ML14042A115

License Amendment Request - Increase of the Allowable Gas Treatment System Maximum Flow Rates from Secondary Containment

ML14042A115
Person / Time
Site:	Susquehanna
Issue date:	01/29/2014
From:	Franke J A Susquehanna
To:	Document Control Desk, Office of Nuclear Reactor Regulation
References
PLA-7131, TAC MC8730, TAC MC8731
Download: ML14042A115 (50)

Text

I, Jon A. Franke Site Vice President PPL Susquehanna, LLC 769 Salem Boulevard Berwick, PA 18603 Tel. 570.542.2904 Fax 570.542.1504 jfranke@pplweb.com No-ppI&#*TM JAN 2 9 2014 U. S. Nuclear Regulatory Commission Attn: Document Control Desk Washington, DC 20555 SUSQUEHANNA STEAM ELECTRIC STATION LICENSE AMENDMENT REQUEST -INCREASE OF THE ALLOWABLE GAS TREATMENT SYSTEM MAXIMUM FLOW RATES FROM SECONDARY CONTAINMENT PLA-7131 Docket No 50-387 and No. 50-388

Reference:

1.NRC Safety Evaluation, "Susquehanna Steam Electric Station, Units I and 2 -Issuance of Amendment re: Implementation of Alternative Radiological Source Term (TAC Nos.MC8730 and MC8731), " dated January 31, 2007 (ADAMS Accession ML070080301).

Pursuant to 10 CFR 50.90, PPL Susquehanna, LLC (PPL) hereby requests approval of the following proposed amendments to the Susquehanna Steam Electric Station (SSES) Unit 1 and Unit 2, Facility Operating Licenses NPF-14 and NPF-22, as described in the Enclosure.

The proposed amendments would change the current licensing basis (CLB) for one postulated accident analysis described in the SSES Final Safety Analysis Report (FSAR), and as previously reviewed by the NRC in Reference

1. Specifically, this request would implement changes in calculations for the Control Room Habitability Envelope and Offsite Post LOCA Doses to permit higher allowable Standby Gas Treatment System (SGTS)exhaust maximum flow rates from secondary containment as described in the Technical Specification Bases.The proposed changes use available margins in the current analysis of record to establish a new dose of record. This change will be used to implement the proposed higher allowable SGTS maximum flow rates to remove an adverse operational impact on SSES Secondary Containment ventilation zones. Specifically, this change improves the dose margins for the control room operator where very little margin currently exists. Additionally, this change maintains the existing substantial DBA LOCA offsite dose margin, which allows SSES to remain aligned with others in the industry.Justification for the change is based on the evaluation presented in the Enclosure.

The License Amendment Request (LAR) has been evaluated in accordance with 10 CFR 50.91(a)(1) using criteria in 10 CFR 50.92(c). Document Control Desk PLA-7131 As demonstrated in the enclosed evaluation, the proposed amendment does not involve a significant hazard consideration and offsite and control room doses will remain within 10 CFR 50.67 limits.PPL requests that the approved amendment be issued to be effective immediately upon approval with the implementation to be completed within 60 days of the NRC approval date.Attachment 1 contains a markup of the proposed changes in the FSAR.Attachment 2 contains a markup of the proposed TS Bases changes.These changes have been reviewed by the SSES Plant Operations Review Committee and by the Susquehanna Review Committee.

In accordance with 10 CFR 50.91(b), PPL Susquehanna, LLC is providing the Commonwealth of Pennsylvania with a copy of this proposed License Amendment Request.There are no regulatory commitments associated with this response.If you have any questions or require additional information, please contact Mr. John L. Tripoli (570) 542-3100.I declare under penalty of perjury that the foregoing is true and correct.Executed on: 0 1 / -Zý /Z0) 4 Sincer J. ranke

Enclosure:

Evaluation of a License Amendment Request -Increase of the Allowable Gas Treatment System Maximum Flow Rates from Secondary Containment Attachments:

Attachment 1 -FSAR Markups (Units 1 and 2)Attachment 2 -Technical Specification Bases Markups (Units 1 and 2)Attachment 3 -Calculations

/ Control Room Habitability Envelope and Offsite Post LOCA Doses, and DBA-LOCA Total Control Room Dose (compact disk) Document Control Desk PLA-7131 Copy: NRC Region I Mr. J. Greives, NRC Sr. Resident Inspector Mr. J. Whited, NRC Project Manager Mr. L. Winker, PA DEP/BRP Enclosure to PLA-7131 PPL Susquehanna Units 1 and 2 Evaluation of a License Amendment Request -Increase of the Allowable Gas Treatment System Maximum Flow Rates from Secondary Containment

3.0 BACKGROUND

4.0 TECHNICAL

EVALUATION

5.0 REGULATORY

SAFETY ANALYSIS 5.1 NO SIGNIFICANT HAZARDS CONSIDERATION

5.2 APPLICABLE

REGULATORY REQUIREMENTS/CRITERIA

6.0 ENVIRONMENTAL

CONSIDERATIONS Enclosure to PLA-7131 Page 1 of 13 PPL EVALUATION

Subject:

Evaluation of a License Amendment Request -Increase of the Allowable Gas Treatment System Maximum Flow Rates from Secondary Containment

1.0 DESCRIPTION

Pursuant to 10 CFR 50.90, PPL Susquehanna, LLC (PPL) hereby requests approval of the following proposed amendments to the Susquehanna Steam Electric Station (SSES) Unit 1 and Unit 2, Facility Operating Licenses NPF-14 and NPF-22. The proposed amendments would change current licensing basis (CLB) for one postulated accident analysis described in the SSES Final Safety Analysis Report (FSAR), and as previously reviewed by the NRC in Reference 1.Specifically, this request would implement changes in calculations for the Control Room Habitability Envelope (CRHE) and Offsite Post LOCA Doses to permit higher allowable Standby Gas Treatment System (SGTS) exhaust maximum flow rates from secondary containment as described in the Technical Specification (TS) Bases.The proposed changes use new inputs/assumptions to the analysis of record that both reduce the allowed control structure unfiltered inleakage that is assumed in the Design Basis Accident (DBA) LOCA dose analysis and increase the assumed maximum allowable SGTS exhaust flow rate from secondary containment.

The proposed changes use the available margins by decreasing an assumption for control room inleakage to establish a new dose of record and increasing the design value for secondary containment flow rate expressed as a percentage of free air volume. These changes increase the allowable SGTS exhaust maximum flow rate for acceptance criteria used in surveillance requirements for the secondary containment.

The changes to the analysis of record will be implemented with the described changes to FSAR and TS Bases pages in Attachments 1 and 2. The TS Bases pages for Secondary Containment, SGTS maximum flow rate acceptance criteria of the Secondary Containment Surveillance Requirement (SR) 3.6.4.1.5 are proposed to be changed.

2.0 PROPOSED CHANGE

NRC review and approval is requested for the following two changes associated with the LOCA parameters for postulated accident analysis.2.1 This proposed change reduces the assumption for the amount of unfiltered control structure inleakage used in the DBA LOCA dose analysis from 500 to 350 cfm.Control structure unfiltered inleakage of 500 cfm is a LOCA parameter for postulated accident analysis which has been used to evaluate control room dose consequence.

The change reduces that assumed inleakage flow to 350 cfm. The other accident dose analyses will continue to use 500 cfm of unfiltered control structure inleakage.

The unfiltered control structure inleakage testing acceptance criterion with Control Room Emergency Outside Air Supply System (CREOASS) in operation (emergency mode) will be reduced to Enclosure to PLA-7131 Page 2 of 13 350 cfm for this change. The accident dose model descriptions in FSAR, Appendix 15B describe the Control Room Habitability Envelope (CRHE) Dose Model.The FSAR markups shown in Attachment 1 reflect these changes on pages 6.4-2, 15B-3, Table 15.6-21 and Table 15.6-22.2.2. The proposed activity increases the SGTS exhaust flow rate from 100 percent to 115 percent of the secondary containment free air volume per day.The existing FSAR describes the reactor building as designed to limit the inleakage to 100 percent of the secondary containment free volume per day at -1/4 in. wg, while operating the SGTS. Secondary Containment Ventilation Zones I, II, and III are described by the proposed changes to show that leak rate at post-accident pressure is 115 percent per day. This change increases this parameter to 115 percent, which requires an increased SGTS flow following drawdown.

These changes increase the maximum flow rate for acceptance criteria in TS Basis for Secondary Containment SR 3.6.4.1.5.

The markups shown in Attachment 1 to the FSAR from this change are on pages for Table 6.2-17 and 6.5-7, Table 15.6-13, Table 15.6-14, Table 15.6-16, Table 15.6-17, Table 15.6-18, Table 15.6-21, and Table 15.6-22. Markups are also shown in Attachment 2 for the Units 1 and 2 TS Bases pages B 3.6-89 and B 3.6-88, for Units I and 2, respectively.

The aggregate effects to the analysis of record from the changes that are described in the markups above are shown in the markup of Table 15.6-18, and Table 15.6-21. The proposed change to the Loss-of-Coolant Accident Summary of Offsite Doses is described by the following:

Existing FSAR Table 15.6-18: Rem / TEDE Regulatory Limit 2 Hour Site Boundary:

10.7 25 30 Day Low Population Zone (LPZ): 4.2 25 Revised FSAR Table 15.6-18: Rem / TEDE Regulatory Limit 2 Hour Site Boundary:

11.2 25 30 Day Low Population Zone (LPZ): 4.3 25 Enclosure to PLA-7131 Page 3 of 13 The proposed change to the Loss-of-Coolant Accident Summary of Control Room Operator Doses is described by the following:

Existing FSAR Table 15.6-21: Rem / TEDE Regulatory Limit 30 Day Operator Dose (Design):

4.79 5 30 Day Operator Dose (Realistic):

0.065 5 Revised FSAR Table 15.6-2 1: Rem / TEDE Regulatory Limit 30 Day Operator Dose (Design):

4.13 5 30 Day Operator Dose (Realistic):

0.050 5

3.0 BACKGROUND

SSES Event Number 49565, dated November 20, 2013, describes a secondary containment failure of drawdown surveillance testing performed on the Unit I Railroad Bay. This configuration was untested and the failure was due to a maximum flow rate for the TS SR 3.6.4.1.5 exceeding the allowable value that is in the TS Bases. Upon failure of the surveillance, secondary containment ventilation was realigned to a previously tested alignment and operability was restored.The discovery that the Unit 1 Railroad Bay was an untested configuration, and the subsequent failure of that configuration to pass its inleakage requirements during the event described above, have identified that very little margin exists for allowable secondary containment inleakage in that configuration.

The proposed changes use available margins in the current analysis of record to establish a new dose of record which will be used to implement the proposed higher allowable SGTS maximum flow rates from secondary containment.

As described above, PPL has determined this proposed change will be beneficial.

Many other licensees have inleakage limits substantially less conservative than the new limits proposed by SSES. For example, Limerick Generating Station has an allowable inleakge of 200 percent of free air volume, Fermi, Clinton and Duane Arnold have no limit on inleakage.

At SSES, the proposed new DBA LOCA analysis will remain well within the acceptance limits specified in 10 CFR 50.67. Although the LOCA offsite dose consequence at the 2 hour2.314815e-5 days <br />5.555556e-4 hours <br />3.306878e-6 weeks <br />7.61e-7 months <br /> site boundary and 30 day low population zone (LPZ) would increase in the analysis, the resulting decreasing available margin to the acceptance limits remains minimal at nominally 5 percent and 2 percent of the existing margins, respectively.

The Control Room 30 Day Operator dose is shown in the new analysis to decrease by more than 13 percent for design, and by 23 percent for the realistic case. Specifically, this change improves the dose margins for the control room operator where very little margin currently exists. Additionally, this change maintains the existing substantial DBA LOCA offsite dose margin, which allows SSES to remain aligned with others in the industry.PPL has concluded that the aggregate effect of these proposed changes reduce the calculated dose consequence to the control room and it yields only a minimal increase in calculated offsite dose that remains well below the regulatory limit. In this way, PPL has concluded that this change would reasonably support better margins for secondary containment operability at SSES.

Enclosure to PLA-7131 Page 4 of 13 4.0 TECHNICAL EVALUATION The implementation of alternative radiological source term (AST) for SSES Units 1 and 2 was evaluated by the NRC for issuance of SSES Units 1 and 2 Amendments 239 and 216, respectively.(')

More recent revision of calculations of the CRHE and Offsite Post LOCA Doses were evaluated by the NRC in a safety evaluation dated March 18, 2009.(2) The current analysis and the new analysis supporting this change are performed with the RADTRAD computer code that is approved for use at SSES by the NRC in Reference

1. The Figure in 4.2 below provides a graphic representation of the release paths for the LOCA and shows where the relevant change in dose consequences results from this proposal.Under LOCA conditions, habitability for the Control Structure is provided by the Control Room Emergency Outside Air Supply System (CREOASS).

This system provides habitability zone isolation and a positive pressure for the Control Room Habitability Envelope (CRHE). The CRHE is defined for SSES as six separate floors of the control building.

The CRHE dose acceptance criterion is 5 Rem total effective dose equivalent (TEDE).Sections 4.1 and 4.2 describe the details of the evaluation for the proposed changes.4.1 Unfiltered control structure inleakage:

The current design basis of SSES allows control structure unfiltered inleakage (see TS 5.5.14 and Reference 1).Habitability systems are designed to ensure habitability inside the control structure pressurization envelope during all normal and abnormal station operating conditions including the post LOCA requirements, in compliance with GDC 19 of 10 CFR 50, Appendix A and 10 CFR 50.67 for dose limits. There are no plant modifications that are performed to reflect this change in the assumed unfiltered control structure inleakage.

Consequently, there are not adverse impacts from this change to the ability for these systems to achieve their required design functions.

The 10 CFR 50.67 dose limit is 5 Rem TEDE to the control room operators.

The control structure unfiltered inleakage is leakage into the control structure habitability boundary that bypasses the emergency filtration system (CREOASS).

The DBA LOCA analysis assumes a value of control structure unfiltered inleakage that must be verified by testing per TS SR 3.7.3.4 and 5.5.14. There are no required changes to the TS pages that will result from this proposed amendment.

The control structure unfiltered inleakage only impacts the control room operator dose and does not affect the Exclusion Area Boundary (EAB) and Low Population Zone (LPZ) dose.The proposed 350 cfm of unidentified unfiltered inleakage is conservatively assumed in the proposed revised analysis.

The Control Room Envelope (CRE) inleakage testing data) NRC Safety Evaluation, "Issuance of Amendment re: Implementation of Alternative Radiological Source Term (TAC Nos. MC8730 and MC8731)," dated January 31, 2007 (ADAMS Accession No. ML070080301).

(2) NRC Safety Evaluation, "Issuance of Amendment re: Technical Specification Change to Technical Specification 3.6.1.3 to Increase the Maximum Allowable Secondary Containment Bypass Leakage Limit," dated March 18, 2009 (ADAMS Accession ML090500233).

Enclosure to PLA-7131 Page 5 of 13 report, dated October 4, 2011, from the tracer gas air inleakage tests performed at SSES during September 2011 conservatively bounds the use of 350 cfm. The data from the September 2011 tests show the inleakage at well under 350 cfm, at 89 scfm.The dose due to intake or infiltration of radioactive material contained in the effluent plume was re-assessed to show how the CRHE DBA LOCA dose is going to be reduced due to the unfiltered control structure inleakage.

