhanced to provide more rigor and to provide a streamlined process for cross-connect operations, in the unlikely event a SFP heat exchanger becomes unavailable post core off-load. These changes provide for promptly initiating cross-connect operations with the other unit's SFHX, if required, and ensure the temperature of both SFPs is monitored, and the available SFHX is alternated between both units as required to maintain both SFPs below 1800F.

0 If unavailability of the outage unit's SFHX occurs post core off load, SI(2).OP-AB.SF-0001 will promptly initiate SFP cooling cross-connect operations and temperature trending for both SFPs. If required, the procedure will then direct the system alignments needed to alternately cool both SFPs.

0 Appropriate cautions and notes have been incorporated to account for the time required to perform cross-connect and/or restoration operations in conjunction with the monitored heat-up rates. This will ensure actions will be. initiated early enough that neither SFP will exceed 180 0F.

0 Various enhancements have been incorporated throughout to expedite the overall response strategy during a loss of Unit 2 Spent Fuel Pool Cooling.

Due to time constraints associated with cross-connecting cooling, the

..steps-to-perform -the-necessary-system--alignment -are included--in-the..

abnormal operating procedure rather than reference a separate procedure.

In the event the available SFHX must be alternated between SFPs the abnormal operating procedure also contains the necessary steps to align the system to alternately cool one SFP and then the other. The time between the alternating cycles is based on actual SFP heat up rate.

The procedure directs the operators to reference the calculations performed as part of the IDHM program, which are provided to the operators prior to the outage. These calculations will provide the operators with expected heat up rates for both the outage and non-outage SFPs.

In addition, the Salem Unit 1 and 2 SFP high temperature alarm setpoint (125 0F) is an entry condition for abnormal operating procedure S1(2).OP-AB.SF-0001(Q),

Loss of Spent Fuel Pool Cooling. With shorter core offload start times during refueling outages, there is a higher probability that the SFP could reach 1250 F. If the peak SFP temperature exceeds 125°F for a refueling outage, as predicted by the IDHM calculation, then exceeding the alarm setpoint is an expected condition, and the alarm would not be indicative of an actual loss or degradation of SFP cooling. The Alarm Response Procedure instructs the setpoint to be increased to allow refueling activities to continue. The IDHM calculation provides 18

Attachment I LAR S08-01 LR-NO8-0046 the temporary alarm setpoint that bounds the calculated peak SFP temperature in case the normal 1250F setpoint is reached.

4.1.5 Fuel Handling Building Temperature Sensitivity The Fuel Handling Building (FHB) ventilation design basis has been reviewed to ensure there is no potential impact on 1 R19 and subsequent refueling operations for both Unit 1 and 2. An ambient temperature sensitivity study was performed (included in S-C-SF-MDC-1 800, Revision 6, provided in PSEG letter dated October 17, 2007), which plots a comparison of the SFP Cross-connect temperatures based on ambient air temps of 105'F, 110°F and 120'F (see Chart 5 below).

For normal cooling, the results show that a 50F increase in ambient temperature raises the bulk temperature of the SFP by 0.2 0F (corresponding to a 4-minute decrease in the time to reach 125 0F), and a 10'F increase in ambient temperature raises the bulk temperature of the SFP by 0.4 0 F (corresponding to a 7-minute decrease in the time to reach 1250F). This has a negligible impact on the overall results of the SFP heat-up.

For a loss of cooling the results show that a 50F increase in ambient temperature raises-the-bulk -temperature -of the._SFP-by..1

. 0_E_(corresponding-to-_a_2-minute.

decrease in the time to reach 180°F), and a 100 F increase in ambient temperature raises the bulk temperature of the SFP by 0.2°F (corresponding to a 4-minute decrease in the time to reach 180°F). This has a negligible impact on the overall results of the SFP heat-up.

As discussed in Section 4.1.1, the IDHM heat-up rates calculated in S-C-SF-MDC-1800 use a conservative FHB ambient temperature value 1106F, versus the design value is 1050F.

Additional information on the FHB ventilation system was provided in PSEG letter dated January 11, 20084, in response to an NRC Request for Additional Information (RAI) on the October 17, 2007 PSEG letter:

The total heat loss from the Spent Fuel Pool (SFP) to the ambient air calculated by the CROSSTIE program consists of three parts - evaporative heat loss (Qevap), convective heat loss (Qconv) and radiation heat loss (Qrad), as discussed in Letter from PSEG to NRC: "Response to Request for Additional Information on License Amendment Request S07-06, One-Time Change to Technical Specifications, Refueling Operations -Decay Time (TAC No, 7027), Salem Nuclear Generating Station, Unit 2, Facility Operating License DPR-75, Docket No. 50-311 ", dated January 11, 2008 19

Attachment I LAR S08-01 LR-N08-0046 HOLTEC's Verification and Validation (VN) document 5. The evaporative heat is calculated based on mass transfer principles as follows:

Qevap = m

As

hfg, where m is the mass evaporation rate (Ibm/hr-ft 2), A, is the pool surface area (ft2),

and hfg is the latent heat of vaporization (BTU/Ibm). The mass evaporation rate is calculated as follows:

m = hD(AT) * (Wps - Ws) where Wps is the humidity ratio of moist air at the pool surface temperature (Ibm-vapor/lbm-dry air), Ws is the humidity ratio of moist air at the ambient air temperature, and hD(AT) is the mass transfer coefficient at the pool surface as a function of the temperature difference between the pool and ambient air.

The convective heat loss is calculated as follows:

Qconv = hc

As

AT where hc is the convective heat transfer coefficient at the pool surface (BTU/hr-ft2 'F). The radiation heat loss is calculated as follows:

Qrad = C-*-

As-5 -(-m oI

_T_4 air) where s is the emissivity of water (0.94) and Y is the Stefan-Boltzmann constant (0.1713*10-8 BTU/hr-ft2-°F).

The insensitivity of SFP temperature to ambient air temperature is consistent with the CROSSTIE model, and is a realistic representation of the expected temperature response of the pool to an actual loss of cooling event, as discussed below.

The evaporative heat loss is the dominant component in the total heat loss to the ambient air. Any differences in convective or radiation heat loss between an ambient air temperature of 105'F and 120°F is insignificant. The dominant parameter in calculating Qevap is the difference in vapor pressure between the pool surface and the ambient air (APv). (This is equivalent to the difference in humidity ratio provided in Holtec's VN documentation as humidity ratio is a function of vapor pressure). A sensitivity study on the effect of ambient temperature on the peak SFP temperature for both normal cooling and loss of forced cooling is provided in Appendix B of Calculation S-C-SF-MDC-1800, Revision 6 (Attachment 4 of LAR S07-06, PSEG letter dated October 17, 2007).

For the loss of forced cooling scenario, the peak SFP temperature is about 5

The Holtec VN document was previously docketed by PSEG letter LR-N02-0331, dated October 2, 2002. See Page 2-4 of Appendix 2 of Attachment 2 of the letter for heat loss to ambient air methodology.

20

Attachment I LAR S08-01 LR-N08-0046 205'F. The vapor pressure at an ambient air temperature of 105°F (1.10 psia) or 120'F (1.69 psia) is small compared to the vapor pressure at 205'F (12.78 psia).

The difference in the ambient air vapor pressure between 105'F and 120'F is also small such that difference between APv for 105'F versus 205°F and APv for 120°F versus 205°F is small (5.1%). Furthermore, the change in vapor pressure with respect to temperature at 205'F is significantly greater than that at 105'F and 120°F. Therefore, an ambient air temperature of 120'F versus 105'F has relatively minor impact on the peak SFP temperature for a loss of forced cooling.

The requirement for the Fuel Handling Ventilation (FHV) System to be operating to maintain the ambient design air temperature is based on Calculation S-2-FHV-MDC-0706, "FHV System Heating/Cooling Load & Air Flow Determination Calculation - Unit 2". The original calculation was performed prior to the rerack of the SFPs (1992-1993), when the peak SFP temperature during a refueling was 1500F. That calculation was based on design outside air conditions. With the rerack of the SFPs, the peak SFP temperature during a refueling increased to 1800F. The calculation was revised, crediting outside air conditions that would exist during refueling months, and concluded that the design (Fuel Handling Building) FHB room temperature of 1050F would be maintained with the FHV System operating and a SFP temperature of 180'F.

The FHB exhaust air conditions resulting from a SFP temperature of 180'F do not impact the ability of the ventilation system to perform its design function.

There-are-no-active components in--the SFP room.- -The exhaust filtration-units and exhaust fans, which are downstream of the filtration units, are located outside the FHB in the Mechanical Penetration Area. The air from the SFP room mixes with air from other rooms in the FHB before entering the exhaust filtration units (refer to Salem UFSAR Figures 9.4-3A and B). With a SFP temperature of 180°F, the dew point of the air mixture is above the ambient air temperature in the Penetration Area, and thus there is a potential for condensation. However, with natural convection on the outside of the duct and filtration units, the dominant thermal resistance term between the ambient air and exhaust air flow is the convective resistance between the ambient air and the outside wall. Thus the difference in temperature between the inside wall and bulk fluid flow will be small, and condensation will be minimal. Any condensation would form on the walls of the duct and inlet plenum, collect on the floor, and then be directed to equipment drains beneath the filtration units.

The HEPA and charcoal filters are designed to perform under 100% relative humidity. Some condensation may form initially on the fan blades, but will eventually dissipate as the blades heat up. Also, the heatup of the SFP to 180 OF is gradual and of limited duration, and thus the heatup of the exhaust air will be gradual. As such any condensation on the fan blades will be minimal, and will not impact fan operation.

Therefore, the FHV System will remain functional in the FHB ambient air environment at SFP temperatures up to 180 °F.

21 LR-N08-0046 LAR S08-01 Chart 5 - Unit 2 Loss of Cooling Based on 3111/08 Start Date at 0 hr - With Core Offloaded 210 200 190 j" 180 C-E I-(L 170 U-m 160 150 140 130 unit 2 - SFP ambient air @1 20TF, 11 0°F, I& 105TF and 1 OC% RH, 99F CCWand 90F SW i

ooý unit2-S unit 2 - S

ýP ambient air @12.0'F, 1 10°F FP ambient air @120*F, 110°F

& 105*Fand 100% RH, 80F

,&105TFand 100% RH, 75F CCW and 71FSW CCW and 66F SW unit 2-S4P ambient air @120TF, 1 10, & 105F and 100% RH,. 70F CCW and 61 F SW 10 0

2 4

6 8

time (hrs) 22

Attachment I LAR S08-01 LR-NO8-0046 4.2 Relocation of Decay Time TS The proposed change will relocate TS 3/4.9.3 Limiting Condition for Operation (LCO), associated action, and surveillance requirement to the Salem UFSAR (or TRM).

The proposed change is justified by the criteria of 10 CFR 50.36, "Technical Specifications." The four criteria described in 10 CFR 50.36(c)(2)(ii), that delineate the requirements for including a LCO in the TS, are evaluated below for LCO 3.9.3.

LCO 3.9.3 Statement The reactor shall be subcritical for at least:

a.

80 hours9.259259e-4 days 0.0222 hours 1.322751e-4 weeks 3.044e-5 months (From October 1 5 th through May 1 5 th, during movement of irradiated fuel in the reactor pressure vessel).

b.

168 hours0.00194 days 0.0467 hours 2.777778e-4 weeks 6.3924e-5 months (From May 1 6 th through October 14 th, during movement of irradiated fuel in the reactor pressure vessel).

Discussion The-TS--LCO estaiblish-es-a minimuun time requiren-ent for ieactor subcriticality before movement of irradiated fuel in the reactor pressure vessel to ensure sufficient time has elapsed to allow the radioactive decay of short-lived fission products. The minimum requirement for reactor subcriticality is consistent with the assumptions used in the fuel handling accident analyses and the resulting dose calculations using the Alternative Source Term described in Reg. Guide 1.183. It also ensures that the decay time is consistent with that assumed in the Spent Fuel Pool cooling analysis. In order not to exceed the analyzed Spent Fuel Pool cooling capability to maintain the water temperature below 1800 F, two decay time limits are provided, based on seasonal river water temperature. In addition, PSEG has developed and implemented a Spent Fuel Pool Integrated Decay Heat Management Program as part of the Salem Outage Risk Assessment. This program requires a pre-outage assessment of the Spent Fuel Pool heat loads and heat-up rates to assure available Spent Fuel Pool cooling capability prior to offloading fuel.

The period for decay following subcriticality will continue to be met for a refueling outage due to program and procedural controls on operations required before moving irradiated fuel in the reactor pressure vessel (e.g., validation of required SFP cooling capability, completion of required actions to establish plant configuration).

23 LAR S08-01 LR-N08-0046 10 CFR 50.36(c)(2)(ii) Criteria Criterion 1 -Installed instrumentation that is used to detect, and indicate in the control room, a significant abnormal degradation of the reactor coolant pressure boundary.

The movement of irradiated fuel into the SFP does not involve a reactor coolant pressure boundary or control room instrumentation that is used to detect a significant degradation of the reactor coolant pressure boundary.

Therefore, TS 3/4.9.3 does not fall within or satisfy this criterion.

Criterion 2 -A process variable, design feature, or operating restriction that is an initial condition of a design basis accident or transient analysis that either assumes the failure of or presents a challenge to the integrity of a fission product barrier.

The FHA is the related design basis accident and inherently postulates a radiological release from a dropped fuel assembly during refueling. The decay time assumption in the FHA analysis (24 hours2.777778e-4 days 0.00667 hours 3.968254e-5 weeks 9.132e-6 months ) defines the nature of the radiological products released from the breached fuel rods.

TS 3/4.9.3 restricts movement of fuel until the requisite decay time assumed in the FHA analysis has elapsed. The FHA does not assume any-further delay-in fuel movement beyond the initial hold time. -Since-the- -

administrative controls as well as the inherent delay associated with completing the required preparatory steps for moving fuel in the reactor vessel will ensure that the proposed 80 hour9.259259e-4 days 0.0222 hours 1.322751e-4 weeks 3.044e-5 months decay time (from October 15th through May 15th) will be met for a refueling outage, this TS is not needed to uphold the FHA analysis assumption. During the development of NUREG-1431, the industry/NRC agreed that this LCO could be relocated to a licensee controlled document, since it is not required to be in TS to provide adequate protection of the health and safety of the public.

Therefore, this specification should be relocated to the TS Bases document, consistent with NUREG-1431.

Criterion 3 -A structure, system, or component that is part of the primary success path and which functions or actuates to mitigate a design basis accident or transient that either assumes the failure of or presents a challenge to the integrity of a fission product barrier.

The specified decay time ensures that sufficient time has elapsed to comply with the source term assumptions of the FHA prior to transferring fuel from the reactor pressure vessel to the spent fuel storage pool. The specified decay time also ensures that the temperature limits of the SFP are not exceeded during a refueling outage. The transfer of fuel to the SFP does not provide a primary success path in accident mitigation.

Therefore, TS 3/4.9.3 does not fall within or satisfy this criterion.

24

Attachment I LAR S08-01 LR-N08-0046 Criterion 4 -A structure, system, or component which operating experience or probabilistic risk assessment has shown to be significant to the public safety.

The minimum decay time requirement will continue to ensure that, if a FHA were to occur, any radiological release would remain below 10 CFR 50.67 and RG 1.183 limits. Public health and safety is not affected by the timing of the fuel movement after the 80 hour9.259259e-4 days 0.0222 hours 1.322751e-4 weeks 3.044e-5 months decay time (from October 1 5 th through May 15th) has elapsed. The SFP heat load associated with an 80 hour9.259259e-4 days 0.0222 hours 1.322751e-4 weeks 3.044e-5 months decay time has also been shown not to be significant to public health and safety. Therefore, TS 3/4.9.3 does not fall within or satisfy this criterion.

The above evaluation demonstrates that the TS LCO for decay time does not fall within or satisfy the criteria for retention in the TS. Therefore, the decay time requirement can be relocated to the UFSAR (or TRM). Changes to the UFSAR (and TRM) are subject to the criteria of 10 CFR 50.59. Given these controls, the UFSAR (or TRM) is an appropriate document to control the decay time requirement. The proposed change is similar to changes previously approved in Amendment 137 for Hope Creek Generating Station, dated January 17, 2002, and Amendment 240 for Millstone Station Unit 2, dated February 10, 2000. In the SER for Amendment 137 for Hope Creek the NRC stated:

"The NRC staff has reviewed the licensee's submittal and agrees with the licensee's conclusion that the Decay.Time LCO does not meet the criteria in 10 CFR 50.36(c)(2)(ii) requiring inclusion of this item as a TS LCO.

Based on this review and the preceding evaluation, the staff finds it acceptable to relocateTS 3/4.9.4 to the HCGS UFSAR. Any changes to these requirements after relocation to the UFSAR will require a 10 CFR 50.59 evaluation. Under 10 CFR 50.59, sufficient regulatory controls exist to ensure continued protection of public health and safety."

25

Attachment I LAR S08-01 LR-N08-0046 5.0 Regulatory Safety Analysis 5.1 Basis for proposed no significant hazards consideration determination As required by 10CFR50.91(a), PSEG provides its analysis of the no significant hazards consideration. According to 10 CFR 50.92(c), a proposed amendment to an operating license involves no significant hazards consideration if operation of the facility in accordance with the proposed amendment would not:

1.

Involve a significant increase in the probability of occurrence or consequences of an accident previously evaluated;

2.

Create the possibility of a new or different kind of accident from any previously analyzed; or

3.

Involve a significant reduction in a margin of safety.

The determinations that the criteria set forth in 1 OCFR50.92 are met for this amendment request are indicated below:

1. Does the change involve a significant increase in the probability of occurrence or consequences of an accident previously evaluated?

... Response:...--- No...-

The proposed license amendment would allow fuel assemblies to be removed from the reactor core and be stored in the Spent Fuel Pool in less time after subcriticality than currently allowed by the TSs. Decreasing the decay time of the fuel affects the radionuclide make-up of the fuel to be offloaded as well as the amount of decay heat that is present from the fuel at the time of offload. The accident previously evaluated that is associated with the proposed license amendment is the fuel handling accident. Allowing the fuel to be offloaded in less time after subcriticality using actual heat loads does not impact the manner in which the fuel is offloaded. The accident initiator is the dropping of the fuel assembly. Since earlier offload does not affect fuel handling, there is no increase in the probability of occurrence of a fuel handling accident. The time frame in which the fuel assemblies are moved has been evaluated against the 10CFR50.67 dose limits for members of the public, licensee personnel and control room. Additionally, the guidance provided in Reg. Guide 1.183 was used for the selective application of Alternative Source Term. All dose limits are met with the reduced core offload times; and significant margin is maintained, as the minimum decay time prior to movement of fuel for the FHA analysis is 24 hours2.777778e-4 days 0.00667 hours 3.968254e-5 weeks 9.132e-6 months .

The Decay Time LCO 3.9.3 does not meet the criteria in 10 CFR 50.36(c)(2)(ii) requiring inclusion of this item as a TS LCO. Therefore it is acceptable to relocate TS 3/4.9.3 to the Salem UFSAR (or TRM). Any changes to these requirements after relocation to the UFSAR (or TRM) will require a 10 CFR 50.59 26 LAR S08-01 LR-NO8-0046 evaluation. Under 10 CFR 50.59, sufficient regulatory controls exist to ensure continued protection of public health and safety.

Therefore, the proposed license amendment does not significantly increase the probability of occurrence or the consequences of accidents previously evaluated.

2. Create the possibility of a new or different kind of accident from any accident previously evaluated.

Response

No.

The proposed license amendment would allow core offload to occur in less time after subcriticality which affects the radionuclide makeup of the fuel to be offloaded as well as the amount of decay heat that is present from the fuel at the time of offload. The radionuclide makeup of the fuel assemblies and the amount of decay heat produced by the fuel assemblies do not currently initiate any accident. A change in the radionuclide makeup of the fuel at the time of core offload or an increase in the decay heat produced by the fuel being offloaded will not cause the initiation of any accident. The accident previously evaluated that is associated with fuel movement is the fuel handling accident; no new accidents are introduced. There is no change to the manner in which fuel is being handled

.or-in the equipment used-to offload or store-the_ fuel. The.effects of the additional decay heat load have been analyzed. The analysis demonstrates that the existing Spent Fuel Pool cooling system and associated systems under worst-case circumstances would maintain licensing limits and the integrity of the Spent Fuel Pool.

The Decay Time LCO 3.9.3 does not meet the criteria in 10 CFR 50.36(c)(2)(ii) requiring inclusion of this item as a TS LCO. Therefore it is acceptable to relocate TS 3/4.9.3 to the Salem UFSAR (or TRM). Any changes to these requirements after relocation to the UFSAR (or TRM) will require a 10 CFR 50.59 evaluation. Under 10 CFR 50.59, sufficient regulatory controls exist to ensure continued protection of public health and safety.

Therefore, the proposed license amendment does not create the possibility of a new or different kind of accident from any accident previously evaluated.

