Text

Cooper Nuclear Station - License Amendment Request for Application of the Alternative Source Term for Calculating Loss-of-Coolant Accident Dose Consequences
	ML082910768
Person / Time
Site:	Cooper
Issue date:	10/13/2008
From:	Minahan S B; Nebraska Public Power District (NPPD)
To:	; Document Control Desk, Office of Nuclear Reactor Regulation
References
	NLS2008014
	Download: ML082910768 (145)
	v • d • e

N Nebraska Public Power District"Always there when you need us" 50.12 50.90 NLS2008014 October 13, 2008 U.S. Nuclear Regulatory Commission Attention:

Document Control Desk Washington, D.C. 20555-0001

Subject:

License Amendment Request for Application of the Alternative Source Term for Calculating Loss-of-Coolant Accident Dose Consequences Cooper Nuclear Station; Docket No. 50-298, DPR-46

References:

1. NUREG-1465, "Accident Source Terms for Light-Water Nuclear Power Plants," February 1995.2. Regulatory Guide 1.183, "Alternative Radiological.

Source Terms for Evaluating Design Basis Accidents at Nuclear Power Reactors," July 2000.

Dear Sir or Madam:

The purpose of this letter is for the Nebraska Public Power District (NPPD) to request Nuclear Regulatory Commission (NRC) approval for adopting the Alternative Source Term (AST), in accordance with 10 CFR 50.67, for use in calculating the Loss-of-Coolant Accident (LOCA)dose consequences at Cooper Nuclear Station (CNS). The requested approval of the LOCA AST involves changes to the CNS Technical Specifications (TS) to effectively increase the Main Steam Isolation Valve allowable leakage by establishing a higher leakage limit for the newly defined Main Steam (MS) Pathway, and revise the Standby Liquid Control (SLC) system TS to reflect crediting it for LOCA mitigation.

This License Amendment Request (LAR) for the LOCA is a selective scope application of the AST, as provided for in Reference 2.In conjunction with this LAR, NPPD is requesting that the NRC grant CNS an exemption from 1) the requirements of 10 CFR 50, Appendix J, Option B, Paragraph III.A to allow exclusion of the MS Pathway leakage, including the leakage from the MS inboard drain line, from the overall integrated leakage rate measured when performing a Type A test, and 2) the requirements of 10 CFR 50, Appendix J, Option B, Paragraph III.B, to allow exclusion of the MS Pathway 7(0) 7 COOPER NUCLEAR STATION P.O. Box 98 / Brownville, NE 68321-0098 A m(Telephone:

(402) 825-3811 / Fax: (402) 825-5211 wwwv.nppd.com NLS2008014 Page 2 of 4 leakage, including the MS inboard drain line, from the combined leakage rate of the penetrations and valves subject to Type B and C tests.Attachment 1 provides a description of the LAR, the basis for the amendment, the No Significant Hazards Consideration evaluation pursuant to 10 CFR 50.91 (a)(1), and the environmental impact evaluation pursuant to 10 CFR 51.22. Appendix A to Attachment 1 describes the conformance of this LAR to Reference

2. Attachment 2 provides responses to standard NRC Requests for Additional Information on crediting the use of SLC for Suppression Pool pH control requested of other licensees.

Attachment 3 is a request for exemption from the testing requirements of 10 CFR 50, Appendix J, Option B, Sections III.A and III.B, for the Main Steam Pathway.Attachment 4 identifies the proposed changes to the current CNS TS on marked up pages.Attachment 5 provides the revised TS pages in final typed format. Attachment 6 provides the corresponding changes to the current TS Bases on marked up pages for information.

NPPD has completed a new LOCA dose analysis using the guidelines detailed in References 1 and 2. This analysis, provided as Enclosure 1, demonstrates that the radiological dose consequences of a LOCA using the AST methodology are within regulatory limits. Enclosure 1 contains information considered to be proprietary to Alion Science and Technology Corporation.

Therefore, in accordance with the provisions contained in 10 CFR 2.390, it is requested that the NRC withhold Enclosure 1 from public disclosure.

As required by 10 CFR 2.390(b)(1)(ii), Enclosure 2 is an affidavit supplied by Alion Science and Technology Corporation supporting this request for withholding Enclosure 1 from public disclosure.

The non-proprietary version of the LOCA dose analysis is provided as Enclosure

3. The calculation of the Suppression Pool pH, an input into the LOCA dose calculation, is provided as Enclosure 4.NRC approval is requested by September 15, 2009, with a 30-day implementation period.Implementation of the LOCA AST has a high importance for future CNS performance, including conduct of the Cycle 25 refueling outage scheduled to begin in September 2009.These proposed TS changes have been reviewed by the necessary safety review committees (Station Operations Review Committee and Safety Review and Audit Board). Amendments to the CNS Facility Operating License through Amendment 231 issued June 30, 2008, have been incorporated into this request. NPPD has concluded that the proposed changes do not involve a significant hazards consideration and that they satisfy the categorical exclusion criterion of 10 CFR 51.22(c)(9).

This request is submitted under oath pursuant to 10 CFR 50.30(b).By copy of this letter and its attachments and enclosures, the appropriate State of Nebraska official is notified in accordance with 10 CFR 50.91(b)(1).

Copies to the NRC Region IV office and the CNS Resident Inspector are also being provided in accordance with 10 CFR 50.4(b)(1).

NLS2008014 Page 3 of 4 Should you have any questions concerning this matter, please contact David W. Van Der Kamp, Licensing Manager, at (402) 825-2904.I declare under penalty of perjury that the foregoing is true and correct.Executed on /6 //?"/)(Date)Sincerely, Stewart B. Minahan Vice President

-Nuclear and Chief Nuclear Officer/mb Attachments

1. License Amendment Request for Application of the Alternative Source Term for Calculating.

Loss-of-Coolant Accident Dose Consequences.

2. Response to Standard NRC Questions Regarding Crediting the Standby Liquid Control System for Loss of Coolant Accident (LOCA) Alternative Source Term (AST) Suppression Pool pH Control.3. Proposed Exemption to 10 CFR 50 Appendix J.4. Proposed Technical Specifications, Markup Format.5. Proposed Technical Specifications, Final Typed Format.6. Proposed Technical Specifications Bases Revisions, Markup Format.Enclosures

1. NEDC 07-082, Rev. 2, "Radiological Dose Analysis for a Loss of Coolant Accident (LOCA)at Cooper Nuclear Station," (Proprietary Version).2. ALION Science and Technology Corporation Affidavit Required by 10 CFR 2.390.3. NEDC 07-082, Rev. 2, "Radiological Dose Analysis for a Loss of Coolant Accident (LOCA)at Cooper Nuclear Station," (Non-proprietary Version).4. NEDC 07-071, Rev. 0, "Review of Alion Calculation ALION-CAL-NPPD-3236-003, Cooper Nuclear Station Post-Accident Suppression Pool pH Analysis."

NLS2008014 Page 4 of 4 cc: Regional Administrator w/attachments, enclosures USNRC -Region IV Cooper Project Manager w/attachments, enclosures USNRC -NRR Project Directorate IV-1 Senior Resident Inspector w/attachments, enclosures USNRC -CNS Nebraska Health and Human Services w/attachments, enclosures Department of Regulation and Licensure NPG Distribution w/o attachments, enclosures CNS Records w/attachments, enclosures NLS2008014 Attachment 1 Page 1 of 75 ATTACHMENT 1 License Amendment Request for Application of the Alternative Source Term for Calculating Loss-of-Coolant Accident Dose Consequences Cooper Nuclear Station, Docket 50-298, DPR-46 1.0 Description

2.0 Proposed

Change 2.1 Licensing Basis Change for Loss-of-Coolant Accident 2.2 Technical Specification Changes 3.0 Background

4.0 Technical

Analysis'4.1 Radiological Consequences of the Loss-of-Coolant Accident 4.2 Atmospheric Dispersion (X/Q)4.3 Post-Accident Access to Vital Areas 4.4 Proposed Technical Specification Changes 5.0 Regulatory Safety Analysis 5.1 No Significant Hazards Consideration

5.2 Applicable

Regulatory Requirements/Criteria

6.0 Environmental

Consideration

7.0 References

7.1 General

References

7.2 Precedent

Appendix A Regulatory Guide 1.183 Comparison NLS2008014 Attachment 1 Page 2 of 75

1.0 DESCRIPTION

The proposed change revises the radiological assessment calculational methodology for the design basis accident.(DBA)

Loss-of-Coolant Accident (LOCA) at Cooper Nuclear Station (CNS) through application of the Alternative Source Term (AST), in accordance with the provisions of 10 CFR 50.67, "Accident Source Term." The Nebraska Public Power District (NPPD) requests Nuclear Regulatory Commission (NRC) review and approval of the AST LOCA methodology for application to CNS. This application represents a selective scope application of AST, as provided for in Regulatory Guide (RG) 1.183 (Reference 7.1-1).The LOCA AST dose calculation, upon which this License Amendment Request (LAR)is based, is provided as Enclosure

1. This calculation was developed using the NRC-approved RADTRAD Version 3.03 software (Reference 7.1-2). Approval of this LAR will replace the current design basis source term assumptions and radiological criteria for the LOCA. In accordance with the AST LOCA analysis results, revisions to the CNS Technical Specifications (TS) and TS Bases are proposed based on the revised safety analysis assumptions for a postulated LOCA.The proposed TS changes consist of (1) revising the TS definition for DOSE EQUIVALENT 1-131 to adopt Federal Guidance Report (FGR) 11 dose conversions factors, (2) changing the Standby Liquid Control (SLC) system TS to require operability of the system in Mode 3 to reflect its credit in the LOCA analysis, (3) establishing a Main Steam (MS) Pathway leakage limit that effectively increases the previous MSIV leakage limit, and (4) reflecting in TS Section 5.5.12 the requested permanent exemption from the requirements of Appendix J, Option B, Paragraph III.A, to allow exclusion of MS Pathway leakage from the overall integrated leakage rate measured during the performance of a Type A test, and from the requirements of Appendix J, Option B, Paragraph III.B, to allow exclusion of the MS Pathway leakage from the combined leakage rate of the penetrations and valves subject to Type B and C tests.The implementation of these changes would reduce radiological dose to personnel during the conduct of future planned outages while maintaining an adequate safety margin.

2.0 PROPOSED CHANGE

NPPD is proposing to modify (a) the licensing basis of the LOCA as described in the CNS Updated Safety Analysis Report (USAR), and (b) Technical Specifications 3.1.7,"Standby Liquid Control System," and 3.6.1.3, "Primary Containment Isolation Valves." 2.1 Licensing Basis Change for Loss-of-Coolant Accident NPPD is revising the licensing basis of the LOCA described in Section XIV-6.3 of the CNS USAR. The proposed licensing basis change is the use of AST NLS2008014 Attachment 1 Page 3 of 75 methodology for dose consequences analysis in accordance with 10 CFR 50.67, RG 1.183 (Reference 7.1-1), and NUREG-1465 (Reference 7.1-3). Sections 3,4, and 5 of this attachment provide the background, the technical analysis, and the regulatory safety analysis respectively, in support of the licensing basis change.The USAR will be revised as an implementing action following issuance of the license amendment, with revised USAR pages submitted pursuant to 10 CFR 50.71(e).2.2 Technical Specification Changes The following TS changes are proposed: 1. TS 1.1, "Definitions," is revised to change the definition of DOSE EQUIVALENT 1- 131 to reflect the dose conversion factors contained in Federal Guidance Report (FGR) 11, "Limiting Values of Radionuclide Intake and Air Concentration." 2. TS 3.1.7, "Standby Liquid Control (SLC) System" is revised by adding MODE 3 to the APPLICABILTY, and by adding Required Action C.2 for CNS to be in MODE 4, with a Completion Time of 36 hours4.166667e-4 days 0.01 hours 5.952381e-5 weeks 1.3698e-5 months .3. In TS 3.6.1.3, "Primary Containment Isolation Valves (PCIVs)," Surveillance Requirement (SR) 3.6.1.3.10 currently states: "Verify combined main steam leakage rate is < 46 scth when tested at > 29 psig." This SR is revised to state: "Verify leakage rate through each MSIV line is < 106 scfh when tested at > 29 psig." 4. Also in TS 3.6.1.3, new SR 3.6.1.3.12 is added to state: "Verify leakage through the Main Steam Pathway is < 212 scfh when tested at > 29 psig." The Frequency is specified as "In accordance with the Primary Containment Leakage Rate Testing Program." 5. In TS 5.5.12, "Primary Containment Leakage Rate Testing Program," paragraph "a," exception numbers 4 and 5 discuss exemptions from Section III.A and III.B, respectively, from 10 CFR Part 50, Appendix J, Option B. These exceptions are revised by replacing "MSIV" with "Main Steam Pathway (Main Steam lines and the Main Steam inboard drain line)." Also, the date of the approved exemption in parentheses is to be revised from "October 30, 2006" to the date that the exemption is approved.

NLS2008014 Attachment 1 Page 4 of 75 Conforming revisions to the associated TS Bases are needed and will be made in accordance with TS 5.5.10, Technical Specifications Bases Control Program, as part of the implementation of the amendment following issuance.

These TS Bases revisions are provided for information in Attachment 6 of this submittal.

3.0 BACKGROUND

In December 1999, the NRC issued a new regulation, 10 CFR 50.67, "Accident Source Term," which provided a mechanism for licensed power reactors to voluntarily replace the traditional accident source term used in their DBA analyses with an AST. Regulatory guidance for the implementation of the AST is provided in RG 1.183, "Alternative Radiological Source Terms for Evaluating Design Basis Accidents at Nuclear Power Reactors." 10 CFR 50.67 requires a licensee seeking to use an alternative source term to apply for a license amendment and requires that the application contain an evaluation of the consequences of DBAs.This LAR addresses the applicable requirements and guidance in proposing selectively to use an AST in evaluating the offsite and control room radiological consequences of a LOCA. This reanalysis involves several changes in selected-analysis assumptions including different atmospheric dispersion values for the Control Room outside air intake. As part of the implementation of the AST, the Total Effective Dose Equivalent (TEDE) acceptance criterion of 10 CFR 50.67(b) replaces the previous whole body and thyroid dose guidelines of 10 CFR 100.11. This will also replace the whole body (and its equivalent to any part of the body) dose criteria of 10 CFR 50, Appendix A, GDC 19.4.0 TECHNICAL ANALYSIS 4.1 Radiological Consequences of the Loss-of-Coolant Accident The objective of analyzing the radiological consequences of a LOCA is to evaluate the performance of various plant safety systems intended to mitigate the postulated release of radioactive materials from the plant to the environment, and ensure calculated doses are within regulatory limits. In accordance with RG 1.183 guidance, NPPD determined the inventory of fission products in the reactor core based on the Original Licensed Thermal Power (OLTP) plus a 2% margin for power measurement uncertainty using an appropriate isotope generation and depletion computer code.1 Fission products from the damaged fuel are released into the Reactor Coolant System (RCS) and then into the Primary Containment By letter dated June 30, 2008, the NRC issued Amendment No. 231 to the CNS Operating License, which authorized an increase in the Maximum Power Level from the OLTP level of 2381 MWt to 2419 MWt. This increase, allowed by 10 CFR 50, Appendix K, was based on increased Feedwater System flowrate measurement accuracy achieved via installation of improved flow instrumentation.

Therefore, analysis of the LOCA using the AST methodology at 1.02 % of OLTP is consistent with the current licensing basis and 10 CFR 50 Appendix K.

NLS2008014 Attachment 1 Page 5 of 75 (i.e., drywell and wetwell).