The offsite doses are unchanged as a result of reducing the parameter for assumed unfiltered control structure inleakage.

This changed parameter also lowers the DBA LOCA dose consequences to the control room operator since less radioactive material is assumed to enter the control structure boundary.

Without considering the other effects to the analysis of record from increasing the reactor building leak rate and SGTS flow rates, the DBA LOCA dose to the control room operator would be reduced from 4.79 Rem TEDE to 4.04 Rem TEDE. The Realistic LOCA dose to the control room operator is reduced from 0.065 Rem TEDE to 0.050 Rem TEDE.The unfiltered control structure inleakage testing acceptance criterion with CREOASSS in operation (emergency mode) will be reduced to 350 cfm. Other accident dose analyses (e.g., Fuel and Equipment Handling Accident) will conservatively use 500 cfm of unfiltered control structure inleakage.

Therefore, other Chapter 15 events are not affected by the proposed change. It should be noted that other analyses (e.g., toxic chemical habitability analysis, control rod drop accident) do not rely on CREOASS actuation and normal control structure ventilation systems remain in operation.

In these scenarios the pressurization flow bypasses the CREOASS filtration system and all flow into the control structure is unfiltered.

This mode of operation has a separate test and acceptance criterion and is not affected by the proposed activity.

The only other analyses that rely on CREOASS actuation is the Equipment and Fuel Handling Accidents.

These analyses will still use a conservative control structure unfiltered inleakage value of 500 cfm which conservatively bounds the 350 cfm and the tested value.Equipment dose for qualification purposes can be assessed as reduced since less radioactive material is assumed to enter the control structure.

Therefore the existing values provided for in FSAR Table 3.11-1 are still conservatively bounding.In summary the only potential effect of this proposed change to the assumed value of control structure unfiltered inleakage is a reduction in the calculated control room operator dose.4.2 Reactor building leak rate (percent/day) increase: SGTS exhaust flow rate from secondary containment is a conservatively assumed input in the DBA LOCA dose analysis.

The change will increase this reactor building leak rate parameter from 100 percent/day of the secondary containment free air volume to 115 percent/day.

The Figure 1 shows the release paths used in the offsite and CRHE Dose Calculation, and shows which portions of those release paths are affected by this change.

Enclosure to PLA-7131 Page 6 of 13 ENVIRONMENT Figure 1: Release Paths in the Offsite and CRHE Dose Calculation (The change effects are shown in the clouded regions)

Enclosure to PLA-7131 Page 7 of 13 Technical Specifications Bases Surveillance Requirement 3.6.4.1.5 specifies the SGTS exhaust flow rate from secondary containment.

Flow rates would change to the following values: Railroad Bay Aligned to New Existing Secondary Containment: (cfri) (cf1)Zones I, II and III 4575 4000 Zones I and III 3275 2885 Zones II and III 3350 2960 Railroad Bay Aligned as New Existing a No Zone: (cfin) (cfm)Zones I, II and III 4475 3910 Zones I and Ill 3175 2800 Zones I and III 3275 2875 The new offsite doses are 11.2 Rem TEDE for the 2 Hour EAB, 4.3 Rem TEDE for the LPZ dose, and the new control room operator dose is 4.125 Rem TEDE. The Realistic LOCA activity releases change slightly but the dose consequences to the public and control room operator will remain the same. The new offsite and control room operator doses are less than 10 percent of the available margin to the applicable regulatory limit for the aggregate change.Due to the assumed increase in release rate, the dose consequences will increase for the EAB, LPZ and control room operator.

As stated in the FSAR Section 6.5.1.1.1, the only design basis accidents that SGTS is required to mitigate are the DBA LOCA (FSAR Section 15.6.5) and Fuel and Equipment Handling Accident (FSAR Section 15.7.4). The Main Steam Line Break Outside Containment and the Control Rod Drop Accident (the two other accidents discussed in current licensing basis, and in Reference

1) do not credit SGTS operation for dose contributions from their respective activity release pathways.

The Fuel and Equipment Handling Accident assumes a non-mechanistic activity release over a two hour period. So even though SGTS is credited for this accident, the SGTS exhaust flow rate is not representative of the physical system and a specific SGTS exhaust flow rate that correlates to the actual system performance is not assumed in the analysis.

Therefore, the only design basis dose analysis that is impacted is the DBA LOCA analysis.

The only impact to DBA LOCA dose analysis will be those that have been summarized above.The change is within the capacity of SGTS so there is not an impact to SGTS equipment due to the increased flow rate expected.

The other impact of the proposed change is the ability of SGTS to drawdown the reactor to -1/4 in. wg post-accident.

Per TS SR 3.6.4.1.4, the secondary containment

-1/4 in. wg pressure must be re-established within 5 minutes, which bounds the 10 minute drawdown assumed in the dose analysis.

The existing SGTS drawdown analysis verifies that the drawdown time can be achieved.

Additionally, the TS surveillance requirement will also verify that the drawdown time will be met. Plant operating experience has demonstrated that even with SGTS exhaust flow rate from Enclosure to PLA-7131 Page 8 of 13 secondary containment in excess of the limits (5033 cfm for 3 zone isolation, Railroad Bay as a No-Zone), the drawdown time was achieved in 91 seconds which still has substantial margin to the 5 minute TS Bases limit. The drawdown time is an assumed value in the DBA LOCA dose consequence analysis that is not changing.The final impact to consider due to the proposed change is the additional equipment dose from the additional radioactive material on the SGTS filter train. The SGTS exhaust flow rate from secondary containment that is assumed for DBA LOCA analysis affects equipment on elevation 806' of the control structure.

The potential for Equipment Qualification impact is assessed.

The additional equipment dose was evaluated in the radiation qualified life calculation.

Therefore, all required equipment on elevation 806' of the control structure, where the SGTS filters are located, is appropriately qualified for the additional equipment dose, which is less than a 2 percent increase over that in the existing analysis.The mission doses for control room operators are provided in FSAR Table 18.1-4, Vital Areas. This table lists the resulting areas considered vital for post-accident operations at SSES. It provides the total integrated doses to personnel in continuously occupied vital areas. This table is for contained sources only and therefore is not impacted by the proposed change, which impacts airborne dose rates and not the dose from contained sources.Regulatory Guide 1.183, "Alternative Radiological Source Terms for Evaluating Design Basis Accidents at Nuclear Power Reactors" provides guidance for implementation of 10 CFR 50.67, including assumptions and methods that are acceptable to the NRC for performing design basis radiological analyses using an AST. Regulatory Guide 1.183, Table 6, for the DBA LOCA, provides for the dose limit at the EAB (worst 2 hour2.314815e-5 days <br />5.555556e-4 hours <br />3.306878e-6 weeks <br />7.61e-7 months <br />) and LPZ of 25 Rem TEDE. The control room dose limit is 5 Rem TEDE per 10 CFR 50.67.The proposed changes result in dose consequences that would still remain within the regulatory limits of the current licensing basis. The aggregate of effects from implementing both of the evaluated changes in the items 4.1 and 4.2 above are also within 10 percent of the available margin to the regulatory and current licensing basis limits for the existing offsite and CRHE dose analysis for LOCA. Consequently, this proposed change represents only a minimal increase in consequences of the previously evaluated conditions in the FSAR.

Enclosure to PLA-7131 Page 9 of 13 5.0 REGULATORY SAFETY ANALYSIS 5.1 No Significant Hazards Consideration PPL Susquehanna, LLC (PPL) proposes amendments to Facility Operating License Nos.NPF-14 and NPF-22, for the Susquehanna Steam Electric Station Units 1 and 2 respectively.

The proposed amendments would change the current licensing basis that is described in the Final Safety Analysis Report for the use of certain inputs and assumptions in the design basis accident (DBA) radiological consequence analysis of record. The proposed amendments would change this analysis for a postulated DBA Loss of Coolant Accident (LOCA). Approval of the amendments will implement changes in the use of maximum acceptance criteria for Standby Gas Treatment System (SGTS) exhaust flow rate from secondary containment.

PPL Susquehanna, LLC has evaluated whether or not a significant hazards consideration is involved with the proposed change by focusing on the three standards set forth in 10 CFR 50.92, "Issuance of amendment," as discussed below: 1. Does the proposed amendment involve a significant increase in the probability or consequences of an accident previously evaluated?

Response:

No.The proposed change to the current licensing basis (CLB) is an analysis change and does not involve any physical changes. No new equipment is being installed nor is any installed equipment being operated in a new or different manner. As such, the proposed change does not increase the probability of an accident previously evaluated.

Based on the revised analysis, the proposed change does revise the performance requirement; however, the proposed change does not involve a revision to the parameters or conditions that could contribute to the initiation of a DBA discussed in Chapter 15 of the FSAR.The new DBA LOCA analysis with an increased SGTS exhaust flow rate from secondary containment is well within the acceptance limits specified in 10 CFR 50.67. The DBA LOCA dose analysis results in a 2 hour2.314815e-5 days <br />5.555556e-4 hours <br />3.306878e-6 weeks <br />7.61e-7 months <br /> EAB dose of 11.2 Rem TEDE and LPZ dose of 4.3 Rem TEDE. The calculated control room dose of 4.125 Rem TEDE is well within the acceptance limits of 10 CFR 50.67.The calculated doses are well within the acceptance limits and, therefore, do not represent a significant increase in consequences of a DBA LOCA.Plant specific radiological analysis has been performed using the proposed changes. This analysis demonstrates that the dose consequences meet the regulatory guidance provided for use with the Alternative Source Term (AST), and the offsite doses are well within acceptable limits (10 CFR 50.67, Regulatory Guide (RG) 1.183, and Standard Review Plan Section (SRP)15.0.1). The aggregate changes are within 10 percent of the margin to these limits.

Enclosure to PLA-7131 Page 10 of 13 Therefore, the proposed amendment does not result in a significant increase in the consequences of any previously evaluated accident.2. Does the proposed amendment create the possibility of a new or different kind of accident from any accident previously evaluated?

Response:

No.The proposed changes do not involve a physical alteration of the plant. No new or different types of equipment are installed and there will there be no changes in methods governing normal plant operation.

The potential for the loss of plant systems or equipment to mitigate the effects of an accident is not altered. The proposed changes do not require any new operator response or introduce any new opportunities for operator error not previously considered.

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

3. Does the proposed amendment involve a significant reduction in a margin of safety?Response:

No.The results of the revised accident analysis are subject to the acceptance criteria in 10 CFR 50.67. The revised allowable standby gas treatment system flow rate from secondary containment and input parameters to the analysis have been conservatively used. Safety margins and analytical conservatisms have been evaluated and have been found acceptable.

The analyzed LOCA event has been carefully selected and margin has been retained to ensure that the analysis adequately bounds postulated event scenarios.

The dose consequences of the limiting event is within the acceptance criteria presented in 10 CFR 50.67, RG 1.183, and SRP 15.0.1. The effect of the revision to the analysis, and changes to the FSAR and TS Bases that result has been analyzed and the analysis shows doses from the pertinent design basis accident have been found to remain within regulatory limits. The change continues to ensure that the doses at the exclusion area and low population zone boundaries, as well as the control room, are within the corresponding regulatory limits. Therefore, the proposed change does not involve a significant reduction in a margin of safety.Based on the above, PPL Susquehanna concludes that the proposed amendment presents no significant hazards consideration under the standards set forth in 10 CFR 50.92(c), and, accordingly, a finding of "no significant hazards consideration" is justified.

Enclosure to PLA-7131 Page 11 of 13 5.2 Applicable Regulatory Requirements/Criteria SSES FSAR Sections 3.1 and 3.13 provide detailed discussion of SSES compliance with the applicable regulatory requirements and guidance.

The proposed license amendment:

a) Does not alter the design or function of any reactivity control system;b) Does not result in any change in the qualifications of any component; and c) Does not result in the reclassification of any component's status in the areas of shared, safety related, independent, redundant, and physically or electrically separated.

10 CFR 50.67, "Accident source term," establishes acceptable radiation dose limits resulting from design basis accidents for an individual located at the exclusion area boundary or low population zone, and for occupants of the control room. These analyses performed by PPL demonstrate that the calculated radiological consequences of a design basis LOCA meet the radiation dose limits specified in 10 CFR 50.67.5.2.1 General Design Criteria: The following applicable General Design Criteria (GDC) for the primary containment require that the primary containment be designed and maintained so that offsite doses remain below the regulatory guidelines and the containment penetrations can be isolated and tested: GDC 16 -Containment Design GDC 16 "Containment Design" requires that secondary containment and the associated safety systems are designed and maintained so that offsite doses, which could result from postulated design basis accidents, remain below the guideline values stated in 10 CFR 100 when calculated by the methods of Regulatory Guide 1.3 (Rev. 2, 6/74).The proposed change will maintain offsite and control room doses below regulatory limits.The proposed change will maintain the Primary Containment as an essentially leak tight barrier against the uncontrolled release of radioactivity to the environment since the overall Primary Containment leak rate (1%/day) is not changing.GDC 50 -Containment Design Basis The proposed change does not change the containment design pressure or temperature to any postulated loss of coolant accident.

Additionally, the proposed change does not alter the total containment leak rate of 1.0%/day for any postulated loss of coolant accident.Based on this information, the proposed change does not impact the commitment to this GDC.GDC 54 -Piping Systems Penetrating Containment The proposed change does not impact the isolation capability of primary containment penetrations.

Therefore, the proposed change does not impact this commitment.

Enclosure to PLA-7131 Page 12 of 13 GDC 55 -Reactor Coolant Pressure Boundary Penetrating Containment GDC 56 -Primary Containment Isolation The proposed change does not impact the isolation capability or valve arrangement of pipes penetrating primary containment.

The following applicable General Design Criteria (GDC) for the Standby Gas Treatment System requires that containment atmosphere cleanup systems reduce the amount of radioactive material released to the environment following a postulated design basis accident.GDC 41 -Containment Atmosphere Cleanup GDC 42 -Inspection of Containment Atmosphere Cleanup Systems GDC 43 -Testing of Containment Atmosphere Cleanup Systems As the proposed changes do not modify any SSES structures, systems or component design functions, and the changes are limited to analyses associated with the dose consequences of the postulated accidents, there can be no changes that can impact the requirements of GDC 41, 42 and 43.5.2.2 Applicable Regulatory Guides The following applicable Regulatory Guides are for the primary containment and the offsite and control room dose calculations:

Regulatory Guide 1.163 is titled "Performance Based Containment Leak Test Program." The proposed change will not change the testing frequency or the testing method for primary containment penetrations.

Therefore, the proposed change does not impact this requirement commitment.

Regulatory Guides 1.145, 1.183 and 1.194: The Regulatory Guide 1.145 is titled "Atmospheric Dispersion Models for Potential Accident Consequence Assessment at Nuclear Power Plants." Regulatory Guide 1.183 is titled "Alternative Radiological Source Terms for Evaluating Design Basis Activities at Nuclear Power Reactors" and Regulatory Guide 1.194 is titled "Atmospheric Relative Concentrations for Control Room Radiological Assessments at Nuclear Power Plants." The dose analysis and atmospheric dispersion values were calculated using the guidance presented in these Regulatory Guides as described in PPL's AST submittal dated October 13, 2005 and approved by the NRC in SER dated January 31, 2007.Conformance with GDC 16, 50, 54, 55 and 56, as well as conformance with Regulatory Guides 1.163, 1.145, 1.183 and 1.194 are not affected by these proposed changes.Thus, the proposed changes do not change the conformance with the above General Design Criteria and regulatory guidance.