3. Does the change involve a significant reduction in a margin of safety?

Response

No.

The margin of safety pertinent to the proposed changes is the dose consequences resulting from a fuel handling accident. The shorter decay time prior to fuel movement has been evaluated against 10CFR50.67 and all limits continue to be met. All dose limits are met with the reduced core offload times; and significant margin is maintained, as the minimum decay time prior to 27

Attachment I LAR S08-01 LR-N08-0046 movement of fuel for the FHA analysis is 24 hours2.777778e-4 days 0.00667 hours 3.968254e-5 weeks 9.132e-6 months . Decay heat-up calculations performed prior to the refueling outage as part of the IDHM program ensure that planned spent fuel transfer to the SFP will not result in maximum SFP temperature exceeding the design basis limit of 149 OF (with both heat exchangers available) or 180 OF (with one heat exchanger alternating between the two pools). As stated above, the changes in radionuclide makeup and additional heat load do not impact any safety settings and do not cause any safety limit to not be met. In addition, the integrity of the Spent Fuel Pool is maintained.

The time frame in which the fuel assemblies are moved has been evaluated against the 10CFR50.67 dose limits for members of the public, licensee personnel and control room. Additionally, the guidance provided in Reg. Guide 1.183 was used. Calculations performed conclude that expected dose limits following a Fuel handling Accident are met with the proposed decay time prior to commencing fuel movement.

The Decay Time LCO 3.9.3 does not meet the criteria in 10 CFR 50.36(c)(2)(ii) requiring inclusion of this item as a TS LCO. Therefore it is acceptable to relocate TS 3/4.9.3 to the Salem UFSAR (or TRM). Any changes to these requirements after relocation to the UFSAR (or TRM) will require a 10 CFR 50.59 evaluation. Under 10 CFR 50.59, sufficient regulatory controls exist to ensure continued protection of public-health -and safety....

Therefore, the proposed changes do not involve a significant reduction in a margin of safety.

Based on this review, it is concluded that the three standards of 10 CFR 50.92(c) are satisfied. Therefore, PSEG proposes that a finding of "no significant hazards consideration" is justified.

5.2 Applicable Reciulatory RequirementslCriteria 10 CFR 50 Appendix A, General Design Criteria 5--Sharinq of structures, systems, and components.

GDC 5 requires that structures, systems, and components important to safety shall not be shared among nuclear power units unless it can be shown that such sharing will not significantly impair their ability to perform their safety functions, including, in the event of an accident in one unit, an orderly shutdown and cooldown of the remaining units.

In the unlikely event that a SFP heat exchanger becomes unavailable after core off-load begins, the sharing of one SFP heat exchanger between units has been adequately evaluated. In the worst case scenario, Operations has adequate time to cross-connect the available heat exchanger to the refuel pool before 180'F is 28 LAR S08-01 LR-N08-0046 reached; the time to make the cross-connect has been appropriately addressed with procedural controls. When the HX is aligned to the non-refuel pool, temperature is rapidly brought down, so the time needed to swap back again to the non-refuel pool is much greater. In addition, Operations has the expected pool heat-up rates prior to fuel off-load; they monitor actual heat-up rates, so they can anticipate required actions.

10 CFR 50 Appendix A, General Design Criteria 61--Fuel storage and handling and radioactivity control.

The fuel storage and handling, radioactive waste, and other systems which may contain radioactivity shall be designed to assure adequate safety under normal and postulated accident conditions. These systems shall be designed (1) with a capability to permit appropriate periodic inspection and testing of components important to safety, (2) with suitable shielding for radiation protection, (3) with appropriate containment, confinement, and filtering systems, (4) with a residual heat removal capability having reliability and testability that reflects the importance to safety of decay heat and other residual heat removal, and (5) to prevent significant reduction in fuel storage coolant inventory under accident conditions.

The changes proposed by the amendment request do not reduce the existing UFSAR. requirements for meeting.GDC 61. The.heat removal capability of the.

SFP cooling system is maintained for normal, abnormal and accident conditions.

10 CFR 50 Appendix A, General Design Criteria 19, Control Room PSEG has applied the guidelines provided by 10 CFR 50.67 and RG 1.183, which is consistent with the current requirements of GDC 19 for the Fuel Handling Accident.

NRC Regulatory Guide 1.183, "Alternative Radiological Source Terms for Evaluating Design Basis Accidents at Nuclear Power Reactors".

The NRC's traditional methods for calculating the radiological consequences of design basis accidents are described in a series of regulatory guides and SRP chapters. That guidance was developed to be consistent with the TID-14844 source term and the whole body and thyroid dose guidelines stated in 10 CFR 100.11. Many of those analysis assumptions and methods are inconsistent with the ASTs and with the total effective dose equivalent (TEDE) criteria provided in 10 CFR 50.67. This guide provides assumptions and methods that are acceptable to the NRC staff for performing design basis radiological analyses using an AST. This guidance supersedes corresponding radiological analysis assumptions provided in other regulatory documents when used in conjunction with an approved AST and the TEDE criteria provided in 10 CFR 50.67.

This application and the supporting analyses comply with this guidance as it applies to a Fuel Handling Accident.

29

Attachment I LAR S08-01 LR-N08-0046 Title 10, Code of Federal Regulations, Part 50 Section 67, "Accident Source Term".

10CFR50.67 permits licensees to voluntarily revise the accident source term used in design basis radiological consequences analyses. This document is part of a 10CFR50.90 license amendment application and evaluates the consequences of a design basis fuel handling accident as previously described in the Salem UFSAR.

USNRC Branch Technical Position ASB 9-2, Residual Decay Heat for Light-Water Reactors for Long-Term Cooling, Revision 2 of July 1981.

BTP ASB 9-2 uses a conservative approach for calculating fuel element decay heat, and is applied to this amendment without scaling factors or other adjustments.

Regulatory Guide 1.25, "Assumptions Used for Evaluating the Potential Radiological Consequences of a Fuel Handling Accident in the Fuel Handling and Storage Facility for Boiling and Pressurized Water Reactors".

RG 1.183 supersedes corresponding radiological assumptions provided in other

_-regulatory guides and standard review plan -chapters when used in-conjunction.,

with an approved alternative source term and the TEDE provided in 10 CFR 50.67.

1OUFR100, "Determination of Exclusion Area, Low Population Zone and Population Center Distance".

10CFR100.11 provides criteria for evaluating the radiological aspects of reactor sites. A footnote to 1 OCFR1 00.11 states that the fission product release assumed in these evaluations should be based on a major accident involving substantial meltdown of the core with subsequent release of appreciable quantities of fission products. A similar footnote appears in 10CFR50.67. In accordance with the provisions of 10CFR50.67(a), PSEG applied the dose reference values in 10CFR50.67 (b) (2) in the analyses in lieu of 10CFR100 for the Fuel Handling Accident.

NUREG-0800, Standard Review Plan, Section 15.7.4, "Radiological Consequences of Fuel Handling Accidents".

The SRP Section 15.7.4 describes the radiological effects of a postulated Fuel Handling Accident. The SRP does not directly refer to the guidance of RG 1.183 or 10CFR50.67. Instead, it refers to regulatory documents, which are superseded by the selective application of the Alternative Source Term for the Fuel Handling Accident.

30

Attachment I LAR S08-01 LR-N08-0046 5.3 Conclusion The FHA dose analyses were performed in accordance with AST and TEDE guidelines provided in Regulatory Guide 1.183 and 10CFR50.67. The assumptions and design inputs are listed in Engineering Calculations provided in the reference section. The SFP Cooling Capacity calculations were performed applying acceptable NRC guidance and conservatism aspects resulting in assurance that the design basis limits for SFP heat removal are maintained. Use of the IDHM Program will ensure that SFP temperature limits are not exceeded.

The doses shown in Attachment 5 of this application are less than the TEDE criteria set forth in RG 1.183 and are a small fraction of the dose criteria in 1 OCFR50.67.

In conclusion, based on the considerations discussed above, (1) there is reasonable assurance that the health and safety of the public will not be endangered by operation in the proposed manner, (2) such activities will be conducted in compliance with the Commission's regulations, and (3) the issuance of the amendment will not be inimical to the common defense and security or to the health and safety of the public.

6.0

'ENVIRONMENTAL CONSIDERATION.-

Pursuant to 1 OCFR51.22(b), an evaluation of this license amendment request has been performed to determine whether or not it meets the criteria for categorical exclusion set forth in 10CFR51.22(c)(9) of the regulations.

PSEG has concluded that implementation of this amendment will have no adverse impact upon the Salem units; neither will it contribute to any significant additional quantity or type of effluent being available for adverse environmental impact or personnel exposure. The change does not introduce any new effluents or significantly increase the quantities of existing effluents. As such, the change cannot significantly affect the types or amounts of any effluents that may be released offsite. The new consequences of the revised Fuel Handling Accident analysis remain well below the acceptance criteria specified in 1 OCFR50.67 and Regulatory Guide 1.183.

It has been determined there is:

1.

No significant hazards consideration,

2.

No significant change in the types, or significant increase in the amounts, of any effluents that may be released offsite, and

3.

No significant increase in individual or cumulative occupational radiation exposures involved.

31

Attachment I LAR S08-01 LR-N08-0046 Therefore, this amendment to the Salem TS meets the criteria of 10CFR51.22(c)(9) for categorical exclusion from an environmental impact statement.

7.0 REFERENCES

7.1 PSEG Calculation S-C-ZZ-MDC-1920, Revision 41R0 7.2 PSEG Calculation S-C-SF-MDC-1810, "Decay Heat-up Rates and Curves", Revision 8 7.3 Critical Software Document S-C-SF-MCS-0113, CROSSTIE 7.4 PSEG Salem Units 1 and 2, Final Safety Analysis Report 7.5 PSEG Salem Units 1 and 2, Technical Specifications 7.6 PSEG, Shutdown Safety Management Program, OU-AA-103(Q) 7.7 PSEG Procedure, Salem Units I and 2, SI(2).OP-AB.SF-0001(Q),

Loss of Spent Fuel Pool Cooling 7.8 PSEG Procedure, SC.OM-AP.ZZ.0001, Shutdown Safety Management Program - Salem Annex 7.9 PSEG Procedure, Salem Units 1 and 2, SI(2).OP-IO.ZZ-0007, Cold Shutdown to Refueling 7.10 10 CFR 50.67, "Accident Source Term" 7.11 Regulatory Guide 1.183, "Alternative Radiological Source Terms for

-Evaluating. Design Basis Accidents at Nuclear Power Reactors"-.

7.12 NRC Branch Technical Position ASB 9-2 Revision 2 of July 1981, USNRC Standard Review Plan 9.2.5, Ultimate Heat Sink, NUREG 0800 32 LR-N08-0046 LAR S08-01 TECHNICAL SPECIFICATION PAGES WITH PROPOSED CHANGES The following Technical Specifications for Facility Operating License DPR-70 and DPR-75 are affected by this change request:

DPR-70, Salem Unit I Technical Specification Page Ix 3/4 9-3 Index 3/4.9.3 DPR-75, Salem Unit 2 Technical Specification Page Ix 3/4 9-3 Index 3/4.9.3

INDEX LIMITING CONDITIONS FOR OPERATION AND SURVEILLANCE REQUIREMENTS SECTION PAGE 3/4.9 REFUELING OPERATIONS 3/4.9.1 BORON CONCENTRATION....................................3/4 9-1 3/4.9.2 INSTRUMENTATION........................................3/4 9-2 3/4.9.3 DELETED......................................3/4 9-3 3/4.9.4 CONTAINMENT BUILDING PENETRATIONS...................3/4 9-4 3/4.9.5 COMMUNICATIONS.........................................3/4 9-5 3/4.9.6 MANIPULATOR CRANE OPERABILITY........................3/4 9-6 3/4.9.7 CRANE TRAVEL -

FUEL HANDLING AREA..

.3/4 9-7 3/4.9.8 RESIDUAL HEAT REMOVAL AND COOLANT 3/-4-.9.

3/4. 9.

3/4.9.

3/4. 10 3/4.10 3/4. 10 3/4.10 3/4.10 CIRCULATION All Water Levels Low Water Level 9-DELETED 10 WATER LEVEL -

REACTOR VESSEL 11 STORAGE POOL WATER LEVEL 12 FUEL HANDLING AREA VENTILATION SYSTEM SPECIAL TEST EXCEPTIONS

.1 SHUTDOWN MARGIN.........

.2 GROUP HEIGHT, INSERTION AND POWER DISTRIBUTION LIMITS................

.3 PHYSICS TESTS

.4 NO FLOW TESTS 3/4 9-8 3/4 9-8a 3/4 9-10 3/4 9-11 3/4 9-12 3/4 10-1 3/4 10-2 3/4 10-3 3/4 10-4 SALEM -

UNIT 1IXnedetNo ix Amendment No.

RLIIING OPERATIONFS LIIIGC TION FOR OPERATION 3.9.3 The react shall be subcritical for at least:.

a.

80 ho s

b. 168 hou APPLICABILITY:

Specificati n 3.9.3.a -

From October 1 5t rough May 1 5th during moveme t of irradiated fuel in t

reactor pressure vessel.

Specification 3.9.

b From 1 6 th through October 1 4 th, during movement of radiate fuel in the reactor pressure vessel.

ACTION:

With the reactor subcritical for ess tthan te required time, suspend all SURVEILLANCE REQUIREsi NTSo apibl iraitduel in the reactor pressure vessel.

DELETED SALEM -

UNIT 1 3/4 9-3 Amendment No.

INDEX LIMITING CONDITIONS FOR OPERATION AND SURVEILLANCE REQUIREMENTS SECTION PAGE 3/4.9 REFUELING OPERATIONS 3/4.9.1 BORON CONCENTRATION 3/4 9-1 3/4.9.2 INSTRUMENTATION 3/4 9-2 3/4.9.3 DELETED......................................

3/4 9-3 3/4.9.4 CONTAINMENT BUILDING PENETRATIONS 3/4 9-4 3/4.9.5 COMMUNICATIONS 3/4 9-5 3/4.9.6 MANIPULATOR CRANE OPERABILITY 3/4 9-6 3/4.9.7 CRANE TRAVEL -

FUEL HANDLING AREA.....................

3/4 9-7 3/4.9.8 RESIDUAL HEAT REMOVAL AND COOLANT 3/4.9.

3/4.9.

3/4.10 3/4.10 3/4.10 3/4.10 3/4.10 9

10 11 12 CIRCULATION All Water Levels.......................................

3/4 Low Water Level.......................................

3/4 DELETED WATER LEVEL -

REACTOR VESSEL..........................

3/4 STORAGE POOL WATER LEVEL 3/4 FUEL HANDLING AREA VENTILATION SYSTEM 3/4 9-8 9-9 9-11 9-12 9-13 I

SPECIAL TEST EXCEPTIONS

.1 SHUTDOWN MARGIN 3/4 10-1

.2 GROUP HEIGHT, INSERTION AND POWER DISTRIBUTION LIMITS....................................

3/4 10-2

.3 PHYSICS TESTS.........................................

3/4 10-4

.4 NO FLOW TESTS.........................................

3/4 10-5 SALEM -

UNIT 2 IX Amendment No.

REED ING OPERATIONS DECAY TI LIMITING CON TION FOR OPERATION 3.9.3 The react shall be subcritical for at least:

a.

80 hours

b. 168 hours0.00194 days 0.0467 hours 2.777778e-4 weeks 6.3924e-5 months APPLICABILITY:

Specificati 3.9.3.a -

From October through May 1 5 th, during moveme of irradiated fuel In the reactor pressure vessel.

Specification 3.9.3.

Fr May 1 6 th through October 1 4th, during movement of ia ated fuel in the reactor pressure vessel.

ACTION:

With the reactor subcri cal for less than the r quired time, suspend all operations involving vement of irradiated fuel the reactor pressure vessel. The provisio of Specification 3.0.3 are not a licable.

DE4.9.L3hET e reactor shall be determined to have been subcritical aDrequired byrification of the date and time of subcriticality prior to mov ent of a ite ul in the reactor pressure vessel.

DELETED SALEM - UNIT 2 3/4 9-3 Amendment No.

2 LAR S08-01 LR-N08-0046 PROPOSED CHANGES TO TS BASES PAGES The following Technical Specifications Bases for Salem Unit 1 and Unit 2, Facility Operating License No. DPR-70 and DPR-75 are affected by this change request:

Salem Unit I Technical Specification Page Index XV [need to check this page in DCRMS]

B 3/4.9.3 B 3/4.9.1b and lc Salem Unit 2 Technical Specification Paqe Index XV [need to check this page in DCRMS]

B 3/4.9.3 B 3/4.9.1b and lc

INDEX BASES SECTION 3/4.9 3/4.9.1 3/4.9.2 3/4.9.3 3/4.9.4 3/4.9.5 3/4.9.6 3/4.9.7 3/4.9.8 3/4.9.9 3/4.9.10 and 3/4.9.11 3/4.9.12 PAGE BASES REFUELING OPERATIONS BORON CONCENTRATION.........

INSTRUMENTATION...........

DELETED...........

B 3/4 9 CONTAINMENT BUILDING PENETRATIONS.........

COMMUNICATIONS MANIPULATOR CRANE...........

CRANE TRAVEL -

SPENT FUEL STORAGE BUILDING RESIDUAL HEAT REMOVAL AND COOLANT CIRCULATION CONTAINMENT PURGE AND PRESSURE-VACUUM RELIEF ISOLATION SYSTEM WATER LEVEL -

REACTOR VESSEL AND STORAGE POOL.

FUEL HANDLING AREA VENTILATION SYSTEM.

B 3/4

.B 3/4

-lb B 3/4 B 3/4 B 3/4 B 3/4 B 3/4 9-1 9-lb 9-1c 9-3 9-3 9-3 9-3 B 3/4 9-4 B 3/4 9-4 B 3/4 9-4 3/4.10 SPECIAL TEST EXCEPTIONS 3/4.10.1 SHUTDOWN MARGIN......

3/4.10.2 GROUP HEIGHT, INSERTION AND POWER DISTRIBUTION LIMITS....

3/4.10.3 PHYSICS TESTS........

3/4.10.4 NO FLOW TESTS........

B 3/4 10-1

.B 3/4 B 3/4 B 3/4 10-1 10-1 10-1 SALEM -

UNIT I XV Amendment No.

3/4.9 REFUELING OPERATIONS BASES In addition to immediately suspending CORE ALTERATIONS and positive reactivity additions, boration to restore the concentration must be initiated immediately.

In determining the required combination of boration flow rate and concentration, no unique Design Basis Event must be satisfied. The only requirement is to restore the boron concentration to its required value as soon as possible.

In order to raise the boron concentration as soon as possible, the operator should begin boration with the best source available for unit conditions.

Once actions have been initiated, they must be continued until the boron concentration is restored.

The restoration time depends on the amount of boron that must be injected to reach the required concentration.

The Surveillance Requirement (SR) ensures that the coolant boron concentration in the RCS, and connected portions of the refueling canal, the fuel storage pool and the refueling cavity, is within the COLR limits.

The boron concentration of the coolant in each required volume is determined periodically by chemical analysis.

Prior to reconnecting portions of the refueling canal, the fuel storage pool or the refueling cavity to the RCS, this SR must be met per SR 4.0.4. If any dilution activity has occurred while the cavity or canal was disconnected from the RCS, this SR ensures the correct boron concentration prior to communication with the RCS.

A minimum frequency of once every 72 hours8.333333e-4 days 0.02 hours 1.190476e-4 weeks 2.7396e-5 months is a reasonable amount of time to verify the boron concentration of representative samples.

The frequency is based on operating experience, which has shown 72 hours8.333333e-4 days 0.02 hours 1.190476e-4 weeks 2.7396e-5 months to be adequate.

3/4.9.2 INSTRUMENTATION The OPERABILITY of the source range neutron flux monitors ensures that redundant monitoring capability is available to detect changes in the reactivity condition of the core.

3/4c u i.3 DECAY TIME The minmmr

.rment for reactor subcriticality prior to ent of/*

irradiated fuel ass ies in the reactor pressure ves *ensures that suffcient time has elaps og allow the raia edcay of the short lived calculations using the Alt ive Source Te described in Reg. Guide 1.183.

SALEM -

UNIT 1 B 3/4 9-1b Amendment No.

3/4.9 REFUELING OPERATIONS BASES The inimum requirement for reactor subcriticality also ensures that the decay t is consistent with that assumed in the Spent Fuel Pool coo ig analysis. be are River water average temperature between Octo 1 5th and May 15th is determ d from historical data taken over 30 rs.

The use of 30 years of data to se t maximum temperature is co stent with Reg.

Guide 1.27, "Ultimate Heat Sink Nuclear Power P A core offload has the potential to r during both applicability time frames.