The gap inventory release phase begins two minutes after the event starts and is assumed to continue for 30 minutes. As a conservatism, the CNS LOCA analysis excludes this two-minute delay. As the core continues to degrade, the gap inventory release phase ends and the in-vessel release phase begins. This phase continues for 1.5 hours5.787037e-5 days 0.00139 hours 8.267196e-6 weeks 1.9025e-6 months . Tables 1, 4, and 5 of RG 1.183 define the source term used for these two phases.The inventory in each release phase is released at a constant rate over the duration of the phase, starting at the onset of the phase. Once dispersed in the Primary Containment, the release to the environment via Secondary Containment or directly to the environment is assumed to occur through four pathways: 1. Leakage of Primary Containment atmosphere to the environment during a brief period in which the Secondary Containment pressure is positive;2. Leakage of Primary Containment atmosphere to Secondary Containment and subsequent release to the atmosphere via Standby Gas Treatment (SGT)System (i.e., design leakage);3. Leakage of Primary Containment atmosphere via design leakage to the environment through the Main Steam Pathway (i.e., proposed TS leakage limit);4. Leakage from Emergency Core Cooling Systems (ECCS) that recirculate Suppression Pool water outside of the Primary Containment to Secondary Containment (i.e., design leakage).Primary Containment (the drywell and suppression chamber) is enclosed in Secondary Containment (the Reactor Building).

Under accident conditions, leakage from the Primary Containment that enters the Secondary Containment is ultimately processed by the SGT system and released from an elevated stack. The Reactor Building exhaust fans, which normally maintain the building at a negative pressure, trip at accident initiation.

A brief period of time may exist prior to the SGT System establishing a negative pressure condition in the Secondary Containment.

A CNS pressurization analysis indicates that the Reactor Building may have a positive pressure for 210 seconds in the worst case of a failure of the intake damper to close. The dose analysis conservatively assumes that the drywell releases directly to the environment for 5 minutes as a ground release in accordance with RG 1.183, Appendix A, "Assumptions on Dual Containments No. 4.2." According to RG 1.183, Table 4, the onset of gap release does not occur until two minutes into the accident.

The LOCA dose analysis conservatively ignores this two-minute delay.The potential for Secondary Containment Bypass Leakage (SCBL) has been considered.

NPPD performed an engineering evaluation to identify potential NLS2008014 Attachment 1 Page 6 of 75 SCBL paths. This evaluation determined that the only SCBL pathway is the assumed Main Steam Pathway leakage to the condenser.

Table 1 provides the key LOCA analysis assumptions used in Enclosure 1.Table 2 provides the calculated LOCA radiological consequences.

4.1.1 Suppression

Pool Post-LOCA pH Control According to NUREG- 1465 (Reference 7.1-3), the iodine entering the containment from the RCS during an accident would be composed of at least 95% cesium iodide (CsI). Upon deposition on interior surfaces and dissolution in the Suppression Pool of a Boiling Water Reactor (BWR), the predominant form of the iodine would be the iodide ion (1). At pH less than 7.0 a large fraction of the iodide could be converted by irradiation into elemental (gaseous) iodine (12) and released into the containment atmosphere.

If the pH was maintained above 7.0, however, the fraction of F converted into 12 would be less than 1%. Since the pH of the Suppression Pool is not normally controlled, 12 may be released during a LOCA as the acids, which are produced due to the radiation effects of the LOCA, lower the pH.One way to minimize this release is to add an alkaline chemical capable of buffering the pH at a value above 7.0. As described in Enclosure 4, NPPD proposes to do this by adding sodium pentaborate (Na 2 B1001 6 *I 0H 2 0)from the Standby Liquid Control (SLC) system during a LOCA. Although the SLC system was designed as a backup method to maintain the reactor subcritical without control rods after an Anticipated Transient Without Scram (ATWS), it can be used for pH control. NPPD proposes to use the SLC system to inject sodium pentaborate into the Reactor Pressure Vessel (RPV), where it will mix with ECCS flow and spill over to the drywell and then to the Suppression Pool. Sodium pentaborate, a base, will neutralize acids generated in the post-accident Primary Containment environment.

NPPD used a combination of known parameters and conservative assumptions as inputs to calculate pH at discrete times for 30 days following the postulated accident.

Credit for the SLC system in the radiological analyses is based on operation of one SLC pump, initiated within 6 hours6.944444e-5 days 0.00167 hours 9.920635e-6 weeks 2.283e-6 months after the event starts, with injection completed within 8 hours9.259259e-5 days 0.00222 hours 1.322751e-5 weeks 3.044e-6 months . This credit assumes the injection of the entire contents of the SLC system sodium pentaborate solution storage tank with a solution concentration that meets the limits specified in CNS Technical Specifications 3.1.7. As an implementing action following issuance of the amendment, CNS operating procedures will be revised to direct operators to manually initiate the SLC system upon detection of symptoms NLS2008014 Attachment 1 Page 7 of 75 indicating that a LOCA with core damage is occurring.

The ability of the SLC system to perform this function is presented in Attachment 2.4.1.1.1 Nitric Acid Nitric acid (HNO 3) is produced by the irradiation of water and air following a LOCA. The amount of nitric acid was calculated using a water radiolysis model. Nitric acid production is proportional to the integrated value of a time-dependent radiation dose rate in the Suppression Pool. Enclosure 4 provides the predicted amount of nitric acid that would be produced over the 30-day period.4.1.1.2 Hydriodic Acid Hydriodic Acid is an aqueous solution of hydrogen iodide (HI), and is formed when iodine is released from the core as fuel failure occurs. NUREG- 1465 (Reference 7.1-3) specifies that 5% of the core halogen inventory is released during the gap release phase while an additional 25% is released during the early in-vessel phase. Consistent with Section 4.5 of NUREG-1465, no more than 5% of the iodine exiting the reactor coolant system will be composed of elemental iodine (12) and hydriodic acid (HI). It is conservatively assumed that the entire 5% of this release is in the form of HI in order to maximize the acid generation.

This release is assumed to occur at a constant rate over the release period (i.e., 30 and 90 minutes for the gap and early in-vessel release phases, respectively).

The inventory includes the stable 1127 isotope to maximize acid production.

Enclosure 4 provides the predicted amount of hydriodic acid that would be produced over the release period.4.1.1.3 Hydrochloric Acid Hydrochloric acid (HCI) is generated by the irradiation of chloride-bearing cable jacketing.

Its amount is proportional to the radiation energy absorbed by the jacketing.

The methodology developed for calculating production of hydrochloric acid is based on NUREG/CR-5950 (Reference 7.1-4) and NUREG 1081, "Post-Accident Gas Generation from Radiolysis of Organic Materials." The rate of production of HC1 is given by the following equations:

R(t) = G. S -(t)

A Where: R -rate of generation of HC1 NLS2008014 Attachment 1 Page 8 of 75 G -radiation "G" value for jacket material S -surface area of cable y -incident radiation energy flux A -absorption fraction of energy flux in the jacket The resulting HCI generated is described in Enclosure 4.4.1.1.4 Sodium Pentaborate Buffering In order to counter the effect of the increasing nitric, hydriodic, and hydrochloric acids in the Suppression Pool, sodium pentaborate solution from the SLC system would be added within six hours to buffer the pH in the alkaline range, with injection completed within eight hours. The injection would be accomplished from the Control room with a keylock switch manipulation, which is an existing operator action intended for reactivity control in the core.The sodium pentaborate inventory required by TS 3.1.7 is sufficient to maintain the Suppression Pool pH above 7.0 for the 30-day accident duration.4.1.1.5 Standby Liquid Control System Considerations The Suppression Pool pH calculation credits the manual initiation of SLC within six hours following initiation of the accident.Review of recent precedent has revealed a set of standard NRC questions used to judge the acceptability of crediting SLC as an accident mitigation feature. The responses to these questions are provided in Attachment 2 of this LAR.. In summary, the non-essential SLC system meets the design and performance requirements necessary to credit its operation post-LOCA.

4.1.2 Alternate

Leakage Treatment RG 1.183, Appendix A, provides assumptions that are acceptable for the evaluation of radiological consequences of the design basis LOCA using AST. For BWR MSIV leakage, RG 1.183 allows credit for reducing MSIV releases due to holdup and deposition in the Main Steam piping downstream of the MSIVs and in the Main Condenser if the components and piping systems used in the release path are capable of performing their safety functions during and following a safe shutdown earthquake (SSE).The LOCA AST dose calculation credits an Alternate Leakage Treatment (ALT) strategy that reduces the MSIV leakage dose consequences due to the holdup and deposition provided by the downstream components.

The ALT pathway (hereafter referred to as the "Main Steam Pathway" or "MS NLS2008014 Attachment 1 Page 9 of 75 Pathway")

includes the 1) the Main Steam lines, originating at the reactor vessel, continuing through the MSIVs to the Main Steam equalizing header, down the drop to the Turbine Bypass Valves' strainers, and through the strainer drain line to the Main Condenser, and 2) the inboard Main Steam drain lines (which tap off the MSIVs inside Primary Containment) to their junction with the outboard Main Steam drain lines (which tap off the Main Steam lines just downstream of the outboard MSIVs), and continuing to the Main Condenser.

In CNS License Amendments 196 and 206 (References 7.1-5 and 7.1-6), the NRC accepted the NPPD evaluation that demonstrated the seismic ruggedness of the Turbine Building, the Main Condenser, and the Main Steam Pathway, as well as the manual actions needed to configure the pathway. In a response to an NRC request for additional information in support of CNS License Amendment 206 (Reference 7.1-7), NPPD discussed the manual actions required to align the Main Steam Pathway and validation of the times required to complete these alignments.

As discussed in Reference 7.1-7 and documented in Amendment 206, NPPD's evaluation of the manual actions resulted in a 2.5 hour5.787037e-5 days 0.00139 hours 8.267196e-6 weeks 1.9025e-6 months limit for manually aligning 14 of 16 valves, and a time limit of 30 hours3.472222e-4 days 0.00833 hours 4.960317e-5 weeks 1.1415e-5 months for manually aligning the two additional valves and installing the shaft adjustment tools (shaft sealing mechanisms) on the Turbine Stop valves.These time limits were based on an evaluation of the times expected for the accident source term to travel down the Main Steam Pathway and pose a radiological hazard to the personnel performing the alignments following a LOCA and assuming fuel damage. NPPD implemented these actions in CNS Procedure 5.2FUEL, establishing completion times of 2.5 hours5.787037e-5 days 0.00139 hours 8.267196e-6 weeks 1.9025e-6 months to align the 14 of 16 valves, and 7.5 hours5.787037e-5 days 0.00139 hours 8.267196e-6 weeks 1.9025e-6 months to align the two additional valves and install the Turbine Stop valve sealing mechanisms.

NPPD has re-evaluated these times using the AST LOCA conditions.

That evaluation has resulted in no change to the 2.5 hour5.787037e-5 days 0.00139 hours 8.267196e-6 weeks 1.9025e-6 months limit for aligning the 14 of 16 valves discussed above. However, implementation of AST LOCA will result in a reduction of the allowed completion time for alignment of the two additional valves and installation of the Turbine Stop valve sealing mechanisms to 5 hours5.787037e-5 days 0.00139 hours 8.267196e-6 weeks 1.9025e-6 months . As discussed in Reference 7.1-7, alignment of the valves would normally be performed by an on-shift plant operator.

Installation of the valve sealing mechanisms would be performed by maintenance personnel deployed out of the Operations Support Center following activation of the emergency response organization (ERO). Walkdowns estimated that the time required for a single individual to install the pre-staged Turbine Stop valve sealing mechanisms would be 30 minutes upon personnel entry into the Turbine Building.

While the allowed time frame for installing the Turbine Stop NLS2008014 Attachment 1 Page 10 of 75 valve sealing mechanisms is less than the time previously allowed, it is still well within a reasonable time frame, even assuming activation of the ERO, briefing, and dress-out, if required.The changes described and requested in the AST LOCA amendment request do not invalidate the previous basis for NRC acceptance of the MS Pathway seismic evaluation or underlying basis for acceptance of the credited manual actions. Therefore, NPPD credits the previously approved seismic evaluation and manual actions for alignment of the MS Pathway in this LAR.4.1.3 Containment Leakage Pathway The AST analysis assumes that the Primary Containment leaks at its design leakage rate of 0.635% of its contents by weight per day for the first 24 hours2.777778e-4 days 0.00667 hours 3.968254e-5 weeks 9.132e-6 months of a LOCA and then at 0.3175% by weight per day for the remainder of the 30-day accident duration.

RG 1.183, Appendix A, Section 3.7 states that for BWRs, Primary Containment leakage may be reduced after the first 24 hours2.777778e-4 days 0.00667 hours 3.968254e-5 weeks 9.132e-6 months , if supported by plant configuration and analysis, to a value not less than 50% of the TS leak rate.The drywell leakage into the reactor building is set at 0.635% per day at the design basis LOCA maximum peak accident pressure Pa. The CNS LOCA containment analysis that gives the worst case leakage at 24 hours2.777778e-4 days 0.00667 hours 3.968254e-5 weeks 9.132e-6 months and beyond shows that there is only one minute before the pressure within the drywell falls to 42.3 psia and the temperature falls below 300'F. After 24 hours2.777778e-4 days 0.00667 hours 3.968254e-5 weeks 9.132e-6 months , the drywell leakage into the reactor building is reduced because the containment analysis shows that under worst case cooling conditions, the long term peak drywell pressure is about 22 psig (240'F) and occurs in approximately eight hours post-accident.

Note that the steam line wall temperature chosen is the design value; following a LOCA this temperature will decrease as soon as the MSIVs close. Similarly, the Main Condenser temperature chosen is based on the design flows and temperatures entering the condenser during normal operation which will also decrease once the MSIVs close.When the ratio of the drywell pressure to the downstream pressure (atmospheric) is greater than approximately two, (i.e., critical pressure ratio of 0.55 for air and steam), the resulting mass flow rate, W 2 can be determined from:

NLS2008014 Attachment 1 Page 11 of 75 rA7g 2(Eq. 1)-V Where: AF = leakage flow area 7 = ratio of the specific heats, (1.28 for steam)P1 = drywell pressure (lb/&)gC = gravitational constant (ft/sec 2)v1 = drywell specific volume (ft 3/lb)It can be shown that by accounting for the change in drywell pressure and steam specific volume over time that the leakage can be reduced to 34% of the original value at 24 hours2.777778e-4 days 0.00667 hours 3.968254e-5 weeks 9.132e-6 months . The leakage in this analysis is conservatively reduced to 50% of the original value, or 0.3175%/day at 24 hours2.777778e-4 days 0.00667 hours 3.968254e-5 weeks 9.132e-6 months in accordance with RG 1.183, Appendix A, Item No. 3.7 under"Assumptions on Transport in Primary Containment." The release mixes homogeneously with the drywell air space. No credit is taken for mixing with the wetwell air space. Credit is taken for natural deposition inside the Primary Containment.

NPPD models this deposition using the 10-percentile model described in NUREG/CR-6189, "A Simplified Model of Aerosol Removal by Natural Processes in Reactor Containments" (i.e., the "Powers Model") (Reference 7.1-8). No credit is taken for Suppression Pool scrubbing.

Leakage from the Primary Containment will collect in the free volume of the Secondary Containment and be released to the environment following treatment in the SGT System. During the period of Secondary Containment positive pressurization, leakage is assumed to be released from the Reactor Building directly to the environment as a ground level release. This unfiltered release phase is conservatively modeled to occur for the first 5 minutes of the event, with additional conservatism provided by assuming the gap release phase begins immediately (the two-minute gap release delay not credited).

To maximize release to the environment, both SGT System trains are assumed to be in operation for the first hour, with an assumed failure of a filter heater in one of the trains. This heater failure is the limiting single failure for determining the AST LOCA dose consequence.

The assumed heater failure results in a reduced SGT System filter efficiency for the train with the failed heater. An assumed failure of filter heater power in one train results in manually securing the faulted SGT System train at one hour.

NLS2008014 Attachment 1 Page 12 of 75 4.1.4 MS Pathway Leakage The four Main Steam lines, which penetrate Primary Containment, are automatically isolated by closure of the MSIVs in the event of a LOCA.There are two MSIVs on each steam line, one inside containment and one outside containment.