Enclosure to PLA-7131 Page 13 of 13 Conclusion Based on the analysis provided in Section 4.0: (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 10 CFR 51.22(c)(9) identifies certain licensing and regulatory actions, which are eligible for categorical exclusion from the requirement to perform an environmental assessment.

A proposed amendment to an operating license for a facility does not require an environmental assessment if operation of the facility in accordance with the proposed amendment would not: (1) involve a significant hazards consideration; (2) result in a significant change in the types or significant increase in the amounts of any effluents that may be released offsite; or (3) result in a significant increase in individual or cumulative occupational radiation exposure.

PPL Susquehanna, LLC has evaluated the proposed change and has determined the analysis demonstrates that the consequences from a DBA LOCA will remain well within acceptance limits, and that the proposed change meets the eligibility criteria for categorical exclusion set forth in 10 CFR 51.22(c)(9).

Accordingly, pursuant to 10 CFR 51.22(b), no environmental impact statement or environmental assessment needs to be prepared in connection with issuance of the amendment.

The basis for this determination, using the above criteria, follows: Basis: The proposed change does not involve any physical alteration of the plant (no new or different type of equipment will be installed) or change in methods governing normal plant operation.

Therefore, no significant change in the types or significant increase in the amounts of any effluents that may be released offsite can occur. The new DBA LOCA analysis is well within the acceptance limits specified in 10 CFR 50.67. Although the LOCA offsite dose consequence at the 2 hour2.314815e-5 days <br />5.555556e-4 hours <br />3.306878e-6 weeks <br />7.61e-7 months <br /> site boundary and 30 day low population zone (LPZ) would increase in the analysis, the resulting decrease in available margin to the acceptance limits remains minimal at nominally 5 percent and 2 percent of the existing margins, respectively.

The Control Room 30 Day Operator dose is shown in the new analysis to decrease by more than 13 percent for design, and by 23 percent for the realistic case. The DBA LOCA dose analysis results in a 2 hours2.314815e-5 days <br />5.555556e-4 hours <br />3.306878e-6 weeks <br />7.61e-7 months <br /> EAB dose of 11.2 Rem TEDE and LPZ dose of 4.3 Rem TEDE. The calculated control room dose of 4.125 Rem TEDE is well within the acceptance limits of 10 CFR 50.67. The calculated control room doses are well within the acceptance limits and, therefore, do not represent a significant increase in consequences of a DBA LOCA. No new significant increase in individual or cumulative occupational radiation exposure can occur, and as demonstrated in the "No Significant Hazards Consideration" evaluation, the proposed amendment does not involve a significant hazards consideration.

Attachment 1 to PLA-7131 FSAR Markups (Units 1 and 2)

SSES-FSAR Text Rev. 74 6.2.3 SECONDARY CONTAINMENT FUNCTIONAL DESIGN The secondary containment comprises the exterior structure of reactor building and the interior walls and floors that separate the three ventilation zones.Zones I and II are the portions of the reactor building below elevation 779 ft. 1 in. surrounding the Unit 1 and Unit 2 primary containments, respectively.

Zone III consists of the portion of the reactor buildings above elevation 779 ft. I in. with the exception of the HVAC equipment rooms which are not part of the secondary containment.

The secondary containment houses the refueling and reactor servicing equipment, the new and spent fuel storage facilities, and other reactor auxiliary or service equipment, including the Reactor Core Isolation Cooling System, Reactor Water Cleanup System, Standby Liquid Control System, Control Rod Drive System equipment, the Emergency Core Cooling System, and electrical equipment components.

6.2.3.1 Desigqn Bases The functional capability of the ventilation system to maintain negative pressure in the secondary containment with respect to outdoors is discussed in Subsections 6.5.1.1 and 9.4.2.The conditions that could exist following a LOCA require the establishment of a method of controlling the leakage from the primary into the secondary containment.

6.2.3.2 System Design 6.2.3.2.1 Secondary Containment Desiqn The reactor building is designed and constructed in accordance with the design criteria outlined in Chapter 3. The base mat, floor slabs and exterior walls below the refueling floor are constructed of reinforced concrete.

Above the refueling floor at elevation 818 ft. 1 in., the building consists of a structural steel frame supporting an insulated metal roof deck and insulated siding wall panels.Joints in the superstructure paneling are designed to ensure leaktightness.

Penetrations of the reactor building are designed with leakage characteristics consistent with leakage requirements of the entire building.

The reactor building is designed to limit the inleakage to 1,gop P .neRt1 15 Irerdent of the secondary containment free volume per day at -% in. wg, while operating the SGTS. The building structure above the refueling floor is also designed to contain a negative interior pressure of 0.25 in. wg.Following a loss-of-coolant accident, all affected volumes of the secondary containment will be maintained at a negative pressure of 0.25 in. w.g. All these volumes are identified on Figures 6.2-24, 6.2-25, 6.2-26, 6.2-27, 6.2-28, 6.2-29, 6.2-30, 6.2-31, 6.2-32, 6.2-33, 6.2-34, FSAR Rev. 65 6.2-38 SSES-FSAR Text Rev. 74 6.2-35, 6.2-36, 6.2-37, 6.2-38, 6.2-39, 6.2-40, 6.2-41, 6.2-42, and 6.2-43 as Ventilation Zones I, II and Ill.An analysis of the post LOCA pressure transient in the secondary containment has been performed to determine the length of time following LOCA signal that the pressure in the secondary containment would exceed -A in. wg. The analysis assumed that the normal ventilation system was operating at the design pressure of -A in. w.g. until the E.S.F. signal isolated the system and initiated SGTS startup. An inleakage rate of 400%-115%

of secondary containment per day was used. A single failure of one SGTS train was assumed as well as a loss of offsite power to maximize the drawdown time. Heat loads from operating equipment and the heat transferred through the drywell head were considered.

Each SGTS fan has a rated capacity of 10,500 CFM at a 17 in. w.g. pressure.

Figure 6.2-60 shows the secondary containment pressure vs time for the drawdown under worst conditions.

The secondary containment pressure recovers to -% in. w.g. within 5 minutes. The completion of the leakage path resulting from the activity release mechanisms inside the containment, leakage through the primary containment and possible leakage through the secondary containment would require a significantly greater period of time than would exist until the -1/4 in. w.g. was restored.Entrance to the reactor building is through the turbine building with air locks provided for separation.

Access doors between building ventilation zones and into the control structure are provided with airlocks.

Secondary containment access doors which are not provided with airlocks are administratively controlled to maintain secondary containment integrity.

The railroad access shaft, provided in Unit I only, is accessible to Zones I and Ill through access hatches that are normally kept closed and will not be opened without proper controls to maintain secondary containment integrity during normal plant operation.

Ventilation supply and return ducting to the railroad access shaft is provided with manual isolation dampers to provide for opening the exterior railroad access door after closing the dampers, thus converting to an airlock and retaining secondary containment integrity.

Operation of these dampers and the railroad access doors and hatches is administratively controlled.

Doors within the secondary containment may be used for personnel ingress and egress during normal plant operation.

The truck bay is part of Zone I1. The truck bay access hatch will be normally closed. Opening of this hatch and the truck bay door (No. 102) will be administratively controlled.

The boundaries of the three zones of the secondary containment are shown on Figures 6.2-24, 6.2-25, 6.2-26, 6.2-27, 6.2-28, 6.2-29, 6.2-30, 6.2-31, 6.2-32, 6.2-33, 6.2-34, 6.2-35, 6.2-36, 6.2-37, 6.2-38, 6.2-39, 6.2-40, 6.2-41, 6.2-42 and 6.2-43.The secondary containment design data can be found in Table 6.2-17.A simplified air flow diagram for the secondary containment normal plant operation is shown on Figure 6.2-53. Figure 6.2-52 shows the simplified air flow diagram when Zone I or II and Zone Ill are isolated.

An air flow diagram for Zone Ill isolation is shown on Figure 6.2-54.6.2.3.2.2 Secondary Containment Isolation System Isolation dampers and the plant protection signals that activate the secondary containment isolation system are described in Subsection 9.4.2.1.3.

FSAR Rev. 65 6.2-39 SSES-FSAR TABLE 6.2-17 INFORMATION FOR THE SSES SECONDARY CONTAINMENT I 1. Secondary Containment Ventilation Zones 1, II and III A. Approximate Free Volume, ft" -Zone I 1,488,600 Zone II 1,598,600 Zone III 2,668,400 B. Pressure, inches of water, gage 1. Normal Operation

-2. Post-accident

-1/4/C. Leak Rate at Post-Accident Pressure 0 QO 115% per day D. Exhaust Fans -common 1. Number- 2 2. Type -Centrifugal, SISW E. Filters -common 1. Number- 2 2. Type- prefilter, HEPA, charcoal, HEPA II. Transient Analysis A. Initial Conditions

1. Pressure, -% in. wq 2. Temperature

-104°F 3. Outside Air Temperature

-92 0 F 4. Thickness of Secondary Containment Wall- 36 in.5. Thickness of Primary Containment Wall -72 in.B. Thermal Characteristics

1. Primary Containment Wall a. Thermal Conductivity, BtuLhr-ft-°F

-.5 b. Thermal Capacitance, Btu/ft 3 -OF -25 2. Secondary Containment Wall a. Thermal Conductivity, Btu/hr-ft-°F

-.5 b. Thermal Capacitance, Bturft 3-OF -25 3. Heat Transfer Coefficients

a. Primary Containment Atmosphere to Primary Containment Wall, Btu/hr-ft 2 --1.46 b. Primary Containment Wall to Secondary Containment Atmosphere, Btu/hr-ft 2 -OF -1.46 c. Secondary Containment Wall to Secondary Containment Atmosphere, Btu/hr-ft 2 -OF -1.46 d. Primary Containment Emissivity, Btu/hr-ft 2-_F -.9 e. Secondary Containment Emissivity, Btu/hr-ft 2-_F -.9 Rev. 35, 07/84 Page 1 of 1 SSES-FSAR Text Rev. 60 h) Radiation monitors, and smoke detectors continuously monitor the outside air at the control structure envelope outside air intakes. The detection of high radiation, smoke is alarmed in the Control Room and related protection functions are simultaneously initiated for high radiation.

The operator may isolate the control structure on smoke alarm at his discretion.

i) In the event of a Control Room evacuation, an Alternate Control Structure HVAC Control Panel provides for manual operation of the required HVAC components from outside the Control Room.6.4.2 SYSTEM DESIGN 6.4.2.1 Control Structure Envelope Habitability system boundaries for Susquehanna SES is the control structure envelope.a) An independent HVAC system is provided for the Control Room area. This includes: Control Room, TSC, OSC, kitchen, toilet and locker, office, conference room, document Control Room, electrical room, vestibule and storage space. All areas on plan floor EL 728'-0" and 741'-0" are served by this system. A detailed description of this redundant system is provided in Subsection 9.4.1.b) Two independent HVAC systems are provided for the remaining areas. One system serves the computer room, lower relay rooms, computer maintenance room, office, and UPS rooms. The other system serves the lower cable spreading room, upper relay rooms, upper cable spreading rooms, electrician's office, battery rooms, cold instrument repair shop, equipment rooms, and HV equipment room. Each of these systems is described in Subsection 9.4.1.There are eleven exterior doors in the control structure envelope.

These doors are gasketed to minimize leakage and will be tested to 1/8" w.g. differential pressure to assure tightness.

Another leakage path across the ventilation barrier between the control structure envelope and outside environment is through the isolation damper blades. Isolation dampers are listed in Table 6.4-1.Tests on the isolation dampers indicate a leakage rate as shown on Table 6.4-1 at test differential pressures ranging from 3 to 21 in. wg. The analysis for Control Room habitability given in Chapter 15 and Appendix 15B assumed a leakage of 10 cfm of outside air for ingress/egress and an additional 50 --cfm of unidentified, unfiltered inleakage to the Control Structure Envelope (Note that all analyses other than the DBA LOCA analysis assumes an unidentified inleakaqe of 500 cfm). Makeup air to the envelope is also filtered, so the makeup air to the Control Structure Envelope would not be at outside air concentrations.

The environment of the Control Structure Envelope is maintained to ensure the integrity of the contained safety related controls and equipment during all operating conditions.

Technical Specification

3.7.3 discusses

maintaining a positive pressure of >0.125 inches water gauge relative to the outside atmosphere during the pressurization mode of operation.

FSAR Rev. 64 6.4-2 SSES-FSAR TABLE 6.5-7 ZONE VOLUMES AND THEIR ESTIMATED RECIRCULATION AIRFLOW RATES ZONE SUBSYSTEM FLOW ESTIMATED DESIGN AIR FLOW RATES(3)VENT VOLUME PATH ZONE FT 3 (ASSOCIATED FANS) MODE A MODE B MODE C MODE D NO.(1) (2) (4) (5) (6) (7)22700240 I 1,488,600 Supply (1V202) 28530 202804O I 28530 2A020" R (1V205,1V206)

_____ 2140 Return293

__29730 .21470. " 30000 200 11 1598600 Supply (2V202)29820 21i780 Return (2V205,2V206) 31100 23060 111 2,68,00Suply8) 5440.0 So0o() 3GQQ0 891400 III 2,668,400 Supply(8)

.51150 49770 36350 :80870 (1V217,2V217 53.00 52000 "8500 8 JReturn 1V213,2V213) 53280 51900 38480 83000 (1) Section 9.4.2.1 defines the boundaries of the ventilation system.(2) Associated fans are listed to identify the zone supply and return subsystems but are assumed not to operate. Only a single OV201A or B recirculation fan plus a single OV109A or OVI09B SGTS fan is assumed to operate in the recirculation modes.(3) Differences between recirculation return air and supply air flows represent the maximum estimated design air flows exhausted through the SGTS system (OV109) in order to maintain negative pressure In the affected zone(s), assuming in leakage of one volume of the affected zone(s) per day.(4) Isolation of Zone land III (5) Isolation of Zone II and III (6) Isolation of zone I, II and III (7) Isolation of Zone III only (8) Separate ducting is provided from the recirculation system (OV201) discharge plenum to the common refueling floor. It is not connected to the normal Zone III supply fan system (1V212& 2V212).Rev. 54,10/99 Page I of I SSES-FSAR INo, Changes5 n This Page Text Rev. 55 -For Information-I APPENDIX 15B ACCIDENT DOSE MODEL DESCRIPTIONS 15B.1 OFFSITE DOSE MODEL This discussion describes the models used to calculate offsite radiological doses that would result from releases of radioactivity due to various postulated accidents.

The following assumptions are used for offsite dose evaluations:

a) The direct dose contribution offsite from any post-accident onsite source point is negligible compared with the direct dose due to immersion in the post-accident effluent cloud.b) All radioactivity releases are treated as ground level releases regardless of the point of discharge.

c) Isotopic data including decay constants and dose conversion factors are listed in Table 15B-2. The isotopic data listed in Table 15B-2 is obtained from the RADTRAD (Reference 15B-4) computer code which is used to evaluate the radiological consequences of accidents.