In order not to exce he analyz Spent Fuel Pool cooling capability to maintain water temperature be 180'F, two decay time limits are provid

.In addition, PSEG has develope nd implemented a Spent Fuel Pool I rated Decay Heat Management Program as pa~r f the Salem Outage sk Assessment.

This program requires a pre-outage as ý;sment of the nt Fuel Pool heat loads and heatup rates to assure available Spe Fuel Pool cooling capability prior to offloading fuel.

3/4.9.4 CONTAINMENT BUILDING PENETRATIONS During movement of irradiated fuel assemblies within containment the requirements for containment building penetration closure capability and OPERABILITY ensure that a release of fission product radioactivity within containment will not exceed the guidelines and dose calculations described in Req. Guide 1.183, Alternative Radiological Source Term for Evaluating Design Basis Accidents at Nuclear Power Reactors.

In MODE 6, the potential for containment pressurization as a result of an accident is not likely.

Therefore, the requirements to isolate the containment from the outside atmosphere can be less stringent.

The LCO requirements during movement of irradiated fuel assemblies within containment are referred to as "containment closure" rather than containment OPERABILITY.

For the containment to be OPERABLE, CONTAINMENT INTEGRITY must be maintained.

Containment closure means that all potential containment atmosphere release paths are closed or capable of being closed.

Closure restrictions include the administrative controls to allow the opening of both airlock doors and the equipment hatch during fuel movement provided that:

1) the equipment inside door or an equivalent closure device installed is capable of being closed with four bolts within 1 hour1.157407e-5 days 2.777778e-4 hours 1.653439e-6 weeks 3.805e-7 months by a designated personnel; 2) the airlock door is capable of being closed within 1 hour1.157407e-5 days 2.777778e-4 hours 1.653439e-6 weeks 3.805e-7 months by a designated personnel,

3) either the Containment Purge System or the Auxiliary Building Ventilation System taking suction from the containment atmosphere are operating and 4) the plant is in Mode 6 with at least 23 feet of water above the reactor pressure vessel flange.

Administrative requirements are established for the responsibilities and appropriate actions of the designated personnel in the event of a Fuel Handling Accident inside containment.

These requirements include the responsibility to be able to communicate with the control room, to ensure that the equipment hatch is capable of being closed, and to close the equipment hatch and personnel airlocks within 1 hour1.157407e-5 days 2.777778e-4 hours 1.653439e-6 weeks 3.805e-7 months in the event of a fuel handling accident inside containmient.

These administrative controls ensure containment closure will be established in accordance with* and not to exceed the dose calculations performed using guidelines of Regulatory Guide 1.183.

SALEM -

UNIT 1 B 3/4 9-1c Amendment No.

INDEX BAS ES SECTION 3/4.9 3/4.9. 1 3/4.9.2 3/4. 9.3 3/4.9. 4 3/4.9.5 3/4.9. 6 3/4.9.7 3/4.9.8 3/4.9. 9 3/4. 9. 10 and 3/4.9.11 3/4.9. 12 PAGE REFUELING OPERATIONS BORON CONCENTRATION...............................

INSTRUMENTATION....................................

DELETED...................................B 3/4 9-1 CONTAINMENT BUILDING PENETRATIONS.....

COMMUNICATIONS MANIPULATOR CRANE.................................

CRANE TRAVEL -

SPENT FUEL STORAGE BUILDING RESIDUAL HEAT REMOVAL AND COOLANT CIRCULATION CONTAINMENT PURGE AND PRESSURE-VACUUM RELIEF ISOLATION SYSTEM WATER LEVEL -REACTOR VESSEL AND STORAGE POOL FUEL HANDLING AREA VENTILATION SYSTEM.........

B 3/4 B 3/4 b

B 3/4 B 3/4 B 3/4 B 3/4 B 3/4 9-1 9-lb 9-1c 9-3 9-3 9-3 9-3 B 3/4 9-4 B 3/4 9-4 B 3/4 9-4 3/4.10 SPECIAL TEST EXCEPTIONS 3/4.10.1 SHUTDOWN MARGIN.......................

3/4.10.2 GROUP HEIGHT, INSERTION AND POWER DISTRIBUTION LIMITS...........

3/4.10.3 PHYSICS TESTS..............

3/4.10.4 NO FLOW TESTS..............

B 3/4 10-1 B

.. B3/4

.B3/4

.B3/4 10-1 10-1 10-1 SALEM -

UNIT 2 X

mnmn o

XV Amendment No.

3/4.9 REFUELING OPERATIONS BASES In addition to immediately suspending CORE ALTERATIONS and positive reactivity additions, boration to restore the concentration must be initiated immediately.

In determining the required combination of boration flow rate and concentration, no unique Design Basis Event must be satisfied. The only requirement is to restore the boron concentration to its required value as soon as possible.

In order to raise the boron concentration as soon as possible, the operator should begin boration with the best source available for unit conditions. Once actions have been initiated, they must be continued until the boron concentration is restored.

The restoration time depends on the amount of boron that must be injected to reach the required concentration.

The Surveillance Requirement (SR) ensures that the coolant boron concentration in the RCS, and connected portions of the refueling canal, the fuel storage pool and the refueling cavity, is within the COLR limits.

The boron concentration of the coolant in each required volume is determined periodically by chemical analysis.

Prior to reconnecting portions of the refueling canal, the fuel storage pool or the refueling cavity to the RCS, this SR must be met per SR 4.0.4.

If any dilution activity has occurred while the cavity or canal was disconnected from the RCS, this SR ensures the correct boron concentration prior to communication with the RCS.

A minimum frequency of once every 72 hours8.333333e-4 days 0.02 hours 1.190476e-4 weeks 2.7396e-5 months is a reasonable amount of time to verify the boron concentration of representative samples.

The frequency is based on operating experience, which has shown 72 hours8.333333e-4 days 0.02 hours 1.190476e-4 weeks 2.7396e-5 months to be adequate.

3/4.9.2 INSTRUMENTATION The OPERABILITY of the source range neutron flux monitors ensures that redundant monitoring capability is available to detect changes in the reactivity condition of the core.

3/4.9.3 DECAY TIME The minimum uirement for reactor subcriticality prior to movement irradiated fuel emblies in the reactor pressure vessel ensur at sufficient time has e sed to allow the radioactive deca the short lived fission products.

The 80-r decay time is consist with the assumptions used in the fuel handling acci analyses and e resulting dose calculations using the Alternative rce m described in Reg. Guide 1.183.

The minimum requirement f reactor subcriticality so ensures that the decay time is consi nt with that assumed in the Spen el Pool cooling analysis. Del e River water average temperature between ber 15th and May 15th,determined from historical data taken over 30 years.

e use of 30 ars of data to select maximum temperature is consistent with Reg.

ide 1.27, "Ultimate Heat Sink for Nuclear Power Plants".

SALEM -

UNIT 2 B 3/4 9-lb Amendment No.

.Writers Note: Text relocated from Page B 3/4 9-ic

3/4.9 REFUELING OPERATIONS BASES A core offload has e po entiaI to occur during both applicab tme mes.

In order not to exceed the analyzed Spent Fuel Pool cooling capa ity to maintain the water temperature below 180'F, two decay ti limits a rovided.

In addition, PSEG has developed and implemen a Spent Fuel Pool Int rated Decay Heat Management Program as part of e Salem Outage Risk Asses ent. This program requires a pre-out assessment of the Spent Fuel Pool heat ads and heat-up rates to ass available Spent Fuel Pool cooling capability p or to offloading fue During movement of irradiated fuel assemblies within containment the requirements for containment building penetration closure capability and OPERABILITY ensure that a release of fission product radioactivity within

.containment willla not exceed the guidelines and d-ose calculations described in Reg Guide 1.183, Alternative Radiological Source Term for Evaluating Design Basis Accidents at Nuclear Power Plants.

In MODE 6,

the potential for containment pressurization as a

result of an accident is not likely.

Therefore, the requirements to isolate the containment from the outside atmosphere can be less stringent.

The LCO requirements during movement of irradiated fuel assemblies within containment are referred to as containment closure" rather than containment OPERABILITY.

For the containment to be

OPERABLE, CONTAINMENT INTEGRITY must be maintained.

Containment closure means that all potential release paths are closed or capable of being closed.

Closure restrictions include the administrative controls to allow the opening of both airlock doors and the equipment hatch during fuel movement provided that: 1) the equipment inside door or an equivalent closure device installed is capable of being closed with four bolts within 1 hour1.157407e-5 days 2.777778e-4 hours 1.653439e-6 weeks 3.805e-7 months by a designated personnel;

2) the airlock doors are capable of being closed within 1 hour1.157407e-5 days 2.777778e-4 hours 1.653439e-6 weeks 3.805e-7 months by designated personnel,

3) either the Containment Purge System or the Auxiliary Building Ventilation System taking suction from the containment atmosphere are operating and 4) the plant is in Mode 6 with at least 23 feet of water above the reactor pressure vessel flange.

Administrative requirements are established for the responsibilities and appropriate actions of the designated personnel in the event of a

Fuel Handling Accident inside containment.

These requirements include the responsibility to be able to communicate with the control

room, to ensure that the equipment hatch is capable of being

closed, and to close the equipment hatch and personnel airlocks within 1 hour1.157407e-5 days 2.777778e-4 hours 1.653439e-6 weeks 3.805e-7 months in the event of a fuel handling, accident inside containment.

These administrative controls ensure containment closure will be established in accordance with and not to exceed the dose calculations performed using guidelines of Regulatory Guide 1.183.

SALEM -

UNIT 2 B 3/4 9-1c Amendment No.

LAR S08-01 LR-N08-0046 PSEG Calculation S-C-SF-MDC-1810, Revision 8 Decay Heat-up Rates and Curves

USER RESPONSIBLE FOR VERIFYING REVISION, STATUS AND CHANGES PRINTED 20080206 CC-AA-309-1001 Revision 3 ATTACHMENT I Design Analysis Major Revision Cover Sheet Design Analysis (Major Revision)

Analysis No.:

S.C.-SI-MDC-1810

Title:

3 Decay Heat-up Rates and Curves EC/ECR No.: 4 80095072 Station(s); "7 Salem Unit No.:

U Li.*ts 1 &2 Discipline:.

Mechanical Deserip. Code/Keyword, to Safety/QA Classi; Safety Related System Cadet 1-Spent Fuel (SF)

Structure: U Last Page No. 6 page 2 of 2 Revislon B

8 Revision: 0 0

1 Component(s). 14 CONTROLLED DOCUMENT REFERENCES z Document No.:

FromnTo Document No.:

From/To S-C-SFýMCS-O1 13 From S-C-SP-MEE-1302 From SC.OM-AP.ZZ-O00 I From S-C-SF-MDC-,1780 From S-2-FHV-MDC-0705 From Is this Design Analysis Safeguards Information? 16 Yes El No I If yes, see SY-AA-l1l -106 Does this Design Analysis contain Unverified Assumptions? 17 Yes El No 0 If yes, ATI/AR#

This Design Analysis SUPERCEDES: Is NONE in Its entirety.

Description of Revision (list affected pages far partials): t'-

The purpose of this calculation is to provide heat-up times, temperatures and curves for the Salem Generating Station (SUS) Unit I Spent Fuel Pool (SFP) for Refueling Outage 1 RI9, as directed by Reference 4.7, Tho calculation is being revised to analyze tho spent fuel pool temperature as a result of additional spent fuel transferred to the pool during the upcoming refueling outage.

Affected pages - See page revision index on page 2, Preparer. 20 Randy Smith / MLEA Sign Name 2/4/2008 Date Print Name Method of Review: --

Detailed Review Alternate Calculations (attache)

Testing Reviewer: 22 Nick Santoleri I MLEA

,_ ______d_

2/4/2008 Print Name Sign Name Date Review Notes: a Independent review [

Peer review El (For External Analyses Only) External Approver: 4 N/A N/A N/A looir 5 PrintName o

Sign Ni

.n**

/(

Date

.Ex,,evio.*oewer.'

4m, ia

/, /q'm W 6 1"

)4"

" A Z 4 Print Name Sign Name C

Date Independent 3i Party Review Reqd? I Yes [I No Excelon Approver: 27 Alan Johnson Print Name_'\\SIM

_Name Print Natae.....

".tlatae Date

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R. Smith 2/4/08 N. Santolerl 2/4/08 REVISION HISTORY Revision Issue Date Revision Description 0

4/6/99 Initial Issue Provides heat-up times for Unit I as of 12/31/99 to support heat exchanger 1

1/20/00 service and valve repairs to provide realistic heat-up times based on current conditions.

2 3/2/01 Analyzes SFP temperature as a result of additional fuel transferred during upcoming refueling outatge.

3 916/02 Analyzes the spent fuel pool temperature as a result of additional spent fuel transferred to the pool during the upcoming refueling outage.

Provides additional heat-up curves for CC temperatures of 70F and 75F to better 4

1012/02 represent expected CC temperatures during 1 R15. This supports LCR S02-03 to revise minimum time from shutdown from 168 hours0.00194 days 0.0467 hours 2.777778e-4 weeks 6.3924e-5 months to 100 hours0.00116 days 0.0278 hours 1.653439e-4 weeks 3.805e-5 months for fuel movement.

5 2/24/04 Analyzes the spent fuel pool temperature as a result of additional spent fuel transferred to the pool during the upcoming refueling outage.

69/8105 Analyzes the spent fuel pool temperature as a result of additional spent fuel transferred to the pool during the upcoming refueling outage.

7 3Analyzes the spent fuel pool temperature as a result of additional spent fuel 3//0 transferred to the pool during upcoming refueling outage I RI8.

Analyzes the spent fuel pool temperature as a result of additional spent fuel 8

See Cover transferred-to thepool during upcoming refueling outage 1-R1 9 and-evaluates.

Page reduction of shutdown time prior to fuel movement from 100 hours0.00116 days 0.0278 hours 1.653439e-4 weeks 3.805e-5 months to 80 hours9.259259e-4 days 0.0222 hours 1.322751e-4 weeks 3.044e-5 months .

PAGE REVISION INDEX PAGE REV PAGE REV PAGE REV PAGE REV 1

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8 8

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R. Smith 2/4/08 N. Santolerl 2/4/08 TABLE OF CONTENTS REVISION HISTORY......................................................

2 PAGE REVISION INDEX 2

TABLE OF CONTENTS

.................................................... 3 1.0 PURPOSE

........................................................ 4 2.0 SCOPE..........

4 3.0 ASSUMPTIONS / INPUTS / CONDITIONS

................................... 4

4.0 REFERENCES

5 5,0 ANALYSIS........................................................

6 5.1 M etho do lo gy..........................................................................................................................

6 5.2 Discussion

.6 5.3 SFP Inventory Data Files...................................................................................................

7 5.4 SFP Water Volume

................................................. 8 5.6 Discussion of Input Data File "Rfile".

..................................... 8 5.6 Param eters Inputted at Run Time.....................................................................................

9 5.7 Run the CROSSTIE Program..........................................................................................

10 5.8 Im port the O utput Files....................................................................................................

10 6.0 C O N C LU S IO N S........................................................................................................................

10 7.0 IMPACT TO STATION PROCEDURES:.............................................................................

13 8.0 DOCUM ENTS AFFECTED:................................................................................................

13 9.0 D ES IG N M A R G IN :...................................................................................................................

13 10.0 CROSS REFEREN CE S:.....................................................................................................

13 APPENDIX A - Sensitivity Analysis for Worst Case Heat Load ATTACHMENT 1 - Salem 1R19 Executive Summary Schedule ATTACHMENT 2 - Nuclear Fuels Letter NF 07-058 "Salem 1 Refueling 19 Assembly Bumup Data for SFP Heat Load Analyses", Rev 0 ATTACHMENT 3 - SFP Heat-up Curves ATTACHMENT 4 - CROSSTIE Input and Output Files (Electronic files on CD)

ATTACHMENT 5 - Salem Verification of Decay Heat Removal for Core Off-load

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R. Smith 12/4/08 1N. Santoleri 2/4/08 1-_

1.0 PURPOSE The purpose of this calculation is to provide heat-up times, temperatures and curves for the Salem Generating Station (SGS) Unit I Spent Fuel Pool (SFP) for Refueling Outage 1 RI 9, as directed by Reference 4.7.

2.0 SCOPE This calculation is being performed for the SFP Cooling System for SGS Unit 1. The following cases are analyzed:

Case 1: Cross-connect Operation with one SFHX unavailable Case 2: Normal SFP cooling with no cross-connect operation Case 3: Loss of SFP cooling In Unit 1 Case 4: Loss of SFP cooling in Unit I - Post Outage 3.0 ASSUMPTIONS / INPUTS I CONDITIONS 3.1 The computer program CROSSTIE is used in this analysis to predict the SFP temperatures (when the unit 1 or unit 2 heat exchanger is out of service) and to evaluate the SFP heat-up rates and equilibrium temperatures without forced cooling. The CROSSTIE program is cdtical software as defined by ND.DE-AP.ZZ-0052(Q), designated CROSSTIE, Reference 4.1. Validation of the software against plant data is included in Appendix A of Reference 4.8.

3.2 The CROSSTIE program does not use the first two digits of the year to specify the date (i.e., 1980 uses "80", 1995 uses "95"). In order to manipulate the program such that a "delta time" from initial spent fuel discharges through 1 R1 9 discharge could be determined, year 2000 is represented as year "100", and years following are represented sequentially from "100". This format for the year is

.changed in the input files,-and".dcy" files that document spent fuel discharged to the SFP; 3.3 The Fuel Handling Building (FHB) ambient temperature design value is 1051F (Reference 4.4, Section 9.4.3.1); however, the calculation conservatively uses 1 100F for all cases. This assumption bounds the conclusion of Reference 4.6, Attachment 8, which determines that the FHB ambient temperature is less than the design value of 1050F with the SFP temperature at its design limit of 180°F. The FHB humidity Is assumed to be the design value of 100% (Reference 4.4, Section 9.4.3.1). To maintain the ambient design temperature with the SFP temperature over 1503F, the Fuel Handling Ventilation (FHV) system must be operating. The basis for this assumption was analyzed in Reference 4.6, Attachment 8. See Appendix B of Reference 4.8 for a sensitivity study comparing FHB ambient temperatures of 1050F, 110°F, and 1201F.

3.4 Cases are run with various Component Cooling (CC) supply temperatures. The "projected" cases are run with CC temperatures of 700F, 75°F, and 800F, representing the projected temperature range during the outage time period. Also, "procedural limit" cases will be run at 990F, representing the maximum temperature allowed by System Operating Procedure SI.OP-SO.CC-0002. This provides an upper bound useful for interpolation of the maximum allowable CC temperature for the outage.

3.5 The net water volume (55536 ft3) Includes the SFP and transfer pool volumes (minus the volume displaced by the fuel assemblies and racks, see Section 5.4). The volume is assumed to be that for Unit 1, since this Is the Unit of concern (CROSSTIE automatically applies the water volume to both unit pools). The net volume is valid for Refueling Outage (1 RI 9) only, as the volume displaced by the fuel assemblies Is dependent on the total number of fuel assemblies in the SFP. The post outage volume will be slightly greater due to the two-thirds core fuel assemblies being reloaded Into the vessel. However, the difference would have a minor impact on the heat-up rate, and will conservatively be Ignored.

3.6 The surface area of the SFP water volume includes the surface area of the transfer pool In the program code.

3.7 Service Water (SW) temperatures determine CC temperatures. For conservatism, the difference between the CC and SW temperatures is assumed to be 90F as determined in Reference 4.10, which is based on a higher SFP heat load, It Is also assumed that only one CCHX is available, and

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R. Smith 2/4/08 N. Santoleri 2/4/08 there is no parallel SFHX operation (i.e., the Unit 1 SFP is not aligned to both SFHXs). Therefore, the corresponding SW temperatures for CC temperatures of 70 0F, 75OF, 80°F and 990F, are 61°F, 660F, 71OF and 90°F respectively.

3.8 The Technical Specification states that fuel cannot be moved for a minimum of 100 hours0.00116 days 0.0278 hours 1.653439e-4 weeks 3.805e-5 months from October 15 through May 15 through the year 2010 (Reference 4.3 Section 3.9.3). This calculation evaluates and supports a proposed change to the Technical Specifications to allow fuel movement after 80 hours9.259259e-4 days 0.0222 hours 1.322751e-4 weeks 3.044e-5 months of shutdown, which is sooner than the 100-hr Technical Specification limit. The analysis for 1 R1 9 based on 80 hours9.259259e-4 days 0.0222 hours 1.322751e-4 weeks 3.044e-5 months bounds the current scheduled offload start time of 86 hours9.953704e-4 days 0.0239 hours 1.421958e-4 weeks 3.2723e-5 months .

Appendix A provides a sensitivity analysis for worst case SFP decay heat load with a full capacity SFP.

3.9 The duration of core offload is assumed to be the current schedule of 41 hours4.74537e-4 days 0.0114 hours 6.779101e-5 weeks 1.56005e-5 months .