The MSIVs are functionally part of the Primary Containment boundary.

The MS Pathway includes the Main Steam lines, the MSIVs, and the MS inboard and outboard drain lines. The MS Pathway is a path for transporting fission products, via MSIV and inboard MS line leakage to the Main Condenser, and thus bypassing Secondary Containment.

Leakage from the condenser is assumed to occur as a ground release from the Turbine Building.

NPPD conservatively assumes that the fission products released from the core are dispersed equally throughout the drywell. Following the initial blowdown of the RPV, the fuel heats up with some fuel damage before the core is cooled.Subsequently, the steaming in the RPV carries fission products to the Primary Containment.

When core cooling is restored, steam is rapidly generated in the core. This steam and the ECCS flow carry fission products from the core to the Primary Containment via the severed recirculation line, resulting in well-mixed RPV dome and Primary Containment fission product concentrations.

However, in this analysis, the fission products are assumed to be available for release via leakage through the MSIVs and the MS inboard drain lines.Per RG 1.183, Regulatory Position 6.3, credit is allowed for aerosol and elemental iodine holdup and deposition in the Main Steam lines and the Main Condenser if these piping systems and components have been demonstrated to be capable of performing their function following an SSE.The assumptions for crediting holdup and plate out in the Main Condenser and Main Steam lines are justified based on the discussion provided in Section 4.1.2 regarding the seismic ruggedness of the Main Steam Pathway. Credit was taken for holdup and deposition in the condenser, however holdup and deposition in the Main Steam lines were conservatively neglected.

The LOCA AST calculation models two of the four Main Steam lines with a total leakage of 300 scfh at accident pressure (equivalent to the aggregate leakage limit of 212 scfh that would be allowed by the proposed changes to the CNS TS when tested at reduced pressure).

The calculation conservatively assumes a flow of 150 scfh at accident pressure in each of two Main Steam lines (equivalent to the per-line leakage limit of 106 scfh that would be allowed by the proposed changes to the CNS TS when tested at reduced pressure).

This treatment in the analysis was assumed because it resulted in a conservative determination of deposition NLS2008014 Attachment 1 Page 13 of 75 efficiencies in the Main Steam lines, although these were conservatively not credited.The condenser model was developed based on the methodology presented in NEDC-31858P-A, "BWROG Report for Increasing MSIV Leakage Rate Limits and Elimination of Leakage Control Systems," (Reference 7.1-9).That model calculates an effective filter efficiency based on deposition and holdup in the condenser, and is based on the condenser free volume above the highest Main Steam Pathway condenser penetration.

The development of this model is presented in Enclosure 1.4.1.5 Leakage From ECCS During the progression of a LOCA, some fission products released from the fuel will be carried to the Suppression Pool via spillage from the broken loop of the Reactor Recirculation System. During the later stages of LOCA mitigation, the Suppression Pool is a source of water for ECCS.Since portions of these systems are located outside of the Primary Containment, leakage from these systems is evaluated as a potential radiation exposure pathway. For the purposes of assessing the consequences of leakage from the ECCS, it is assumed that 100% of the radioiodines released from the fuel are transported to the Suppression Pool as they are released.

This source term assumption is conservative in that 100% of the radioiodines released from the fuel are assumed to be available for ECCS leakage, Primary Containment leakage, and MSIV leakage. In actuality, the radioiodines in the Primary Containment atmosphere would relocate to the Suppression Pool over time. Noble gases released from the fuel are assumed to remain in the drywell atmosphere.

Since aerosols and particulate radionuclides are not expected to become airborne on release from the ECCS, they are not included in the ECCS source term. These assumptions follow the guidance of RG 1.183.The analysis considers ECCS leakage of 45,000 cubic centimeters per minute (cc/min) starting at the onset of a LOCA. It is assumed that 10%of the iodine in the ECCS leakage becomes airborne and is available for release. Use of 10% is consistent with the guidance of RG 1.183 as the peak Suppression Pool water temperature does not exceed 212'F. No credit was assumed for holdup and dilution in the Secondary Containment.

As was assumed for the Primary Containment leakage pathway, the leakage enters the environment via the SGT System as a filtered elevated release, with a percentage that bypasses the filter.

NLS2008014 Attachment 1 Page 14 of 75 4.1.6 Offsite Doses NPPD evaluated the maximum two-hour TEDE to an individual located at the Exclusion Area Boundary (EAB) and the 30-day TEDE to an individual at the outer boundary of the Low Population Zone (LPZ). The resulting doses are less than the requirements of 10 CFR 50.67 (see Table 2).4.1.7 Control Room Doses To calculate the dose in the Control Room, two air intake flow rates were used, representing operation of the Control Room Air Conditioning System (normal ventilation system) and the Control Room Emergency Filtration System (CREFS). Upon receipt of a LOCA signal (Reactor Vessel Water Level -Low Low, Level 2 or Drywell Pressure -High) the CREFS is automatically initiated, and achieves isolation within 11 seconds. However, this analysis conservatively assumes that the Control Room Air Conditioning System remains in operation for the first minute of the event. Therefore, a normal air intake flowrate of 3235 cubic feet per minute (cfm), plus an assumed unfiltered inleakage of 400 cfm for a total unfiltered intake flow of 3635 cfm for the first minute of the event. The unfiltered inleakage assumption is conservative relative to the inleakage values reported in the NPPD response to Generic Letter 2003-01 (Reference 7.1-10). After the first minute, the CREFS is assumed to be in operation and draws in 900 cfm +/- 10% for the duration of the accident.For this analysis, a value of 810 cfm was conservatively assumed to minimize radionuclide removal from the Control Room, with 400 cfm of unfiltered inleakage also assumed during CREFS operation.

The post-LOCA 30-day gamma shine dose to Control Room personnel was calculated.

The shine contributors to Control Room dose were from the most significant sources: outside cloud, the Reactor Building, a Core Spray line, the CREFS filter, and the Primary Containment.

Microshield Version 5.05 was used in this calculation.

This software is a Point Kernel Integration code used for general purpose gamma shielding analysis in safety-related applications by many nuclear plants in the United States.The resulting 30-day TEDE to an individual in the Control Room is less than the 1OCFR50.67 criteria (see Table 2).4.2 Atmospheric Dispersion (X/Q)The Control Room intake X/Q values used in this analysis were developed using ARCON96 and are based on meteorological data previously submitted to and NLS2008014 Attachment 1 Page 15 of 75 accepted by the NRC. The X/Q values were previously submitted by NPPD and accepted by the NRC for use at CNS as part of the current LOCA calculation of record (Reference 7.1-5).2 The existing LOCA analysis models an elevated release from Secondary Containment, with fumigation assumed for the first 30 minutes, and a diffuse ground level release from the Turbine Building from the wall closest to the Control Building air intake for modeling the Main Condenser release. The AST LOCA analysis assumes the 30 minute fumigation conditions occur at 1.3 hours3.472222e-5 days 8.333333e-4 hours 4.960317e-6 weeks 1.1415e-6 months into the event, during the worst two-hour exposure period, in accordance with RG 1.183, Regulatory Position 5.3. Additionally, the AST LOCA analysis models a 5 minute ground release from the Reactor Building at the beginning of the event. The 5 minute ground release is based on a pressurization analysis that assumes, in the worst case, a 210 second positive Reactor building pressure following event initiation, assuming a failure of the air-operated Reactor Building Ventilation supply damper to close.The X/Qs used for the EAB and LPZ were calculated using site specific inputs and methodology described in RG 1.3. The Turbine Building release to the offsite locations was corrected for building wake effects. These X/Q values were previously reviewed and accepted by the NRC in License Amendment 196 for the current LOCA analysis of record (Reference 7.1-5). However, the X/Q values for the EAB and the LPZ were also adjusted to model for 30 minute fumigation during the worst two-hour interval, beginning at 1.3 hours3.472222e-5 days 8.333333e-4 hours 4.960317e-6 weeks 1.1415e-6 months into the event.4.3 Post-Accident Access to Vital Areas In response to Item II.B.2 of NUREG-0737, "Clarification of TMI Action Plan Requirements," NPPD identified vital areas in the station where personnel access was required post-accident, and determined that GDC 19 limits would not be exceeded by personnel in these areas. The CNS vital area analyses are based on application of TID-14844.

Regulatory Position 1.3.2 of RG 1.183 discusses an evaluation of AST versus TID-14844 performed by the NRC for three representative operating reactors, and determined that the TID-14844 and its associated whole body and thyroid methodology generally bound the results from analyses based on the AST and TEDE methodology.

The RG also states that licensees may use the applicable conclusions of that evaluation in addressing the impact of the AST on design basis radiological analyses.

Therefore, the existing analyses will be bounding over an AST-based analysis.

The NRC recently accepted this position in Reference 7.2-1, with Vermont Yankee having provided a bounding analysis comparing the time-dependent gamma radiation emission 2. The X/Q calculation (NEDC 99-031, Revision 4) was submitted in a letter from J. Swailes (NPPD),"Proposed License Amendment Related to the Design Basis Accident Radiological Assessment Methodology," dated September 14, 2001. As discussed in that submittal, the revised X/Q values for the elevated release were not incorporated into the DBA calculations, as the previous values in Revision 3 of the calculation were conservative.

NLS2008014 Attachment 1 Page 16 of 75 characteristics of the isotopic mix assumed in the AST to the isotopic mix of the TID-14844 source term. Notwithstanding this position, Regulatory Position 1.3.2 of RG 1.183 states that licensees are not exempt from evaluating the remaining radiological and nonradiological impacts of the AST implementation and the impacts of the associated plant modifications.

NPPD is proposing no plant modifications as part of this LAR. The Main Steam Pathway analysis leakage is increased from 46 scfh when tested at 29 psig to 212 scfh when tested at 29 psig; however, that is offset by the increased Decontamination Factors (DFs) due to holdup and plateout in the Main Condenser.

These increased DFs are a function of the acceptability of the Main Steam Pathway rather than from AST characteristics.

As part of the implementation of the LAR, NPPD will update the vital area analyses to reflect these conclusions.

4.4 Proposed

Technical Specification Changes This LAR proposes the following TS changes: a. TS 1.1, "Definitions," would be revised to change the definition of DOSE EQUIVALENT 1-131 to reflect the dose conversion factors contained in Federal Guidance Report (FGR) 11, "Limiting Values of Radionuclide Intake and Air Concentration." This change is proposed pursuant to the guidance contained in R.G. 1.83, Regulatory Position 4.1.2.b. TS 3.1.7, "Standby Liquid Control System," would be revised by adding MODE 3 to the APPLICABILITY section, and adding Required Action C.2 to be in MODE 4 within 36 hours4.166667e-4 days 0.01 hours 5.952381e-5 weeks 1.3698e-5 months . These changes are needed since a LOCA could occur in MODES 1, 2, and 3. These TS changes make the SLC System consistent with other SSCs contained in the CNS Technical Specifications LOCA mitigation.

c. TS Surveillance Requirement (SR) 3.6.1.3.10 would be revised by replacing the combined leakage rate with a limit for each MSIV line. The new allowable MSIV leakage rate for each line would be "< 106 scfh when tested at 29 psig." New SR 3.6.1.3.12 is proposed that would establish a new allowable aggregate Main Steam Pathway leakage limit, increasing the current SR 3.6.1.3.10 limit of"< 46 scfh when tested at > 29 psig" to "< 212 scfh when tested at > 29 psig." Maintaining leakage within these values ensures that the analyzed dose contribution via this pathway remains bounding.The inclusion of leakage limits (< 106 scfh for each MSIV line and < 212 scfh for the MS Pathway) at > 29 psig serves to identify leakage requirements at a NLS2008014 Attachment 1 Page 17 of 75 reduced test pressure (the MSIVs are tested at the reduced pressure, and not at Pa). The reduced test pressure leakage rates of< 106 scfh and < 212 scfh were determined based on the methodology as designated in the American Society of Mechanical Engineers (ASME) Operating and Maintenance Code, Section ISTC-3600 (Reference 7.1-11). The relationship between the leakage rates at accident pressure to the specified leakage rates at the test pressure are discussed in the TS Bases for clarification, and provided for information in Attachment 6.d. TS 5.5.12.a.4 would be revised to reflect an exemption from Section III.A of 10 CFR 50, Appendix J, Option B, to allow the leakage contribution from the MS Pathway (Main Steam lines and the Main Steam inboard drain line)leakage to be excluded from the overall integrated leakage rate from Type A tests. CNS currently has an exemption for the MSIV leakage. This change adds the leakage through the MS inboard drain line.e. TS 5.5.12.a.5 would be revised to reflect an exemption from Section III.B of 10 CFR 50, Appendix J, Option B, to allow the contribution from the MS Pathway (Main Steam lines and the Main Steam inboard drain line)leakage to be excluded from the sum of the leakage rates from Type B and Type C tests. CNS currently has an exemption for the MSIV leakage. This change adds the leakage through the MS inboard drain line.

NLS2008014 Attachment 1 Page 18 of 75 Table 1 LOCA Analysis Assumptions (Page 1 of 3)Reactor power Core inventory Core release fractions and timing Iodine species fraction (%)Particulate/aerosol Elemental Organic Drywell volume, ft 3 Suppression Pool water volume, ft 3 Peak Accident Containment Pressure (Pa), psig Primary Containment leakage, volume %/day 0 -24 hours> 24 hours2.777778e-4 days 0.00667 hours 3.968254e-5 weeks 9.132e-6 months MSIV allowable leakage (per valve), scfh (at Pa)Main Steam Pathway Leakage (total), scfh (at Pa)0 -24 hours> 24 hours2.777778e-4 days 0.00667 hours 3.968254e-5 weeks 9.132e-6 months Duration of release, days Drywell natural deposition, %Main Steam line deposition analysis-No credit taken for Main Steam line deposition Condenser deposition elemental

& particulate efficiency 0-24 hours24-720 hours 2429 Mwt (102% Power)*See Table 5-1 of Enclosure 1 RG 1.183, Tables 1 and 4 95 4.85 0.15 132,250 96,445 58 0.635 0.3175 150 300 150 30 10 (Powers Model)94.91%97.39%MSIV leakage is released from Main Condenser to environment without dilution or holdup in Turbine Building ECCS leak rate (flashed), cfm 0.159 ECCS iodine species fraction, %Particulate/aerosol 0 Elemental 97 Organic 3*2429 Mwt is 102% of the OLTP of 2381 Mwt NLS2008014 Attachment 1 Page 19 of 75 Table 1 LOCA Analysis Assumptions (Page 2 of 3)ECCS iodine flash fraction 0.1 SGT System Elevated Release flow, cfm Active train 1492 Idle train (0 -1 hr) 1492 (1 -720 hr) 288 SGT System filter efficiency (Active), %Particulates 98 Elemental 94 Organic 94 SGT System filter efficiency (Idle), %Particulates 98 Elemental 89 Organic 29 Control room volume, ft 3 141,860 Normal ventilation unfiltered fresh intake, cfm 3,235 CREFS filtered intake, cfm 810 CREFS start delay time, minutes 1 Unfiltered outside inleakage, cfm 400 CREFS filter efficiency, %Particulates 98 Elemental 89 Organic 89 Control room occupancy factors 0-24 hr 1.0 24-96 hr 0.6 96-720 hr 0.4 Control room breathing rate, m 3/sec 3.5E-4 Offsite breathing rate, m 3/sec 0-8 hr 3.5E-4 8-24 hr 1.8E-4 24-720 hr 2.3E-4 NLS2008014 Attachment 1 Page 20 of 75 Table 1 LOCA Analysis Assumptions (Page 3 of 3)Atmospheric dispersion factors, sec/mr 3 A. SGT System Elevated Release Period (hours) EAB 0-.083 1.6E-05 0.083-1.3 1.6E-05 1.3-1.8 1.2E-04 1.8-2.0 1.6E-05 2.0-8.0 8.0-24 24-96 96-720 LPZ 4.OOE-05 4.OOE-05 1.40E-04 4.OOE-05 4.OOE-05 1.60E-05 5.80E-06 1.70E-06 Control Room 4.15E-03 1.00E-10 3.03E-04 1.OOE- 10 8.58E-10 1.41E-08 5.62E-09 5.69E-09 Control Room 8.64E-04 4.66E-04 2.32E-04 1.53E-04 1.25E-04 B. Turbine Building Ground Level Diffuse Release Period (hours) EAB LPZ 0.0-2.0 5.20E-04 2.90E-04 2.0-8.0 2.90E-04 8.0-24 7.30E-05 24-96 2.50E-05 96-720 5.20E-06 NLS2008014 Attachment 1 Page 21 of 75 Table 2 Calculated LOCA Radiological Consequences TEDE (rem)Calculated results Primary Containment Main Steam Pathway ESF Leakage Shine EAB 0.458 0.375 0.170 N/A 1.002 LPZ 1.559 2.311 1.727 N/A 5.596 Control Room 0.374 2.401 0.102 0.319 3.196 Total Dose acceptance criteria (10 CFR 50.67)25 25 5 NLS2008014 Attachment 1 Page 22 of 75 5.0 REGULATORY SAFETY ANALYSIS 5.1 No Significant Hazards Consideration 10 CFR 50.91 (a)(1) requires that licensee requests for operating license amendments be accompanied by an evaluation of significant hazard posed by issuance of an amendment.