These dose conversion factors are used to calculate immersion and inhalation doses and are derived from Federal Guidance Report Nos. 11 and 12 (References 15B-6 and 15B-7).The acceptance criteria for the offsite doses is in terms of Rem TEDE. The determination of TEDE doses takes into account the committed effective dose equivalent (CEDE) dose resulting from the inhalation of airborne activity (the long-term dose accumulation in the various organs)as well as the effective dose equivalent (EDE) dose resulting from immersion in the cloud of activity.

The definition of these doses is given in 1OCFR20.1003.

The models used to evaluate offsite doses for accidents are as follows: Immersion Dose (Effective Dose Equivalent)

Assuming a semi-infinite cloud, the immersion doses are calculated using the equation: Dim = DCFi Rij (X/Q)j (EQ. 15B-1)where: Dim = Immersion (EDE) dose (rem)DCFI = EDE dose conversion factor for isotope i (rem-m 3/Ci-sec)RUj = Amount of isotope i released during time period j (Ci)(x/Q)j = Atmospheric dispersion factor during time period j (sec/mr 3)FSAR Rev. 64 1513-1 SSES-FSAR No. Changes on'This Page Text Rev. 55 -For Information-Inhalation Dose (Committed Effective Dose Equivalent)

The CEDE doses are calculated using the equation: DcmE. DCFi Rij (BR)j Q /Q )j (EQ. 15B-2)where: DCEDE = CEDE dose (rem)DCFj = CEDE dose conversion factor (rem per curie inhaled) for isotope i Rij = Amount of isotope i released during time period j (Ci)(BR)j = Breathing rate during time period j (m 3 lsec)(X/Q)j = Atmospheric dispersion factor during time period j (sec/ms)Total Dose (Total Effective Dose Equivalent)

The TEDE doses are the sum of the EDE and the CEDE doses.15B.2 CONTROL ROOM HABITABILITY ENVELOPE DOSE MODEL This discussion describes the models used to calculate control room habitability envelope (CRHE) radiological doses that would result from releases of radioactivity due to various postulated accidents.

The acceptance criteria for CRHE doses is in terms of Rem TEDE. The determination of TEDE doses takes into account the committed effective dose equivalent (CEDE) dose resulting from the inhalation of airborne activity (the long-term dose accumulation in the various organs) as well as the effective dose equivalent (EDE) dose resulting from immersion in the cloud of activity.

The definition of these doses is given in 10CFR20.1003.

The total CRHE TEDE dose is the sum of the EDE and the CEDE doses for all CRHE post-accident radiation sources.The design basis for the CRHE is to provide adequate radiation protection to permit access to and occupancy of the control room under accident conditions without personnel receiving radiation exposures in excess of 5 rem total effective dose equivalent (TEDE) for the duration of the accident.

This basis is consistent with 10CFR50.67.

Radiation protection for the CRHE is provided by radiation shielding and by an emergency ventilation system.The CRHE radiation shielding is designed to reduce gamma radiation shine from both normal and post-accident radiation sources to levels consistent with the requirements of 10CFR20 or 1 OCFR50.67.

The post-accident emergency ventilation system is designed to preclude entrance of unfiltered air to the control room and to maintain outleakage of air from this zone with respect to other plant ventilation zones and the air outside the plant.Details of control room emergency ventilation system design and instrumentation are discussed in Subsection 9.4.1 and Section 6.4.FSAR Rev. 64 15B-2 SSES-FSAR Text Rev. 55 During emergency operation, 5810 +/- 10% cfm filtered outside air is supplied to the control structure.

In addition to the intake of air through the filter system, some air will enter the control building due to ingress/egress of personnel and via infiltration from other identified leakage paths. An infiltration rate of 10 scfm has been assumed for ingress/egress and 500-4m350 cfm for the other unidentified leakage for the DBA LOCA analysis.

Note that all analyses other than the DBA LOCA analysis assume an unidentified, unfiltered inleakage of 500 cfm. Credit for operation of the CRHE emergency ventilation system is only taken for the DBA-LOCA and fuel handling/equipment handling accidents.

Under accident conditions, radiation doses to control room personnel may result from several sources. While in the control room, personnel are exposed to beta and gamma radiation from gaseous fission products that enter after an accident via the ventilation system or from unfiltered air entering the control room. In addition, personnel may be subject to gamma shine dose from fission products in the containment and reactor building, from contained system sources and from fission products in the atmosphere outside the control room.To evaluate the capability of the control room ventilation system and radiation shielding to keep doses within the specified criteria, control room doses are evaluated for each of these dose contributors.

This analysis includes control room doses from the following radiation sources: Contamination of the control room atmosphere by the intake or infiltration of the radioactive material contained in the radioactive plume released from the facility,* Radiation shine from the external radioactive plume released from the facility, Contamination of the control room atmosphere by the intake or infiltration of airborne radioactive material from areas and structures adjacent to the control room envelope, Radiation shine from radioactive material in buildings adjacent to the control structure; includes containment, reactor building and turbine building, Radiation shine from radioactive material in systems and components inside or external to the control room envelope, e.g., piping, components and radioactive material buildup in HVAC filters.The short term accident X/Q's for the SSES Control Room Habitability Envelope (CRHE) were calculated using the methodology provided in NUREG/CR-6331

-ARCON96 (Reference 15B-1)and Regulatory Guide 1.194 (Reference 155B-3). The ARCON96 code uses hourly meteorological data and recently developed methods for estimating X/Q's in the vicinity of buildings to calculate relative concentrations at control room air intakes that would be exceeded no more than five percent of the time. These concentrations are calculated for averaging periods ranging from one hour to 30 days in duration.The specific locations requiring ARCON96 X/Qs for use in the applicable radiological evaluations were: Turbine Building Unit 1 exhaust vent.Turbine Building Unit 2 exhaust vent.Standby Gas Treatment System exhaust vent.FSAR Rev. 64 1513-3 SSES-FSAR Replace data with Insert 3 and 4 Table Rev. 56 TABLE 15.6-13 LOSS OF COOLANT ACCIDENT ACTIVITY AIRBORNE IN REACTOR BUILDING (curies)(Design Basis Accident)Isotop Reactor Building Airborne Activity As a Function of Time Post-Accident

_______ ________ ________ _______ (curies)____

____0.1667 hr 0.5 hr 1 hr 2 hr 8 hr 24 hr 96 hr 240 hr 480 hr 720 hr Co-58 4 64R-02~ 3ý63r,-04.

14R4-01 1 369-0151R-4 8 -1~ 4 F-04 I- OIE-94 83,0 Co-60 P 2y,-02 4-Or. 94 P-77FO49 7.40ra 02 269r 04 a 20r-as &5fi-05 Aar_-0 Kr-85 6,2369.

,4EQ 8S'- 6-4EO Q42,O3 ~W 3.6 .3439&,93 X.?E9 R Kr-85r ~4-i459'4-4.24F92 4 59tE'.0 64E43 6 A4PO 942-76-49&

-24&49E. 9.09S, 00 GAG, 9O 4,09-00 Kr-87 .290 3 :&E 45g: ý1460 5.626,03 424 S-2 626 0 r*Q Q-~0 Q0Q=-OQQ QO0F-QQ 0 ,gF-go Rb-86 I ISF- 1, .6 S 4 3AGý 4 MR-4 I -QrS04- 4 94r-00 4 74R 02 A446 -03 2439-O3 :440 Ora Sr-09 ;.8F!0.4-6.65E-9O2 Z2 92. 4.23C-04 9-629-044

-.2O-49 4-0E94 Q.gn Q QQ Sr-9O 6&7om.14 &4~.9964 J4

!Egigp 6 sa-04 .004. 0 Q:O-O QQ-Q ~409-09 ~ 00F.40 Yr-902 &~S'47r 0 4 7~ 40;5S 0 a s8~-O-g a~ ~O jar, ~ g2 :27)r )2 a 7. 6r O R V-91 s3.-1 MR I r 57.10 .! *4Or=kQ* R~S'O 97R, i-0 I S .3:319 3 Pr Q3- I 46F=-93 Yc-92 m 05O.-G 044E.44 2.5426,9 5.664-0 2.2302< 0.99600 0 -04 0 Wr-09 Yu-93 4404 5G890. 4-4A4O0~ a ar9 j .4 Q4-, a4 I-0* 4 40~0 ~o~-0oa QJ-QQ&*Zr-105 &~I r"0 84E- ,052 190 1-8~.4 22G0 .-004 RG 2PQz aAr ~42- o.4Q-$G-Sb-95 27 I4E-Qd 4or0- i.4&w0 3604 &2gr,0 g <&,OQ 3 ra 4 244E-03 Mo-929 4 Sr~r4-j :2 r,-08 : 1,O- 2 Gfi-G 646- 4-946-014 2~00 70S0~ 04 :1.26 Te-129mA A0~04-, R8A7PQ4 Q 4r-04 ,S'4 4 A -4 I 7G.04 .,2 n.04 ~47r.or ~4 47Pgo Te-137irr-o 1ri .4,22 G4 E-O~ I5~0 A 049~0 P A"r.03 1.~0603 4 316 4 Sb-132 449 k4.&~o &4.G0Fa.9 3,08420 ~4.6?m02 3.3-942r 67~i 4h.E&.000 9.099'0 Te-1347 1.459*01 44F-94~0 .60E,~02 444E-04 G~rQ01 :1.68E. 02 1r,02#0 44,E43'Te-127r 2. 71n-9 4 ,946-94 246 -4 7.366- 02 6 6-0A0541-42502 6.-005 03 4.24 6 g Xe-i 33 4AS0.4~750 94O0A6S0S04 IMP= 2fifr-4 3.44904 340566r9 a 64 r- w. A44&m02 449E-"O Xe-i 35n 4 93Wi01 2A6Fa"Q4 fi,&8S206 0450129 1.12E5*01 3.a2i.O2 14854 a ~ oos: ii r Cs-i 34 n 2 44;r-04 P &440

• 3.1,4E-402

& 405.'. 24,0 9 44 0 6660- 04 4 2alz-02 3:90r-0go Cs-I 362 4,5*5*,02 Afir5.*G3 5&0r*03 6.785-92 1.199,99 41302Fs4 9.5W03ý 6.956 4 1-131 Revq.w 64 Page0 1.2;9 ofr4 .8d0 .5-.4 .3-3 23P62 676 2 3V: FSAR Rev. 64 Page 1 of 2 SSES-FSAR Table Rev. 56 Replace data with Insert 3 and 4 _r TABLE 15.6-13 LOSS OF COOLANT ACCIDENT ACTIVITY AIRBORNE IN REACTOR BUILDING (curies)________ ~(Design Basis Accident)

_____IsolopekReactor Building Airborne Activity As a Function of Time Post-Accident

________ ________ ________ ________ (curies)_________

___ ____0.1667 hr 0.5 hr I hr 2hr 8 hr 24 hr 96 hr 240 hr 480 hr

• 720 hr Cs-137 9;2-07:4J6 3-QE-02 4j& :2:)FEý03

-4 Qg= *41-5994 4 79r=* 04 .25F.01 2G4r-Be-139 34~O 5Q"; 24.E4361i ,oar- Q1 o~gc , g Q9& :4O.44&h~

9 OOri.O9 Ba-140 4 =E&..0:2 QQ 930 4 r,43- 3 45ý -0 ýSg 4099Q4. 474S-92 La-140 2 Ga r0O 2 Qfio4- 44a.~ 4 jA.kQ2 I4O4RO0 2J5MRO01 4-27&O 04 , iE-02 La-I 41 0 2iZr04- 6.42Fe'O 3.2-0 4AW! -2 74W 14. O'0go-O, 90r-o~ 0 ~r La-142 6.94604. 3.99 3.3;694- 6.66C-Q6 0.099 OQ.QC o~ .9.99,99'O OQ 99r, G Ce-143 2.35E.609 4 ArW 94- 2.63E-9 7.049-00 24 rm 2 0 A.966 03 4.34 6 W.Pr-143 I-,O0*OO .94E.,Q0 14494,i 2AS0 42 48IIng02 44R-03. 1.31 9 5 52F-04 Nd-147 :4 50F,04- 3.54i".-OO 2 9 .4 276-09 4.096.03 556- 04 3.40604 435r. 04 Np-239 ;.26F!!.Q4 P 346FM-G12-14 8 4* 6499- 9a 244E 94 4-Q Pu-239 7.53904 5 g3r, Q2 ~42-O2 2fE-, 03 6846,- 74-R as pa, ~OR 54r.-06 Pu-240 *4-24-=-Cp 95Q 3 4.i 3 61 R44-02 4,45 4 r & 4.34 r Q ~2 Qr 0 2,44E-09-Pu-241 2.98rO4- 2.5F0 &ME-044G 94P 94~ 3,4~O 4AOO-O 2 i.699 792 4 6.01 re 0 Arm-241 4 r;A-94 424r--02 :1.77E 03- 4 MR 94 1005 O -O8 9 & -07 Z 7 ~,2 A=07 Cm-242 4.14-4-02

~2 Mi. 4 .4 624E- 4 239-9 4 74r- 4 4-A44.re 4 4-.0 1 r-4 :74&-SUM-244 " 2..@9 9; 4944n 02 P;4g.w o ;.4S-9; P. 142W ~ O 96 8.0E9 -6. r!FSAR Rev. 64 Page 2 of 2