3.10 The core reload is scheduled to start 264 hours0.00306 days 0.0733 hours 4.365079e-4 weeks 1.00452e-4 months after the offload is completed, and be completed in 45 hours5.208333e-4 days 0.0125 hours 7.440476e-5 weeks 1.71225e-5 months (see Attachment 1).

3.11 Cross-connect operation, with one SFHX unavailable, is assumed to begin Immediately after core offload is completed. A nominal minimum value of 1 hour1.157407e-5 days 2.777778e-4 hours 1.653439e-6 weeks 3.805e-7 months after core offload is used for the analysis.

3.12 The CC flow to the SF Heat Exchanger (SFHX) Is assumed to be the design value of 3000 gpm (Reference 4.9).

3.13 The SF flow to the SFHX is assumed to be 2500 gpm (Reference 4.2).

3.14 The current scheduled reactor shutdown date and time for Unit I is 10/14/2008 at 2000 hours0.0231 days 0.556 hours 0.00331 weeks 7.61e-4 months (Attachment 1). However, for the CROSSTIE "Rfile" Reactor shutdown date as stated in Section 5.5 Line 2, a shutdown date and time of 10/1512008 at 0000 hours0 days 0 hours 0 weeks 0 months will be used.

3.15 The current SFP inventory (i.e., pre-offload) for Units I and 2 are contained in files "UnitI.dcy" and "Unit2.doy" for-Units 1 and 2, respectively.

3.16 The Unit I core parameters - fuel assembly burnups and average assembly Ura niumr weight-are included In Attachment 2.

3.17 The SFP cooling system will maintain pool temperatures at or below 149°F provided one SFP heat exchanger is available for each pool, and at 180OF if only one HX is available between both pools (Reference 4.4, Section 9.1.3.2). These design base limits are used as acceptance criteria in the model. Per Operations, cross-connect manipulations take less than one hour to complete. To ensure that Operators have sufficient time to manipulate cross-connect valves and maintain SFP temperatures less than 180°F, the cross-connect temperature limit will be set at a lower value based on the heat-up rate in the Unit 1 SFP.

3.18 There are -1221 fuel assemblies in the Unit 1 SFP prior to the start of the 1R19 outage (per Salem Reactor Engineering and Reference 4.12), and Is used to calculate the SFP net water volume after core offload. The actual number may vary slightly, but would not impact the calculation results.

3.19 The Core Rated Thermal Power has been increased from 3411 MWt to 3459 MWt (due to the 1.4%

power uprate) per Reference 4.3 Section 1.25. This new value applies to all fuel assemblies transferred to the SFP after June 2001 (for both Units 1 & 2), and Is conservative for the fuel assemblies that have been radiated at both power levels.

4.0 REFERENCES

4.1 Critical Software, S-C-SF-MCS-01 13, "CROSSTIE" A. Sheet 1, Critical Software Document, Revision 1 B. Sheet 1, Software Media, Revision 0 4.2 Calculation S-C-SF-MDC-1 780, "Capability Of Salem Spent Fuel Pool Heat Exchanger To Maintain 149°F Pool Temperature", Revision 0 4.3 Salem Technical Specifications 4.4 Salem Updated Final Safety Analysis Report (UFSAR) 4.5 Vendor Document, 316748, "Pool Layout - (Region I & II) for Spent Fuel Pool Storage Racks, Revision 1 4.6 Calculation S-1-FHV-MDC-0705, "FHV Sys Htg/Clg Load & Airflow Determination Calcs Unit 1",

Revision 5

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R. Smith 214108 N. Santoleri 2/4/08 4.7 Administrative Procedure SC.OM-AP.ZZ-0001(Q), "Shutdown Safety Management Program -

Salem Annex ", Revision I 4.8 Calculation S-C-SF-MDC-1 800, "Decay Heat-up Rates and Curves", Revision 6.

4.9 Westinghouse's Letter PSE-89-744 (11/8/89) to M. F. Metcalf (PSE&G), "Salem CCW Calculation Summaries" 4.10 Calculation SC-SF-MEE-1679, "Spent Fuel Pool Cooling System Capability with Core Offload Starting 100 Hours After Shutdown", Revision I 4.11 Exelon Procedure OU-AA-103, "Shutdown Safety Management Program", Revision 9 4,12 Nuclear Fuels Calculation DN2.6-0018, "Salem 2 Scoping Study", dated 5/17/2005 4.13 Procedure SI.OP-AB.SF-0001, "Loss of Spent Fuel Pool Cooling", Revision 14 4.14 Calculation S-C-SF-MDC-1240, SFP Thermal - Hydraulic Calculation, Revision 1 5.0 ANALYSIS 5.1 Methodology The purpose of this calculation is to provide heat-up times, temperatures and curves for the SGS Unit 1 Refueling Outage #19 (1R19). This analysis Is required by Reference 4.7. The calculation is performed using Holtec's computer program CROSSTIE (Reference 4.1), for the following cases:

Case 1: Cross-connect Operation with one SFHX unavailable (see note 1)

Case 2: Normal SFP cooling with no cross-connect operation Case 3: Loss of SFP cooling in Unit I (see note 2)

Case 4: Loss of SFP cooling in Unit I - Post-Outage Note I Removing a SFHX from service Is a manual action, and would never be scheduled during or following a core offload. The one exception where this could occur is a tube leak in one of the SFHXs, and would be a management decision. This case predicts when swapovers between the two pools would be required with one available SFHX, if required, to maintain the design limit of 1 800F for this potential but unlikely condition.

Note 2 Due to the upgrade of the SF Cooling Systems, a loss of cooling due to a seismic event no longer needs to be postulated, However, for a shutdown condition, a loss of cooling is postulated as follows. During shutdown modes 5 and 6, an EDG can be removed from service with no LCO (Reference 4.3 Section 3.8.1.12). As such, on a loss of offsIte power (LOOP), another single failure needs to be considered. If the EDG out of service powers one of the SF pumps, and single failure occurs on the other pump, or its EDG, no SF pumps would remain. The loss of cooling case, then, is performed. The results would show how long operators would have to re-establish forced cooling prior to the pool reaching the design limit of 1800F.

The following methodology was used to evaluate each of the above cases:

APPROACH Step 1:

Provide discussion of CROSSTIE program, and Its application for each case Step 2: Establish SFP inventory data files.

Step 3:

Determine SFP net water volume.

Step 4: Establish the outage related input data file.

Step 5: Determine the remaining parameters Inputted at run time Step 6:

Run the program, Step 7:

Import the Output file plot.dat into EXCEL and generate the heat-up curves.

5.2 Discussion

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R. Smith 2/4/08 N. Santoleri 2/4/08 I

The CROSSTIE code was designed to model cross-connect operation during an outage with one SFHX unavailable. To model other scenarios, the input data files need to be manipulated to get the meaningful results, The program requires the following inputs:

(1) Pre-offload SFP inventory burnup data for both units. This is contained in files "Unit1.dcy" and "Unit2.dcy" for Units I and 2, respectively. The program requires these specific file names to be used.

(2) A user defined input data file (called Rfile) related to the specific outage. This includes the core offload burnup data, outage start date, unit In the outage, core offload start time and duration, amongst other Inputs. This is to allow the transient condition of offloading a "hot" core to be modeled.

(3) Miscellaneous data inputted at run time, including the time to start cross-connect operation and the pool temperature limit.

The CROSSTIE program automatically starts offloading the core to the unit specified at the time specified. At the time specified to start cross-connect operation, the program automatically isolates cooling to the unit not in the outage first (since this has the lower heat load), and then swaps cooling to the isolated pool when It reaches the specified pool temperature limit. The cycle continues between the two pools until the specified end time is reached.

To model the first case, cross-connect operation, the inputs and program execution are straightforward. To model the second case, normal cooling with no cross-connect operation, the Atim-e to start cross-connect Is simply set to the end time for the model run or greater, such that no swapping takes place.

To model the third and fourth cases, loss of cooling, the opposite unit must be specified as the outage unit, since the program automatically Isolates cooling to the unit not in the outage first at the specified time to start the cross-connect. Thus, to model a loss of cooling for the Unit I SFP, Unit 2 needs to be specified, However, CROSSTIE also automatically applies the core offload data in "Rfile" to the unit specified as being in the outage, Thus, it would add the Unit 1 core to the Unit 2 pool, while Isolating cooling to the Unit I pool, As such, the core offload data in "Rfile" is set to 0, and Is Included in the SFP inventory data file "Unit1.dcy. The reactor shutdown date then has to be changed to coincide with the offload completion date, since the analysis. starts with a full "hot" core offload already in the pool. Time t = 0 in this case, then, corresponds to the time offload is completed for both the cross-connect and normal cooling cases. The initial pool temperature starting with the full core in the pool will tend to be slightly higher than the temperature corresponding to offload complete time for the cross-connect and normal cooling cases. This temperature difference Is minor and is conservative in nature; therefore, the analysis will be used as-is.

5.3 SFP Inventory Data Files The current (pre-offload) SFP Inventory burnup data is contained within data files "Unit1.dcy" and "Unit2.dcy" for Units 1 and 2, respectively. These files currently contain the inventory up through Cycle 18 for Unit I and the projected inventory up through Cycle 16 for Unit 2 (for Cases I & 2), as developed in Reference 4.8. The Cycle 19 offload burnup data was provided by Fuels per, including both the full core offload and the fuel to remain in the pool after core reload.

For the first two cases, cross-connect operation and normal cooling, "Unitl.dcy" remains as Is with the core offload data inputted under a separate input file discussed in Section 5.5 below. The current "Unit1.dcy" file Is included as an electronic file on CD In Attachment 4 (listed as "unitl-18.dcy"),

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DATE VE R. Sith2/4/08 N, Santolerl 2//0 For the loss of cooling case (Case 3), as discussed in Section 5.2, "Unitl.dcy" Is updated to include the Unit I full core offload for Cycle 19. The core offload data included with "Rfile", as discussed in Section 5.5, Is inputted as three batches, with an average burnup for each batch. The updated "Uniti.dcy" file includes these three batches, as shown in Attachment 4 (included as an electronic file on CD, and listed as "unitl-19FC.dcy").

For the loss of cooling case post-outage (Case 4), "Unitl.dcy" is updated to reflect the fuel assemblies permanently discharged to the SFP after the Reactor vessel is reloaded for Cycle 20.

The updated "Unitl.dcy" file is shown in Attachment 4 (included as an electronic file on CD, and listed as "unit1-19POdcy").

5.4 SFP Water Volume The net water volume includes the SFP and transfer pool volumes at an elevation of 23 feet above the fuel assemblies, minus the volume displaced by the fuel assemblies and racks, The volume is calculated based on the methodology from Section 3.1 of Reference 4.1A.

Total volume of SFP and transfer pool at 23 feet above the fuel assemblies: 62148 ft3 Rack volume: 564 ft3 Volume/fuel assembly: 4.277 ft3

fuel assemblies in Unit I pool after core offload: 1221 (Assumption 3,18) + 193 1414 assemblies Total fuel assembly volume: 1414 4 4.277 = 6048 ft3 Net-water volume: 62148-564 - 6048 = 55536 ft3 (net SFP volume) 5.5 Discussion of Input Data File "Rfile" This file contains seven lines of input. The following provides a breakdown of the Input Data File:

Line 1: Description of job (freeform comments).

Line 2: Reactor shutdown date.

" Cases 1 & 2:10/1512008 (Actual shutdown @ 20:00 on 10/14/2008; rounded forward to next day)

" Case 3:10/20/2008 (From current schedule - 80 hours9.259259e-4 days 0.0222 hours 1.322751e-4 weeks 3.044e-5 months + 41 hours4.74537e-4 days 0.0114 hours 6.779101e-5 weeks 1.56005e-5 months , and rounded forward to the next day). Note: Reactor shutdown date in uniti.dcy also rounded forward to the next day (10/20/2008)

" Case 4:11/1/08, 12/1/08, 1/1/09, 2/1/09 then quarterly beginning on 3/1/09 through 3/1/10 Note: For loss of cooling (case 3), this is actually the start date of CROSSTIE run. The total elapsed time is 5 days and 1 hour1.157407e-5 days 2.777778e-4 hours 1.653439e-6 weeks 3.805e-7 months ; however, CROSSTIE cannot model time into the shutdown date, and an elapsed time of 5 days will conservatively be used.

Line 3: Unit in outage.

" Cases1&2: 1

" Cases 3&4: 2 Line 4: CC flow, SF flow, SFP water volume: 3000, 2500, 55536 (Design values for flows used -- see Assumptions 3,12 & 3.13)

Line 5: batch 1 # assemblies, batch 2 # assemblies, batch 3 # assemblies, decay time before fuel transfer, total transfer time for offload

(NC,DE-AP.ZZ-0002(Q), Rev. 12, Form 2)

CALCULATION CONTINUATION SHEET SHEET:

9 of 13 CALC NO.:

S-C-SF-MDC-1810 REV:

8 REF:

CONT'D ON SHEET:

ORIGINATOR:

DATE:

REVIEWER:

DATE:

VERIFIER:

DATE:

R. Smith 2/4/08 N. Santoleri 2/4/08 The first 3 numbers are the number of assemblies in each discharge batch. The average burnup for each batch Is included in the next line. CROSSTIE has the core unload In 3 batches, with the assumption that about 113 has a one-cycle burnup, 1/3 has a two-cycle burnup and the remaining 1/3 has a three-cycle burnup (the 1/3 that will remain in the pool after reload). The 1R19 core doesn't quite fit those percentages. The batches are grouped based on the bumups rather than In 3 equal groups. From Attachment 2, the number of assemblies per batch, in descending order of fuel burnup, are taken as: 32, 77, 84.

The 4th number Is the decay time before fuel transfer. From Section 3.8, the value is 80 hours9.259259e-4 days 0.0222 hours 1.322751e-4 weeks 3.044e-5 months . The 5th value is the total transfer time for the offload. From Section 3.9, the current schedule is 41 hours4.74537e-4 days 0.0114 hours 6.779101e-5 weeks 1.56005e-5 months .

The input lines, then, are:

" Cases 1 & 2: 32,77,84,80,41

" Cases 3 & 4: 0,0,0,0,0 Line 6 1st #: Reactor rated power: 3459 MWt 2 "d #: Capacity Factor for last 4 months - assume 1.0 for conservatism 3 rd, 4 h & 5th #'s: Average burnups for 3 batches in line 4 - 51662, 46398, 25639 (based on )

6 th #: Average uranium weight - 455.8 (Attachment 2)

The Input lines, then, are:

" Cases I& 2: 3459,1.0, 51662,46398,25639,455.8

" Cases 3 & 4: 3459,0,0,0,0,0 Line 7 Ambient air temperature (110°F) and RH (100%) in Fuel Handling Building (FHB). From Section 3.3:

Cases 1,2,3&4:110,1.0 These files are included as electronic files on CD In Attachment 4, and are saved as the following input files:

Cases I & 2: 1 R1 9clg.dat

" Cases 3 & 4: 1Rl91oc.dat (case 4 will adjust the shutdown date in line 2) 5.6 Parameters Inputted at Run Time Input 1: Rflle (*.dat) from Section 5.5 Input 2: Time after shutdown to start CROSSTIE (hrs):

Case 1:122(80+41+1)

(Assumptions 3.8, 3.9, 3.11)

Case 2: 500 (Section 5.2 - bounds core reload)

Cases 3 & 4:1 (Section 5.2 and Assumption 3.11)

Input 3: Pool water temperature limit for swltchover:

Case 1:170 &180 (limit with one SF heat exchanger unavailable - 2nd run at 170°F includes 1 0°F margin for Cross-connect valve manipulations)

" Cases 2, 3 & 4: 210 (high enough to prevent swapover)

Input 4: CCW coolant temperature (Assumption 3.4):

" Cases 1, 2 & 3: 70, 75, 80 and 99

" Case 4:80 & 99 Input 5: Ending time for integration:

Cases 1 & 2: 500 (bounds core reload)

(NC.DE-AP.ZZ-0002(Q), Rev. 12, Form 2)

CALCULATION CONTINUATION SHEET SHEET:

10 of 13 CALC NO,:

S-C-SF-MDC-1810 REV:

8 REF:

CONTrD ON SHEET:

ORIGINATOR:

DATE:

REVIEWER:

DATE.

VERIFIER:

DATE:

R. SmIth 2/4/08 N. Santoleri 2/4/08 Cases 3 & 4: 250 - 500 (high enough to reach equilibrium temperature for Unit I pool or boiling) - determined by trial and error These inputs are also shown on the "result.tem" files on CD in Attachment 4.

5.7 Run the CROSSTIE Program The model was run for the following scenarios. The output files "result.tem", "unit1.htl" and "plotdat" are Included as electronic files on CD in Attachment 4.

Case 1: Cross-connect Operation with one SFHX unavailable Case 2: Normal SFP cooling with no cross-connect operation Case 3: Loss of SFP cooling in Unit 1 Case 4: Loss of SFP cooling in Unit 1 - Post-outage 5.8 Import the Output Files The PLOT.DAT file for each unit was imported into EXCEL, and a temperature vs. time graph was plotted for each unit. The graphs can be found in the Attachment 3.

6.0 CONCLUSION

S

-The Unit 1 SFP analysis for 1 RI 9-was performed-for the following cases, with one heat exchanger available and Cross-connect swapover at 1 80°F (design bases limit) and at 1 70OF (to assure that Operators have sufficient time to manipulate Cross-connect valves and maintain SFP temperatures less than 1800F):

Case 1: Cross-connect Operation with one SFHX unavailable Case 2: Normal SFP cooling with no cross-connect operation Case 3: Loss of SFP cooling in Unit 1 Case 4: Loss of SFP cooling in Unit 1 - Post-outage Cases 1, 2 and 3 were run with CC supply temperatures of 70°F, 75°F, 80°F and 990F, which

.correlate to maximum SW temperatures of 61 OF, 66°F, 71°F and 90°F respectively. Case 4 was run with CC supply temperatures of 800F1 and 990F. Plots showing "SFP temperature vs time" for each case are included in Attachment 3. A summary of the results is as follows:

Case 1: A first run was performed for cross-connect operation at the licensing basis limit of 1880°F.

For cross-connect operation with the Unit I SFP aligned to the Unit 2 SFHX, the isolated Unit 2 peak SFP temperature is shown to reach the licensing basis limit of 180°F with one SFHX isolated, and thus swapping of SFPs is required. A summary of the results Is included in the table below. The worst-case scenario would be a loss of one SFP HX just upon completing a full core offload. The results show that, after the initial swap of cooling to the non-outage (Unit 2) SFP, Operators would have a minimum of 5.8 hrs at CC supply temperature of 800F before Cross-connect valve manipulations would be required to swap cooling back to the recently offloaded SFP, and 79.1 hours1.157407e-5 days 2.777778e-4 hours 1.653439e-6 weeks 3.805e-7 months at CC supply temperature of 80°F to swap cooling back to the non-outage SFP. Additional Cross-connect manipulations (if required) are bounded by these initial times.

'The CC temperature cases of 75OF and 70OF were not performed for Case 4 since these are post-outage cases with low heat load conditions; also, the CC temperature would likely be set between 80*F and 990F.

(NC.DE-AP.ZZ-0002(Q), Rev. 12, Form 2)

CALCULATION CONTINUATION SHEET SHEET:

11 of 13 CALC NO.:

S-C-SF-MDC-1810 REV:

8 REF:

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ORIGINATOR:

DATE:

REVIEWER:

DATE:

VERIFIER:

DATE:

R. Smith 2/4108 N. Santoleri 2/4/08 Minimum time between swapover (hrs)

Case

. Sequence 99*

86**

80 75 70 1

From U2 to U1 3.5 5.0 5.8 6.4 7.0 1

From Ul to U2 60.8 73.5 79.1 83.6 88.2 Loss of cooling heatup rate (OF /hr)

" 8.4 8.4 8.4 8.4 8.4

Not a valid pre-existing condition due to Unit 1 SFP temperature of 1600F > 149"F design bases limit

- 865F Is the maximum CC temperature that ensures SFP temperature < 149°F design bases limit Operations may choose to swap cooling between the two pools prior to 180°F as a precautionary measure. A second run was performed for cross-connect operation at 170'F, based on a Unit I SFP heatup rate of 9.1 *F/hour and a one-hour duration to complete cross-connect manipulations (Assumption 3.17). A summary of the results is included in the table below. The worst-case scenario would be a loss of one SFP HX just upon completing a full core offload. The results show that, after the initial swap of cooling to the non-outage (Unit 2) SFP, Operators would have a minimum of 3.8 hrs at CC supply temperature of 80°F before Cross-connect valve manipulations would be required to swap cooling back to the recently offloaded SFP, and 37.9 hours1.041667e-4 days 0.0025 hours 1.488095e-5 weeks 3.4245e-6 months at CC supply temperature of 80OF to swap cooling back to the non-outage SFP. Additional Cross-connect manipulations (If required) are bounded by these initial times.