In accordance with 10 CFR 50.92, a proposed change to the operating facility involves no "significant hazards" if operation of the facility, in accordance with the proposed change, would not 1) involve a significant increase in the probability or consequences of any accident previously evaluated, 2) create the possibility of a new or different kind of accident from any accident previously evaluated, or 3)involve a significant reduction in a margin of safety.The Nebraska Public Power District (NPPD) is requesting an amendment of Operating License No. DPR-46 for Cooper Nuclear Station (CNS). The amendment consists of revisions to the CNS Technical Specifications (TS): (1) TS Section 1.1, "Definitions," is revised to change the definition of DOSE EQUIVALENT 1- 131 to reflect the dose conversion factors contained in Federal Guidance Report (FGR) 11, "Limiting Values of Radionuclide Intake and Air Concentration," as recommended in NRC RG 1.183.(2) TS Section 3.1.7, "Standby Liquid Control (SLC) System," is revised by adding Mode 3 to the Applicability, and adding a Required Action to be in Mode 4 within 36 hours4.166667e-4 days 0.01 hours 5.952381e-5 weeks 1.3698e-5 months in the event that one SLC subsystem is inoperable greater than 7 days or both SLC subsystems are inoperable greater than 8 hours9.259259e-5 days 0.00222 hours 1.322751e-5 weeks 3.044e-6 months .(3) TS Section 3.6.1.3, "Primary Containment Isolation Valves," Surveillance Requirement (SR) 3.6.1.3.10 is revised by specifying anew limit for leakage through each Main Steam line, and new SR 3.6.1.3.12 is added to specify a new aggregate leakage limit for the Main Steam Pathway.(4) TS Section 5.5.12, "Primary Containment Leakage Rate Testing Program," paragraph "a," exception numbers 4 and 5 is revised by replacing "MSIV" with "Main Steam Pathway (Main Steam lines and the Main Steam inboard drain line) to reflect the requested exemptions from 10 CFR Part 50, Appendix J, Option B, Sections III.A and III.B, respectively.

The proposed amendment would revise the accident source term used in the radiological consequence analyses for the design basis loss-of-coolant accident (LOCA) in accordance with Title 10 of the Code of Federal Regulations (10 CFR)

NLS2008014 Attachment 1 Page 23 of 75 Part 50.67, "Accident Source Term." The revised accident source term replaces the current methodology that is based on Technical Information Document (TID) -14844, "Calculation of Distance Factors for Power and Test Reactor Sites," with the Alternative Source Term (AST) methodology described in Nuclear Regulatory Commission (NRC) Regulatory Guide (RG) 1.183,"Alternative Radiological Source Terms for Evaluating Design Basis Accidents at Nuclear Power Reactors." The proposed license amendment is for implementation of the AST as described in RG 1.183 for the design basis LOCA.The proposed amendment includes a revision of two current exemptions from the requirements of 10 CFR Part 50, Appendix J. The current exemptions authorize exclusion of the Main Steam Isolation Valve (MSIV) leakage from the Primary Containment leakage required to be measured and accounted for under 10 CFR 50, Appendix J, "Primary Reactor Containment Leakage Testing for Water-Cooled Power Reactors." The technical basis for the current exemption is that the MSIV leakage would bypass Primary Containment following a LOCA, and this leakage is accounted for separately in the current LOCA dose calculation of record. The proposed amendment would change this technical basis only by defining and establishing the Main Steam (MS) Pathway, which includes leakage from the MSIVs and the inboard MS drain line. Leakage through the newly defined MS Pathway (including leakage contribution from the inboard MS drain line) is accounted for separately from the Primary Containment leakage in the AST LOCA calculation submitted with this License Amendment Request (LAR).Therefore, accompanying this LAR is a request for exemption from the requirements to include the measured MS Pathway leakage in the determination of Primary Containment leakage required in accordance with 10 CFR 50 Appendix J.The exemptions being revised are reflected in TS 5.5.12, Primary Containment Leakage Rate Testing Program.NPPD has evaluated whether the proposed amendment involves a significant hazards consideration based on the three standards set forth in 10 CFR 50.92,"Issuance of amendment." The evaluation concludes that the proposed amendment involves no significant hazards considerations.

The following is that evaluation.

1. Do the proposed changes involve a significant increase in the probability or consequences of an accident previously evaluated?

Response:

No For the postulated design basis accident (DBA) LOCA, the AST is an input to the calculations that evaluate the radiological consequences of a LOCA. The AST and the requested Appendix J exemption do not affect the design of the plant or the manner in which the plant is normally operated.

Adoption of the NLS2008014 Attachment 1 Page 24 of 75 AST and the requested Appendix J exemption do not affect the initiators of a DBA. Neither the AST nor the requested Appendix J exemption affect the response to the DBA LOCA, or the pathway of the radiation released from the nuclear fuel. Rather, the AST better represents the physical characteristics of the radiation release.Because the initiators of a DBA are not affected by adoption of the AST for LOCA dose assessment, the probability of an accident are not increased by the proposed amendment or requested Appendix J exemption.

The AST is an input to calculations used to evaluate the radiological consequences of the LOCA. Use of the AST does not affect the plant response to the accident, or the pathways to the environment for the radiation and activity released from the fuel. The LOCA radiological analyses have been performed using the AST. Adoption of the AST methodology revises the acceptance criteria for the accident to the limits specified in 10 CFR 50.67.The results of those analyses demonstrate that the dose consequences are within the acceptance criteria presented in 10 CFR 50.67 and in NRC RG 1.183.Implementation of the AST for the LOCA involves the use of the SLC System to control the pH of the suppression pool during mitigation of a LOCA. As a result the proposed amendment revises the CNS TS for the SLC System.These changes do not require any physical modification of the plant, nor result in any change in normal plant operation.

This additional use of the SLC system does not compromise or adversely affect the function of the SLC system as a means of shutting down the reactor in addition to the control rods.Therefore, it is concluded that adoption of AST and granting of the Appendix J exemption do not involve a significant increase in the consequences of an accident previously evaluated.

Based on the above discussion, it is concluded that the proposed changes do not involve a significant increase in the probability or consequences of an accident previously evaluated.

2. Do the proposed changes create the possibility of a new or different kind of accident from any accident previously evaluated?

Response:

No Implementation of the LOCA AST and the requested Appendix J exemption do not involve a physical alteration of the plant or a change in how the plant is normally operated.

No new or different types of equipment will be installed NLS2008014 Attachment 1 Page 25 of 75 and there are no physical modifications to existing equipment associated with the proposed changes. The proposed changes, effectively increasing the allowable MSIV leakage, establishing a leakage limit for the MS Pathway, and crediting the SLC system for LOCA mitigation do not create initiators or precursors of a new or different kind of accident.

New equipment or personnel failure modes that might initiate a new type of accident are not created as a result of the proposed amendment.

Therefore, the proposed changes do not create the possibility of a new or different kind of accident from any previously analyzed.3. Do the proposed changes involve a significant reduction in a margin of safety?Response:

No The proposed amendment involves the implementation of a new licensing basis for the design basis LOCA. Approval of this change from the original source term to an AST, derived in accordance with the guidance of RG 1.183, results in revised acceptance criteria for the LOCA analysis.

For the LOCA, RG 1.183 sets the Exclusion Area Boundary (EAB), Low Population Zone (LPZ), and Control Room limit consistent with 10 CFR 50.67. The AST LOCA radiological analysis has been performed using conservative methodologies, as specified in RG 1.183. Safety margins have been evaluated and confirmed to have not been reduced. Analytical conservatism has been utilized to ensure that the analysis adequately bounds the limiting postulated event. The dose consequences of the DBA LOCA remain within the acceptance criteria presented in 10 CFR 50.67 and RG 1.183.The proposed changes continue to ensure that the doses at the EAB and LPZ boundary, as well as the Control Room, are within the corresponding regulatory limits.Since the proposed amendment continues to ensure the doses at the EAB, LPZ and Control Room are within corresponding regulatory limits, the proposed license amendment does not involve a significant reduction in a margin of safety.Based on the above, NPPD 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.

NLS2008014 Attachment 1 Page 26 of 75 5.2 Applicable Regulatory Regquirements/Criteria The construction of CNS predated the 1971 issuance of 10 CFR 50 Appendix A,"General Design Criteria for Nuclear Power Plants." CNS is designed to be in conformance with the intent of the Draft General Design Criteria (GDC), published in the Federal Register on July 11, 1967, except where commitments have been made to specific 1971 GDCs. The applicable GDCs are: Draft GDC 10 -Containment"Containment shall be provided.

The containment structure shall be designed to sustain the initial effects of gross equipment failures, such as a large coolant'boundary break, without loss of required integrity and, together with other engineered safety features as may be necessary, to retain for as long as the situation requires the functional capability to protect the public." This LAR does not alter CNS commitments to conformance with this Draft GDC, except to the degree that 10 CFR 50.67 replaces 10 CFR 100 as the regulatory requirement for public protection from the consequences of a LOCA.Draft GDC 37 -Engineered Safety Features Basis for Design"Engineered safety features shall be provided in the facility to back up the safety provided by the core design, the reactor coolant pressure boundary, and their protection systems. As a minimum, such engineered safety features shall be designed to cope with any size reactor coolant pressure boundary break up to and including the circumferential rupture of any pipe in that boundary assuming unobstructed discharge from both ends." This LAR proposes crediting the SLC system as an Engineered Safety Feature (ESF). As discussed in Section 4.1.1.4, the SLC system is capable of performing its Suppression Pool pH control function during the design basis LOCA. No modifications to the system are required for it to perform this post-LOCA function.Draft GDC 38 -Reliability and Testability of Engineered Safety Features"All engineered safety features shall be designed to provide high functional reliability and ready testability.

In determining the suitability of a facility for a ,proposed site, the degree of reliance upon and acceptance of the inherent and;engineered safety afforded by the systems, including engineered safety features, will be influenced by the known and the demonstrated performance capability and'reliability of the systems, and by the extent to which the operability of such systems can be tested and inspected where appropriate during the life of the NLS2008014 Attachment 1 Page 27 of 75, plant." This LAR proposes crediting the SLC system as an ESF. As discussed in Attachment 2, the SLC system meets the necessary reliability and testability of an ESF.Draft GDC 40 -Missile Protection"Protection for engineered safety features shall be provided against dynamic effects and missiles that might result from plant equipment failures." This LAR proposes crediting the SLC system as an ESF. Due to the location of SLC system components, the system is not vulnerable to the dynamic effects and missiles resulting plant equipment failures that could be postulated to occur post-LOCA.Draft GDC 41 -Engineered Safety Features Performance Capability"Engineered safety features such as emergency core cooling and containment heat removal systems shall provide sufficient performance capability to accommodate partial loss of installed capacity and still fulfill the required safety function.

As a i minimum, each engineered safety feature shall provide this required safety function assuming a failure of a single active component." This LAR proposes crediting the SLC system as an ESF. As discussed in Attachment 2, the SLC system does not fully meet the single failure criteria.However, the high reliability and excellent performance history of the susceptible components justifies an exception to this criterion.

Draft GDC 42 -Engineered Safety Features Components Compatibility"Engineered safety features shall be designed so that the capability of each component and system to perform its required function is not impaired by the effects of a loss-of-coolant accident." This LAR proposes crediting the SLC system as an ESF. As discussed in Attachment 2, the SLC system can perform its required function under the loadings of a Safe Shutdown Earthquake and under the environmental conditions experienced during a LOCA.Draft GDC 43 -Accident Aggravation Prevention"Engineered safety features shall be designed so that any action of the engineered safety features which might accentuate the adverse after-effects of the loss of NLS2008014 Attachment 1 Page 28 of 75 normal cooling is avoided." This LAR proposes crediting the SLC system as an ESF. The SLC system is i credited with acting in a positive manner in mitigating the consequences of a LOCA without adverse after-effects.

Operation of other ESFs does not accentuate the adverse after-effects of the loss of normal cooling using AST methodology.

Draft GDC 54 -Containment Leakage Rate Testing"Containment shall be designed so that integrated leakage rate testing can be conducted at design pressure after completion and installation of all penetrations and the leakage rate measured over a sufficient period of time to verify its conformance with required performance."'The proposed amendment does not alter the design of the containment.

As a result, the ability to conduct leakage rate testing at design pressure would not be adversely impacted.

Thus, the requirements of this criterion will continue to be met with the proposed exclusion of MS Pathway leakage.Draft GDC 55 -Containment Periodic Leakage Rate Testing i"The containment shall be designed so that integrated leakage rate testing can be done periodically at design pressure during plant lifetime." The proposed amendment does not alter the design of the containment.

As a'result, the ability to perform periodic testing of the containment would not be* adversely impacted.

Thus, the requirements of this criterion will continue to be met with the proposed exclusion of the MS Pathway leakage.'Draft GDC 57 -Provisions for Testing Isolation Valves I "Capability shall be provided for testing functional operability of valves and associated apparatus essential to the containment function for establishing that no failure has occurred and for determining that valve leakage does not exceed ,acceptable limits." ,Excluding MS Pathway leakage rates does not affect CNS conformance to this draft GDC. Leakage rate testing will still provide assurance that no failure has occurred and that valve leakage does not exceed acceptable limits.1971 GDC 19 -Control Room NPPD is committed to the provisions of 1971 GDC 19 as it applies to the ,radiological exposures to Control Room occupants:

NLS2008014 Attachment 1: Page 29 of 75"Adequate radiation protection shall be provided to permit access and occupancy of the control room under accident conditions without personnel receiving radiation exposures in excess of 5 rem whole body, or its equivalent to any part of the body, for the duration of the accident." With adoption of AST, the dose of 5 rem whole body (or equivalent dose to any part of the body) for the LOCA will be replaced with the 5 rem TEDE limits of 10 CFR 50.67 (as provided for in GDC 19).10 CFR 50.67, "Accident Source Term" 10 CFR 50.67 permits licensees to voluntarily revise the accident source term used in design basis radiological consequence analyses.

NRC issuance of the license amendment will result in a revision to the evaluation and consequences of a design basis LOCA currently reported in the Updated Safety Analysis Report.10 CFR 50.67 will supersede 10 CFR 50, Appendix A, GDC 19 as the dose*requirements for Control Room exposure limits, and 10 CFR 100 for EAB and LPZ exposure limits.,10 CFR 50 Appendix E, Paragraph IV.E.8 Paragraph IV.E.8 requires adequate provisions be made for an onsite Technical Support Center (TSC) and near-site Emergency Operations Facility (EOF) from which effective direction can be given and effective control can be exercised'during an emergency.