-Ce~ rBA RAc~r 1wjLA.1667hr 0.5hr 1 hr 2hr 8hr 24hr 96hr 240hr 480hr 720hr Ci cl ci ci Ci Ci Ci ci ci ai Co-58 4.60E-02 3,60E-01 4.79E-01 1.04E-01 2.65E-04 1.29E-04 8.74E-05 5.92E-05 Co-60 2.47E-02 1.94E-01 2.59E-01 5.65E-02 1.48E-04 7,63E-05 5.68E-05 4.23E-0 Kr-85 8.23E-01 7.41 E+00 6.87E+01 4.58E+02 3.00E+03 3.08E+03 2.98E+03 2.84E+03 2.70E+03 Kr-85m 1.452+01 1.24E+02 1.07E+03 6.09E+03 1.58E+04 2.45E+03 1.98E-02 0.00E+00 0.00E+00 0.OOE+00 Kr-ý87 2.72E+01 2.04E+02 1.44E+03 5.58E+03 1.39E+03 4.19E-01 0.00E+00 0.OOE+00 0.00E+00 0.00E+00 Kr-88 3.98E+01 3.30E+02 2.71E+03 1.42E+04 2.15E+04 7.99E+02 1.04E-05 0.OOE+00 0.00E+00 0.002+00 Rb-86 1.16E-01 9.59E-01 3.85E+00 1.51E+01 1.77E+01 3.77E+00 8.73E-03 3.59E-03 1.85E-03 9.54E-04 Sr-89 6.38E+01 5.00E+02 6.64E+02 1.44E+02 3.61E-01 1.72E-01 1.12E-01 7.30E-02 Sr-90 8.14E+00 6.38E+01 8.50E+01 1.86E+01 4.86E-02 2.52E-02 1.88E-02 1.40E-02 Sr-91 7.56E+01 5.51E+02 4.74E+02 3.23E+01 4.42E-04 6.25E-09 0.00E+00 0.00E+00 Sr-92 6.69E+01 4.06E+02 1.17E+02 4.26E-01 1.95E-02 2.34E-02 0.00E+00 0.00E+00 Y-90 1.17E-01 1.35E+00 6.90E+00 4.24E+00 1.21E-02 1.76E-08 1.88E-02 1.41E-02 Y-91 .._ 8.37E-01 6.63E+00 9.53E+00 2.27E+00 5.96E-03 2.87E-03 1.91E-03 1.26E-03 Y-92 5.832+00 9.00E+01 1.98E+02 427E+00 1.14E-08 1.44E-10 0.002+00 0.00E+00 Y-93 6.19E-01 4,53E+00 4.00E+00 2,92E-01 5.46E-06 0.00+E00 0.00E+00 0.00E+00 Zr-95 1.19E+00 9.33E+00 1.24E+01 2.69E+00 6.82E-03 3.31E-03 2.22E-03 1.49E-03 Zr-97 1.13E+00 8.522+00 8.88E+00 1.01E+00 1.38E-04 1.94E-07 7.68E-12 O.OO+00 Nb-95 1.19E+00 9.34E+00 1.25E+01 2.72E+00 7.112E-03 3,65E-03 2.65E-03 1.90E-03 Mo-99 1.55E+01 1.20E+02 1.51E+02 2.78E+01 3.42E-02 3.90E-03 2.34E-04 1.41E-05 Tc-99m 1.39E+01 1.09E+02 1 Al E+02 2.79E+01 3.51E-02 4.00E-03 2.40E-04 1.44E-05 Ru-103 1.34E+01 1.05E+02 1.39E+02 3.002E01 7.452-02 3.47E-02 2.17E-02 1.36E-02 Ru-lO1 ____ ____ 7.902+00 5.302+01 2.77E+01 4.982-01 1.712E-08 1.622-02 0.002+00 0.002+00 Ru-.106 ._. 5.32E+00 4.17E+01 5.56E+01 1.21E+01 3.16E-02 1.21E-08 1.18E-02 8.68E-03 Rh-105 8.60E+00 6.73E+01 8.48E+01 1.41E+01 9.05E-03 2.78E-04 1.88E-06 1.27E-08 Sb-127 1.45E+01 1.13E+02 1.44E+02 2.78E+01 4.24E-02 7.46E-03 9.20E-04 1.14E-04 Sb-129 4.59E+01 3.08E+02 1.56E+02 2.62E+00 6.59E-08 1.46E-02 0.00E+00 0.00E+00 Te-127 1.44E+01 1.13E+02 1.49E+02 3.11E+01 5.51E-02 1.09E-08 6.19E-03 3.84E-03 Te-127m 2.47E+00 1.93E+01 2.58E+01 5.63E+00 1.47E-02 7.39E-03 5.21E-03 3.66E-03 Te-129 4.88E+01 3.53E202 2.40E+02 2.38E+01 4.95E-02 2.26E-02 1.37E-02 8.36E-03 Te-129m I 1.03E+01 8.10E+01 1.08E+02 2.33E+01 5.72E-02 2.62E-02 1.59E-02 9.66E-03 Te-131m 3.26E+01 2.50E+02 2.90E+02 4.38E+01 2.17E-02 4.03E-04 1.18E-06 3.43E-09 Te-132 2.37E+02 1.842+03 2.33E+03 4.42E+02 6.11E-01 8.82E-0 7.85E-03 6.98E-04 1-131 5.79E+01 4.83E+02 2.11E+03 9.01E+03 1.39E+04 9.16E+03 3.67E+03 2,02E+03 7.562+02 2.86E+02 1-132 8.14E+01 6.35E+02 2.60E+03 9.78E+03 4.87E+03 6.38E+02 4.792+01 1.122+01 9.93E-01 8.64E-02 1-133 1.20E+02 9.91E+02 4.28E+03 1.782+04 2.42E2+04 1.07+04 5.052+02 3.85.00 1.15-03 -3.42E-07 m a)X

.1667 hr 0.5 hr 1 hr 2hr 8 hr 24 hr 96 hr 240 hr 480 hr 720 hr-_ ci ci ci ci ci ci ci ci ci ci 1-134 1.17E+02 7.49E+02 2.22E+03 4.31.E+03 6.25E+01 1.51E-04 0.00E+00 0.OOE+00 0.OOE+00 0.00E+00 1-135 1.12E+02 9.08E+02 3.78E+03 1.46E+04 1.30E+04 1.83E+03 5.00E-01 1.28E-07 0.00+E00 0.00+E00 Xe-133 1.18E+02 1.07E+03 9.89E+03 6.60E+04 4.40E+05 8.19E+05 3.89E+05 1.70E+05 4.11E+04 1.00E+04 Xe-135 4.08E+01 4.02E+02 3.74E+03 2.55E+04 1.58E+05 1.55E+05 1.28E+03 2.04E-02 1.78E-10 0.00E+00 Cs-134 1.23E+01 1.02E+02 4.10E+02 1.62E+03 2.04E+03 5.43E+02 2.10E+00 6.30E-01 4.67E-01 3.45E-01 Cs-136 3.91E+00 3.25E+01 1.31 E+02 5.15E+02 6.39E+02 1.65E+02 5.44E-.01 1.20E-01 5.27E-02 2.32E-02 Cs-137 9.22E+00 7.67E+01 3.09E+02 1.22E+03 1.53E+03 4.09E+02 1.58E+00 4.79E-01 3.57E-01 2.67E-01 Ba-139 7.34E+01 3.50E+02 2.43E+01 2.08E-03 0.00E+00 0.O0E+00 0.00E+00 0.00E+00 Ba-140 1.22E+02 9.56E+02 1.34E+03 3.45E+02 1.14E+00 2.52E-01 1.09E-01 4.74E-02 La-140 2.08E+00 2.67E+01 1.63E+02 1.18E+02 1.01E+00 2.85E-01 1.27E-01 5.50E-02 La-141 9.27E-01 6.12E+00 3.02E+00 4.79E-02 5.71E-10 0.00+E00 0.00E+00 0.OOE+00 La-142 6.91E-01 3.47E+00 3.32E-01 6.66E-05 0.00E+00 0.OOE+00 0.00E+00 0.00E+00 Ce-141 2.80E+00 2.20E+01 3.11E+01 8.20E+00 3.00E-02 8.07E-03 4.87E-03 2.94E-03 Ce-143...

2.54E+00 1.96E-+1 2.46E+01 4.68E+00 4.03E-03 5.98E-05 2.89E-07 1.40E-09 Ce-144 2.35E+00 1.85E+01 2.63E+01 7.01E+00 2.71E-02 8.17E-03 5.96E-03 4.34E-03 Pr-143 1.00E+00 7.91E+00 1.14E+01 3.13E+00 1.18E-02 2.75E-03 1.23E-03 5.532-04 Nd-147 4.50E-01 3.54E+00 4.95E+00 1.27E+00 4.09E-03 8.55E-04 3.40E-04 1.35E-04 Np239 3.26E+01 2.53E+02 3.35E+02 7.34E+01 1.18E-01 6.19E-03 2.44E-04 9.59E-06 Pu-238 7.09E-03 5.58E-02 7.93E-02 2.12E-02 8.26E-05 2.53E-05 1.89E-05 1.41 E-05 Pu-239 7.53E-04 5.93E-03 8.43E-03 2.25E-03 8.84E-06 2.71E-06 2.03E-06 1.51E-06 Pu-240 ..... 1.21 E-03 9.53E-03 1.35E-02 3.61E-03 1.41 E-05 4.31 E-06 3.22E-06 2.41 E-06 Pu-241 2.98E-01 2.35E+00 3.34E+00 8.91E-01 3.48E-03 1.06E-03 7.92E-04 5.91E-04 Am-241 .........

1.58E-04 1.24E-03 1.77E-03 4.76E-04 1.90E-06 6.09E-07 4.90E-07 3.92E-07 Cm-242 ._. _4.14E-02 3.26E-01 4.62E-01 1.23E-01 4.74E-04 1.41 E-04 1.01E-04 7.24E-05 Cm-244 2.43E-03 1.91E-02 2.71E-02 7.24E-03 2.82E-05 8.63E-06 6.44E-06 4.80E-06 SUM 7.55E+02 6.12E+03 3.57E+04 1.84E+05 7.10E+05 1.01E+06 3.99E+05 1.75E+05 4.52E+04 1.35E+04 ET.CD 0 ca SSES-FSAR Replace data wi.th Inserts 5 an 6 Table Rev. 56 TABLE 15.6-14 LOSS OF COOLANT ACCIDENT ACTIVITY RELEASED TO ENVIRONMENT (curies)_________(Desigqn Basis Accident)Total Activity Released To The Environment As a Function of Time Post-Accident Isotope 0.5 hr 2 hr 8 hr 24 hr 96 hr 240 hr 480 hr 720 hr Co-58 &GE00Q 89.79E03 -249E 02- 2.589-02 2.72992- ~2.7-2F 42.7299 2.-~2F02.Cr-60 4 -Q99 44 4. 9 3 46 9 4.39E~~1 92~Q Q 4,4- -7P 0 4 47P~Q 02 14-=0 4.79 E402 Kr-85m 4.23&-Q! 6.04 6-1-9 9.59.4-93 2.22§!!04 29+44 '2.364-04 2 2F4,044 -9~Kr-87 2141-~*Q4

@409-42 a &4-7.-;- + Q -2 36 0 78 -F-4; -0 271=+0 7P-0 Kr-88 3EQ1 4 46rý0 I -7=Q4 2,4r=Q4 2 RQ4 ý'37-444 :22r0 2-7-94 Rb-86 2.4-E 92 3.8c- 94 HU &-4. 9.915 -4.9@C-499 4.99E-49 4.42C-4-9 443E-99 Sr-89 0.906+9 1-2=V§4: ~012 4-*Q 2 4AO4 2 77"(31 ~~E~4 2 377991 77-+EQl Sr-90 0. EA 0 0 96P.6 ARQQ 4.6&G 4.82C-49 A-9 0 &- 492-- 4.92E-49 Sr-91 4~§0 21IZ+Q4 Q,2§40- 34&,4940-3.9,48Q 2 4AW404- 2 2-48Q4'O 2 QW48E'Sr-92 0.QQE40Qg 442 04. 2.49E-sQ4 2.22*01 2.229+- 2.22910 2.2E--.4 2.E*-04, Y-92 GQGE09§ 4~.98-i0 642E+0Q 6.59940Q 6.696+. 6 AGE40 a g6-A9-Q §§;E*-0 Y-93 a Mr0kan 44414 01 -.41 1 -2.E-l6 4904 2.9E01 2899 01. BR-0IZ 04 t sE=04-Zr-95 Q040Q-Q 2.28F 04 5.~676-Q 01:R6qgrQ-04 7-,04991 7.049-Q4 7,04E,91-2-44904 Nb-95 0QQr=QQ :2:2v-Q4 .5,69r=04 a ~.70r= 4 wQaE-44 ~Q6F=-W4 oar. 4 ;z 6r=Q4-Mo-99 GGGE499 ;1,05F-0 :7-p2XF;QQ

&43P40 84n4 9.74449 8 .:749+9 74lZ*.9§Tc-99m 0,Q0 ,6Q 6.5EjQQ 7- QF=QQ 9:09900 9.039*99 .,03E*Q0 &W&G~O Ru-103 pQ'9 25r.*0Qn 2rF-i.0Q :7.4994-9 7-A9+9 7A1E+9 7 A-Q k--0 Ru-1O05 ~ 4~QQ 2~Q &Q*O 3.06C=+00 3,9§O4+QG 346*09 ;9.96E-00 Ru-I 06 2.594R1.,'.Q 2 94 P gQ-i 3.459*90 045E:14 Q4~0 I 5F-Rh-105 o mr.-Q 4 4,Q 44Z.Qgj -4 ;Qi-Q§ 4~4W9* 4 84F.hQQ 4. R-9 444"&Q Sb-I 27 0Q='Q0Q p~EQ 7ap-Er) 6'4 44 7A2*-QQ A=R00- -S Xg;=ýQ A____8;2E+0 Sb-i 29 77F+-Q &0~9Q 1~g. 4 04 I 77F.+01 1.779+- 177P-4-Q1-1~0 Te-I 27 -.0.0F=+9 7617=40 648E*00 '0.9E-9 8.6n9 ý8.41@990 ,g~g &.46E+GGo Te-1 27m O0.9&ý9 .7&4 40- 4~49C4-99 4.9E44 46"GO~ 4.40=Q9 -4.4649 4 4-4F-QQ Te-1 29 94990b0Q &940 QQ=Q! -24 401i-9 2~4 a F=- 0 1 2 1-3zk 244E-+4 2.449*4-0 Te-1 29m 4) 9Q -=-- 9 4.99E=:Q 4. 09 &O5 0 AQF0-QQ 6 -O -QQQ 644-0 W0~AQ Te-1 31 m 9.0E0 6469499Q 1 4A+Q4~4 0 1.494-4 4 :4F-4+0- d,94~0 4S Te-I 32 Q'94.E-+W 4'541-+0Q4 4.4 1E-+2 4.29E-2. 1.3EQ 4.36C-+2 4.95C=+0 4-35E 1-132 4-76Fý04-2AG§E49.2 5181XPL2 &6 0: 26=G 42R*02 A 421-A02 6..429G2 6. 4.2& Q 2 1-134 _____ ;IQ F,;2Fý0;2

.2S2FiQ p_____I-I135 2.459+91 2 84P+02 9.329+02 1.2-,*03 1.444-0: A 4r=+02 4-34E*9g 14.4E-+9 Xe-I 33 .1 2rQp. 6.9EA9 I 94054Q 1.26E4 4 90QPF+ Q 74FCQa 4 4F-9 4440R.W Xe-I 35 3469 247&-9 6.11&4 2 21606 5.399-'Qý

4 41 -#.& 44-*Qý 5

4j&(5 Cs-I 34 2.666-1,9 4.;3-94 9.8!49 R42___ __.4 _ 4.___ _____ _2______4-4-19 4 ý69 Cs-I 36 48-1 -2+0 2.4~2&94.