Minimum time between swapover hrs)

Case Sequence 99*

86**

80 75 70 1.FromU2 to U!

1.7 3.1 3.8 4.4 5.0 1

From Ul to U2 19.2 32.2 37.9 42.5-.

47.4 Loss of cooling heatup rate (OF/hr) 9.2 9.1 9.1 9.1 9.0

Not a Valid pre-existing condition due to Unit I SFP temperature of 160OF > 1491F design bases limit

- 86OF is the maximum CC temperature that ensures SFP temperature < 149OF design bases limit Case 2: A summary of the results is included in the table below. For normal SFP cooling, the licensing basis limit of 1497F is exceeded for the "procedural limit" case with 99°F CC temperature, and an 80-hr offload start would not be permitted. However, interpolating between the 800F CC temperature case and 99OF CC temperature case results, a peak SFP temperature of 148.30F Is reached with a CC temperature of 860F, correlating to a maximum SW temperature of 77 0F. Since this will be higher than the SW temperature at the time of core offload, the SFP temperature limit of 149"F will not be exceeded.

Case Offload Peak Temp (OF)

Time to I12511F (hr)*

Heatup (OF/hr)

Start (hr) 99 180 j75 170 99 I80 7 5 170 99 80 ]75 1-70 2

80 160 143 138 133 17.8 30.9 34.51 38.3 1.2 1.3_1 1.3 1.3

Time from "start of off-load" Also, through linear interpolation of the results, the SFP high temperature alarm setpolnt of 125°F will be reached with a CC temperature of 62°F for an offload start time of 80 hours9.259259e-4 days 0.0222 hours 1.322751e-4 weeks 3.044e-5 months (see Attachment 5). Alarm Response Procedure SI.OP-AR,ZZ-0003 allows the setpoint to be increased to allow refueling activities to continue. Temporary alarm setpoints as a function of CC temperature, If required, are provided in the table below. The setpoints are set to a value 5°F higher than the calculated peak SFP temperature. This accounts for a 2.5°F instrument uncertainty (Reference SAP ICD screen for FLOC S2SF -2TIC651) plus provides a 2.5°F margin above the peak temperature.

Data points are tabulated below, and a graph included in Attachment 5 for inclusion in SC.OM-AP.ZZ-0001(Q) Attachment 5.

(NC.DE-AP.ZZ-0002(Q), Rev. 12, Form 2)

CALCULATION CONTINUATION SHEET SHEET:

12 of 13 CALC NO.:

S-C-SF-MDC-1 810 REV:

8 REF:

CONTOD ON SHEET:

ORIGINATOR:

DATE:

REVIEWER:

DATE:

VERIFIER:

DATE:

R. Smith 2/4108 N, Santoleri 2/4/08 CC temperature (F)

Alarm Setpoint (F) 80 148 75 143 70 138 Case 3: This case pertains to loss of both SFP pumps such that cross-connect operation Is not available. A summary of the results is included in the table below. On a loss of cooling to the Unit 1 SFP, the maximum design limit of 180°F will be reached in a range of 1,2 hours2.314815e-5 days 5.555556e-4 hours 3.306878e-6 weeks 7.61e-7 months to 3.3 hours3.472222e-5 days 8.333333e-4 hours 4.960317e-6 weeks 1.1415e-6 months after core offload is complete. This is the time operators have to complete contingency actions to re-establish forced cooling. The heat-up rate for Unit 1 is within a range of 11.1 F/hr to 11.6°F/hr. The Unit I SFP will not boil if cooling is not restored.

Case Offload Peak Temp (IF)

Time to reach 180°F Heatup (IF/hr)

......._(hr)*

Start 99 80 75 70 99 80 75 70 99 80 75 70 (hr) 3 80 206 206 206 206 1.2 2.6 2.9 3.3 11.1 11.4 11.6 11.6

Time from "loss of cooling" after core offload complete Case 4: Heat-up rates In the event of Unit 1 SFP loss of cooling post-outage, See curves for details.

Maximum CC/SW Temperatures for Proposed 80-hr Limiting Core Offload The maximum river temperature is based on a maximum CC supply temperature, to ensure that the SFP will not exceed the licensing basis limit of 1490F. For the proposed revisedTech Spec.

minimum offload start of 80 hours9.259259e-4 days 0.0222 hours 1.322751e-4 weeks 3.044e-5 months , the maximum SW and CC temperatures are 77°F and 860F, respectively (see Attachment 5).

Worst Case SFP Decay Heat Load II Appendix A demonstrates that additional background decay heat from discharges beyond 1 RI19 until the SFP reaches full capacity will have little impact on SFP temperatures and heatup rates with the core offload start time at 80 hours9.259259e-4 days 0.0222 hours 1.322751e-4 weeks 3.044e-5 months after shutdown. Thus the proposed change in the core offload start time to 80 hours9.259259e-4 days 0.0222 hours 1.322751e-4 weeks 3.044e-5 months is justified.

(NC.DE-AP.ZZ-0002(Q), Rev, 12, Form 2)

CALCULATION CONTINUATION SHEET SHEET!

13 of 13 CALC NO,:

S-C-SF-MDC-1 810 REV:

8 REP:

CONTI2 ON SHEETL ORIGINATOR:

DATE:

REVIEWER:

DATE:

VERIFIER:

I DATE:

R. Smith 2/4/08 N. Santolerl 2/4/08 7.0 IMPACT TO STATION PROCEDURES:

None 8.0 DOCUMENTS AFFECTED:

None 9.0 DESIGN MARGIN:

This calculation is used to determine heat-up rates for the SFP during refueling outages. It provides a planning tool for the Central Outage Group (COG) and Operations to plan fuel moves to ensure SFP temperature are manageable, and allows contingency planning in the event that a pump and/or heat exchanger is lost. Design margin is not applicable to this calculation.

10.0 CROSS

REFERENCES:

Cross-References - Critical Software S-C-SF-MCS-01 13 and Design Calculations S-C-SF-MDC-1780 and S-2-FHV-MDC-0706 were used as Input for development of this calculation, There are no output documents resulting from this calculation,

S-C-SF-MDC-1810 / R8 APPENDIX A Sensitivity Analysis for Worst Case Heat Load DISCUSSION The main calculation was performed for the IRI 9 refueling outage, which is the upcoming outage as of Revision 8 to this calculation. It provides the methodology for determining SFP heatup rates and temperatures for future outages. Revision 8 is based on a proposed change to minimum core offload start time after shutdown of 80 hours9.259259e-4 days 0.0222 hours 1.322751e-4 weeks 3.044e-5 months for implementation in IRI 9. This appendix provides a. sensitivity analysis for the estimated worst-case decay heat load, based on a core offload start time of 80 hours9.259259e-4 days 0.0222 hours 1.322751e-4 weeks 3.044e-5 months . The worst-case decay heat load will occur when the SFP reaches full capacity following a full core offload. This analysis will be performed for each unit.

During a refueling outage, the SFP decay heat load is dominated by the "fresh" full core offload.

The background decay heat load from the pre-existing assemblies is a small percentage of the total heat load. For a given offload start time and discharge rate from the core into the SFP, the core decay heat will be relatively constant, Since the background decay heat is small in comparison, the increase in background decay heat from future outage discharges will have a small impact on the total SFP decay heat load during an outage, Thus it is expected that the impact on SFP heatup rates and peak temperatures for future outages, due to the increase in background decay heat, will be small as compared to those predicted for 1R19, INPUTS/ASSUMPTIONS

1. SFP Storage Capacity / Number of Remaining Outages From Reference 4.14, Section 1.0, the SFP has a total storage capacity of 1632 cells. This is for the current storage rack configuration, which includes nine Holtec high density racks and three Exxon Nuclear Corporation (ENC) racks. Reference 4.14 also provides a total storage capacity of 1776 cells based on a potential for replacing the remaining ENC racks with Holtee racks. Since there is an ongoing project to implement Dry Cask Storage at Salem, there are no plans to replace the remaining ENC racks, Thus the total storage capacity for the future is the current capacity of 1632 cells. Each SFP currently has approximately 100 unusable cells due to inaccessibility, damage or non-fuel items. For the sensitivity analysis, it is conservatively assumed that all the cells are usable, to maximize the SFP heat load.

Thus the outage in which each SFP will reach its net maximum capacity (accounting for unusable cells) will be earlier than that predicted to reach full capacity (assuming all cells are usable).

From Input 3.18, there are 1221 fuel assemblies in the Unit 1 SFP prior to 1R19. From, 72 assemblies will be permanently discharged into the Unit 1 SFP during 1R19, increasing the number of assemblies in the Unit 1 SFP after 1R19 to 1293 (1221 + 72),

The total number of open cells for discharges beyond 1R19 then is 339 (1632 - 1293). A full core consists of 193 assemblies. Thus the number of assemblies that can be discharged after iR19, to allow a full core discharge the following outage, is 146 (339 - 193). The number of Page 1 of 4

S-C-SF-MDC-1810 / R8 APPENDIX A Sensitivity Analysis for Worst Case Heat Load assemblies permanently discharged into the SFP during an outage is typically about 70-80.

Two outages beyond 1RI19 with 73 permanently discharged assemblies apiece equates to 146 assemblies. Therefore, the outage in which the Unit 1 SFP will reach full capacity, following a full core offload, will be IR22.

From Reference 4.8, Input 3.18, there are 1040 fuel assemblies in the Unit 2 SFP prior to 2R16. From Attachment 2 of Reference 4.8, 81 assemblies will be permanently discharged into the Unit 2 SFP during 2R1 6, increasing the number of assemblies in the Unit 2 SFP after 2R16 to 1121 (1040 + 81). The total number of open cells for discharges beyond 2R16 then is 511 (1632 - 1121)., A full core consists of 193 assemblies. Thus the number of assemblies that can be discharged after 2R1 6, to allow a full core discharge the following outage, is 318 (511 - 193). The number of assemblies permanently discharged into the SFP during an outage is typically about 70-80. Two outages beyond 2R16 with 79 permanently discharged assemblies apiece plus two outages with 80 permanently discharged assemblies apiece equates to 318 assemblies. Therefore, the outage in which the Unit 2 SFP will reach full capacity, following a full core offload, will be 2R2 1.

2. Background Decay Heat:

The "unitl.dcy" and "unit2.dcy" files will-be modified to include future-assemblies beyond IR1 9 and 2R1 6. The average fuel bumup for the discharged assemblies is assumed to be 50000 MWD/MTU, which bounds the average fuel bum-up from past discharges. The average uranium weight is assumed to be the same as that for the 1R19 discharge, or 455.8 kg/assembly. Reactor power is assumed to be the same. The reactor shutdown dates are assumed to be on 4/15 for spring outages and 10/15 for fall outages as follows:

Unit I Case (based on 1R22 occurring on 4/15/2013) 1R20:4/15/2010 1R21: 10/15/2011

2R17: 10/15/2009

2R18:4/15/2011 Unit 2 Case (based on 2R21 occurring on 10/15/2015) 1R20:4/15/2010 1R21: 10/15/2011 I 1R22: 4/15/2013*

1R23: 10/15/2014*

2R17: 10/15/2009

2R18:4/15/2011

2R19: 10/15/2012

2R20:4/15/2014 Page 2 of 4

S-C-SF-MDC-1810 / R8 APPENDIX A Sensitivity Analysis for Worst Case Heat Load

Assemblies will need to be removed and placed in dry cask storage to allow room for full core offload during the subsequent refueling outages. For this analysis, no assemblies will be deleted from the "uniti.dcy" file for conservatism. Furthermore, for the purposes of this analysis, the effect will be negligible, since Unit I is the non-outage unit, 3, Rfile Data All Rfile and run time data is assumed to be the same as those used for the 1R19 analysis, except for the following A. Reactor Shutdown Date (see Input 2): 1R22 - 4/15/2013 2R21 - 10/15/2015 B. SFP Volume; Total volume of SFP and transfer pool at 23 feet above the fuel assemblies: 62148 ft3 Rack volume: 564 ft3 Volume/fuel assembly: 4.277 ft3

fuel assemblies in Unit 1 pool after core offload: 1632 (Input 1)

Total-fuel assembly volume: l1632-* 4;2-77 = 6980 ft 3 Net water volume: 62148 - 564 - 6980 = 54604 ft (net SFP volume)

C. CC temperature: Sensitivity will only be performed at 80'F CC, which corresponds to a SW temperature of 71IF.

RESULTS First, the normal cooling case is run. The peak SFP temperature for Unit 1 is 143.47F. This is only 0.70F above the peak SFP temperature determined for IR19 of 142.7°F. The peak SFP temperature for Unit 2 is 143.7°F. This is only 0.3°F above the peak Unit 1 SFP temperature at full capacity. Thus the difference between the two units with the SFP at full capacity is insignificant. Based on these results, the other cases were not performed as it is evident the additional background decay heat up through 1R22 and 2R21 has minimal impact.

These results are based on a CC temperature of S07 CC, which corresponds to a SW temperature of 71 OF. This bounds the historical maximum SW temperature for the period of October 15 through May 15. Therefore, with core offload starting within the period of October 15 through May 15, a full core offload start time after shutdown of 80 hours9.259259e-4 days 0.0222 hours 1.322751e-4 weeks 3.044e-5 months is acceptable for the worst case decay heat load, and thus is acceptable for all future outages on both units.

Page 3 of 4

S-C-SF-MDC-1810/ R8 Appendix A Page 4 of 4 150 145 140 135 130 125 120 Unit I SFP Normal Cooling 0

50 100 150 200 250 300 350 400 450 Time after shutdown (hrs) 500

0 L40.

HOJ-1CE5 0*

1 40ItB33WiC 15=

n STRIP REACTO

SHI4TOW~TC inrEC I& TURBINE - COMM OOLDOWN TO MOmE 1 SPEC HOLD TO FU PLACE.K NCE OUTAGE 5 WINDOW L M0VEMET

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/2 "op C. f e 2-v -f 2

NUCLEAR FUELS TRANSMITTAL OF DESIGN INFORMATION G2 SAFETY RELATED Originating Organization NF ID#

NFS07-058

[I NON-SAFETY RELATED Z Nuclear Fuels Revision#

0

[I REGULATORY RELATED ED Other (specify)

SRRS #

Page 1 of 0 Station:

Salem Unit:

I Cycle:

19 Generic:

Subject:

Salem Unit 1 Refueling 19 Assembly Bumup Data for SFP Heat Load Analysis To:

Alan Johnson (Salem Design Engineering Manager)

FCP#: 80092683 Joe Dascanlo

_//2_1_A_

Prepared by natffre' Date Keith Robinson

~

A -343z.

Reviewed by Signature Date Tom Ross Approved by

-ignature Date Status of Information:

ED Verified El Unverified El Engineering Judgment Acti6nrTrackl ng-#forMethod and-Schedule of-Verification for Unverified DESIGN INFORMATION:

Description o : Salem I Refueling 19 Assembly Burnup Data Note the Cycle 19 Information Is based on an end-of-cycle best estimate burnup of 20820 MWD/MTU. An EXCEL spreadsheet was used In summarizing the Attachment I data which has the following name and date/time stamp:

NFS07-058_SIRIgCROSSTIE.xIs 11126f2007 2:27 PM The spreadsheet is located under (M:) Entdata on

'Njnbufp20':Shared\\DCP List (Exelon Process)\\Salem\\!009xxxx Series (Salem)\\B0092583 Salem I Cycle 19 Cycle Management, Purpose of Information:

To provide Salem Design Engineering with Inputs to be utilized In the I RI9 SFP heat load calculation using the CROSSTIE (or Equivalent),

code.

Source of Information:

I) DNI.6-0045, Salem I Cycle 20 Energy Utilization Plan (EUP),

11/212007.

2) SC.OM-AP.ZZ-0001-Rev 1, Shutdown Safety Management Program-Salem Annex.

Supplemental Distribution:

E-Ml:

Hard Co R; Down

~1r~.#~~r~TZ 7e..1 f2-

NrFS07-058 Page 2 of 6 Cycle 19 Loading = 87.973 MTU Average Assembly Loading = 0.45582 MTU/Assembly Assembly Srnu Disoharaed 1

MWD/MTU1 6ssemblies AG21 51751 X

AG22 51705 X

AG23, 51705 X

AG24 51751 X

AG25 51751 X

AG26 61705 X

AG27 51705 X

AG2B 51751 X

AG29 51161 X

AG30 52079 X

AG31 51158 X

AG32 51158 X

AG33 52041 X

AG34 51161 X

AG35 52079 X

AG36 52079 X

AG37 52079 X

AG38 52041 x

AG39 51158 X

AG40 51161 X

AG41 51158 X

AG42 52041 X

AG43 52041 X

AG44 51181 X

AG45 49316 X

AG47 52205 X

A348 62205 X

AG49 51195 X

AG50 51195 X

AG51 52205 X

AG54 49316 X

AG56 49315 X

AG56 51195 X

AG57 52205 X

AG63 51195 X

AG70 49316 X

AH0I 45921 X

AH02 46852 X

AH03 48943 X

AH04 46852 X

AHO&

48943 X

AH06 38886 AH07 47479 X

AHOB 47861 X

Rev 82 1,o.je

2.

It r,

NFS07-058 Attachment I Page 3 of 6 ID AH09 AHIO AH1 1 AH12 AH13 AH14 AH15 AH16 AHI7 AH18.

AH19 AH20 AH21 AH22 AH23 AH24 AH25 AH26 AH27 AH28 AH29 AH30 AH3I AH32 AH33 AH34 AH36 AH36 AH37 AH38 AH39 AH40 AH41 AH42 AH43 AH44 AH45 AH46 AH47 AH4B AH49 AH5D AH5I AH52 AH53 AH54 AH55 AH56 47674 46852 48102 49066 46852 48102 47851 47479 47479 47674 49066 47851 49066 47479 48102 48321 47851 48102 48321 38906 38886 38905 49066 38905 48321 48321 38886 47674 47674 48943 38905 48943 38886 44662 46148 44704 44704 46129 44662 46148 46148 46148 46129 44704 44662 44704 46129 46129 Discharmed Assmblies x

x x

X x

x x

x Q -J 8 47_fe 3

NFS07-058 Attachme~nt I Page 4 of 6 IQ AH57 AH88 AH59 AHSO AH51 AH62 AH6U AH64 AH65 AH66 AH87 AH6B AH69 AH70 AH71 AH72 AH73 AJOI AJ02 AJO3 AJ04

_AJO5 AJO6 AJO7 AJOS AJO9 AJIO AJI1 AJ12 AJ13 AJ14 AJ15 MJIB AJ17 AJI8 AJ19 AJ20 AJ21 AJ22 AJ23 AJ24 AJ25 AJ26 AJ27 AJ28 AJ29 AJ30 AJ31 44662 456886 47023 47023 45886 4B601 45886 47133 47133 47023 48601 45886 47133 47023 48601 48601 47133 2B137 28137 28137 27708 28018 27708 27708 28018 27617 28018 27054 27054 27054 27054 27617 27708 27617 28137 28018 27617 22370 22405 22405 22370 22370 22405 22405 22370 24430 25117 25117 ODa-charae Assemblies X

See Note 1 x

x See Note I See Note I i

4-

2.

1d-7e I/ v

(

]NPS07-058 Page 5 of 6 AsSemblv BurnuD 2kqhqarpd D

(MWDIMTU)

Aisemlies AJ32 24430 AJ33 25115 AJ34 24430 AJ35 25117 AJ36 25115 AJ37 25115 AJ3B 25115 AJ39 25117 AJ40 24430 AJ41 27382 AJ42 28333 AJ43 23559 AJ44 27026 AJ45 19353 AJ46 28333 AJ47 23597 AJ48 19353 AJ49.

27382 AJSO 26999 AJ51 23597 AJ52 24192 A.J53.

28357 AJ54 23597 AJ56 28333 AJ5S 23597 AJ57 28357 AJ58 27420 AJ59 19353 AJ6O 19353 AJ6I 24238 AJ62 27420 AJ63 28999 AJ64 24238 AJ65 27026 AJd5 24192 AJ87 27026 AJ68 23559 AJ69 27026 AJ70 28357 AJ71 24238 AJ72 23559 AJ,73 28367 AJ74 27382 AJ75 27420 AJ76 27420 AJ77 28333 AJ78 27382 AJ79 24238 S-C-:SF-M*bc-/s to Fe~v 8 S-&F-f~bL7-' lare7e S-o-(

NFS07-058 Attachment I Page 6 of 6 Assembly Burnun Dlacharmed I2 (MWY2D/MTU)

Assemblies AJ8O 26999 AJ8I 24192 AJ82 24192 AJ83 26999 AJ84 23559 Note 1: Assemblies AJl 1, A1l13, and AllI4 will be used in fhture cycles as center assemblies in the core.