Adoption of the AST for the LOCA will not affect CNS ,conformance to the NUREG-0654, "Criteria for Preparation and Evaluation of Radiological Emergency Response Plans and Preparedness in Support of Nuclear Power Plants," as described in the CNS Emergency Plan.10 CFR 100, Paragraph 11, "Determination of Exclusion Area, Low Population Zone and Population Center Distance"'This paragraph provides criteria for evaluating the radiological aspects of reactor sites. A footnote to 10 CFR 100.11 states that the fission product release assumed in these evaluations should be based on a major accident involving substantial meltdown of the core with subsequent release of appreciable quantities of fission;products.

A similar footnote appears in 10 CFR 50.67.'In accordance with the provisions of 10 CFR 50.67(a), the radiation dose reference values in 10 CFR 50.67(b)(2) were used in these analyses in lieu of'those prescribed in 10 CFR 100. (Refer to footnote 5 on page 1.183-7 of Reference 7.1 -1).

NLS2008014 Attachment 1: Page 30 of 75 NUREG-0737 Commitments

1. Post-Accident Access Shielding (NUREG-0737, II.B.2) -As discussed in Section 4.3, the existing analyses will remain bounding for the AST methodology.

2. Post-Accident Sampling Capability (NUREG-0737, II.B.3) -Adoption of the AST for the LOCA does not adversely affect NPPD commitments to this Action Item.3. Accident Monitoring Instrumentation (NUREG-0737, II.F.1) -Adoption of the AST for the LOCA does not adversely affect NPPD commitments to this Action Item.4. Leakage Control (NUREG-0737, III.D.1.1)

-Adoption of the AST for the LOCA does not adversely affect NPPD commitments to this Action Item.5. Emergency Response Facilities (NUREG-0737, III.A.1.2)

-For CNS, 10 CFR 50 Appendix E Paragraph IV.E.8 subsumes the commitments to this Action Item. Adoption of the AST for the LOCA does not adversely affect NPPD compliance with these requirements.

6. Control Room Habitability (NUREG-0737, III.D.3.4)

-This LAR replaces NPPD commitments to GDC- 19 radiological protection for Control Room personnel (see 1971 GDC 19, above), with 10 CFR 50.67 as the applicable radiological regulatory requirement.

In conclusion, based on the considerations discussed above, (1) there is reasonable

'assurance that the health and safety of the public will not be endangered by operation in the proposed manner, (2) such activities will be conducted in compliance with the Commission's regulations, and (3) the issuance of the amendment will not be inimical to the common defense and security or to the health and safety of the public.6.0 ENVIRONMENTAL CONSIDERATION I A review has determined that the proposed amendment would change a requirement with respect to installation or use of a facility component located within the restricted area, as defined in 10 CFR 20. However, the proposed amendment does not involve (i) a significant hazards consideration, (ii) a significant change in the types or significant increase in the amounts of any effluent that may be released offsite, or (iii) a significant increase in individual or cumulative occupational radiation exposure.

Accordingly, the proposed amendment meets the eligibility criterion for categorical exclusion set forth in 10 CFR 51.22(c)(9).

Therefore, pursuant to 10 CFR 51.22(b), no environmental impact NLS2008014 Attachment 1 Page 31 of 75 statement or environmental assessment need be prepared in connection with the proposed amendment.

7.0 REFERENCES

7.1 General

References

1. USNRC Regulatory Guide 1.183, "Alternative Radiological Source Terms for Evaluating Design Basis Accidents at Nuclear Power Reactors," July 2000.2. NUREG/CR-6604, "RADTRAD:

A Simplified Model for Radionuclide Transport and Removal and Dose Estimation," S. L. Humphries et al, dated December 1997, including Supplement 1 (dated June 8, 1999) and Supplement 2 (dated October 2, 2002).3. NUREG-1465, "Accident Source Terms for Light-Water Nuclear Power Plants," L. Soffer et al., February 1995.4. NUREG/CR-5950, "Iodine Evolution and pH Control," dated December 1992.5. Letter from M. Thadani (NRC) to C. Warren (NPPD), dated February 21, 2003, "Cooper Nuclear Station -Issuance of Amendment Regarding Design Basis Accidents' Radiological Dose Assessment Methodologies, and Revision to License Condition 2.C.(6) (TAC No. MB4654)."[Amendment No. 196]6. Letter from M. Honcharik (NRC) to R. Edington (NPPD), dated September 1, 2004, "Cooper Nuclear Station (CNS) -Issuance of Amendment On Loss-of-Coolant Accident (LOCA) Dose Methodology and Resolution of Remaining License Condition 2.C.(6) Issues (TAC No.MC 1572)." [Amendment 206],7. Letter from R. Edington (NPPD) to USNRC, dated June 8, 2004,"Response to Request for Additional Information Regarding Loss-of-Coolant Accident (LOCA) Dose Calculation Methodology and Resolution of Remaining License Condition 2.C.(6) Issues, Cooper Nuclear Station, NRC Docket No. 50-298, DPR-46." 8. NUREG/CR-6189, "A simplified Model of Aerosol Removal by Natural Processes in Reactor Containments," July, 1996.

NLS2008014 Attachment 1 Page 32 of 75 9. NEDC-31858P-A, "BWROG Report for Increasing MSIV Leakage Rate Limits and Elimination of Leakage Control Systems," August, 1999.10. Letter from R. Edington (NPPD) to USNRC, dated September 30, 2004,"Initial Actions Summary Report and Response to NRC Generic Letter 2003-01, 'Control Room Habitability"' (NLS2004105).

11. ASME Operating and Maintenance Code, 2001 Edition through 2003 Addenda.7.2 Precedent There are numerous NRC-approved BWR precedents for use of AST for the LOCA. The following are approved LOCA AST amendments that were considered in the development of this License Amendment Request.1. Letter from R. Ennis (NRC) to M. Kansler (Entergy Nuclear Operations), dated March 29, 2005, "Vermont Yankee Nuclear Power Station -Issuance of Amendment Re: Alternative Source Term (TAC No.MC0253)." 2. Letter from E. Brown (NRC) to K. Singer (Tennessee Valley Authority), dated September 27, 2004, "Browns Ferry Nuclear Plant, Units 1, 2, and 3-Issuance of Amendments Regarding Full-Scope Implementation of Alternative Source Term (TAC Nos. MB5733, MB5734, MB5735, MC0156, MC0157 and MC0158) (TS-405)." 3. Letter from B. Mozafari (NRC) to J. Keenan (Carolina Power & Light), dated May 30, 2002, "Brunswick Steam Electric Plant, Units 1 and 2 -Issuance of Amendment Re: Alternative Source Term (TAC Nos. MB2570 and MB2571)." 4. Letter from B. Mozafari (NRC) to G. Van Middlesworth (Nuclear Management Company) dated July 31, 2001, "Duane Arnold Energy Center -Issuance of Amendment Regarding Alternative Source Term (TAC No. MB0347)."

NLS2008014 Attachment 1 Page 33 of 75 Appendix A Regulatory Guide 1.183 Comparison This Appendix provides a comparison of the Cooper Nuclear Station LOCA AST calculation (Enclosure

1) to the criteria of RG 1.183-that pertain to-BWR guidance-for LOCA-AST selective scope analysis development-and License Amendment Request content.Regulatory Position No. Regulatory Position Statement CNS Analysis 1. IMPLEMENTATION OF AST 1.1.1 The proposed uses of an AST and the associated proposed facility modifications Conforms.

Adequate safety margins are Safety and changes to procedures should be evaluated to determine whether the maintained, as discussed in the No Margins proposed changes are consistent with the principle that sufficient safety margins Significant Hazards Consideration.

are maintained, including a margin to account for analysis uncertainties.

The Future changes will be evaluated under safety margins are products of specific values and limits contained in the the provisions of 10 CFR 50.59.technical specifications (which cannot be changed without NRC approval) and other values, such as assumed accident or transient initial conditions or assumed safety system response times. Changes, or the net effects of multiple changes, that result in a reduction in safety margins may require prior NRC approval.Once the initial AST implementation has been approved by the staff and has become part of the facility design basis, the licensee may use 10 CFR 50.59 and its supporting guidance in assessing safety margins related to subsequent facility modifications and changes to procedures.

NLS2008014 Attachment 1 Page 34 of 75 Appendix A Regulatory Guide 1.183 Comparison Regulatory Position No. Regulatory Position Statement CNS Analysis 1.1.2 (a) The proposed uses of an AST and the associated proposed facility Conforms.

Adoption of the AST for the Defense in modifications and changes to -procedures should be evaluated to determine LOCA makes possible increased Depth whether the proposed changes are consistent with the principle that adequate allowable leakage from the MS defense in depth is maintained to compensate for uncertainties in accident Pathway. No changes are proposed that progression and analysis data. Consistency with the defense-in-depth affect system redundancy, philosophy is maintained if system redundancy, independence, and diversity are independence, and diversity.

preserved commensurate with the expected frequency, consequences of Compliance with the GDCs are challenges to the system, and uncertainties.

In all cases, compliance with the maintained (both the 1967 draft GDCs General Design Criteria in Appendix A to 10 CFR Part 50 is essential.

and the 1971 GDCs to which NPPD has Modifications proposed for the facility generally should not create a need for committed).

While SLC System compensatory programmatic activities, such as reliance on manual operator initiation will be a licensed manual actions. action to stay within the assumptions of the LOCA AST calculation, no modifications are proposed that rely on compensatory programmatic actions (including manual operator actions) to maintain adequate defense-in-depth.

1.1.2 (b) Proposed modifications that seek to downgrade or remove required Conforms.

NPPD proposes to increase Defense in engineered safeguards equipment should be evaluated to be sure that the the MS Pathway allowable leakage rate Depth modification does not invalidate assumptions made in facility PRAs and does as part of the LOCA AST LAR. This not adversely impact the facility's severe accident management program. action has no impact on the assumptions of the CNS Probabilistic Safety Assessment and does not adversely affect the CNS severe accident management program.

NLS2008014 Attachment 1 Page 35 of 75 Appendix A Regulatory Guide 1.183 Comparison Regulatory Position No. Regulatory Position Statement CNS Analysis 1.1.3 (a) The design basis accident source term is a fundamental assumption upon Conforms.

See RG Section 1.3.1 Integrity of which a significant portion of the facility design is based. Additionally, many discussions.

Facility aspects of facility operation derive from the design analyses that incorporated Design Basis the earlier accident source term. Although a complete re-assessment of all facility radiological analyses would be desirable, the NRC staff determined that recalculation of all design analyses would generally not be necessary.

Regulatory Position 1.3 of this guide provides guidance on which analyses need updating as part of the AST implementation submittal and which may need updating in the future as additional modifications are performed.

1.1.3 (b) This approach would create two tiers of analyses, those based on the Conforms.

The CNS LOCA AST Integrity of previous source term and those based on an AST. The radiological acceptance License Amendment Request is a Facility criteria would also be different with some analyses based on whole body and selective scope AST application.

Design Basis thyroid criteria and some based on TEDE criteria.

Full implementation of the Approval of this License Amendment AST revises the plant licensing basis to specify the AST in place of the previous Request will supersede the previous accident source term and establishes the TEDE dose as the new acceptance design basis source term assumptions criteria.

Selective implementation of the AST also revises the plant licensing and radiological criteria for the LOCA.basis and may establish the TEDE dose as the new acceptance criteria.

Future revisions of this analysis, if any, Selective implementation differs from full implementation only in the scope of will use the updated approved the change. In either case, the facility design bases should clearly indicate that assumptions and criteria.the source term assumptions and radiological criteria in these affected analyses have been superseded and that future revisions of these analyses, if any, will use the updated approved assumptions and criteria.1.1.3 (c) Radiological analyses generally should be based on assumptions and inputs Conforms.

The LOCA analysis is a Integrity of that are consistent with corresponding data used in other design basis safety selective scope application of the AST.Facility analyses, radiological and nonradiological, unless these data would result in It relies on assumptions and inputs that Design Basis nonconservative results or otherwise conflict with the guidance in this guide. do not create a conflict with, or render non-conservative, other design basis safety analyses.

NLS20080 14 Attachment 1 Page 36 of 75 Appendix A Regulatory Guide 1.183 Comparison Regulatory Position No. Regulatory Position Statement CNS Analysis 1.1.4 (a) Requirements for emergency preparedness at nuclear power plants are set NRC statement of fact.-Emergency forth in 10 CFR 50.47, "Emergency Plans." Additional requirements are set Preparedness forth in Appendix E, "Emergency Planning and Preparedness for Production Applications and Utilization Facilities," to 10 CFR Part 50. The planning basis for many of these requirements was published in NUREG-0396, "Planning Basis for the Development of State and Local Government Radiological Emergency Response Plans in Support of Light Water Nuclear Power Plants." This joint effort by the Environmental Protection Agency (EPA) and the NRC considered the principal characteristics (such as nuclides released and distances) likely to be involved for a spectrum of design basis and severe (core melt) accidents.

No single accident scenario is the basis of the required preparedness.

The objective of the planning is to provide public protection that would encompass a wide spectrum of possible events with a sufficient basis for extension of response efforts for unanticipated events. These requirements were issued after a long period of involvement by numerous stakeholders, including the Federal Emergency Management Agency, other Federal agencies, local and State governments (and in some cases, foreign governments), private citizens, utilities, and industry groups. __________________

1.1.4 (b) Although the AST provided in this guide was based on a limited spectrum Conforms.

NPPD is not proposing any Emergency of severe accidents, the particular characteristics have been tailored specifically exemption from the requirements of 10 Preparedness for DBA analysis use. The AST is not representative of the wide spectrum of CFR 50.47 or Appendix E to 10 CFR Applications possible events that make up the planning basis of emergency preparedness.

Part 50 with this License Amendment Therefore, the AST is insufficient by itself as a basis for requesting relief from Request.the emergency preparedness requirements of 10 CFR 50.47 and Appendix E to 10 CFR Part 50.1.1.4 (c) This guidance does not, however, preclude the appropriate use of the NRC statement of fact.Emergency insights of the AST in establishing emergency response procedures such as Preparedness those associated with emergency dose projections, protective measures, and Applications severe accident management guides.

NLS2008014 Attachment 1 Page 37 of 75 Appendix A Regulatory Guide 1.183 Comparison Regulatory Position No. Regulatory Position Statement CNS Analysis 1.2.2 Selective Implemen-tation Selective implementation is a modification of the facility design basis that (1) is based on one or more-of-the characteristics of the AST or (2)-entails re-evaluation of a limited subset of the design basis radiological analyses.

The NRC staff will allow licensees flexibility in technically justified selective implementations provided a clear, logical, and consistent design basis is maintained.

An example of an application of selective implementation would be one in which a licensee desires to use the release timing insights of the AST to increase the required closure time for a containment isolation valve by a small amount. Another example would be a request to remove the charcoal filter media from the spent fuel building ventilation exhaust. For the latter, the licensee may only need to re-analyze DBAs that credited the iodine removal by the charcoal media. Additional analysis guidance is provided in Regulatory Position 1.3 of this guide. NRC approval for the AST (and the TEDE dose criterion) will be limited to the particular selective implementation proposed by the licensee.

The licensee would be able to make subsequent modifications to the facility and changes to procedures based on the selected AST characteristics incorporated into the design basis under the provisions of 10 CFR 50.59.However, use of other characteristics of an AST or use of TEDE criteria that are not part of the approved design basis, and changes to previously approved AST characteristics, would require prior staff approval under 10 CFR 50.67. As an example, a licensee with an implementation involving only timing, such as relaxed closure time on isolation valves, could not use 10 CFR 50.59 as a mechanism to implement a modification involving a reanalysis of the DBA LOCA. However, this licensee could extend use of the timing characteristic to adiust the closure time on isolation valves not included in the original approval.Conforms.