3---4E--" 3-7-F-4 3.79*4i 3-74&Q*94 2-~FSAR Rev. 64 Page I of 2 SSES-FSAR IReplace data with Inserts 5an Table Rev. 56 4 TABLE 15.6-14 LOSS OF COOLANT ACCIDENT ACTIVITY RELEASED TO ENVIRONMENT (curies)(Desian Basis Accident)Total Activity Released To The Environment As a Function of Time Post-Accident (curies______

Isotope 0.5 hr 2 hr 8 hr 24 hr 98 hr 240 hr 480 hr 720 hr Ba-I 39 -4DE99IQQ R+01 IA P F,ý Q 4.92E*4 Q 2c*4 4,82.21H-*G*01 4.82E*01 .812E4=Q4 Ba-i 40 a ~=Q 22,R,04Q-5.-69j94 zzr=&n4- : rZ44 Q .7 11 i-4-Q1- :7 14 F44 7.-A4 4. =ý.La-I 40 0 QQF'QQ 446;V.Qg : 42111-0 4.PF-00 5 5FQQ 8amjQ 5 R6F.iggQg 5 rW_La-I 41 O.GOE*0 1-.62F=-01

-3337E 0- & ,4.7E-§4 3,47E-QI-

'347--E 3 47o-E 01 24F 4 La-i 42 04F59 4(1F -O4I '-. a- I-e0Q4 i,.9-1-F5-04

,4~.91E; G4 '4 MR-Q4. 4.94-1 EOW IF414~4 Ce-14I 9 OQE0~ 5.349.01 1 226=40 147-E'QO 145E*90. I 651&7*0 1.659+9 14.6*OO Ce-143 9.094.9 4 RO-Q 4 44. i7--49 4.349&9O 24r-QQ 4'.37+0 4.~37E+99 4.3-&4Q Ce-144 QQGO&O 4.49EO 914.2E*QQ 4.32&49 4.99E+99 4.39949 4.349E-+ 4 P-Pr-I 43 QQGQ i W-rQ4 4 ;.gi;Q w S 7F 0 5QRQ4 5 gjQR_04. 5,8-. Pa=0 9F Pu-238 9499+99 1 -r.-Qg :3,g~zF5 3.98E90.-.Q 9 3 4 49 93 4. 99! ______9!_Pul-239 0.919 4A4 P-04 3,48E-04 4-22--F.Q--4 45P-04 4. a Er Q4 A4.45- 4 4 45F-Q4 Pu-240 ________g 2.34-Fa-§4 4-04. 6-79F.-04 7.4 5E-O4 7-15F-Q4 :7 45 F. 4 1 15P§4 Pu-241 4.44&__2 '44-42"4 G-R8,-Q4 4-74&" A 4Z61ZQ4 I -6r-04 4.76S Q.1 Cm-242 Q9Q-99E.9

0=92Q 1.WE 0 2.326.92 2.44Fm-O2 2.44 E-2 2.449-Q2 2;44E 9.2 FSAR Rev. 64 Page 2 of 2 Insert-5.

DBA Environment Insert 5: DBA Environment Time 0.5 hr 2 hr 8 hr 24 hr 96 hr 240 hr 480 hr 720 hr Isotope CI Ci Ci Ci Ci Ci Ci Ci Co-58 0.OOE+00 8.82E-03 2.23E-02 2.62E-02 2.72E-02 2.72E-02 2.72E-02 2.72E-02 Co-60 0.OOE+00 4.75E-03 1.20E-02 1.41E-02 1.47E-02 1.47E-02 1.47E-02 1.47E-02 Kr-85 7.34E-01 4.51E+01 1.32E+03 9.74E+03 4.06E+04 1.10E+05 2.51E+05 4.08E+05 Kr-85m 1.25E+01 6.44E+02 1.05E+04 2.44E+04 2.58E+04 2.58E+04 2.58E+04 2.58E+04 Kr-87 2.16E+01 7.16E+02 3.80E+03 4.12E+03 4.12E+03 4.12E+03 4.12E+03 4.12E+03 Kr-88 3.37E+01 1.56E+03 1.89E+04 3.09E+04 3.12E+04 3.12E+04 3.12E+04 3.12E+04 Rb-86 2.52E-02 3.81E-01 8.50E-01 1.00E+00 1.06E+00 1.09E+00 1.12E+00 1.13E+00 Sr-89 0.00E+00 1.22E+01 3.09E+01 3.63E+01 3.77E+01 3.77E+01 3.77E+01 3.77E+01 Sr-90 0.00E÷00 1.56E+00 3.94E+00 4.63E+00 4.82E+00 4.82E+00 4.82E+00 4.82E+00 Sr-91 0.OOE+00 1.40E+01 3.25E+01 3.51E+01 3.52E+01 3.52E+01 3.52E+01 3,52E+01 Sr-92 0.00E+00 1.12E+01 2.21E+01 2.25E+01 2.25E+01 2.25E+01 2.25E+01 2.25E+01 Y-90 0.00E+00 2.55E-02 1.07E-01 2.04E-01 2.62E-01 2.62E-01 2.62E-01 2.62E-01 Y-91 0.00E+00 1.61E-01 4.13E-01 4.94E-01 5.17E-01 5.17E-01 5.17E-01 5.17E-01 Y-92 0.00E+00 1.38E+00 5.86E+00 6.74E+00 6.75E+00 6.75E+00 6.75E+00 6.75E+00 Y-93 0.00E+00 1.15E-01 2.68E-01 2.90E-01 2.92E-01 2.92E-01 2.92E-01 2.92E-01 Zr-95 0.OOE+00 2.28E-01 5.76E-01 6.77E-01 7.04E-01 7.04E-01 7.04E-01 7.04E-01 Zr-97 0.00E+00 2.12E-01 5.11E-01 5.69E-01 5.75E-01 5.75E-01 5.75E-01 5.75E-01 Nb-95 0.00E+00 2.29E-01 5.77E-01 6.79E-01 7.06E-01 7.06E-01 7.06E-01 7.06E-01 Mo-99 0.00E+00 2.96E+00 7.38E+00 8.53E+00 8.78E+00 8.78E+00 8.78E+00 8.78E+00 Tc-99m 0.00E+00 2.66E+00 6.69E+00 7.80E+00 8.05E+00 8.05E+00 8.05E+00 8.05E+00 Ru-103 0.00E+00 2.56E+00 6.46E+00 7.58E+00 7.88E+00 7.88E+00 7.88E+00 7.88E+00 Ru-105 0.OOE+00 1.40E+00 2.99E+00 3,10E+00 3.10E+00 3.10E+00 3.10E+00 3.10E+00 Ru-106 O.OOE+00 1.02E+00 2.58E+00 3.03E+00 3.15E+00 3.15E÷00 3.15E+00 3.15E+00 Rh-105 0.00E+00 1.65E+00 4.13E+00 4.76E+00 4.88E+00 4.88E+00 4.88E+00 4.88E+00 Sb-127 0.00E+00 2.77E+00 6.92E+00 8.04E+00 8.29E+00 8.29E+00 8.29E+00 8.29E+00 Sb-129 0.00E+00 8.10E+00 1.73E+01 1.79E+01 1.79E+01 1.79E+01 1.79E+01 1.79E+01 Te-127 0.00+E00 2.77E+00 6.99E+00 8.18E+00 8.47E+00 8.47E+00 8.47E+00 8.47E200 Te-127m 0.00E+00 4.73E-01 1.19E+00 1.40E+00 1.46E+00 1.46E+00 1.46E+00 1.46E+00 Te-129 0.00E+00 8.98E+00 2.03E+01 2.16E+01 2.18E+01 2.18E+01 2.18E+01 2.18E+01 Te-129m 0.00E+00 1.98E+00 5.00E+00 5.87E+00 6.10E+00 6.10E+00 6.10E+00 6.10E+00 Te-131m 0.OOE+00 6.18E+00 1.52E+01 1.72E+01 1.76E+01 1.76E+01 1.76E+01 1.76E+01 Te-132 0.OOE+00 4.52E+01 1.13E+02 1.31E+02 1.35E+02 1.35E+02 1.35E+02 1.35E+02 1-131 1.29E+01 2.25E+02 6.61E+02 1.51E+03 4.47E+03 8.42E+03 1.16E+04 1.29E+04 1-132 1.77E+01 2.81E+02 5.96E+02 6.46E+02 6.51E+02 6.51E+02 6.51E+02 6.51E+02 1-133 2.66E+01 4.48E+02 1.25E+03 2.33E+03 3.62E+03 3.74E+03 3.74E+03 3.74E+03 1-134 2.27E+01 1.84E+02 2.53E+02 2.54E+02 2.54E+02 2.54E+02 2.54E+02 2.54E+02 1-135 2.46E+01 3.85E+02 9.52E+02 1.30E+03 1.38E+03 1.38E+03 1.38E+03 1.38E+03 Xe-133 1.05E+02 6.45E+03 1.86E+05 1.33E+06 5.07E+06 9.82E+06 1.31E+07 1.41E+07 Xe-135 3.54E+01 2.33E+03 6.76E+04 3.47E+05 5.72E+05 5.73E+05 5.73E+05 5.73E+05 Cs-134 2.67E+00 4.05E+01 9.03E+01 1.07E+02 1.14E+02 1.18E+02 1.23E+02 1.26E+02 Cs-136 8.52E-01 1.29E+01 2.87E+01 3.38E+01 3.58E+01 3.67E+01 3.75E+01 3.78E+01 Cs-137 2.01E+00 3.05E+01 6.79E+01 8.05E+01 8.56E+01 8.88E+01 9.26E+01 9.51E+01 Ba-139 0.00E+00 1.09E+01 1.83E+01 1.83E+01 1.83E+01 1.83E+01 1.83E+01 1.83E+01 Ba-140 0.00E+00 2.332+01 5.86E+01 6.86E+01 7.12E+01 7.12E+01 7.12E+01 7.12E+01 In~sert -6: DB5A Environment

' I Time 0.5 hr 2 hr 8 hr 24 hr 96 hr 240 hr 480 hr 720 hr Isotope Ci Ci Ci Ci Ci Ci Ci Ci La-140 O.OOE+00 4.69E-01 2.19E+00 4.30E+00 5.47E+00 5.47E+00 5.47E+00 5.47E+00 La-141 O.OOE+00 1.62E-01 3.41E-01 3.51E-01 3.51E-01 3.51E-01 3.51E-01 3.51E-01 La-142 O.OOE+00 1.06E-01 1.82E-01 1.83E-01 1.83E-01 1.83E-01 1.83E-01 1.83E-01 Ce-141 O.OOE+O0 5.35E-01 1.35E+00 1.59E+00 1.65E+00 i.65E+00 1.65E+00 1.65E+00 Ce-143 O.OOE+00 4,81E-01 1.19E+00 1,35E+00 1.38E+00 1.38E+00 1.38E+00 1.38E+00 Ce-144 O.OOE+00 4.50E-01 1.14E+00 1.34E+00 1.39E+00 1.39E+00 1.39E+00 1.39E+00 Pr-143 O.OOE+00 1.92E-01 4.86E-01 5.74E-01 5.98E-01 5.98E-01 5.98E-01 5.98E-01 Nd-147 O.OE+00 8.61E-02 2.17E-01 2.54E-01 2.63E-01 2.63E-01 2.63E-01 2.63E-01 Np-239 O.OOE+00 6.20E+00 1.54E+01 1.78E+01 1.83E+01 1.83E+01 1.83E+01 1.83E+01 Pu-238 O.OOE+00 1.36E-03 3.43E-03 4.03E-03 4,19E-03 4.19E-03 4.19E-03.

4.19E-03 Pu-239 O.OOE+00 1.44E-04 3.64E-04 4.28E-04 4.45E-04 4.45E-04 4.45E-04 4.45E-04 Pu-240 O.OOE+00 2.32E-04 5.85E-04 6.88E-04 7,15E-04 7.15E-04 7.16E-04 7.16E-04 Pu-241 O.OOE+00 5.71E-02 1.44E-01 1.70E-01 1.76E-01 1.76E-01 1.76E-01 1.76E-01 Am-241 O.OOE+00 3.03E-05 7.64E-05 8.99E-05 9.36E-05 9.36E-05 9.36E-05 9.36E-05 Cm-242 O.OOE+00 7.92E-03 2.OOE-02 2.35E-02 2.44E-02 2.44E-02 2.44E-02 2,44E-02 Cm-244 O.OOE+00 4.64E-04 1.17E-03 1.38E-03 1.43E-03 1.43E-03 1.43E-03 1.43E-03 SUM 3.t9E+02 1.35E+04 2.92E+05 1.75E+06 5.75E+06 1.06E+07 1.40E+07 1.52E÷07 SSES-FSAR Replace-da with Insert 2. ! !Table Rev. 56 TABLE 15.6-16 LOSS OF COOLANT ACCIDENT ACTIVITY AIRBORNE IN REACTOR BUILDING (curies)___________(Realistic Analysis)Reactor Building Airborne Act"vt As a Function of Time Post-Accident

___________ (curies)Isotope 2 hr 8 hr 24 hr 96 hr 720 hr 1-131 95-O14 2.-474-90O 441-E+O 9-596 4-499-0 1-132 .4 ý5F14 97&19 p8-~§ &aý =~~-2 'Q9Q4Fi 1-134 4-399 4-,242-92 g-85E-99 0 &=+Q -0.~O O__=QE00 Xe-135m 4.5 (A. 6,59-14 9-44E-E94QO-4-. --2 r Q 4 -26&9 Xe-138 3-~ -29E4 3-44~94J9 4_39&-9" -344g Kr-835m .249-E-§1-4469 ~2-.E

-6.69E-Q 9 .g E g Kr-85m .2 4 r=-0 4. 7499-92 3~-26-9~4 0A5-99&9 G.QGE*0O Kr-88 2.16F40 14-&E4QQ &64-9-ýQ 7-49E-49 ~~~Q Kr-85 .2-49E-49 9439E-4942 4_,57E--9&-E0 46 -= 7-9A49-f Co-60 _________

4-212E-94_

2-&&4__-_8E-04___E*

Sr-89 4,46__E_94_

449-4-__6_2__Q_164E94___

GE4 Sr-90 ________ _______ __________F=Q 2 -2 4F .Q4 PQQ.Q Sr-91 ________ -. 4TE4W Rý Or;___ 94&-Q.iQQ Sr-92 4I8E-4 00 1446&-9 339E-02 C) 0l;+OQ 0-9Go19 Zr-95 4.05F= w 31F=Q .a 28ý-02 429QE-Q 05 QQE4.Q9 Zr-97 4.99E___3 2-7E-4-4 2-46-93 44-Fý -Q QE+Qg-.NbJ-S5 4§.-QE 9-Q ý3QE.96&3

.4-43E-9&.