S'-C S-5F

/bDC

- /8,- 10 B1e A rr/*cdv&eA.,r 4,qe 4 o{' C

Design Calculation S-C-SF-MDC-1 810, Rev. 8 page 1 of 21 Unit I - Cross-connect Operation Swapover at 180°F (CCW 3000 gpm, SF 2500 gpm)

Plant Shutdown (10114108) - 20:00 hr, Off!oad Start - 80 h r, Offload Complete - 121 hr 3.5 hrs to swap cooling backto Unit 1 180

_60.8 irs to swap ctooling back t Unit 2 unit1 -S Pambienta r@ 110F End

/

100% RH 99F CCW Ind 9OF SW

/

/"

160j

/

E-140 IL 11 0

a I_

120 2 SFP ambient airý1 0 aný 100% RH, 9F CCW a4d 80 1________

0 50 100 150 200 250 300 350 400 450 500 time (hrs)

Case 1 a - Procedural Limit

Design Calculation S-C-SF-MDC-1 810, Rev. 8 page 2 of 21 Unit I - Cross-connect Operation Swapover at 180°F (CCW 3000 gpm, SF 2500 gpm)

Plant Shutdown (10114108) - 20:00 hr, Offload Start - 80 hr, Offload Complete - 121 hr 5.0 hrs to swap cxofng back to lIt1 1 7

73.5 hrs to cooling ck to Unit 2 T

i ostsj"a*i ii-180 bc

/'-i

/,I

.s '

unit 1 -

FPambientair@110°F

'T and 100 RH, 86F C CW and 77F SW 160.I i

at.

f 120

I i

unit I!-SFP arný ient air @ 1' 0°F and 10(,% RH, 86F CCW and 7 IF SW 100 J

I;

jII, 80 I1 50 100 150 200

.250 300 350 400 450 500 time (hrs)

Case lb - Projected

Design Calculation S-C-SF-MDC-1810, Rev. 8 page 3 of 21 Unit I - Cross-connect Operation Swapover at 180TF (CCW 3000 gpm, SF 2500 gpm)

Plant Shutdown (10114108) - 20:00 hr, Offload Start - 80 hr, Offload Complete - 121 hr 180 160 E 140 LL 120 100 80 0

50 100 "150 200 250 300 350 400 450 time (hrs) 500 Case lc - Projected

Design Calculation S-C-SF-MDC-1810, Rev. 8 page 4 of 21 Unit I - Cross-connect Operation Swapover at 1800 F (CCW 3000 gpm, SF 2500 gpm)

Plant Shutdown (10114108) - 20:00 hr, Offload Start - 80 hr, Offload Complete - 121 hr 190 6.4 hrs to swap

,oling p

back to UnitI 83.6 hco t:) swap oolin back to

/:1

/

//

170I

/

I 150 unit 1 -

FP ambient air @110 0T

,and 1O 6 RI

/"75FC Wand66 SE 1

ISW U.I 130 -

110; J uni ent air @H, 75F C and 6 FSW 70-n A

nn 1A 9fl 250 300 30 400 450 500 time (hrs)

Case ld 4 Projected

Design Calculation S-C-SF-MDC-1810, Rev. 8 page 5 of 21 Unit I - Cross-connect Operation Swapover at 180OF (CCW 3000 gpm, SF 2500 gpm)

Plant Shutdown (10114108) - 20:00 hr, Offload Start - 80 hr, Offload Complete - 121 hr 190 TG hrs to swap Pcholing back to Unit 1 88.2 hrs t swap cooliHn back to Unit 2

/!

170 I

/i 150_

unit I-FP ambient air @110°F i

and 100 RH,70F C W andr61F/

S sw

/

I 130-IL 110-90 9oI I

unit 2-SFPam ientair@ !J0°F and 10%RH, 70F CW and 61;F SW 71 1

II t-I.

0 50 100 150 200 250.

300 350 400 450 500 time (hrs)

Case le -Projected

Design Calculation S-C-SF-MDC-1 810, Rev. 8 page 6 of 21 Unit I - Cross-connect Operation Swapover at 170OF (CCW 3000 gpm, SF 2500 gpm)

Plant Shutdown (10114108) - 20:00 hr, Offload Start - 80 hr, Offload Complete - 121 hr 180 170 160 150 C,

140 LL (13.

m 130 120 110 100 0

50 100 150 200 250 300 350 400 450 500 time (hrs)

Case 1f-Procedural Limit

Design Calculation S-C-SF-MDC-1 810, Rev. 8 page 7 of 21 Unit I - Cross-connect Operation Swapover at 170°F (CCW 3000 gpm, SF 2500 gpm)

Plant Shutdown (10/14108) - 20:00 hr, Offload Start - 80 hr, Offload Complete - 121 hr 1 3.1 hrs to s, yap cooling bkto Unit 1I 321.2 hrs to swap cooling back to Unit 2 170 1I Ii I

160

] "

1" 1_

I *

.1 I'

I 1

/1

.l i

i I

150 --

E 130/

I__

i I'

/

I ILI LL,

,, 140o

/

K-.rl u-ni I -

air

@. I 1 1

"L li/

1 m 120 I

J_

uni 1 SFambient air t

unit 2 - SF ambient air

@110°Fanl 100%RH,'

100 86F CCW a id 77F SW

. I....

go-1 80 o 5

1 1

0 2

80 I

050 100 150 200 250 300 350 400 450 500 time (hrs)

Case Ig

- Projected

Design Calculation S-C-SF-MDC-1810, Rev. 8 page 8 of 21 Unit I - Cross-connect Operation Swapover at 170OF (CCW 3000 gpm, SF 2500 gpm)

Plant Shutdown (10114/08) - 20:00 hr, Offload Start - 80 hr, Offload Complete - 121 hr 180" 3.8 hrs t swap coolin back toUni I I 7 hrstoswa icktoUnit2 170 JI t

I/

160 140I 0Eii

/

//

/

_'I I

I t'

i 130 ii

@ II V

j... 120 I

110 ON crWa f

IFiv 110~11 0Fad1ORH J

unit 2 - SFR ambient air

@ 110-FanU 100% RH.10-80F CCW a d 71 F SW___

90

,,[

unit I F

m ietarI

_I__

80 0

50 100 150 200 250 300 350 400 450 500 tim~e (hrs)

Case 1h!- Projected

Design Calculation S-C-SF-MDC-1 810, Rev. 8 AttaGhMent 3 page 9 of 21 Unit I - Cross-connect Operation Swapover at 170TF (CCW 3000 gpm, SF 2500 gpm)

Plant Shutdown (10114108) - 20:00 hr, Offload Start - 80 hr, Offload Complete - 121 hr 180 4.4 hrs to 4wap cooling t ack to Unit I 1 1.42.5 hrs tb swap cooling! back to Unit I

II 160_

150 130

' 120-75F CC in 6

@1-IO°F an~d 100% RHk F C n75F CdW and 66F !3!

90 80 1i 25 0

0 50 100 150 200 250 300 350 400 450 500 time (hrs)

Case Ii'- Projected

Design Calculation S-C-SF-MDC-1 810, Rev. 8 page 10 of 21 Unit I - Cross-connect Operation Swapover at 170'F (CCW 3000 gpm, SF 2500 gpm)

Plant Shutdown (10114108) - 20:00 hr, Offload Start - 80 hr, Offload Complete - 121 hr 170 150 a1.

E 130 a.

110 90 70 0

50 100 150 200 250 300 350 400 450 timel (hrs) 500 Case 1j -Projected

Design Calculation S-C-SF-MDC-1810, Rev. 8 page 11 of 21 Unit I - Normal Cooling - No Cross-connect (CCW 3000 gpm, SF 2500 gpm)

Plant Shutdown (10114108) - 20:00 hr, Offload Start - 80 hr, Offload Complete - 121 hr 170 I

I I

160-uniO 1-SFP am bent air @1 T0F and 100% RH, 99F CCW and 90F SW E

120 M 110 unitI-SFPambi Iair @110Tand 100'/

RH, 8OF:C Wand7 SW 100 I

ur1 P

e1 FI7nd 100¶'

}5F-O Wiia6F 9

unit I - SFPambientair@110 F and 1000/c RH, 8OFC CW and 61 SW 80 1

70-0 50 100 150 200 1250 300 350 400 450 500 time (hrs)

Case 2

Design Calculation S-C-SF-MDC-1810, Rev. 8 page 12 of 21 Unit I - Loss of Cooling Based on 10120108 Start Date at 0 hr - With Core Offloaded 210 200 190

.180 EL 170 E

160 150 140 130 120 unit 1 - SS ambient a'r @1 10F anid 100% RH, 1',

99F CCW ýnd 90F SW I

/

1~

unit 1 - SFP ambient air Cc unit 1 - SI:P ambientO ir (C

~110*Fa

~11 OT a rid 100% RH Sd 100% RH

ýd 100% RH

80F CCW 6nd 71 F SW 75F CCW Lnd 66F SW

'75F6

, 70F CCW *nd 61F SW unit 1 - SFP ambient !dir @110°F a

/

'I!

I I

I 0

2 4

6 8

110 tmse (hrs) 12 14 16 18 20 Case 3

Design Calculation S-C-SF-MDC-1810, Rev. 8 page 13 of 21 Unit I-Loss of Cooling Based on loss of cooling on 1111108 - Transfer Pool Communicates with SFP 200 190 180.

160.

150 S

,*110°F SFP ambient air @ 100% RH -99°F CCW I. 1 4 0 14

. 110°F SFP ambient air @ 100% RH - 80°F CCW 130 I

120- 7 z*

m

.]

110

"'/-

100 90i 80 i

0 10 20 30 40 50 60 tim~e (hr.)

Case 4a

Design Calculation S-C-SF-MDC-1810, Rev. 8 page 14 of 21 Unit 1-Loss of Cooling Based on loss of cooling on 1211108 - Transfer Pool Communicates with SFP 200 180 170,-_

7" 160

/

."Z 8L10 4

0 0

-20 30 40 50.60

.//

i*110°F SFP ambient air @ 100% RH 1F CCW 1340.

./......

° 120 ---.

110 0 90 0

10 20 i30 40 50 60 timbe (his)

Case4b

Design Calculation S-C-SF-MDC-1 810, Rev. 8 page 15 of 21 Unit 1-Loss of Cooling Based on loss of cooling on 111109 - Transfer Pool Communicates with SFP 200 190-180 170.

160 E

110°F SFP ambient air @ 100% RH-99°F CCW S140

/

110°F SFP ambient air @ 100% RH -80*F CCW V)

~130.

120 100 1

90 80 0

10 20 30 40 50 60 time (hrs).

Case 4c

Design Calculation S-C-SF-MDC;-1 810, Rev. 8 page 16 of 21 Unit 1-Loss of Cooling Based on loss of cooling on 211109 - Transfer Pool Communicates with SFP 4.E I-,

a.

GII 200 190 180 170 160 150 140 130 120 110 100 90 80 0

10 20 30 40

.50 tire (hrs)

Case 4d 60

Design Calculation S-C-SF-MDC-1 810, Rev. 8 page 17 of 21 C-E C1 LL i-a.

m 200 190 180 170 160 150 140 130 120 110 100 90 80 Unit 1-Loss of Cooling Based on loss of cooling on 311109 - Transfer Pool Communicates with SFP 0

10 20

'30 40 50 time (hrs)

Case 4e 60

Design Calculation S-C-SF-MDC-1810, Rev. 8 page 18 of 21 Unit 1-Loss of Cooling Based on loss of cooling on 6/1109 - Transfer Pool Communicates with SFP ItL I-Ii-U, r0 200 190 180 170 160 150 140 130 120 110 100 90 80 L.-

1-4-

p-4-..

4--

.I p

I I

S110°F SFP ambient air @ 100% RH - 99°F CCW

--...... 110°F SFP ambient air @ 100% RH - 800F CCW I

I J

_______I___________

0 10 20

'30 time (hrs) 40 50 60 Case 4f

Design Calculation S-C-SF-MDC-1810, Rev. 8 page 19 of 21 Unit 1-Loss of Cooling Based on loss of cooling on 911109 - Transfer Pool Communicates with SFP 1

Fi Ei 0.

E I-to U) 200 190 180 170 160 150 140 130 120 110 100 90 80 0

10 20 30 40 50 timne (hrs)

Case 4g 60

Design Calculation S-C-SF-MDC-1810, Rev. 8 page 20 of 21 Unit 1-Loss of Cooling Based on loss of cooling on 12/1109 - Transfer Pool Communicates with SFP 200 190--

1 10°F SFP ambient air @ 100% RH - 99°F CCW 180_

110°F SFP ambient air @ 100% RH - 80°F CCW 170 160-150

-- 130 120 110 100 -_

90 -

80

_t 0

10 20 30 40 50 60 time (hrs)

Case 4h

Design Calculation S-C-SF-MDC-1810, Rev. 8 page 21 of 21 Unit 1-Loss of Cooling Based on loss of cooling on 311110 - Transfer Pool Communicates with SFP 200-1 9 0 180 11O0 F SFP ambient air @ 100% RH - 99TF CCW 180 170 160 E

I 140 130-120*

110-100" 90 80 I

0 10 20 30 40 50 60 timne (hrs)

Case 4i

S.C-SF-MDC-1810 Revision 8 page 1 of I CROSSTIE Input and Output Files (Electronic files on CD)

Input files SFP Inventory data files

" Unitl-18.dcy Unitl-19FC.dcy

" Unitl-19PO,dcy

" Unit2-16PO.dcy Input Data File "Rfile" 0 1R19clg.dat a lR19loc.dat a 10-1-08.dat 0 11-1-08.dat

12-1-08.dat 0 1-1-09.dat

2-1-09.dat 0 3-1-09.dat a 6-1-09.dat 0 9-1-09.dat

12-1-09.dat

3-1-10.dat

-Output fil*s -7Th1 dutput files-are included for-each case-for-both the "Procedural Limit'-

and "Projected" sub-case with the following hierarchy:

Cases 1, 2 and 3 Procedural Limit

> Plot.dat

>, Result.tem

>, Unitl.htl

> Unlt2.htl

>, Projected

> Plot.dat

> Result.tem

>, Unitl.htl

)ý- Unlt2.htl Case 4

> 10-1-08 (typical for each date)

> Procedural Limit

> Plot.dat

> Result.tem

> Unitl.htl

>Unit2.htl

> Projected

> Plot.dat

> Result,tem

> Unit2.htl

>Unit2.htl

SC.OM.AP.ZZ-0001(Q) REVISION 2 ATTACHMENT 5 SALEM VERIFICATION OF DECAY HEAT REMOVAL FOR CORE OFF-LOAD Calculation Number:

Attachments:

Calculation cover sheet, conclusion and heatup curves The minimum time In which core off load could be conducted and adequate decay heat removal would exist in the spent fuel pool Is 80 hours9.259259e-4 days 0.0222 hours 1.322751e-4 weeks 3.044e-5 months after sub criticality.

The following parameters are based on the above core off load start time:

" The maximum CC temperature to ensure the SFP will not exceed the licensing basis limit of 1497F is 86 OF. The corresponding maximum river temperature is 77 OF.

" The predicted CC temperature at which the SFP high temperature alarm setpoint of 1250F will be reached is 62 OF. The corresponding river temperature Is 53 OF.

" If the SFP high temperature alarm setpoint Is reached, Alarm Response Procedure S2.OP-ARZZ-0003 (Alarm C-1 9) allows the setpoint to be temporarily increased to allow refueling activities to continue. If necessary to reset, see graph below.

Note: Calculation results assume one SFHX and one CCHX in service.

CCW vs Alarm Setpolnt IS Note: Includes STFmargln above 146,e~ak calculated SFP tempera tu.re 140 125 60 75 CCWtoIem~ratura p) 80 g0 Engineer:

Randall Smith Engineering Supervisor:

Alan Johnson Date: 1/16/08 Date: */.11/08 Copy to:

Salem Outage Manager I

S-C-SF-MDC-1810 Revision 8 Page I of 2

Design Calculation S-C-SF-MDC-1 810, Rev. 8 page 2 of 2 Unit I - Normal Cooling - No Cross-connect (CCW 3000 gpm, SF 2500 gpm)

Plant Shutdown (10114108) - 20:00 hr, Offload Start - 80 hr, Offload Complete - 121 hr 150 4

1-SF'nt/RH 7

u1 ult 1-SFP Jai@

bient air Ca 110 and00%RH, 8R 86 CW 7FW 78FSW 140 S

i 120 130 IZ ntie FPaýnt air @;1 I0°F and 10 % RH, 86F -CCW and *,7F SW unt. -SF a

120 100 9

0 50 100 150 200 250 time (hrs) 300 350 400 450 500 80-hr limiting case

CC-AA-103-1001 Revision 0 FORM 15 COMMENT I RESOLUTION FORM FOR DESIGN DOCUMENT PEER DESIGN REVIEW OR INDEPENDENT DESIGN REVIEW REFERENCE DOCUMENT NO. IREV. S-C-SF-MDC-1810 / R8 COMMENTS - Owner Review

1. Assumptions 3.1 & 3.3 refer to Appendices A & B, respectively. These appendices were added to the latest revision of the Unit 2 calc (S-C-SF-MDC-1800 R6) to address issues raised by the NRC during review of the LCR to change the core offload start time. These appendices are not included in this coal, and do not need to be. As such, make the following changes:

Revise Reference 4.8 to reflect MDC-1800, for consistency with MDC-1 800: Calculation S-C-SF-MDC-1800, "Decay Heat-up Rates and Curves", Revision 6.

Revise Assumptions 3.1 & 3.3 to say "... Appendix A of Reference 4.8" and ".,, Appendix B of Reference 4.8",

respectively,

2. Input 3.8: The proposed LAR is actually for a permanent change to the Tech Specs of 80 hours9.259259e-4 days 0.0222 hours 1.322751e-4 weeks 3.044e-5 months . A one-time change was done for 2R1(6 due to the time crunch. Reword to say "... supports a proposed change to the Technical Specifications..,",

Also add the following statement: "The analysis for IR19 based on 80 hours9.259259e-4 days 0.0222 hours 1.322751e-4 weeks 3.044e-5 months bounds the current scheduled offload start time of 86 hours9.953704e-4 days 0.0239 hours 1.421958e-4 weeks 3.2723e-5 months .

3.

Change Reference 4.11 (OU-AA-103) revision to 9.

-4. -Ref~ie-rie4-13:-Ch-afl~httUnflpt-Ipbcedau-*(SI-.OP h)),Change revision-#tto-i4;................-.............-

5.

Section 5.3: Unlike past revisions, this calc revision is being performed prior to the next Unit 2 outage, As such, revise 2'"

sentence of l? paragraph to read: "These files currently contain the inventory up through Cycle 18 for Unit I and the projected inventory up through Cycle 16 for Unit 2...,

6, Section 5.5: For Line 6/ "Capacity Factor", delete the statement "Note: since the individual capacity factors are also built into the burnup each individual assembly, 1.0 is entered". This was deleted in the last revision to MDC-1800 as it is incorrect.

7.

Section 6.0, Case 1: In the 1" sentence of the first paragraph (180'F case), delete reference to the heatup rate and the one-hour duration to complete the cross-connect, These parameters are only applicable in determining the temperature limit for the 2'" case that is performed at a temperature below the 1800F limit (170TF in this case) - the swapover temperature in the 2Ud case ensures the SFP will not exceed 1801F after the assumed one-hour swapover time (at 9, PF/hour heatup rate, SFP temperature will be 179.1OF at completion of swapover).

8.

Section 6.0, Case 1: In last row of each table, change units for heatup rate to "0F/hr",

9.

Section 6.0, Case 2: Delete the alarm setpoint for the 860F CC temperature. Even though 860F results in a peak temperature just under 149TF, it doesn't look good to have an alarm setpoint higher than 1490F.

10, Section 6.0, last sub-section (Max CC/SW temperatures): The current scheduled offload start time is 86 hours9.953704e-4 days 0.0239 hours 1.421958e-4 weeks 3.2723e-5 months , not 80 hours9.259259e-4 days 0.0222 hours 1.322751e-4 weeks 3.044e-5 months - 80 hours9.259259e-4 days 0.0222 hours 1.322751e-4 weeks 3.044e-5 months is the proposed revised Tech Spec minimum offload start time. Revise "current scheduled offload start" to "proposed revised Tech Spec minimum offload start".

11. The post-outage permanently discharged assemblies include nine additional assemblies not indicated in the Fuels letter:

AH62, AH67, AHT1, AH72, AH64, AH65, A#-69, AH73, AH59. Delete these assemblies, recalculate the average burnup, update Unitl-19PO.dcy, and redo the Case 4 runs.

12. Also for the Case 4 runs, extend the integration end time for the latter cases (i.e., later months) so that the SFP temperature reaches a plateau (e.g.: for 2R16, the I" case was run for 60 hours6.944444e-4 days 0.0167 hours 9.920635e-5 weeks 2.283e-5 months , and the last case was run for 160 hours0.00185 days 0.0444 hours 2.645503e-4 weeks 6.088e-5 months ).