This License Amendment Request is a selective scope application of the AST to change the design basis radiological dose consequence analysis of the LOCA. Subsequent modification or changes based on AST will be made in accordance with 10 CFR 50.59.

NLS2008014 Attachment 1 Page 38 of 75 Appendix A Regulatory Guide 1.183 Comparison Regulatory Position No. Regulatory Position Statement CNS Analysis 1.3.1 Design Basis Radiological Analyses There are several regulatory requirements for which compliance is demonstrated, in part, by the evaluation of the radiological -consequences of -design basis accidents.

These requirements include, but are not limited to, the following.

Environmental Qualification of Equipment (10 CFR 50.49)* Control Room Habitability (GDC 19 of Appendix A to 10 CFR Part 50)* Emergency Response Facility Habitability (Paragraph IV.E.8 of Appendix E to 10 CFR Part 50)* Alternative Source Term (10 CFR 50.67)* Environmental Reports (10 CFR Part 51)* Facility Siting (10 CFR 100.11)'There may be additional applications of the accident source term identified in the technical specification bases and in various licensee commitments.

These include, but are not limited to, the following from Reference 2, NUREG-0737.

Post-Accident Access Shielding (NUREG-0737, II.B.2)* Post-Accident Sampling Capability (NUREG-0737, II.B.3)* Accident Monitoring Instrumentation (NUREG-0737, II.F. 1)* Leakage Control (NUREG-0737, III.D.1.1)

Emergency Response Facilities (NUREG-0737, III.A. 1.2)* Control Room Habitability (NUREG-0737, III.D.3.4)

Conforms.

See Section 5.2 of this LAR-for discussion of affected regulatory requirements/criteria.

NLS2008014 Attachment 1 Page 39 of 75 Appendix A Regulatory Guide 1.183 Comparison Regulatory Position No. Regulatory Position Statement CNS Analysis 1.3.2 Re-Analysis Guidance (a) Any implementation of an AST, full or selective, and any associated facility modification should be supported by evaluations of all significant radiological and nonradiological impacts of the proposed actions. This evaluation should consider the impact of the proposed changes on the facility's compliance with the regulations and commitments listed above as well as any other facility-specific requirements.

These impacts may be due to (1) the associated facility modifications or (2) the differences in the AST characteristics.

The scope and extent of the re-evaluation will necessarily be a function of the specific proposed facility modification 6 and whether a full or selective implementation is being pursued. The NRC staff does not expect a complete recalculation of all facility radiological analyses, but does expect licensees to evaluate all impacts of the proposed changes and to update the affected analyses and the design bases appropriately.

An analysis is considered to be affected if the proposed modification changes one or more assumptions or inputs used in that analysis such that the results, or the conclusions drawn on those results, are no longer valid. Generic analyses, such as those performed by owner groups or vendor topical reports, may be used provided the licensee justifies the applicability of the generic conclusions to the specific facility and implementation.

Sensitivity analyses, discussed below, may also be an option. If affected design basis analyses are to be re-calculated, all affected assumptions and inputs should be updated and all selected characteristics of the AST and the TEDE criteria should be addressed.

The license amendment request should describe the licensee's re-analysis effort and provide statements regarding the acceptability of the proposed implementation, including modifications, against each of the applicable analysis requirements and commitments identified in Regulatory Position 1.3.1 of this guide.Conforms.

See Sections 4.3 and 5.2 of this LAR.

NLS2008014 Attachment 1 Page 40 of 75 Appendix A Regulatory Guide 1.183 Comparison Regulatory Position No. Regulatory Position Statement CNS Analysis 1.3.2 Re-Analysis Guidance (b) The NRC staff has performed an evaluation of the impact of the AST on three representative operating reactors.

This evaluation determined that radiological analysis results based on the TID-14844 source term assumptions and the whole body and thyroid methodology generally bound the results from analyses based on the AST and TEDE methodology.

Licensees may use the applicable conclusions of this evaluation in addressing the impact of the AST on design basis radiological analyses.

However, this does not exempt the licensee from evaluating the remaining radiological and nonradiological impacts of the AST implementation and the impacts of the associated plant modifications.

For example, a selective implementation based on the timing insights of the AST may change the required isolation time for the containment purge dampers from 2.5 seconds to 5.0 seconds. This application might be acceptable without dose calculations.

However, evaluations may need to be performed regarding the ability of the damper to close against increased containment pressure or the ability of ductwork downstream of the dampers to withstand increased stresses.Conforms.

This LAR uses the AST to revise the design basis radiological analysis of the LOCA. However, there are no plant modifications that are planned to implement the LOCA AST analysis.

Thus, ancillary radiological and nonradiological effects from modifications are not applicable.

NLS2008014 Attachment 1 Page 41 of 75 Appendix A Regulatory Guide 1.183 Comparison Regulatory Position No. Regulatory Position Statement CNS Analysis 1.3.2 Re-Analysis Guidance (c) A selective implementation of an AST and any associated facility modification based on the AST should evaluate all the radiological and nonradiological impacts of the proposed actions as they apply to the particular implementation.

Design basis analyses are updated in accordance with the guidance in this section. There is no minimum requirement that a DBA LOCA analysis be performed.

The analyses performed need to address all impacts of the proposed modification, the selected characteristics of the AST, and if dose calculations are performed, the TEDE criteria.

For selective implementations based on the timing characteristic of the AST, e.g., change in the closure timing of a containment isolation valve, re-analysis of radiological calculations may not be necessary if the modified elapsed time remains a fraction (e.g., 25%) of the time between accident initiation and the onset of the gap release phase.Longer time delays may be considered on an individual basis. For longer time delays, evaluation of the radiological consequences and other impacts of the delay, such as blockage by debris in sump water, may be necessary.

If affected design basis analyses are to be re-calculated, all affected assumptions and inputs should be updated and all selected characteristics of the AST and the TEDE criteria should be addressed.

Conforms.

This LAR is a selective scope application of the AST to the LOCA. The design basis analysis has been developed consistent with RG 1.183. NPPD is proposing no plant modifications as part of this LAR.

NLS2008014 Attachment 1 Page 42 of 75 Appendix A Regulatory Guide 1.183 Comparison Regulatory Position No. Regulatory Position Statement CNS Analysis 1.3.3 Use of Sensitivity or Scoping Analyses It may be possible to demonstrate by sensitivity or scoping evaluations that existing analyses have sufficient margin and need not be recalculated.

As used in this guide, a sensitivity analysis is an evaluation that considers how the overall results vary as an input parameter (in this case, AST characteristics) is varied. A scoping analysis is a brief evaluation that uses conservative, simple methods to show that the results of the analysis bound those obtainable from a more complete treatment.

Sensitivity analyses are particularly applicable to suites of calculations that address diverse components or plant areas but are otherwise largely based on generic assumptions and inputs. Such cases might include post accident vital area access dose calculations, shielding calculations, and equipment environmental qualification (integrated dose). It may be possible to identify a bounding case, re-analyze that case, and use the results to draw conclusions regarding the remainder of the analyses.

It may also be possible to show that for some analyses the whole body and thyroid doses determined with the previous source term would bound the TEDE obtained using the AST. Where present, arbitrary "designer margins" may be adequate to bound any impact of the AST and TEDE criteria.

If sensitivity or scoping analyses are used, the license amendment request should include a discussion of the analyses performed and the conclusions drawn. Scoping or sensitivity analyses should not constitute a significant part of the evaluations for the design basis exclusion area boundary (EAB), low population zone (LPZ), or control room dose.Not Applicable.

No sensitivity or scoping analyses are utilized as part of this LAR.

NLS2008014 Attachment 1 Page 43 of 75 Appendix A Regulatory Guide 1.183 Comparison Regulatory Position No. Regulatory Position Statement CNS Analysis 1.3.4 This guidance is also applicable to selective implementations to the extent that Conforms.

Subsequent updates to the Updating the affected analyses are within the scope of the approved implementation as LOCA AST analysis after NRC Analyses described in the facility design basis. In these cases, the characteristics of the approval will consider the Following AST and TEDE criteria identified in the facility design basis need to be characteristics of the AST and TEDE Implemen-considered in updating the analyses.

Use of other characteristics of the AST or criteria in the facility design basis.tation TEDE criteria that are not part of the approved design basis, and changes to previously approved AST characteristics, requires prior NRC staff approval under 10 CFR 50.67.1.3.5 Current environmental qualification (EQ) analyses may be impacted by a Not Applicable.

This selective scope Equipment proposed plant modification associated with the AST implementation.

The EQ AST LOCA analysis will not supersede Environ- analyses that have assumptions or inputs affected by the plant modification the current source term used as a basis mental should be updated to address these impacts. The NRC staff is assessing the for the CNS Environmental Qualification effect of increased cesium releases on EQ doses to determine whether licensee Qualification program (TID-14844).

action is warranted.

Until such time as this generic issue is resolved, licensees NPPD is proposing no plant may use either the AST or the TID 14844 assumptions for performing the modifications as part of this LAR.required EQ analyses.

However, no plant modifications are required to address the impact of the difference in source term characteristics (i.e., AST vs.TID 14844) on EQ doses pending the outcome of the evaluation of the generic issue. The EQ dose estimates should be calculated using the design basis survivability period.1.4 (a) The use of an AST changes only the regulatory assumptions regarding the Not Applicable.

No facility Risk analytical treatment of the design basis accidents.

The AST has no direct effect modifications are proposed or planned Implications on the probability of the accident.

Use of an AST alone cannot increase the as implementation actions of the LOCA core damage frequency (CDF) or the large early release frequency (LERF). AST analysis.However, facility modifications made possible by the AST could have an impact on risk. If the proposed implementation of the AST involves changes to the facility design that would invalidate assumptions made in the facility's PRA, the impact on the existing PRAs should be evaluated.

NLS2008014 Attachment 1 Page 44 of 75 Appendix A Regulatory Guide 1.183 Comparison Regulatory Position No. Regulatory Position Statement CNS Analysis 1.4 (b) Consideration should be given to the risk impact of proposed Conforms.

NPPD proposes to increase Risk implementations that seek to remove or downgrade the performance of the MSIV allowable leakage rate as part Implications previously required engineered safeguards equipment on the basis of the of the LOCA AST LAR. This action reduced postulated doses. The NRC staff may request risk information if there has no impact on Core Damage is a reason to question adequate protection of public health and safety. Frequency or Large Early Release Frequency.

1.4 (c) The licensee may elect to use risk insights in support of proposed changes to Not Applicable.

NPPD is not utilizing Risk the design basis that are not addressed in currently approved NRC staff risk insights as a basis for this LAR.Implications positions.

For guidance, refer to Regulatory Guide 1.174, "An Approach for Using Probabilistic Risk Assessment in Risk-Informed Decisions on Plant-Specific Changes to the Licensing Basis." 1.5 (a) According to 10 CFR 50.90, an application for an amendment must fully Conforms.

The LAR is formatted in Submittal describe the changes desired and should follow, as far as applicable, the form accordance with accepted NRC/industry Requirements prescribed for original applications.

Regulatory Guide 1.70, "Standard Format guidance.

The request describes the and Content of Safety Analysis Reports for Nuclear Power Plants (LWR radiological and nonradiological Edition)," provides additional guidance.

The NRC staff s finding that the impacts of the LOCA AST analysis.amendment may be approved must be based on the licensee's analyses, since it Consistent with several of the most is these analyses that will become part of the design basis of the facility.

The recently approved BWR precedents amendment request should describe the licensee's analyses of the radiological (Vermont Yankee, River Bend, and and nonradiological impacts of the proposed modification in sufficient detail to Brunswick), affected USAR pages are support review by the NRC staff. The staff recommends that licensees submit not included in the analyses.

Rather, affected FSAR pages annotated with changes that reflect the revised analyses or this LAR includes the LOCA dose submit the actual calculation documentation.

analysis and Suppression Pool pH analysis calculations.

Approval of this LAR will result in the necessary revisions to the USAR, with revised USAR pages submitted pursuant to 10 CFR 50.71(e).

NLS2008014 Attachment 1 Page 45 of 75 Appendix A Regulatory Guide 1.183 Comparison Regulatory Position No. Regulatory Position Statement CNS Analysis 1.5 (b) If the licensee has used a current approved version of an NRC-sponsored Conforms.

NPPD has used RADTRAD Submittal computer code, the NRC staff review can be made more efficient if the licensee Version 3.03 in the performance of the Requirements identifies the code used and submits the inputs that the licensee used in the LOCA AST analysis.

The inputs made calculations made with that code. In many cases, this will reduce the need for are included with Enclosure 1.NRC staff confirmatory analyses.

This recommendation does not constitute a ARCON96 was used to develop the requirement that the licensee use NRC-sponsored computer codes. Control Room X/Q values.1.6 Requirements for updating the facility's final safety analysis report (FSAR) are Conforms.

Approval of this License FSAR in 10 CFR 50.71, "Maintenance of Records, Making of Reports." The Amendment Request will result in the Requirements regulations in 10 CFR 50.71 (e) require that the FSAR be updated to include all necessary revisions to the USAR, with changes made in the facility or procedures described in the FSAR and all safety revised USAR pages submitted pursuant evaluations performed by the licensee in support of requests for license to 10 CFR 50.71 (e).amendments or in support of conclusions that changes did not involve unreviewed safety questions.

The analyses required by 10 CFR 50.67 are subject to this requirement.

The affected radiological analysis descriptions in the FSAR should be updated to reflect the replacement of the design basis source term by the AST. The analysis descriptions should contain sufficient detail to identify the methodologies used, significant assumptions and inputs, and numeric results. Regulatory Guide 1.70 provides additional guidance.

The descriptions of superseded analyses should be removed from the FSAR in the interest of maintaining a clear design basis.2. ATTRIBUTES OF AN ACCEPTABLE AST 2.1 The AST must be based on major accidents, hypothesized for the purposes of Conforms.

This LAR applies the AST design analyses or consideration of possible accidental events that could result to the design basis LOCA.in hazards not exceeded by those from other accidents considered credible.

The AST must address events that involve a substantial meltdown of the core with the subsequent release of appreciable quantities of fission products.2.2 The AST must be expressed in terms of times and rates of appearance of Conforms.

See Enclosure 1, Section radioactive fission products released into containment, the types and quantities 2.2.of the radioactive species released, and the chemical forms of iodine released.

NLS2008014 Attachment 1 Page 46 of 75 Appendix A Regulatory Guide 1.183 Comparison Regulatory Position No. Regulatory Position Statement CNS Analysis 2.3 The AST must not be based upon a single accident scenario but instead must Conforms.

This selective scope LAR represent a spectrum of credible severe accident events. Risk insights may be provides revised dose consequences to used, not to select a single risk-significant accident, but rather to establish the the DBA LOCA. That event bounds a range of events to be considered.

Relevant insights from applicable severe spectrum of breaks evaluated in the accident research on the phenomenology of fission product release and transport 1 OCFR50.46 ECCS analysis.

Risk behavior may be considered.

insights and insights from accident research were not used.2.4 The AST must have a defensible technical basis supported by sufficient Conforms.

Enclosure 1, developed experimental and empirical data, be verified and validated, and be documented based on NUREG- 1465 and this in a scrutable form that facilitates public review and discourse.

Regulatory Guide, provides the technical basis for the LOCA AST.This basis is summarized in this Attachment 1 to the LAR. The calculation, which utilizes RADTRAD Version 3.03, was developed in accordance with 10 CFR 50 Appendix B, Criterion III.

NLS2008014 Attachment 1 Page 47 of 75 Appendix A Regulatory Guide 1.183 Comparison Regulatory Position No. Regulatory Position Statement CNS Analysis 2.5 The AST must be peer-reviewed by appropriately qualified subject matter Conforms.