QG.QG944 Mo-99m 56A~-1 1 F!-. 9 12.F=4QQ 447-9449 &9E-7& 9,999444 Rul-I03 4-93F. 04 -E -Q- -§+O Ru-I 06 -2.Q 1 A4F4--0 4 3,-77-E-94 2-909-99 P4OOE.ýO Te-1 32 1.221i-+.9 366E4W 5.629+00Q

~2.21F02 o~~Cs-134 _4-199-92 a ,7 4._F --02 &44E4)2 4-94E--04 OOOE49-O Cs-137 699E03 4.93E92. 3.42E 02 *2-49~-94 0.09.919 Ba-139 4O~-6944 24A=-O1- -4.4 4--9 Q4 .O,=O9oO P-99944 Ba- 140 p-2;-.§1-04 0 1. Q I .2~-4 F 9 CF W _4G____Ce-I 41 -&94E-04;

44-3-9~ 447-E98 :3.82E,95 Ce-I 43 Z 4F= 04 -2. 3 9 I=0 34.2&93 4Q~ 4QO.49944 Ce-I 44 9 W, 0-4 294EL-94 .99 -2Q-Q 046 -Q5 0499449 Pr-I 43 9,58E-O4 3ý-03E 03~ 54GE-03 fifF-9 nfi .~i0 Nd-I 47 3Q 59r 4' i.OQ-0a 93. 1 3 114F= Q5 ',O.QOE*Np-239 &4&Q 144- Q-ý5EQ 4ýQ ,Q~4 FSAR Rev. 64 Page 1 of I Insert 2 Realistic Reactor Building Time 2 8 24 86 720 Isotope (hours) (hours) (hours) (hours) (hours)1-131 7.85E-01 2,36E+00 3.72E+00 7.37E-01 3.89E-02 1-132 1.23E+00 3.65E+00 6.03E+00 1.40E-02 0.00E+00 1-133 1.76E+00 4.43E+00 4.34E+00 1.01E-01 4.68E-11 1-134 4.33E-01 1.16E-02 6.19E-08 0.00E+00 O.OOE+00 1-135 1.47E+00 2.41E+00 7.51E-01 1.01E-04 0.OOE+00 Xe-131m 3.69E-02 1.12E-01 1.85E-01 1.00E-01 2.63E-02 Xe-133m 1.84E-01 5.47E-01 8.22E-01 2.57E-01 5.20E-05 Xe-133 6.52E+00 1.97E+01 3.16E+01 1.36E+01 3.63E-01 Xe-135m 3.57E-01 4.80E-01 3.54E-01 4.98E-05 0.OOE+00 Xe-135 5.82E+00 1.40E+01 1.01E+01 5.21E-02 0.OOE+00 Xe-138 1.66E-02 1.09E-09 0.OOE+00 0.OOE+00 0.00E+00 Kr-83m 2.39E-01 7.55E-02 2.95E-04 O.00E+00 0.00F+00 Kr-85m 9.25E-01 1.12E+00 1.58E-01 1.26E-06 O.00E+00 Kr-87 8.22E-01 9.59E-02 2.62E-05 0.00E+00 0.00E+00 Kr-88 2.12E+00 1.51E+00 5.08E-02 6.49E-10 0.00E+00 Kr-85 2.77E-01 8.49E-01 1.42E+00 7.77E-01 6.81 E-01 Co-58 4.80E-04 1.47E-03 2.44E-03 1.02E-05 0.00E+00 Co-60 4.44E-05 1.36E-04 2.28E-04 9.80E-07 0.00E+00 Sr-89 1.54E-01 4.70E-01 7.79E-01 3,22E-03 0.00E+00 Sr-90 5.75E-03 1.77E-02 2.96E-02 1.27E-04 0.OOE+00 Sr-91 1.72E+00 3.42E+00 1.78E+00 4.01E-05 0.00E+00 Sr-92 1.66E+00 1.10E+00 3.07E-02 0.00E+00 0.OOE+00 Zr-95 1.03E-03 3.17E-03 5.26E-03 2.19E-05 0.OOE+00 Zr-97 1.06E-03 2.55E-03 2.21E-03 4.98E-07 O.00E+00 Nb-95 1.05E-03 3.22E-03 5.39E-03 2.32E-05 0.00E+00 Mo-99 5.38E-01 1.55E+00 2.19E+00 4.44E-03 0.00E+00 Tc-99m 5.67E+00 9.43E+00 4.31E+00 4.55E-03 0.OOE+00 Ru-103 4.87E-04 1.49E-03 2.46E-03 1.01 E-05 0.00E+00 Ru-106 6.66E-05 2.04E-04 3.41E-04 1.46E-06 0.00E+00 Te-129m 1.00E-03 3.07E-03 5.06E-03 I 2.05E-05 0.OOE+00 Te-132 1.20E+00 3.50E+00 5.08E+00 1.16E-02 0.00E+00 Cs-134 1.16E-02 3.57E-02 5.97E-02 2.57E-04 0.00E+00 Cs-136 2.76E-03 8.37E-03 1.35E-02 4.97E-05 0.00E+00 Cs-137 6.01E-03 1.85E-02 3.09E-02 1.332-04 0.00E+00 Ba-139 1.58E+00 2.37E-01 1.27E-04 0.00E+00 0.00E+00 Ba-140 2.24E-01 6.78E-01 1.09E+00 4.00E-03 0.00E+00 Ce-141 9.82E-04 3.00E-03 4.95E-03 2.00E-05 0.OOE+00 Ce-143 8.44E-04 2.28E-03 2.73E-03 2.59E-06 0.00E+00 Ce-144 9.17E-04 2.81E-03 4.70E-03 2.01 E-05 0.00E+00 Pr-143 9.46E-04 2.90E-03 4.80E-03 1.86E-05 0.00E+00 Nd-147 3.46E-04 1.05E-03 1.68E-03 5,982-06 0.00E+00 Np-239 5.85E+00 I1.67E+01 2.29E+01 4.09E-02 0.OOE+00 SSES-FSAR Table Rev...%Replace data wih nsert 1 TABLE 15.6-17 LOSS OF COOLANT ACCIDENT ACTIVITY RELEASED TO ENVIRONMENT (curies)___________(Realistic Analysis)Total Activity Released To The Environment As a Function of Time Post-Accident (curies) ________ 'Istp 2 hr 8 hr 24 hr 96 hr 720 hr Xe-I131 m 142*QI .7-4 1* ~ SO§+~1§E§l Xe-I133 5,61i' 0 244.05+9 4.75E+01 6.28E 1 1 2AF*§--Xe-I 31 m 4r2-924QQ

-642E 04 2~Q 1,999+QG 5.9 9 1E9 Xe-I 35 2-44&-§ý 8.45E+91OA 14-4Ei-0-2 9-4Ri= ~AE§Xe-I 38 34*Q I -5*Q a.579-49 Kr-83m 4w,___ 4__-_4___-_.3&0' 8___+QQ .99"O -ýKr-87m 4 ZFi.QQ i~.-2EQQ 17 QAE+OQ! ;3R*04Q 4A42.QQ Kr-85 444E+99O 4 77F=+QQO 1.52F401-4,O2E"- 242Eý=Q Co-58 14. 94. -1§-: w7-4E-09 2.29E-94 2--4&@Q2 2-74~-Q Co-60 4-_____-04

________ 244E-03__5_-_

Sr-89 _______ ________ 0_::__________EQ0

& ý5=-9 9 5-+9 Sr-90 2__32&92_

Q-24&___2

___________9&-94 Sr-91 __-46_____

___________-G 464i91 4-49Q Sr-92 8 -~7-=*Ql 243Eq;ý ar=E*Q 243E'*94_Zr-95 4447 .4~-66-2 4.-4-E-92 9992Q .&.49-99 Zr-97 4 4iiF 0 1.58E 02 ~4RF=Q;?;

&7hO77E2 q-E-Q2 Nb-95 4.24E-93 1.69E 92 6. 91I-E-02 9 6.-93&92 Mo-99 24499+4 9-49&-GQ 2-34E1*9 3 269"01 -692=6~404 Tc-99mn 62-=+4 7 5.9"04 4-27-F=.Q-P ap-+=9 4_929442'Ru-I 03 :A7-= -23 ý7.21i-Q93 2-4&92- 2-74-EQ 2___7E_42 Ru-I 06 2.68F= 04 1,07r=-02 17F=-03 1 -82E-0Q ____Te-I 32 4499-W0 4-99Q-i-4

&2-49*49 6.*Q 1 & a+91 143F=4~1-Cs-134 4 Q F-= Q

• 4W--4 555= al.4 a ~Cs-I 36 4429.-2 .4A4;-k.Q;2 I =F 4-Q1 4~.549-Q4 14-54E4 Cs-i 37 2-43E-Q2 9,G-9-2 '2497-E-9--

345E-9 3-45E-0-Ba-i 39 4-O7-E=9 :1 6E*O-4 149E*91 I 69E*Q1- I 4QF=*Q Ba-I 40 Q-Q 05F 0 2RF44' I 4F-4-Q1- 4-~249494 4-~24E44)4 Ce-I 41 3.96E ;3 1.67-E02 4.64F=02.

&47E-94 )2 7&E-Q Ce-I 43 2 A7r-Q30 --3GE 3-19E-92 3-71-E--Q 31--Ce-I 44 4-74-9 4~-47-E9 437-E-ý 2-9 &,24&42 Pr-l 43 Z42JE-9a :1.62E- 2 4.49E-2' &39E-92 &39E-402 Nd-I 47 4.491i 9 &Qi- 93 1- 61 F-0:2 4ý4-O 41 F1e -Q22 Np-239 2~9-e9941-

@-64449. 4-4944 2_99E__42 2_M___5____Q_

FSAR Rev. 64 Page 1 of 1 Insert 1: Realistic Environment Time 2 8 24 96 720 Isotope (hours) (hours) (hours) (hours) (hours)1-131 2.45E+00 9.65E+00 3.07E+01 3.46E+01 3.90E+01 1-132 4.13E+00 1.79E+01 6.11E+01 6.60E+01 6.60E+01 1-133 .5.66E+00 2.05E+01 5.07E+01 5.28E+01 5.29E+01 1-134 3.16E+00 3.97E+00 3.97E+00 3.98E+00 3.98E+00 1-135 5.07E+00 1.52E+01 2.47E+01 2.47E+01 2.47E+01 Xe-131m 1.52E-01 6.38E-01 1.90E+00 5.01 E+00 1.61E+01 Xe-133m 7.62E-01 3.11E+00 8.63E+00 1.85E+01 2.41E+01 Xe-133 2.70E+01 1.12E+02 3.27E+02 8.04E+02 1.61E+03 Xe-135m 1.43E+00 3.44E+00 6.47E+00 6.77E+00 6.77E+00 Xe-1 35 2.42E+01 8.59E+01 1.64E+02 2.07E+02 2.07E+02 Xe-138 3.66E+00 3.57E+00 3.53E+00 3.57E+00 3.57E+00 Kr-83m 1.44E+00 2.65E+00 2.65E+00 2.79E+00 2.79E+00 Kr-85m 4.44E+00 1,23E+01 1.58E+01 1.75E+01 1.75E+01 Kr-87 5.97E+00 9.02E+00 8.79E+00 9.15E+00 9.15E+00 Kr-88 1.11E+01 2.56E+01 2.79E+01 3.01E+01 3.01E+01 Kr-85 1.14E+00 4.81E+00 1.46E+01 4.03E+01 2.12E+02 Co-58 1.94E-03 7.71E-03 2.52E-02 2.74E-02 2.74E-02 Co-60 1.79E-04 7.13E-04 2.34E-03 2.55E-03 2.55E-03 Sr-89 6.20E-01 2.47E+00 8.05E+00 8.75E+00 8.75E+00 Sr-90 2.32E-02 9.24E-02 3.03E-01 3.30E-01 3.30E-01 Sr-91 7.46E+00 2.42E+01 4.63E+01 4.64E+01 4.64E+01 Sr-92 8.63E+00 1.87E+01 2.13E+01 2.13E+01 2.13E+O1 Zr-95 4.18B-03 1.66E-02 5.43E-02 5.90E-02 5.90E-02 Zr-97 4.46E-03 1.58E-02 3.70E-02 3.77E-02 3.77E-02 Nb-95 4.24E-03 1.69E-02 5.54E-02 6.03E-02 6.03E-02 Mo-99 2.19E+00 8.46E+00 2.53E+01 2.69E+01 2.69E+01 Tc-99m 2.53E+01 7.59E+01 1.31E+02 1.32E+02 1.32E+02 Ru-103 1.97E-03 7.82E-03 2.55E-02 2.77E-02 2.77E-02 Ru-106 2.69E-04 1.07E-03 3.51 E-03 3.82E-03 3.82E-03 Te-129m 4.05E-03 1.61E-02 5.24E-02 5.70E-02 5.70E-02 Te-132 4.89E+00 1.90E+01 5.74E+01 6.13E+01 6.13E+01 Cs-134 4.70E-02 1.87E-01 6.14E-01 6.68E-01 6.68E-01 Cs-136 1.12E-02 4.42E-02 1.42E-01 1.54E-01 1.54E-01 Cs-137 2.43E-02 9.66E-02 3.17E-01 3.45E-01 3.45E-01 Ba-139 1.07E+01 1.66E+01 1.68E+01 1.69E+01 1.69E+01 Ba-140 9.05E-01 3.58E+00 1.15E+01 1.25E+01 1.25E+01 Ce-141 3.96E-03 1.57E-02 5.13E-02 5.57E-02 5.57E-02 Ce-143 3.47E-03 1.30E-02 3.55E-02 3.71E-02 3.71E-02 Ce-144 3.70E-03 1.47E-02 4.83E-02 5.26E-02 5.26E-02 Pr-143 3.82E-03 1.52E-02 4.96E-02 5.39E-02 5.39E-02 Nd-147 1.40E-03 5.53E-03 1.77E-02 1.91E-02 1.91E-02 Np-239 2.39E+01 9.17E+01 2.69E+02 2.85E+02 2.85E+02 SSES-FSAR Table Rev. 57 TABLE 15.6-18 LOSS-OF-COOLANT ACCIDENT

SUMMARY

OF OFFSITE DOSES Dose Dose (Rem / TEDE)Location Regulatory Design Basis Realistic Limit Analysis Analysis THYROID 2 Hour Site Boundary 25 4-.01.12E+

1.7E-02 o- (T 30 Day Low Population Zone 25 4.3E-03 4-_24.3E+00 I1. 4k071.12E+01 4,071. 12 X10+0'FSAR Rev. 64 Page I of I SSES-FSAR Table Rev. 56 TABLE 15.6-21 LOSS-OF-COOLANT ACCIDENT

SUMMARY

OF CONTROL ROOM OPERATOR DOSES (Design Basis Analysis)Control Room Operator Dose Dose (Rem TEDE)Regulatory Limit Design Basis Realistic Basis I Analysis Analysis 30 Day Operator Dose 5 0-0650,050 I FSAR Rev. 64 Page 1 of I SSES-FSAR Table Rev. 57 TABLE 15.6-22 LOSS-OF-COOLANT ACCIDENT PARAMETERS FOR POSTULATED ACCIDENT ANALYSIS Design Realistic Assumptions Basis Assumotions

1. Data And Assumptions Used To Estimate Radioactive Source Term From Postulated Accidents A. Reactor power level(MWt) 4032 4032 B. Fuel damaged(percent) 100 0 C. Activity released to primary containment atmosphere Subsection 15.6.5.5.1.1 Subsection 15.6.5.5.2.1 D. Activity released to suppression pool water Subsection 15.6.5.5.1.1 Subsection 15.6.5.5.2.1 E. Iodine form fractions(percent)

1. Aerosol 95 95 2. Elemental 4.85 4.85 3. Organic 0.15 0.15 II. Data And Assumptions Used To Estimate Activity Released A. Primary containment leak rate(percent/day) 1 (0-24 hr) 1 (0-24 hr)n3 s tl -,ind til n si1-nn rn 115 100(0-10min) 115 44l00(0-10min)

B. Reactor building leak rate(percent/day) 0 > 10 min 0 > 10 min C. Secondary containment bypass leak rate(SCFH) 15 15 D. MSIV Leakage(SCFH for 4 steam lines) 300 300 E. ESF leak rate 20 20 1. Leakage rate inside reactor building(gpm) 10 10 2. Flashing fraction for iodine(percent)

Primary containment free vdume(it3) 239600 239600 Dryell 148590 148590 Wetwell 388190 388190 TOTAL G. Reactor building free volume(ft 1,488,600 1,488,600 1. Zone 1 1,598,600 1,598,600 2. Zone 2 2,668,000 2,668,000 3. Zone 3 4,156,600 4,156,600 TOTAL Used in Analysis (Zone I + Zone 3)H. Suppression pool water volume(ft3) 132,000 132,000 I. Standby Gas Treatment System Parameters