13. Attachment 5: Change engineer's name to Robert Down. Leave date blank for now.

Page 1 of 2

CC-AA-103-1001 Revision 0 FORM 15 COMMENT/ RESOLUTION FORM FOR DESIGN DOCUMENT PEER DESIGN REVIEW OR INDEPENDENT DESIGN REVIEW REFERENCE DOCUMENT NO. /REV. S-C-SF-MDC-1810/ R8 RESOLUTION

1. Incorporated

2.

Incorporated 3, Incorporated

4. Incorporated 5, Incorporated 6, Incorporated 7, Incorporated

8. Incorporated

9. Incorporated

10. Incorporated

11. Incorporated 12, Incorporated

13. Incorporated

.ACCEPTANCE OF-RESOLUTION.

Kevin King 1/29/08 Randall Smith 2/4/08 SUBMITTED BY DATE RESOLVED BY DATE Page 2 of 2 LCR S08-01 LR-N08-0046 Calculation S-C-ZZ-MDC-1920, Rev. 41RO Fuel Handling Accident Radiological Consequences Evaluation

LQDE-AP.Z7AOOOZ(0. Rev. IL

-orm 1)

CALCIULATION COVER SHEET Page I of 45 CALCULATION NUMBER:

S-C-ZZ-fMDC-1920 REVWSION:

4JR0 TITLE:

Fuel Handling Accideints Radiological Consequences 4SHTS CALC:_

45

ATT/#SHTS 2/3'

IDV/50,59/72.48 SIITS:

6/4/0

TOTAL SHTS:

8 CHECK ONE:

o

/

FINAL lf If1ERIM (Proposed Plant Change) 1 VOID El FINAL (Future Confirmation Reqld, enter tracking Notification number:)_

SALEM OR lIOP MREW:

ElQ - LIST N.[MORTANT TO SAFETY LI NON-SAFETY RELATED HOPE CREEK ONLY:

IQ EQs ji]Qsh myF IR ISFSM:.l IMPORTANT TO SAFETY El NOT IMPORTANT TO SAFETY El ARE STATION PROCEDURES IMPACTED? YES I" NO [

IF 'YES', INTERFACE WITH THE SYSTEM ENGINEER &PRQGEDURE SPONSOR. ALL IMPACTED PROCEDURES SHOULD.Be IDENTIFIED IN A SECTION IN THE CAI.CULATION,BODY.1tR.CA 70035194432803.

INCLUDE AN SAP OPERATION.FOR UPDATE AND LIST THE SAP ORDERS HERE AND WITHIN THE BODY OFTHIS CALCULATION.

-i CP and. ADs INCORPORATED (IF ANY):

]DESCRIPTION OF CALCULATION REVISION (IF APPL.:

The analysis is revised to-calculate doses at variousdeciay times in support of an anticipated submittal for a Technical Specificatie+/-2 change, The nature of revision is such that the entire calculation is revised.

Thepurpose ofthis analysis is to determine the Exclusion Area boundary (EAB),t ow Poplatio0 Zone (LPZ)and Control-Room

.(CR) doses-dueto afuel handling accident (FHA)-occxr in 'the containment building. withthe containment-equipment hatch (ll)..

open and in the fuel handling building The FRA 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 1iwme.

CONCLUSIONS; The Sections S. 1 and 8.2 results indicate that the EAB, LPZ, and -CR doses are within their respective allowable limits for th dFHAs occurring in the containment building and fuel handling bmilding. The FRA ocwu.n g in the containment provides basis for cbaoging the following SNGS Technical Specification requirements:

1. The indiated fluel assemblies can be handled in tl.rinrator pressure vessel (IRV).4efr the reactor has beefi sub-crftical 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 minimu"m sub-critic*,lime from 168 hours0.00194 days 0.0467 hours 2.777778e-4 weeks 6.3924e-5 months to 24: boras (Technical Specification Limiing Condition for Operation (WO) 3.9,3)

2. The irradiated fuel assemblies can be moved with the containment *quipment hatch and.*oxinel lomks opened. and all containment penetrations opened in the piping ponetration aeas. with*ut containment integrity (operability) (Technical Specifcation LCO 3.9.4)

3. The core alterations can be performed without coot*nmet integr-t.(Tec*nical Specifc.ion LC 03.9,4).

The FHA occuring in the IFB provide basis fo0r-eJsia*he SNGS Technical Specificafion S m-veilance requirements 4.*,.2.b gnd:

4.9.12 c.

-/

/

ORIGIIIATOR/COMPANY NAME:

Gope1 J. Patel/NUCORE 1/V05/17/2006 REVIEWER/COMPANY NAME:

N/A N/A VERMflR/COMPANY NAME MaikDrucker/NUCORE

,ý051/20 CONTRACrOR SUPERVISOR f.*,hi=bl)

N/A PSEG SUPERVISOR APPROVAL: (Always requied)

Alan A. Johns.

SEGioZ12I

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REFERENCE:

G. Patel/NUCORE, ORIGINATOR, DATE REV:

05/17/2006 4

M. Dnicker/NUCORE, REVIEWEIUVERIFIER, 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 LAW 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-Teclmical-Specification-change...The nature.of revision-is.....

such that the entire calculation is revised.

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

REFERENCE:

G.

.a.

N.C, 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

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2]

4 38 4

5 4

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42 4

9 1

4 26 4

43 4

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45 4

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29 4

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30 4 3.1 4

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31 4 3.2 4

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-tSHEET 4 of 45 CALC. NO.: S-C-ZZ-MDC-1920

REFERENCE:

G. Patel/NUCORE, ORIGINATOR, DATE REV:

05117/2006 4

M. Drucker/NUCORE, REVIEWRIVERN ER, 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 lCal.l.tions 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 Kevis~o~ i~ I I -v,~oiu d"Iv Y n ReVisioni 12

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:

I.

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

2.

The fuel handling building (FBB)

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 TDVIE 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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REFERENCE:

G. Patel/NUCORE, ORIGINATOR, DATE REV:

05/17/2006 4

M. Drucker/NUCORE, REVIEWERIVERIFIER, 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 INS TRhUMENTATION 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-O-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 MWtW The radial peaking factor of 1.7 is conservatively used instead of the 1.65 value I Nuclear Common Revision 12

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REFERENCE:

G. Patel/NUCORE, ORIGINATOR, DATE REV:

05/17/2006 4

M. DrukerfNUCORB, REVIEWER/VERIFIER, DATE 05118/2006 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, pealing factor, and one fuel assembly failed dining 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/br peak rod average power for burnups 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..3-for use-in-thisFEHA-doseanalysis (Table-2).LlThe RADTRAD V3.02code d.

default nuclide inventory file (NIF) Bwr def. NIF is modified based on the normalized Ci/MW, in Table 3. The plant-specific NIF SNGSFHA def 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 Fgrll&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 SALEMFHAFG1l&12 is used in the FHA analyses.

3.1 FLEA 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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REFERENCE:

I G. Patel/NUCORE, ORIGINATOR, DATE REV:

05/17/2006 4

J M. Drucker/NUCORE, REVIEWER1VERIFIER, DATE 05/18/2006 Salem 1 CR Intake X/Qs (s/mý)

Time Unit I Unit I Interval Equip Hatch Plant Vent (hr)

Unit I Unit 1 CR Intake CR Intake 0-2 2.86E2-03 1.78E-03 2-8 2.22E-03 2.31&,03 8-24 9.15E-04 5.22E-04 24-96 6.60E2-04 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 FI-IA 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 FRA 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 FHB 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 FHA 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 FEB 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 cfirn

2.

Leakage through truck bay roll-up door at a rate of 3,883 cfrn

3.

Leakage through gravity damper (that replaced the truck bay exhaust fan) at a rate of 256 efn I Nuclear Common Revision 12 I Nuclear Common Revision 12 1

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

REFERENCE:

G. PateIJNICORE, ORIGINATOR, DATE REV:

05/17/2006 4

M. Drucker/NUCORE, RE VIEWERIVERIFTER, DATE 05/18/2006 The atmospheric dispersion factors (c/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, FEB 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 FHB to calculate the EAB, LPZ, and CR doses. Activity from the FEB is assumed to be released to the environment at a rate of 21,439 cfln (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 CER1 and PV are the release points for the FEA occurring in the containment. As discussed in Section 3.2 above, the plant vent, FBB rollup doors and gravity damper (modeled as the smoke hatch) are the release points for the FHA occurring in the F-TB. 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 CEHl

& 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 FHA 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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REFERENCE:

G. PateINUCORE, ORIGINATOR, DATE REV:

05/17/2006 4

M. Drucker/NUCORE, RE VIEWER/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 There are two radiation monitors in each normal CR air intake duct having the alarm/trip set point of 2.48 x 103 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

........sp-e pme- (1-3 s-cbd) (Ref. 1068)-if the ]oss-of-offsite power is assumed-to-occurat 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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SHEET 11 of 45-CALC. NO.: S-C-ZZ-MDC-]920

REFERENCE:

G. Patel/NUCORE, ORIGINATOR, DATE REV:

05/17/2006 4

M. Dnicker/NUCORE, REVIEWEBIVERIFIER, 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 RU 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 bumup. All the gap activity in tile 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 1 Nuclear Common Revision 12 I

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REFERENCE:

G. Patel/NUCORE, ORIGINATOR, DATE REV:

05/17/2006 4

M. Drucker/NUCORE, REVIEWERJVERIF[ER, 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.

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 elemenltal (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).

uiel Handling Aceidents 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

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REFERENCE:

G. Patel/NUCORE, ORIGINATOR, DATE REV:

05/17/2006 4

M. Drucker/NUCORE, REVIEWER/VERIFIER, DATE 05/1812006 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-FHJA 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 3.4.1). 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 cffn 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 TBDE 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 Rem TEDE 4.10 The breathing rates for persons at offsite locations are given in Reference 10.1, ROP 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 Rem TEDE I Nuclear Common Revision 12 1 Nuclear Common Revision 12

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):

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),

Contamination of the control room atmosphere by the intake or infiltration of airborne radioactive

. iiteial frbir

-a-reas and structures adjageit to-the-cont il-o im env elop-e -(Vi --CR -unfiltr-ed..

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 FHA 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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REFERENCE:

C.PatelNUCORE, ORIGINATOR, DATE REV:

05/17/2006 4

M. Drudcer/NUCORE, REVIEWER/VERIFIER, DATE 05/1812006 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 FEA (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).

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REFERENCE:

G. Patc/NUCORE, ORIGINATOR, DATE REV:

05/17/2006 4.

M. Drucker/NUCORE, REVIEWERNVERIFIER, 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 F-A 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 %/Qs in Reference 10.16 were accepted by the staff in the previous licensing proceedings.

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.CALCULATION CONTINUATION SHEET SH[EET 17 of 45 CALC. NO.: S-C-ZZ-MIDC-1920

REFERENCE:

G. Patel/NUCORE, ORIGINATOR, DATE REV:

05/17/2006 4

M. DnukerI/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 12 1 II Nuclear Common Revision 12 I

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

REFERENCE:

ORIGINATOR, DATE G. PateYiNUCORE, REV:

05/17/2006 4

M. Drucker/NUCORE, REVIE WERVE*RU

, DATE 05/18/2006 Design Input Parameter Value Assigned Reference 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 MW1 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 Activi.

KR-83M 1.20E+07 1-132 1.40E+08 XE-133 2.O0E+08 K.R-85M 2.60E+07 1-133 2.OOE+08 XE-135 5.00E+07 KR-85 1.1OE+06 1-134 2.20E+08 XE-135M 4.OOE+07 KR-87 4.70E+07 1-135 1.90E+08 KE-138 1.60E+08 KR-88 6.70E+07 XE-131M 7.OOE+05 1-131 9.90B+07 XE-133M 2.90E+07 5.3.1.3 Radionuclide Release Fractions (10.1, RGP 3.2, Table 3)

Group-.....

_-action

'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 His used in the analysis Assumed Time 72 I-Irs 60 Hrs 48 Hrs 24 His 5.3.1.8 Radial Peaking Factor 1.65 (1.70 used in the analysis) 10.19 Nuclear Common Revisioni 12

CALCULATION CONTINUATION SHEET SIREET 19 of 45..............

CALC. NO.: S-C-ZZ-MDC-1920

REFERENCE:

G. PatelNUCORE, ORIGINATOR., DATE REV:

05/17/20064 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 ft3 10.11 Free Air Volume 5.3.2.3 Iodine Decontamination Facors (DFs)

Elemental 500 10.1, Appendix B, Section 2 Organic 1

5.3.2.4 Overall Effective Decontamination Factor (DFs) for Iodine Total Iodine 3

200 I 10.1, 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 cfm 10.18.g & 10.18.h From Ring Header 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 f9 Section 7.2.2 Volume 5.3.3.3 Activity release rate 21,439 cfmn 10.18.a, 10.18.d, & 10.21 (19,490 x 1.1 = 21,439 cfm) 5.3.3.4 FHB Charcoal Filter Not credited in the analysis 10.6.10 Efficiencies The remaining FHA occurring in the FHt3 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 f?

10.12, page 33 5.3.4.2 CR Normal Flow Rate 1,320 efin Section 7.3 5.3.4.3 CREACS Design Makeup 2,200 cfin 10.6.13 Flow Rate 5.3.4.4 CREACS Ventilation 8,000 efin +/- 10% cfin 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

~~T CALCULATION CONTINUATION SHEET "

SHE'E T 20 of 45....

CALC. NO.: S-C-ZZ-MDC-1920

REFERENCE:

G. Patel/NUCORE, ORIGINATOR, DATE REV:

05/17/2006 4

M. Drucker/NEJCORE, REVWE WER/VERIFIER, DATE 05/18/2006 Design Input Parameter Value Assigned Reference 5.3.4.6 CREACS IJEPA Filter 99%

Section 7.7 Efficiency 95%

Used in Analysis 5.3.4.7 CR Unfiltered lnleakage 150 cfin (nominal value measured 10.32, Table 1 is less than 100 elm) 5.3.4.8 CR Occupancy Factors Time (Hr) 10.1, RGP 4.2.6 0-24 100 24-96 60 96-720 40 5.3.4.9 CR Breathing Rate) 3.5E-04 M3/sec 10.1, RGP 4.2.6 5.3.4.10 Unit 1 CR X/Qs - Post-FHA Release From Unit I CEH Time (Hir) x/Q (sec/mn) 0-2 2.86E-03 10.5, page 33 2-8 2.22E-03 8-24

9.15F,-04 24-96 6.60E-04 96-720 5.62E-04 5.3.4.11 F.HB 0-2 hr Equivalent

/Q 1.85E-03 s/mn Section 7.5 5.3.4.12 Unit 1 CR x/Qs-Post-FtHA Release From Unit 1 Plant Vent Time (fr)

X/Q (sec/r 3) 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 I CR X/Qs -Post-FHA Release From FHB Rollup Door Time (Hir)

X/Q (sec/m 3) 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

REFERENCE:

G. Patel/NUCORE, ORIGINATOR, DATE REV:

05/17/2006 4

M. Drucker/NUCORE, REVIEWERIVERIF¶ER, DATE 05/18/2006 Design Input Parameter T Value Assigned Reference 5.3.4.14 Unit 1 CR X/Qs - Post-FHA Release From Smoke Hatch Time (Hr)

X/Q (seclms) 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 BAB Atmospheric

]

1.30E-04 110.16, Table 5 Dispersion Factor (X/Q) (sec/h 3) 1 5,3.5.2 LPZ Atmospheric Dispersion Factors (X/Qs)

Time (Hr)

X/Q (sec/r 3) 10.16 Table 5 0-2 1.86E-05 2-8 7.76E-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 5.3.5.4 Offsite Breathing Rate (el/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 CRIntake Monitor Xe-133 6.2 x 107 cpm/LCi/cc 10.13, page 12 Sensitivity 5.3.5.6 CR Intake Monitor 2.48 x 10' cpm 10.6.7 Alert/Tip Setpoint I Nuclear Common Revisiou 12

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

REFERENCE:

G. PatelNUCORE, ORIGINATOR, DATE REV:

05/17/2006 4

M. Drucker/NUCORE, REVIEWER/VERIFIER, 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 FIB exhaust rate of 21,439 cfm and the requirement to remove 99% of the activity in a two hour period.

6.2 Fuel Handling Accident in the FMB with a Failure of an Exhaust Fan

. -The-post-FHA activity releases-through three-different.release-pathsdue topressurization 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 (NWF) For RADTRAD V3.02 Input The parameter Ci/MW, 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/MW, information developed in Table 3. The RADTRAD nuclide inventory file SNGSFtIA-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 =Ao e""'

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

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, REVIEWER/VERIFIER, DATE 05/18/2006 Where; A0 Initial Activity in Source Node A = Final Activity in Source Node X Removal Rate (vol/hr) t = Removal Time (hr) = 2.0 hr Assuming that 99% of activity is released into the environment, AZAo = 0.01 Therefore, A/Ao = e"-t 0.01 = e"21 In (0.01) = - 2,% ln(e)

- 4.605

-2 X

- 4.605/-2 = 2.3 03 volume/lir Containment Building Release Rate = 2.303 1/hr x 2,620,000 ft3 x 1 hr/60 minr 100,600 ft3/min 7.2.2 Fuel Building Fuel building exhaust flow rate= 19,439 cfm (Ref. 10.18.a & 10.18d) x 1.10 = 21,439 cfm 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 = 3.838E-02 vol/min FEB Volume =

21,439 ft3/min

= 558,550 ft3 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 FBB release rate is 558,550 ft3/min is used in the RADTRAD model.

I Revision 12,

II Nucleari rnmmonn eiin1

CALCULATION CONTINUATION SHEET SHEET 24 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 7.3 Control Room Flow Rates Normal Flow Rate Reference 10.20, Note "S" provides the outside air flow rates to Zone I 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 cfm (2,200 clmii outside air + 30,400 efi recirc air)

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

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

= (8,000 clm / 32,600 cfrn) x 2,200 cfm = 0.2454 x 2,200 olm = 540 cfii Use 600 cmin for Zone 1 During Normal Plant Operation Total Amount of Outside Air Flow Rate From Both Intakes = 2 x 600 ein = 1,200 cfin Maximum Amount of Outside Air-Flow R

.te

.1x 1,-200 -friff-1,320-tfin CRBACS Recirculation Flow Rate CREACS ventilation flow rate= 8,000 cin_+ 10% cfin (Ref. 10.6.12)

Minimum CREACS flow rate = 8,000 clm - 0.10 x 8,000 clm = 8,000 clm - 800 efin = 7,200 clin Net CREACS recirculation flow rate = Minimum CREACS flow rate - CREACS makeup flow rate 7,200 cfml-2,200 clim (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 f/min (Ref. 10.18.c & e) x 1

x 2.86E-03 sec/m3 2.62E+06 ft3 60 see/minn

= 1.132 Ci/m3 = 1.132 GCi/cc CR Intake Monitor Xe-133 Sensitivity = 6.2 x i07 cpm/tCi/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 I Nuclear Common Revision 12 1

CALCULATION CONTINUATION SHEET SHEET 25 of 45 CALC. NO.: S-C-ZZ-M.DC-1920 G

acNCR,

REFERENCE:

ORIGINATOR, DATE REV:

05/17/2006 l4 M. Drmcker/NUCORE, REVIEWER/VERIFIER, DATE 05118/2006 1.132 jiCi/cc x 6.2 x 107 cpm/4Ci/cc = 7.02 x 10 7Cpm >> 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/mtn

= 2.104 Ci/m3 = 2.104 pCi/cc CR Intake Monitor Xe-133 Sensitivity= 6.0 x 107 cpm/ýLCi/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-FRA Activity Concentration At CR Intake

= 2.104 I.Ci/cc x 6.2 x 10& cpm/p.Ci/cc = 1.30 x 10C cpm >> 2.48 x 103 cpm

-It is clear that-theCR intake monitorw~illtinstantaneously reach its Alamn/Trip setpoint-following a FHA occurring in the containment or fuel handling building.

7.5 Equivalent 0-2 hr,,/Q For FHB Release Path Plant Vent 0-2 x/Q 1.78E-03 s/m3 (Ref. 10.5, page 34)

FHB 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/M3 (Ref. 10.9, Section 8.4)

1.

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

2.

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

3.

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

= 15,300 efmn x 1.78E-03 s/m 3 + 3,883 cfm x 1.50E-03 s/M3 + 256 cfm x 1.15E-02 s/m3 (15,300 olin + 3,883 cfm + 256 cfin)

= 36.00 cfr.s/m3 = 1.85E-03 s/rn 3 19,439 lmin I

I Nuclear Common

..Revision 12 Nuclear Common Revision 12 1

CALCUATION CONTINUATION SHEIET SHEET 26 of 45..