Enclosure 1 has been experts. The peer-review comments and their resolution should be part of the developed by industry experts and documentation supporting the AST. reviewed and accepted by CNS Engineering.

The calculation was developed in accordance with 10 CFR 50 Appendix B program, Criterion III.3. ACCIDENT SOURCE TERM 3.1 (a) The inventory of fission products in the reactor core and available for release Conforms.

See Enclosure 1, Section Fission to the containment should be based on the maximum full power operation of the 2.2.Product core with, as a minimum, current licensed values for fuel enrichment, fuel Inventory burnup, and an assumed core power equal to the current licensed rated thermal power times the ECCS evaluation uncertainty.8 The period of irradiation should be of sufficient duration to allow the activity of dose-significant radionuclides to reach equilibrium or to reach maximum values.9 The core inventory should be determined using an appropriate isotope generation and depletion computer code such as ORIGEN 2 or ORIGEN-ARP.

Core inventory factors (Ci/MWt)provided in TID 14844 and used in some analysis computer codes were derived for low bumup, low enrichment fuel and should not be used with higher bumup and higher enrichment fuels.

NLS2008014 Attachment 1 Page 48 of 75 Appendix A Regulatory Guide 1.183 Comparison Regulatory Position No. Regulatory Position Statement CNS Analysis 3.1 (b) For the DBA LOCA, all fuel assemblies in the core are assumed to be Conforms.

See Enclosure 1, Section Fission affected and the core average inventory should be used. For DBA events that 2.2.Product do not involve the entire core, the fission product inventory of each of the Inventory damaged fuel rods is determined by dividing the total core inventory by the number of fuel rods in the core. To account for differences in power level across the core, radial peaking factors from the facility's core operating limits report (COLR) or technical specifications should be applied in determining the inventory of the damaged rods.3.1 (c) No adjustment to the fission product inventory should be made for events Conforms.

See Enclosure 1, Section Fission postulated to occur during power operations at less than full rated power or 2.2.Product those postulated to occur at the beginning of core life. For events postulated to Inventory occur while the facility is shutdown, e.g., a fuel handling accident, radioactive decay from the time of shutdown may be modeled.

NLS2008014 Attachment 1 Page 49 of 75 Appendix A Regulatory Guide 1.183 Comparison Regulatory Position No. Regulatory Position Statement CNS Analysis 3.2 Release Fractions1 0 The core inventory release fractions, by radionuclide groups, for the gap release and early in-vessel damage phases for DBA LOCAs are listed in Table 1 for BWRs and Table 2 for PWRs. These fractions are applied to the equilibrium core inventory described in Regulatory Position 3.1.Table 1 BWR Core Inventory Fraction Released Into Containment Conforms.

See Enclosure 1, Section 2.2.Group Noble Gases Halogens Alkali Metals Tellurium Metals Ba, Sr Noble Metals Cerium Group Lanthanides Gap Release Phase 0.05 0.05 0.05 0.00 0.00 0.00 0.00 0.00 Early In-vessel Phase Total 0.95 1.0 0.25 0.3 0.20 0.25 0.05 0.05 0.02 0.02 0.0025 0.0025 0.0005 0.0005 0.0002 0.0002 NLS2008014 Attachment 1 Page 50 of 75 Appendix A Regulatory Guide 1.183 Comparison Regulatory Position No. Regulatory Position Statement CNS Analysis 3.3 (a) Table 4 tabulates the onset and duration of each sequential release phase for Conforms.

See Enclosure 1, Section Timing of DBA LOCAs at PWRs and BWRs. The specified onset is the time following 2.2.Release the initiation of the accident (i.e., time = 0). The early in-vessel phase Phases immediately follows the gap release phase. The activity released from the core during each release phase should be modeled as increasing in a linear fashion over the duration of the phase.m2 For non-LOCA DBAs in which fuel damage is projected, the release from the fuel gap and the fuel pellet should be assumed to occur instantaneously with the onset of the projected damage.Table 4 LOCA Release Phases PWRs BWRs Phase Onset Duration Onset Duration Gap Release 30 sec 0.5 hr 2 min 0.5 hr Early In-Vessel 0.5 hr 1.3 hr 0.5 hr 1.5 hr 3.3 (b) For facilities licensed with leak-before-break methodology, the onset of the Not Applicable.

CNS is not licensed Timing of gap release phase may be assumed to be 10 minutes. A licensee may propose with a leak-before-break methodology.

Release an alternative time for the onset of the gap release phase, based on facility-Phases specific calculations using suitable analysis codes or on an accepted topical report shown to be applicable to the specific facility.

In the absence of approved alternatives, the gap release phase onsets in Table 4 should be used.

NLS2008014 Attachment 1 Page 51 of 75 Appendix A Regulatory Guide 1.183 Comparison Regulatory Position No. Regulatory Position Statement CNS Analysis 3.4 Table 5 lists the elements in each radionuclide group that should be considered Conforms.

See Enclosure 1, Section Radionuclide in design basis analyses.

2.2.Composition Table 5 Radionuclide Groups Group Elements Noble Gases Xe, Kr Halogens I, Br Alkali Metals Cs, Rb Tellurium Group Te, Sb, Se, Ba, Sr Noble Metals Ru, Rh, Pd, Mo, Tc, Co Lanthenides La, Zr, Nd, Eu, Nb, Pm, Pr Sm, Y, Cm, Am Cerium Ce, Pu, Np 3.5 Of the radioiodine released from the reactor coolant system (RCS) to the Conforms.

See Enclosure 1, Section Chemical containment in a postulated accident, 95 percent of the iodine released should 2.2.Form be assumed to be cesium iodide (CsI), 4.85 percent elemental iodine, and 0.15 percent organic iodide. This includes releases from the gap and the fuel pellets.With the exception of elemental and organic iodine and noble gases, fission products should be assumed to be in particulate form. The same chemical form is assumed in releases from fuel pins in FHAs and from releases from the fuel pins through the RCS in DBAs other than FHAs or LOCAs. However, the transport of these iodine species following release from the fuel may affect these assumed fractions.

The accident-specific appendices to this regulatory guide provide additional details.

NLS2008014 Attachment 1 Page 52 of 75 Appendix A Regulatory Guide 1.183 Comparison Regulatory Position No. Regulatory Position Statement CNS Analysis 4. DOSE CALCULATION METHODOLOGY

4.1 Offsite

Dose Consequences 4.1.1 The dose calculations should determine the TEDE. TEDE is the sum of the Conforms.

See Enclosure 1, Section 5.3 committed effective dose equivalent (CEDE) from inhalation and the deep dose and Sections 4.1.7 of this License equivalent (DDE) from external exposure.

The calculation of these two Amendment Request.components of the TEDE should consider all radionuclides, including progeny from the decay of parent radionuclides, that are significant with regard to dose consequences and the released radioactivity.

1 3 4.1.2 The exposure-to-CEDE factors for inhalation of radioactive material should be Conforms.

See Enclosure 1, Section derived from the data provided in ICRP Publication 30, "Limits for Intakes of 5.3.Radionuclides by Workers." Table 2.1 of Federal Guidance Report 11,"Limiting Values of Radionuclide Intake and Air Concentration and Dose Conversion Factors for Inhalation, Submersion, and Ingestion," provides tables of conversion factors acceptable to the NRC staff. The factors in the column headed "effective" yield doses corresponding to the CEDE.4.1.3 For the first 8 hours9.259259e-5 days 0.00222 hours 1.322751e-5 weeks 3.044e-6 months , the breathing rate of persons offsite should be assumed to Conforms.

See Enclosure 1, Section be 3.5 x 1 0.4 cubic meters per second. From 8 to 24 hours2.777778e-4 days 0.00667 hours 3.968254e-5 weeks 9.132e-6 months following the 2.7.accident, the breathing rate should be assumed to be 1.8 x 10.4 cubic meters per second. After that and until the end of the accident, the rate should be assumed to be 2.3 x 10-4 cubic meters per second.

NLS2008014 Attachment 1 Page 53 of 75 Appendix A Regulatory Guide 1.183 Comparison Regulatory Position No. Regulatory Position Statement CNS Analysis 4.1.4 The DDE should be calculated assuming submergence in semi-infinite cloud Conforms.

See Enclosure 1, Section assumptions with appropriate credit for attenuation by body tissue. The DDE is 2.7.nominally equivalent to the effective dose equivalent (EDE) from external exposure if the whole body is irradiated uniformly.

Since this is a reasonable assumption for submergence exposure situations, EDE may be used in lieu of DDE in determining the contribution of external dose to the TEDE. Table III. 1 of Federal Guidance Report 12, "External Exposure to Radionuclides in Air, Water, and Soil," provides external EDE conversion factors acceptable to the NRC staff. The factors in the column headed "effective" yield doses corresponding to the EDE.4.1.5 The TEDE should be determined for the most limiting person at the EAB. The Conforms.

See Enclosure 1, Section maximum EAB TEDE for any two-hour period following the start of the 2.7.radioactivity release should be determined and used in determining compliance with the dose criteria in 10 CFR 50.67.14 The maximum two-hour TEDE should be determined by calculating the postulated dose for a series of small time increments and performing a "sliding" sum over the increments for successive two-hour periods. The maximum TEDE obtained is submitted.

The time increments should appropriately reflect the progression of the accident to capture the peak dose interval between the start of the event and the end of radioactivity release (see also Table 6).4.1.6 TEDE should be determined for the most limiting receptor at the outer Conforms.

See Enclosure 1, Section boundary of the low population zone (LPZ) and should be used in determining 2.7.compliance with the dose criteria in 10 CFR 50.67.4.1.7 No correction should be made for depletion of the effluent plume by deposition Conforms.

Enclosure 1 takes no credit on the ground. for depletion of the effluent by deposition on the ground.

NLS2008014 Attachment 1.Page 54 of 75 Appendix A Regulatory Guide 1.183 Comparison Regulatory Position No. Regulatory Position Statement CNS Analysis 4.2 Control Room Consequences 4.2.1 The TEDE analysis should consider all sources of radiation that will cause Conforms.

The calculated dose results exposure to control room personnel.

The applicable sources will vary from take these sources of radiation into facility to facility, but typically will include: consideration in calculating TEDE.* Contamination of the control room atmosphere by the intake or infiltration of the radioactive material contained in the 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 the external radioactive plume released from the facility," Radiation shine from radioactive material in the reactor containment, Radiation shine from radioactive material in systems and components inside or external to the control room envelope, e.g., radioactive material buildup in recirculation filters.4.2.2 The radioactive material releases and radiation levels used in the control room Conforms.

See Enclosure 1, Section dose analysis should be determined using the same source term, transport, and 2.2.release assumptions used for determining the EAB and the LPZ TEDE values, unless these assumptions would result in non-conservative results for the control room.4.2.3 The models used to transport radioactive material into and through the control Conforms.

See Section 4.1.7 of this room, 1 5 and the shielding models used to determine radiation dose rates from License Amendment Request.external sources, should be structured to provide suitably conservative estimates of the exposure to control room personnel.

NLS2008014 Attachment 1 Page 55 of 75 Appendix A Regulatory Guide 1.183 Comparison Regulatory Position No. Regulatory Position Statement CNS Analysis 4.2.4 Credit for engineered safety features that mitigate airborne radioactive material Conforms.

Credit is taken for the SGT within the control room may be assumed. Such features may include control System and for CREFS in mitigating the room isolation or pressurization, or intake or recirculation filtration.

Refer to LOCA in the AST analysis.

Filter Section 6.5.1, "ESF Atmospheric Cleanup System," of the SRP and Regulatory efficiencies and flow rates are within Guide 1.52, "Design, Testing, and Maintenance Criteria for Post-accident design basis and Technical Specification Engineered-Safety-Feature Atmosphere Cleanup System Air Filtration and requirements for these systems.Adsorption Units of Light-Water-Cooled Nuclear Power Plants," for guidance.The control room design is often optimized for the DBA LOCA and the protection afforded for other accident sequences may not be as advantageous.

In most designs, control room isolation is actuated by engineered safeguards feature (ESF) signals or radiation monitors (RMs). In some cases, the ESF signal is effective only for selected accidents, placing reliance on the RMs for the remaining accidents.

Several aspects of RMs can delay the control room isolation, including the delay for activity to build up to concentrations equivalent to the alarm setpoint and the effects of different radionuclide accident isotopic mixes on monitor response.4.2.5 Credit should generally not be taken for the use of personal protective Conforms.

See Enclosure 1, Section 4.equipment or prophylactic drugs. Deviations may be considered on a case-by-case basis.4.2.6 The dose receptor for these analyses is the hypothetical maximum exposed Conforms.

See Enclosure 1, Section 3.individual who is present in the control room for 100% of the time during the first 24 hours2.777778e-4 days 0.00667 hours 3.968254e-5 weeks 9.132e-6 months after the event, 60% of the time between 1 and 4 days, and 40% of the time from 4 days to 30 days.1 6 For the duration of the event, the breathing rate of this individual should be assumed to be 3.5 x 10-4 cubic meters per second.

NLS2008014 Attachment 1 Page 56 of 75 Appendix A Regulatory Guide 1.183 Comparison Regulatory Position No. Regulatory Position Statement CNS Analysis 4.2.7 Control room doses should be calculated using dose conversion factors Conforms.

See Enclosure 1, Section 3.identified in Regulatory Position 4.1 above for use in offsite dose analyses.

The DDE from photons may be corrected for the difference between finite cloud geometry in the control room and the semi-infinite cloud assumption used in calculating the dose conversion factors. The following expression may be used to correct the semi-infinite cloud dose, DDE, to a finite cloud dose, DDEfinite, where the control room is modeled as a hemisphere that has a volume, V, in cubic feet, equivalent to that of the control room.DDEfinite

= DDEoJV°3 3 8 / 1173 Equation 1 4.3 The guidance provided in Regulatory Positions 4.1 and 4.2 should be used, as Conforms.

See Sections 4.3 and 5.2 of Other Dose applicable, in re-assessing the radiological analyses identified in Regulatory this License Amendment Request.Consequences Position 1.3.1, such as those in NUREG-0737.

Design envelope source terms provided in NUREG-0737 should be updated for consistency with the AST. In general, radiation exposures to plant personnel identified in Regulatory Position 1.3.1 should be expressed in terms of TEDE. Integrated radiation exposure of plant equipment should be determined using the guidance of Appendix I of this guide.

NLS2008014 Attachment 1 Page 57 of 75 Appendix A Regulatory Guide 1.183 Comparison Regulatory Position No. Regulatory Position Statement CNS Analysis 4.4 (a) The radiological criteria for the EAB, the outer boundary of the LPZ, and for Conforms.

See Enclosure 1, Section 6.Acceptance the control room are in 10 CFR 50.67. These criteria are stated for evaluating Criteria reactor accidents of exceedingly low probability of occurrence and low risk of public exposure to radiation, e.g., a large-break LOCA. The control room criterion applies to all accidents.

For events with a higher probability of occurrence, postulated EAB and LPZ doses should not exceed the criteria tabulated in Table 6.Table 6 Accident Dose Criteria EAB and LPZ Accident or Case Dose Criteria Analysis Release Duration LOCA 25 rem TEDE 30 days for containment, ECCS, and MSIV (BWR)leakage 4.4 (b) The acceptance criteria for the various NUREG-0737 items generally Conforms.

As an implementing activity reference General Design Criteriaý 19 (GDC 19) from Appendix A to 10 CFR to the license amendment, the Part 50 or specify criteria derived from GDC 19. These criteria are generally acceptance criteria for the applicable specified in terms of whole body dose, or its equivalent to any body organ. For NUREG-0373 items will be updated to facilities applying for, or having received, approval for the use of an AST, the 10 CFR 50.67 TEDE criteria.applicable criteria should be updated for consistency with the TEDE criterion in 10 CFR 50.67(b)(2)(iii).