1. SGTS flow during drawdown(cfm) 11.110 11,110 FSAR Rev. 64 Page 1 of 2 SSES-FSAR Table Rev. 57 TABLE 15.6-22 LOSS-OF-COOLANT ACCIDENT PARAMETERS FOR POSTULATED ACCDENT ANALYSIS Design Realistic Assumptions Basis Assumptions

2. Drawdown time to reach 0.25 inch of vacuum water 10 10 gage in reactor building (minutes)3. SGTS flow following drawdown(cfm) 2M8853320 2Z8,3320 4. SGTS filter efficiencles(percent) 99 99 lodine(All species)J. Reactor Building Recirculation System Parameters

1. Row rate(cfm) 83,000 83,000 2. Mixing efficlency(percent) 50 50 3. Filter efficiency 0 0 4. Recirculation system actuation(seconds) 10 to 30 10 to 30 K. Post-LOCA activity concentrations in primary containment and Tables 15.6-11, Tables 15.6-15, reactor building 15.6-13 15.6-16 Ill. Data And Assumptions Used To Evaluate Control Room Doses A. Control structure habitability envelope free 518,000 518,000 vt 3 1 B. Control room free volume(ft 3) 110,000 110,000 C. Control structure filtered air Intake fiow(cfrm) 5229- 6391 5229-6391 D. Control structure unfiltered outside air infiltration rate 10 10-Ingress/egress (cfmin)E Control structure unidentified unfiltered outside air 6oo 350 6oo 350 infiltration rate (cfm)99 99 F. Control structure filter efficiency for Iodine( percent)IV. Dispersion Data A. Site Boundary/Low Population Zone distance(meters) 549/4827 549/4827 Table 2.3-92 Table 2.3-92 B. Site Boundary atmospheric dispersion factors (0.5 percentile)

(0.5 percentile)

C. Low Population Zone atmospheric dispersion factors Table 2.3-105 Table 2.3-105 (0.5 percentile)

(0.5 percentile)

Appendix 15B Appendix 15B D. Control room atmospheric dispersion factors Appendix 15BAppendix_15B V. Dose Data A Method of calculation App 15B Appendix 15B B. Isotopic data and dose conversion factors App 15B Appendix 15B C. Activity released to environment Table 15.6-14 Table 15.6-17 D. Offsite doses Table 15.6-18 Table 15.6-18 Table 15.6-21 Table 15.6-21 E.__Control room doses ______________________

FSAR Rev. 64 Page 2 of 2 Attachment 2 to PLA-7131 Technical Specification Bases Markups (Units 1 and 2)

PPL Rev. 9 Secondary Containment B 3.6.4.1 BASES SURVEILLANCE REQUIREMENTS (continued)

SR 3.6.4.1.4 and SR 3.6.4.1.5 The SGT System exhausts the secondary containment atmosphere to the environment through appropriate treatment equipment.

To ensure that all fission products are treated, SR 3.6.4.1.4 verifies that the SGT System will rapidly establish and maintain a pressure in the secondary containment that is less than the pressure external to the secondary containment boundary.This is confirmed by demonstrating that one SGT subsystem will draw down the secondary containment to > 0.25 inches of vacuum water gauge in less than or equal to the maximum time allowed. This cannot be accomplished if the secondary containment boundary is not intact. SR 3.6.4.1.5 demonstrates that one SGT subsystem can maintain > 0.25 inches of vacuum water gauge for at least 1 hour1.157407e-5 days <br />2.777778e-4 hours <br />1.653439e-6 weeks <br />3.805e-7 months <br /> at less than or equal to the maximum flow rate permitted for the secondary containment configuration that is operable.

The 1 hour1.157407e-5 days <br />2.777778e-4 hours <br />1.653439e-6 weeks <br />3.805e-7 months <br /> test period allows secondary containment to be in thermal equilibrium at steady state conditions.

As noted, both SR 3.6.4.1.4 and SR 3.6.4.1.5 acceptance limits are dependent upon the secondary containment configuration when testing is being performed.

The acceptance criteria for the SRs based on secondary containment configuration is defined as follows: SECONDARY MAXIMUM DRAWDOWN TIME(SEC)

MAXIMUM FLOW RATE (CFM)CONTAINMENT (SR 3.6.4.1.4 (SR 3.6.4.1.5 TEST CONFIGURATION ACCEPTANCE CRITERIA)

ACCEPTANCE CRITERIA)Group 1 Zones I, II and III (Unit 1 < 300 Seconds :! 4000 C-FM 4575 CFM Railroad Bay aligned to (Zones I, II, and Ill) (From Zones I, II, and III)Secondary Containment).

Zones I and III (Unit 1 300 Seconds <2899.CM 3275 CFM Railroad Bay aligned to (Zones I and Ill) (From Zones I and III)Secondary Containment).

Group 2 Zones 1, 11 and III (Unit 1 < 300 Seconds < 324,0 CFM 4475 CFM Railroad Bay not aligned to (Zones I, II, and III) (From Zones I, II, and III)Secondary Containment).

Zones I and III (Unit 1 r 300 Seconds < 2 G00 CFMW 3175 CFM Railroad Bay not aligned to (Zones I and Ill) (From Zones I and III)Secondary Containment).

L Only one of the above listed configurations needs to be tested to confirm secondary containment OPERABILITY.(continued)

I SUSQUEHANNA-UNIT 1 TS / B 3.6-89 Revision 5-No Changes This Page PPL Rev. 9-For Information Only -Secondary Containment I~ B3.6.4.1 BASES SURVEILLANCE SR 3.6.4.1.4 and SR 3.6.4.1.5 (continued)

REQUIREMENTS A Note also modifies the Frequency for each SR. This Note identifies that each configuration is to be tested every 60 months. Testing each configuration every 60 months assures that the most limiting configuration is tested every 60 months. The 60 month Frequency is acceptable because operating experience has shown that these components usually pass the Surveillance and all active components are tested more frequently.

Therefore, these tests are used to ensure secondary containment boundary integrity.

The Unit 1 Railroad Bay can be aligned as a No Zone (isolated from secondary containment) or as part of secondary containment (Zone I or Ill).Drawdown testing of the secondary containment shall be performed with the Unit I Railroad Bay aligned in the most limiting configuration.

More specifically, secondary containment drawdown testing will be performed with the Unit 1 Railroad Bay aligned as a No Zone with the Railroad Bay door open or as part of secondary containment.

The specific alignment will be selected based on the alignment that provides the least amount of inleakage and drawdown time margin (on a percentage basis) to the acceptance criteria.

This could result in one alignment (e.g., No Zone with the Railroad Bay door open) being limiting for one criterion (e.g., drawdown time) and the other alignment (e.g., Railroad Bay aligned to secondary containment) being limiting for the other criterion (e.g., inleakage).

It also could result in one alignment being limiting for both criteria.Note that aligning the Railroad Bay to either Zone I or III is acceptable since either zone is part of secondary containment.

It is preferred to align the Railroad Bay to Zone III for testing since Zone III is included in all possible secondary containment isolation alignments.

The most limiting Unit I Railroad Bay alignment shall be established each Surveillance period (60 month). Subsequent drawdown testing during the same Surveillance period only requires testing of the Unit 1 Railroad Bay in the most limiting configuration.

For example, Zone 1, 11, and III Surveillance testing is performed with the Unit 1 Railroad Bay aligned both as a No Zone with the Railroad Bay door open and as Zone Ill. If the Surveillance testing determined the most limiting configuration occurs with the Unit 1 Railroad Bay aligned as Zone Ill, then subsequent Zone I and III drawdown testing during the same Surveillance period only needs to be performed with the Unit 1 Railroad Bay aligned as Zone Ill.(continued)

SUSQUEHANNA

-UNIT 1 TS / B 3.6-90 Revision 3 No Changes This Page..:-For Information Only -PPL Rev. 9 Secondary Containment B 3.6.4.1 BASES SURVEILLANCE REQUIREMENTS SR 3.6.4.1.4 and SR 3.6.4.1.5 (continued)

Since these SRs are secondary containment tests, they need not be performed with each SGT subsystem.

The SGT subsystems are tested on a STAGGERED TEST BASIS, however, to ensure that in addition to the requirements of LCO 3.6.4.3, either SGT subsystem will perform SR 3.6.4.1.4 and SR 3.6.4.1.5.

Operating experience has shown these components usually pass the Surveillance when performed at the 24 month Frequency.

Therefore, the Frequency was concluded to be acceptable from a reliability standpoint.

REFERENCES

1. FSAR, Section 6.2.3.2. FSAR, Section 15.6.3. FSAR, Section 15.7.4.4. Final Policy Statement on Technical Specifications Improvements, July 22,1993 (58 FR 39132).SUSQUEHANNA-UNIT 1 TS / B 3.6-90a Revision 0 PPL Rev. 9 Secondary Containment B 3.6.4.1 BASES SURVEILLANCE REQUIREMENTS (continued)

SR 3.6.4.1.4 and SR 3.6.4.1.5 The SGT System exhausts the secondary containment atmosphere to the environment through appropriate treatment equipment.

To ensure that all fission products are treated, SR 3.6.4.1.4 verifies that the SGT System will rapidly establish and maintain a pressure in the secondary containment that is less than the pressure external to the secondary containment boundary.This is confirmed by demonstrating that one SGT subsystem will draw down the secondary containment to >_ 0.25 inches of vacuum water gauge in less than or equal to the maximum time allowed. This cannot be accomplished if the secondary containment boundary is not intact. SR 3.6.4.1.5 demonstrates that one SGT subsystem can maintain >_ 0.25 inches of vacuum water gauge for at least 1 hour1.157407e-5 days <br />2.777778e-4 hours <br />1.653439e-6 weeks <br />3.805e-7 months <br /> at less than or equal to the maximum flow rate permitted for the secondary containment configuration that is operable.

The 1 hour1.157407e-5 days <br />2.777778e-4 hours <br />1.653439e-6 weeks <br />3.805e-7 months <br /> test period allows secondary containment to be in thermal equilibrium at steady state conditions.

As noted, both SR 3.6.4.1.4 and SR 3.6.4.1.5 acceptance limits are dependent upon the secondary containment configuration when testing is being performed.

The acceptance criteria for the SRs based on secondary containment configuration is defined as follows: SECONDARY MAXIMUM DRAWDOWN TIME(SEC)

MAXIMUM FLOW RATE (CFWR CONTAINMENT (SR 3.6.4.1.4 (SR 3.6.4.1.5 TEST CONFIGURATION ACCEPTANCE CRITERIA)

ACCEPTANCE CRITERIA)Group 1 Zones 1, 11 and III (Unit 1 300 Seconds 4 4000 FM 4575 CFM Railroad Bay aligned to (Zones I, II, and 1II) (From Zones I, II, and Il)Secondary Containment).

Zones II and III (Unit 1 300 Seconds ý2iQ0-FM 3350 CFM Railroad Bay aligned to (Zones II and III) (From Zones II and Ill)Zone Ill).Group 2 Zones I, II and III (Unit 1 < 300 Seconds 3Q1- 75 CFM Railroad Bay not aligned to (Zones 1, 11, and Ill) (From Zones 1, 11, and Ill)Secondary Containment).

Zones II and III (Unit 1 300 Seconds 2756 GFM 3275 CFM Railroad Bay not aligned to (Zones II and Ill) (From Zones II and III)Secondary Containment).

Only one of the above listed configurations needs to be tested to confirm secondary containment OPERABILITY.(continued)

I SUSQUEHANNA-UNIT 2 TS / B 3.6-88 Revision 6-No Changes This Page I PPL Rev. 9-For Information Only -I Secondary Containment I B 3.6.4.1 BASES SURVEILLANCE SR 3.6.4.1.4 and SR 3.6.4.1.5 (continued)

REQUIREMENTS A Note also modifies the Frequency for each SR. This Note identifies that each configuration is to be tested every 60 months. Testing each configuration every 60 months assures that the most limiting configuration is tested every 60 months. The 60 month Frequency is acceptable because operating experience has shown that these components usually pass the Surveillance and all active components are tested more frequently.

Therefore, these tests are used to ensure secondary containment boundary integrity.

The Unit 1 Railroad Bay can be aligned as a No Zone (isolated from secondary containment) or as part of secondary containment (Zone I or Ill).Drawdown testing of the secondary containment shall be performed with the Unit 1 Railroad Bay aligned in the most limiting configuration.

More specifically, secondary containment drawdown testing will be performed with the Unit 1 Railroad Bay aligned as a No Zone with the Railroad Bay door open or as part of secondary containment.

The specific alignment will be selected based on the alignment that provides the least amount of inleakage and drawdown time margin (on a percentage basis) to the acceptance criteria.

This could result in one alignment (e.g., No Zone with the Railroad Bay door open) being limiting for one criterion (e.g., drawdown time) and the other alignment (e.g., Railroad Bay aligned to secondary containment) being limiting for the other criterion (e.g., inleakage).

It also could result in one alignment being limiting for both criteria.Note that aligning the Railroad Bay to either Zone I or III is acceptable since either zone is part of secondary containment when 3 zone testing is performed.

When a Zone II & III test is performed with the Unit I Railroad Bay aligned to Secondary Containment, it must be aligned to Zone III since aligning to Zone I will not allow communication with the isolated zones. The most limiting Unit I Railroad Bay alignment shall be established each Surveillance period (60 month). Subsequent drawdown testing during the same Surveillance period only requires testing of the Unit I Railroad Bay in the most limiting configuration.

For example, Zone I, II, and III Surveillance testing is performed with the Unit I Railroad Bay aligned both as a No Zone with the Railroad Bay door open and as Zone Il1. If the Surveillance testing determined the most limiting configuration occurs with the Unit 1 Railroad Bay aligned as Zone III, then subsequent Zone II and III drawdown testing during the same Surveillance period only needs to be performed with the Unit I Railroad Bay aligned as Zone III.(continued)

SUSQUEHANNA

-UNIT 2 TS / B 3.6-89 Revision 3 No Changes This. Page-For Information Only -PPL Rev. 9 Secondary Containment B 3.6.4.1 BASES SURVEILLANCE REQUIREMENTS SR 3.6.4.1.4 and SR 3.6.4.1.5 (continued)

Since these SRs are secondary containment tests, they need not be performed with each SGT subsystem.

The SGT subsystems are tested on a STAGGERED TEST BASIS, however, to ensure that in addition to the requirements of LCO 3.6.4.3, either SGT subsystem will perform SR 3.6.4.1.4 and SR 3.6.4.1.5.

Operating experience has shown these components usually pass the Surveillance when performed at the 24 month Frequency.

Therefore, the Frequency was concluded to be acceptable from a reliability standpoint.

REFERENCES

1. FSAR, Section 6.2.3.2. FSAR, Section 15.6.3. FSAR, Section 15.7.4.4. Final Policy Statement on Technical Specifications Improvements, July 22, 1993 (58 FR 39132).(continued)

SUSQUEHANNA-UNIT 2 TS / B 3.6-89a Revision 0 Attachment 3 to PLA-7131 Calculations

/ Control Room Habitability Envelope and Offsite Post LOCA Doses, and DBA-LOCA Total Control Room Dose (Attachment 3 is provided as a Compact Disk -Digital Media)