CALC. NO.: S-C-7_Z-MDC-1920

REFERENCE:

ORIGIATOR, DATE REV:

05/17/2006 4

M. Drucker/NUCORE,

[REVEWERNVERIFIER, DATE 05/1 8/2006 7.6 Distance of TSC Air Intake Plant North S480' Unit 1 Containment 181.64' 49.82' South Coordinate of Unit 1 Containment = South Coordinate of Plant + Distance between Centerlines of Plant and Unit I 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 S

$715.88' (Ref. 10.28) +1'-6" (Ref. 10.28) = S717.38' Distance between Column 1B and TSC Air Intake Distance between Columns 1B 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 S $717.38 - 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 12 1

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) -' 1-6" (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 I Containment and CEH = 59.37' (Ref. 10.5, page 26)

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

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

=-E2.2-7' +.Wi.20;0'. + =-.22.27-..

East-west Distance between Centerlines of Unit 1 CFH and TSC Air Intake East-west Distance between Centerlines of Unit I 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 + ( 12 2.2 7 )']"a = 282.28' = 86.06 m Slant Distance between Centerlines of Unit 1 CEI-and TSC Air Intake

= [(204.61)2 + (181. 64)21

= 273.60' = 83.42 m The distance between the source locations (Plant Vent and CEH) and receptor locations (Unit 1 CR and Unit 1

& 2 TSC) are compared in the following table:

I Nuclear Common Revigioin 12 1

T CALCULATION CONTINUATION SEEET SHEET 28 of 45 CALC. NO.: S-C-ZZ-MvDC-1920

REFERENCE:

G. PatVl/NUCORE, ORIGINATOR, DATE REV:

05/17/2006 4

M. Druoker/NUCORE, REVIEWER/VERIFIER, 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 CGE CE H and and and and Unit 1 TSC Intake Unit I TSC Intake CR Intake (Meter)

CR Intake (Meter)

(Meter)

(Meter) 30.25 86.06 46.62 83K42 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 99n02-(Ref1 0.30)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% - r1)/safety factor = (100% - 11)/2 Where rj = charcoal filter efficiency to be credited in the analysis CREACS Charcoal Filter 2.5% = (100% -rj)/2 5% = (100%- 7)

Tr = 100% - 5%

95%

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

Filter Aerosol Elemental

__Organic CREACS 95 95 95 I Nuclear Common Revision 12 1.'

I Nuclear Common Revision 12

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

REFERENCE:

G. PatelNUCORE, ORIGINATOR, DATE REV:

05/17/2006 4

M. Drucker/NUCORE, REVIEWERIVERIFTER, 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 S24FHA15O.oO 0.95 1.05 0.15 S48FBA_150.o0 60 0.89 0.99 0.14 S60FHA 150,7oO 72 0.84 0.93 0.13 S72F11A150.oO 96 0.76 0.84 0.12 S96FHA15O.oO Allowable TEDE 5.0 6.3 6.3 Limits 5.

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 cfm 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. Pate5/NUCORE, ORIGINATORI, DATE GV 05/17/2006 4.

M. Drucker/NUCORE, REVIEWER/VERIFIER, 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 FB24FIRA150.oO 48

0.

0.62 1.05 0.15 EB48FIIA15O.oO 60 0.58 0.99 0.14 FB6OFHA150.o0 72 0.55 0.93 0.13

-FB72FHA15O.oO.-

96 0.49 0.84 0.12 FB96FHA150.oO Allowable TEDE 5.00 6.3 6.3 Limits Significant assumptions used in this analysis:

FEB charcoal filtration is not credited Gap fission product fractions doubled Activity is released to the environment at a rate of 21,439 cfin CR envelope is pressurized with actuation of the CREACS following a FHJA 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 Spent fuel pool overall effective DF = 200 Core thermal power = 3,632 MWt Radial Peaking Factor= 1.70 Revision l~

I Nni'h~r Common Revision 12 [

I

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 FB48PUFF150.o0 60 1.67 1.00 0.14 FB60PUFF15O.oO 72 1.58 0.94 0.13 FB72PUFF15O.oO 1.43 0.85 0.12 FB96PTUFF150.o0 Allowable TEDE 5.00 6.3 6.3 Limits Significant assumptions used in this analysis:

FHB charcoal filtration is not credited Post-FHI-A activity is released to the environment at a rate of one volume/minute (558,550 cfin)

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 cfin 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 1

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, REVIEWERNVERIFIER, DATE 05/18/2006

9.0 CONCLUSION

S 9.1 FjA 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 CEHl, 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 1 & 2 Technical Specification Surveillance requirements 4.9.12.b and 4.9.12.c can be relaxed.

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

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

REFERENCE:

G. Pate/NtUCORE, ORIGINATOR, DATE REV:

05/17/2006 4

M. Drucker/NUCORE, REVIEWERIVERIFIER, 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-i.25,-Rated-ThermalPower...

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.dil, CREACS Ventilation Flow Rate 6.13 Specification Surveillance Requirement 4.7.6.1.d.3, CREACS Design Makeup Flow Rate 6.14 Specification Surveillance Requirement 4.7.6.1.e, REPA 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 I/Unit 2 Containment Configuration I Nuclear Common Revision 12 1

CALCULATION CONTINUATION SHEET SHEET 34 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

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 I 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

[ Nuclear Common Revision 12

CALCULATION CONTINUATION SHEET SHEET 35 of 45 CALC. NO.: S-C-ZZ-MDC-1920

REFERENCE:

0. Pate1/NUCOR.E, ORIGINATOR, DATE REV:

05/17/2006 4

M. Drucker/NUCORE, REVIEWERIVERIFIER, DATE 05/18/2006

21.

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

Subject:

Estimate of Unfiltered Inleakage from FHB 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. a.... 602511,_Re-v 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 IVAC 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 1 Nuclear Common Revision 12 I Nuclear Common Revision 12

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, REVIEWERIVERIFIER, 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 t.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 12400E+08...

1.009..

-1:A412E+08 1-133 2.OOOE+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.OOOE+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.OOOE+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 Nuclear Common Revision 12 ]

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

Bi C

=AC/

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 3.196E+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 XB-135M 4.036E+07 0.05 0.10 8.071B+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 I Nuclear Common Revision 12 1 I Nuclear Common Revision 12 I

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, REVIEWER/VERIFIER, DATE 05/18/2006 Table 3 Post-FRA 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)

(CiIMWt)

(Ci/MWt)

A B

C D

E=A*B*D/C F

G=E/F H=G/3632 I=H1 q K.R-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 4.62 IE+05 1.272E+02

.1272E+03 K.R-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+09 1.70 193 1

1.191E+06 1.0 1.191E+06 3.279E+02

.3279E+03 1-131.

3.196E13408 J1.70-193 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

3.555E1+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 Nuclear Common Revision 12 I f

12.0 FIGURES Figure 1: FRA In Containment Building With Equipment Hatch Open RADTRAD Nodalization 1*w l Cr i'nmmnn Revision 12 MMOD

.[

CALCULATION CONTINUATION SHEET SHEET 40 of'4S CALC. NO.: S-C-ZZ-MDC-1920

REFERENCE:

G. PateI/NUCORE, ORIGINATOR, DATE REV:

05/17/2006 4

M. Dmcker/NUCORE, REVIEWER/VERIFIER, DATE 05/18/2006 Figure 2: FfA Occurring In Fuel Handling Building RADTRAD Nodalization I

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

1,320 cfm < m min u

2,200 cfm> lmin Figure 3: Salem Control Room RADTRAD Nodalization I Nuclear Common Revision 12 1

I Nuclear Common Revision 12 1

CALCULATION CONTINUATION SHEET

[ SHEE T 43 of 45 CALC. NO.: S-C-ZZ-MDC-1920 RE'FERENCE:

IG..P~atel/NUCOILE, ORIGINATOR, DATE REV:

05/17/2006 4

M. DruckeriNUCORE, REVIEWER/VERIFIER, DATE 05/18/2006 Figure 5: Post-FIA CR TEDE Dose Vs Fuel Decay Time (FHB) 0.8 0.7 0.6 0,5 0.4 0.3 A

~%

0.1 0

.0

,.,!: ..*. ::i '::*

!.: *!i*l It Iý 0

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

I Nuclear Common Revision 12 1

CALCULATION CONTINUATION SHEET SHEET 44 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 6: Post-FHA CR TEDE Dose Vs Fuel Decay Time (FHB Puff) 2.5 2.0 S1.5 1.0 0.5 0.0 FjI

FF~i~FFFv.

F II F

.1 F

FFF r

F

,~

FFFF F

F.,

F I

I FF1 I

I 0

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

Nuclear Common Revigion 12 1 Nula omo viin1

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

REFERENCE:

G. PateIINUCORE, O.IGINATOR, 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 SNGSFHA_def.txt Nuclide Release Fraction & Timing File SNGSFHA-rftxt FGR Dose Conversion File SALEMFHAFG1 1&12.txt RADTRAD Input and Output Files for FHA Inside Containment:

S24FHA150.psf and S24FHA 150.oO S48FHA150.psf and S48FHA15O.oO S60FHA150.psf and S60FHA150.o0 S72FHA15O.psf and S72FHA150.oO S96FHA150.psf and S96FHA150.oO RADTRAD Input and Output Files for MEA Inside Fuel-Handling Building:

FB24F-A1 5 0.psf and FB24FHAl 5 0.oO FB48FHA15O.psf and FB48FHA150.o0 FB60FHAI5O.psf and FB60FHAI 50.oO FB72FHA150.psf and FB72FHA150.o0 FB96FHA1 50.psf and FB96FHAI 50.oO RADTRAD Input and Output Files for FHA Inside Fuel Handling Building (Puff Release):

FB24PUFF150.psf and FB24PUFF150.oO FB48PUFF15O.psf and FB48PUFF15O.oO FB60PUFFl 50.psf and FB60PUFF1 50.oO FB72PUFFI 50.psf and FB72PUFF150.oO FB96PUFF15O.psf and FB96PUFF150.o0 13.2 Copy ofReference 10.21 (2 pages) 14.0 AFFECTED DOCUMENTS S-C-ZZ-MDC-1920, Revision 3 will be superseded.

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

3.1 S-C-Z-M C-i1920., Rev. 4 CD With Various Electronic Files 3.2 S-C-ZZ-MDC-1920, Rev. 4 1 of 2 A

I Fmrn the Desk of Paul Woods TO:

Joh' ulfy D'u' f DATE:

12116/96

SUBJECT:

Emimate of Unfiltered Leakage m FHB with One Exhaust Fan ana te Supply Fan Opentlrng.

Initially, the FHV system will bO in ft normal alignmMet rof fuel handling, 2-exhaust fans and tm supply fan operating. Upon loss of one exhaum fan he building pessure controler will atampt to moduieta open to maintain building negfalve preaan. Evmnwally the mcdmum travel stop wig be reacied end no further uxhaiust flow Incrse Is poesible. Building prussure will wrnflnue to Increue due to le Iinmblm*mn betWoen the exhoat flow and the supply flow.

when the bilding msaure reaces tie slrm eiolntt (epprox. 0.10 kvdes of water negaftl, WRT the catmide) fte contr rowm manwclatr vwil

m. Per the Alum Ruspamus procedure tea per*tor is dErectod to s,". dawn the gpvratin supply ton when fth bW1 psMure lam is received. During the period whn tie operator Is evaluating ftw alarm, and tekng action, the fuel handling building may go poiWe.

Pitenta release paints ae e Vuck bay rMOup door an t* west and of te Fuel Handling Bidng, lea kAe tiJhftU dosled Qvit daIper that replaced tle truW bay exhaust fan (a1m on Dwg 207647 and located in the south wal at elevation 124%r; Ir west of the N-N grid oca tIni),

end lealtage thrugh tie 2FHV supply ioW handling unit damper bted In *th norh wg at the 100' elevaton; approziet*l IY east of gld W-eftn We R-R, also in the truck bey area (am show n Pwg 207847).

During r*war* operaloo,te supply fan is set to awp ximately 2000 em less Man the exsMat flow rae. The normal exhaust flow raf Is upprwoxmatey 19,490 Thrm*nore tiMe norms! supply o wound 17,500 dm.

The FHV exhaust tans are Identicl tans opated In paraleli, with back drat dampers on the ekdhat to porm* the hi flow from one fan to be exhatuted if the other fan Is stopped. The pressur vs. flow espo*se for uti arraM*nemet can be modeled by plotting the equivalent fI QVC foe both fans from tm tan curve fore *Qg fan by doublng the flow ae for each constant total pressue poiit When omfa Is touM

e resul Is tat t*e system resistance curve is followed down to *he Iterwsection of t system curve With OWe single fun cr.

The I rW 2 fain curves can be plotted oug ft vendor supplied fat cv The equation of ft system mve is Ogven by

- FRV.

When AP Is the ran tout pressure, R is a omsant, and V is Vte volumflic Iw. The oerating point on the 24an curve is 19,4W0 drn, From tie wrnbned 2 Fan curn hNs cornsepds to 8.2 in W.3. total pressure. Solving for R R-6Pi V R~ aS.2t1(i9490)1 g1pt-rZFo NWll 7671 OgLe Rm 2.1600O) In W.O.cfmha 3.2 S-C-ZZ-MDC-1920, Rev. 4 2 of 2 A

j FronMth. Derk of Paul WoOde This constant is then used to plot the syatem culvO, wtth Interseta the 1-fan curve It 14,500 O'm (refer to the atnche curves).

Theemfore, the flow thmugh the normal exhaust path (filtered) is 14,500 elm, A conservtive esumpton Is thaiten supply fantoultinues to opera at 17.500 cfrn wit 14,500 ct extausted to fte sad, and tht remainder or 3,000 dcn aelng unlntoend to the env ronmmnt The mehodl for this type of evaluation is provided in SeIlon 19.4 of HiNEdbook of Air Condiftonkig nd RetrfgeraUon by Shen K. Wang; McOlrw-H4ll, 1993.

The gravi damper Identifled above Is Class II lea1aOS Per ASME AG-I (Refer to PS9P 317245). Clan sl damm arerated for @fnm persq. ft cf face ar at 1 in WO. For the 41r x 48' gravity Mlief damper, No is equslto 128 dfm 4 t in.

The maximum pr en flte supply fan Is approWIMdtly 3 In W.Q.

theefreto be =nauvam e te leskag from the ampe Is esuimated to be 2 tImes oR 2tS Clm.

Uilng a similar method Wte above, th*e10% and.10% system curve wam plotted, and tew WO lakae for each cass was evaluated, toe resut.e we presented. below.

Dlesgin Akflw Design +10%

Otlgn -10%

2-Exhaust Fans 19,490 21.439 17.541 Supp Fan 17,490 1%,439 15.541 I-Exhaust Fan 14,500 15,300 13,D00 Dolt (Supply-.

2,990 4,139...

41 single exthust)

Leakage at Truck 2,734 3,883 Less Bay Leakage Thlugh 256 256 256 GMV DOamlr Thefr are meal cMnseratliMs built into th aove 9tIMlet. First. The estimated system ourve Ignores U's effec of the supp fan and only looks at the a~e of fte two exhaust fans In parallel. The press~iazaton efct of the supply fa wi tend to push moair #thou the *e*Su fiers Sawed calmialng the leake Uug1 Vi gaty damper ignores the menu demper r*eceny placed In sefes with this damr, The menus temper Aodd be docse duing fuel handlIng an" w pw lndl oWN ree*lene to leakage at em locatio. Td.

te leakag Uough flle wuvty damper Is sassumed to double at.a pmussure of 3 in W.Q. allhough Ole Is square mrt mIUUoIW)ip, and h*e leakage would achfuly a* inrease by a factor of 1.7. Fourth, tha budtin pressure is aSame to to 3 in W.G. pouti.

ie. e. te nMimum stainable by the suppt fan, In fac the building pressure will be less than thIs nmoerTuM, aid may be nearly reutreL Reviewed BYre Date: Z2QL&2.C..

NC.CC-AP.ZZ-oO0O(Q)

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

SUMMARY

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

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.

Markt Design Verifier Assigned By Name of Lead Design Verifier I Date (print name of SupvlManager/Director)*

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

Design Verifier Assigned By Name of Design Verifier" 1 Date (print name of Supv/Manager/Director)"

Design Verifier Assigned By Name of Design Verifier/ IDate (prnt name of SupvlManager/Dlirector)*

Iftho Maonr/Supmvisor s a ti Uasijgn Veiier, the=m onahoim hihaerlave of teWhnoka magethent ii quir*d in 11e lel column, Nuclear Common Roy.,3

NC.CC-AI.Z7-0010(Q)

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

REFERENCE DOCUMENT NO. /REV.

S-C-zz-MDC-1920, Revision 41R0 COMMENTS RESOLUTION ACCEPTANCE OF RESOLUTION 1 General incorporated.

Editorial Comments are being provided separately in the form of a redline/strikeout mark-up. The y/3-~~

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 Versiov 3.02 of the RADTRAI) computer code (R~ef.

10.2) to calculate the potential radioogical Consequences of an FHA. The RADTRAD code is documentd in NUREGICR-6604 (Ref 10.2). The RADTRAD code is maintained as Software ID Number A-0-ZZ-MCS-0225 (Ref, 10.33)..

3.

Section 3.2 Typo corrected.

The text erroneously states that the results of the 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 of the- -

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 deletd.

The text discussion ofthe Reference 10.9 ARCON95 o

t d

e analysis of the smoke hatch is not necessary.

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

5 Design Input 523.4 Incorpprated,

~A Please add a new design input section (5.3.4.14) to document the X0Qs for the smoke hatch release taken 14?-.06

__from Reference 10. 9 page 42.________

Nuclear Commoon Rev. 3

NC.CC-AP.ZZ-OD1O(Q)

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

REFERENCE DOCUMENT NO. /REV.

S-C-ZZ-MDC-1920, Revision 4110R 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 Rlvlý and LPZ X/Q values. In previous calculation 5--127-04, 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-MDC41959 which provides the smoke hatch X/Q values, but which does not provide the offskte XIQ values. Please add a reference for offsite X/Q values.

7 Section 7.2.1, Section 5.3.2.9, and FIHA Inside New release rate is 100,600 cfbi used in the Containment RADTRAD runs analysis and the RADTR.AD runs for the FH The containment building release rate of 99,800 cfmn occurring in the containment building are

- /c¶ -d modeled in the RADTRAD runs is calculated in revised.

Section 7.2.1 for a containment volume of 2.66 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 cflm.

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

,At f

--The Section 7.4 calculations model a CR intake monitor Xe-133 seiftivity of 6.0E7 rather than the 6.2E7 value shown if Design Input 5.3.5.5 (which cites Ref. M 13 page 12), Pleasv revise asnecessary 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 1.14&-2. Per Reference 10.9 Section8.4 the maximum X/Q is 1.15E-2 forthe Ul smoke hatchto Ul 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 8.3 Incorporated.,

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

II RADTRAD nims for IRA in containment incorporated.

4j.

In each run, the CREACS recirculation filter is turned

- 0 on at2 minutes, instead ofthe I minute value specified in Section 3.4 12 RADTRAD rna for FRA in PUB (2 hr release)

Incorporated.

1) In each run, the CREACS recirculation filter is turned on at 2 minutes, instead of the I minute value ýpecifted in Section 3,4

2)

The FB24FHA24 output file name has an extra dot before its extension: "..W_

Nuclear Common Aev. 3

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 cfrn. 150 cfm bounds the nominal inleakage results except for the isolation mode. 4000 cfni 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. 1 OCFR50.59 Screening: A screening is not required if all aspects of the calculation revision support the license change submittal.

J. Duffy 05117/2006 SUBMITTED BY DATE

-Nuclear Common

Page.25 -of.27 Rev. 3

COMMENT I 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 05/1 7/2006 RESOLVED BY DATE ACCEPTANCE OF RESOLUTION J. Duffy 05/1812006 SUBMITTED BY DATE Nuclear Common

-Page-25 -of 27 Rev. 3

NCCC-A1jIZ-00IJ(Q)

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

REFERENCE DOCUMENT NO. /REV. S-C-ZZ-MDC-1920, Revision 41110 COMMENTS RESOLUTION ACCEPTANCE OF RESOLUTION 13 RADTRAD runs for FR A in FHIS (puft)

Incorporated-

&Aý

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

2) In each run, the CREACS filtered and unfiltered iniflow rates are initiated at 2 minutes, instead of the 1 minute value specified in Section 3.4

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

4) In each run, the CR X/Q is modeled as 1.78E-3 (not 1.85E-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 eftm ark Drucker 05/1412006 Gopal J. Patel 17/2008 SUBMITTED BY DATE RESOLVED BY=7 r DA,.TE..

Nuclear CommonR Rev. 3

v t e Letter Sequence Request
TAC:MD8259, (Approved, Closed)
Initiation Request Acceptance... Supplement Administration Questions 12 May 2008 24 June 2008 Answers 23 July 2008 Results Approval, Approval, Approval Other:
MONTHYEAR2008-06-17 [Table View]

LR-N08-0046, Request for Changes to Technical Specifications Refueling Operations - Decay Time, License Amendment Request (LAR) S08-01

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6.0 CONCLUSION

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