NLS2008014 Attachment 1 Page 58 of 75 Appendix A Regulatory Guide 1.183 Comparison Regulatory Position No. Regulatory Position Statement CNS Analysis 5. ANALYSIS ASSUMPTIONS AND METHODOLOGY

5.1 General

Considerations 5.1.1 (a) The evaluations required by 10 CFR 50.67 are re-analyses of the design Conforms.

Enclosures 1 and 4 and Analysis basis safety analyses and evaluations required by 10 CFR 50.34; they are other supporting calculations and Quality considered to be a significant input to the evaluations required by 10 CFR 50.92 evaluations were prepared and accepted or 10 CFR 50.59. These analyses should be prepared, reviewed, and maintained by NPPD under a 10 CFR 50 Appendix in accordance with quality assurance programs that comply with Appendix B, B Quality Assurance program."Quality Assurance Criteria for Nuclear Power Plants and Fuel Reprocessing Plants," to 10 CFR Part 50.5.1.1 (b) These design basis analyses were structured to provide a conservative set of Conforms.

NPPD is not proposing Analysis assumptions to test the performance of one or more aspects of the facility deviations to conformance with this Quality design. Many physical processes and phenomena are represented by Regulatory Guide.conservative, bounding assumptions rather than being modeled directly.

The staff has selected assumptions and models that provide an appropriate and prudent safety margin against unpredicted events in the course of an accident and compensate for large uncertainties in facility parameters, accident progression, radioactive material transport, and atmospheric dispersion.

Licensees should exercise caution in proposing deviations based upon data from a specific accident sequence since the DBAs were never intended to represent any specific accident sequence -- the proposed deviation may not be conservative for other accident sequences.

NLS2008014 Attachment 1 Page 59 of 75 Appendix A Regulatory Guide 1.183 Comparison Regulatory Position No. Regulatory Position Statement CNS Analysis 5.1.2 Credit for Engineered Safeguard Features Credit may be taken for accident mitigation features that are classified as safety-related, are required to be operable by technical specifications, are powered by emergency power sources, and are either automatically actuated or, in limited cases, have actuation requirements explicitly addressed in emergency operating procedures.

The single active component failure that results in the most limiting radiological consequences should be assumed. Assumptions regarding the occurrence and timing of a loss of offsite power should be selected with the objective of maximizing the postulated radiological consequences.

Conforms.

Where credited, safety-related mitigating structures, systems, and components were assumed to operate consistent with single failure, design basis emergency power, and other requirements.

A new safety-related function was credited for the SLC System for Suppression Pool pH control. As shown in Attachment 2 of this License Amendment request, the system is capable of performing this post-LOCA function.

NLS2008014 Attachment 1 Page 60 of 75 Appendix A Regulatory Guide 1.183 Comparison Regulatory Position No. Regulatory Position Statement CNS Analysis 5.1.3 Assignment of Numeric Input Values The numeric values that are chosen as inputs to the analyses required by 10 CFR 50.67 should be selected with the objective of determining a conservative postulated dose. In some instances, a particular parameter may be conservative in one portion of an analysis but be nonconservative in another portion of the same analysis.

For example, assuming minimum containment system spray flow is usually conservative for estimating iodine scrubbing, but in many cases may be nonconservative when determining sump pH. Sensitivity analyses may be needed to determine the appropriate value to use. As a conservative alternative, the limiting value applicable to each portion of the analysis may be used in the evaluation of that portion. A single value may not be applicable for a parameter for the duration of the event, particularly for parameters affected by changes in density. For parameters addressed by technical specifications, the value used in the analysis should be that specified in the technical specifications.18 If a range of values or a tolerance band is specified, the value that would result in a conservative postulated dose should be used. If the parameter is based on the results of less frequent surveillance testing, e.g., steam generator nondestructive testing (NDT), consideration should be given to the degradation that may occur between periodic tests in establishing the analysis value.Conforms.

The numeric values chosen provide a conservative dose result.There are no parameters in the LOCA AST analysis that have conflicting credit within the analysis.

Parameters that are controlled by TS either use the TS values, or are otherwise bounded by them (e.g. CREFS filter efficiency relaxation).

The parameters are not based on surveillance testing results.

NLS2008014 Attachment 1 Page 61 of 75 Appendix A Regulatory Guide 1.183 Comparison Regulatory Position No. Regulatory Position Statement CNS Analysis 5.1.4 The NRC staff considers the implementation of an AST to be a significant Conforms.

The LOCA analysis Applicability change to the design basis of the facility that is voluntarily initiated by the assumptions and methods are of Prior licensee.

In order to issue a license amendment authorizing the use of an AST compatible with the AST and the TEDE licensing and the TEDE dose criteria, the NRC staff must make a current finding of criteria.Basis compliance with regulations applicable to the amendment.

The characteristics of the ASTs and the revised dose calculational methodology may be incompatible with many of the analysis assumptions and methods currently reflected in the facility's design basis analyses.

The NRC staff may find that new or unreviewed issues are created by a particular site-specific implementation of the AST, warranting review of staff positions approved subsequent to the initial issuance of the license. This is not considered a backfit as defined by 10 CFR 50.109, "Backfitting." However, prior design bases that are unrelated to the use of the AST, or are unaffected by the AST, may continue as the facility's design basis. Licensees should ensure that analysis assumptions and methods are compatible with the ASTs and the TEDE criteria.5.2 (a) The appendices to this regulatory guide provide accident-specific Conforms.

See RG 1.183 Appendix A Accident-assumptions that are acceptable to the staff for performing analyses that are Sections of this appendix.Specific required by 10 CFR 50.67. The DBAs addressed in these attachments were Assumptions selected from accidents that may involve damage to irradiated fuel. This guide does not address DBAs with radiological consequences based on technical specification reactor or secondary coolant-specific activities only. The inclusion or exclusion of a particular DBA in this guide should not be interpreted as indicating that an analysis of that DBA is required or not required.

Licensees should analyze the DBAs that are affected by the specific proposed applications of an AST.

NLS2008014 Attachment 1 Page 62 of 75 Appendix A Regulatory Guide 1.183 Comparison Regulatory Position No. Regulatory Position Statement CNS Analysis 5.2 (b) The NRC staff has determined that the analysis assumptions in the Conforms.

See RG 1.183 Appendix A Accident-appendices to this guide provide an integrated approach to performing the Sections of this appendix.Specific individual analyses and generally expects licensees to address each assumption Assumptions or propose acceptable alternatives.

Such alternatives may be justifiable on the basis of plant-specific considerations, updated technical analyses, or, in some cases, a previously approved licensing basis consideration.

The assumptions in the appendices are deemed consistent with the AST identified in Regulatory Position 3 and internally consistent with each other. Although licensees are free to propose alternatives to these assumptions for consideration by the NRC staff, licensees should avoid use of previously approved staff positions that would adversely affect this consistency.

5.2 (c) The NRC is committed to using probabilistic risk analysis (PRA) insights in Conforms.

PRA was not used as a basis Accident-its regulatory activities and will consider licensee proposals for changes in for acceptability of this LOCA AST Specific analysis assumptions based upon risk insights.

The staff will not approve License Amendment Request.Assumptions proposals that would reduce the defense in depth deemed necessary to provide adequate protection for public health and safety. In some cases, this defense in depth compensates for uncertainties in the PRA analyses and addresses accident considerations not adequately addressed by the core damage frequency (CDF)and large early release frequency (LERF) surrogate indicators of overall risk.5.3 (a) Atmospheric dispersion values (X/Q) for the EAB, the LPZ, and the control Conforms.

X/Qs that are used were Meteorology room that were approved by the staff during initial facility licensing or in previously reviewed by the NRC as part Assumptions subsequent licensing proceedings may be used in performing the radiological of the licensing of the existing CNS analyses identified by this guide. Methodologies that have been used for LOCA analysis (see Section 4.2 of this determining X/Q values are documented in Regulatory Guides 1.3 and 1.4, License Amendment Request).Regulatory Guide 1.145, "Atmospheric Dispersion Models for Potential Accident Consequence Assessments at Nuclear Power Plants," and the paper,"Nuclear Power Plant Control Room Ventilation System Design for Meeting General Criterion 19."

NLS2008014 Attachment 1 Page 63 of 75 Appendix A Regulatory Guide 1.183 Comparison Regulatory Position No. Regulatory Position Statement CNS Analysis 5.3 (b) References 22 and 28 should be used if the FSAR X/Q values are to be Not Applicable.

The X/Q values were Meteorology revised or if values are to be determined for new release points or receptor previously reviewed by the NRC as part Assumptions distances.

Fumigation should be considered where applicable for the EAB and of the licensing of the existing CNS LPZ. For the EAB, the assumed fumigation period should be timed to be LOCA analysis (see Section 4.2 of this included in the worst 2-hour exposure period. The NRC computer code License Amendment Request).PAVAN implements Regulatory Guide 1.145 and its use is acceptable to the NRC staff. The methodology of the NRC computer code ARCON961 9 is generally acceptable to the NRC staff for use in determining control room X/Q values. Meteorological data collected in accordance with the site-specific meteorological measurements program described in the facility FSAR should be used in generating accident X/Q values. Additional guidance is provided in Regulatory Guide 1.23, "Onsite Meteorological Programs." All changes in X/Q analysis methodology should be reviewed by the NRC staff.Appendix A to Regulatory Guide 1.183: ASSUMPTIONS FOR EVALUATING THE RADIOLOGICAL CONSEQUENCES OF A LWR LOSS-OF-COOLANT ACCIDENT A-1 Acceptable assumptions regarding core inventory and the release of Conforms.

See discussions in RG radionuclides from the fuel are provided in Regulatory Position 3 of this guide. Section 3 of this matrix.A-2 If the sump or suppression pool pH is controlled at values of 7 or greater, the Conforms.

See Section 4.1.1 of this chemical form of radioiodine released to the containment should be assumed to License Amendment Request, and be 95% cesium iodide (CsI), 4.85 percent elemental iodine, and 0.15 percent Section 2.2 of Enclosure 1.organic iodide. Iodine species, including those from iodine re-evolution, for sump or suppression pool pH values less than 7 will be evaluated on a case-by-case basis. Evaluations of pH should consider the effect of acids and bases created during the LOCA event, e.g., radiolysis products.

With the exception of elemental and organic iodine and noble gases, fission products should be assumed to be in particulate form.

NLS2008014 Attachment 1 Page 64 of 75 Appendix A Regulatory Guide 1.183 Comparison Regulatory Position No. Regulatory Position Statement CNS Analysis A-3.1 The radioactivity released from the fuel should be assumed to mix Conforms.

See Enclosure 1, Sections instantaneously and homogeneously throughout the free air volume of the 2.2 and 2.3.primary containment in PWRs or the drywell in BWRs as it is released.

This distribution should be adjusted if there are internal compartments that have limited ventilation exchange.

The suppression pool free air volume may be included provided there is a mechanism to ensure mixing between the drywell to the wetwell. The release into the containment or drywell should be assumed to terminate at the end of the early in-vessel phase.A-3.2 Reduction in airborne radioactivity in the containment by natural deposition Conforms.

See Enclosure 1, Section within the containment may be credited.

Acceptable models for removal of 2.3.iodine and aerosols are described in Chapter 6.5.2, "Containment Spray as a Fission Product Cleanup System," of the Standard Review Plan (SRP), NUREG-0800 and in NUREG/CR-6189, "A Simplified Model of Aerosol Removal by Natural Processes in Reactor Containments." The latter model is incorporated into the analysis code RADTRAD. The prior practice of deterministically assuming that a 50% plateout of iodine is released from the fuel is no longer acceptable to the NRC staff as it is inconsistent with the characteristics of the revised source terms.A-3.3 (a) Reduction in airborne radioactivity in the containment by containment spray Not Applicable.

Containment sprays are systems that have been designed and are maintained in accordance with.Chapter not credited in the analysis (see 6.5.2 of the SRP may be credited.

Acceptable models for the removal of iodine Enclosure 1, Section 2.3).and aerosols are described in Chapter 6.5.2 of the SRP and NUREG/CR-5966,"A Simplified Model of Aerosol Removal by Containment Sprays." This simplified model is incorporated into the analysis code RADTRAD.

NLS2008014 Attachment 1 Page 65 of 75 Appendix A Regulatory Guide 1.183 Comparison Regulatory Position No. Regulatory Position Statement CNS Analysis A-3.3 (b) The evaluation of the containment sprays should address areas within the Not Applicable.

Containment sprays are primary containment that are not covered by the spray drops. The mixing rate not credited in the analysis (see attributed to natural convection between sprayed and unsprayed regions of the Enclosure 1, Section 2.3).containment building, provided that adequate flow exists between these regions, is assumed to be two turnovers of the unsprayed regions per hour, unless other rates are justified.

The containment building atmosphere may be considered a single, well-mixed volume if the spray covers at least 90% of the volume and if adequate mixing of unsprayed compartments can be shown.A-3.3 (c) The SRP sets forth a maximum decontamination factor (DF) for elemental Not Applicable.

Containment sprays are iodine based on the maximum iodine activity in the primary containment not credited in the analysis (see atmosphere when the sprays actuate, divided by the activity of iodine remaining Enclosure 1, Section 2.3).at some time after decontamination.

The SRP also states that the particulate iodine removal rate should be reduced by a factor of 10 when a DF of 5 0 is reached. The reduction in the removal rate is not required if the removal rate is based on the calculated time-dependent airborne aerosol mass. There is no specified maximum DF for aerosol removal by sprays. The maximum activity to be used in determining the DF is defined as the iodine activity in the columns labeled "Total" in Tables 1 and 2 of this guide multiplied by 0.05 for elemental iodine and by 0.95 for particulate iodine (i.e., aerosol treated as particulate in SRP methodology).

A-3 .4 Reduction in airborne radioactivity in the containment by in-containment Not Applicable.

Containment recirculation filter systems may be credited if these systems meet the guidance recirculation filtration is not credited in of Regulatory Guide 1.52 and Generic Letter 99-02. The filter media loading the analysis (see Enclosure 1, Section caused by the increased aerosol release associated with the revised source term 2.3).___________should be addressed.

NLS2008014 Attachment 1 Page 66 of 75 Appendix A Regulatory Guide 1.183 Comparison Regulatory Position No. Regulatory Position Statement CNS Analysis A-3.5 Reduction in airborne radioactivity in the containment by suppression pool Conforms.

Suppression Pool scrubbing scrubbing in BWRs should generally not be credited.

However, the staff may is not credited in the analysis.

Analysis consider such reduction on an individual case basis. The evaluation should of Enclosure 4 demonstrates pH of the consider the relative timing of the blowdown and the fission product release Suppression Pool liquid is maintained from the fuel, the force driving the release through the pool, and the potential greater than 7.for any bypass of the suppression pool. Analyses should consider iodine re-evolution if the suppression pool liquid pH is not maintained greater than 7.A-3.6 Reduction in airborne radioactivity in the containment by retention in ice Not Applicable.

The AST LOCA condensers, or other engineering safety features not addressed above, should be analysis does not credit engineered evaluated on an individual case basis. See Section 6.5.4 of the SRP. safety features not previously addressed for reducing Primary Containment airborne radioactivity.

A-3.7 The primary containment (i.e., drywell for Mark I and II containment designs) Conforms.

Primary Containment is should be assumed to leak at the peak pressure technical specification leak rate assumed to leak at the peak pressure TS for the first 24 hours2.777778e-4 days 0.00667 hours 3.968254e-5 weeks 9.132e-6 months . For PWRs, the leak rate may be reduced after the first 24 leak rate for the first 24 hours2.777778e-4 days 0.00667 hours 3.968254e-5 weeks 9.132e-6 months , after hours to 50% of the technical specification leak rate. For BWRs, leakage may which, the leakage rate is reduced to be reduced after the first 24 hours2.777778e-4 days 0.00667 hours 3.968254e-5 weeks 9.132e-6 months , if supported by plant configuration and 50% of the technical specification analyses, to a value not less than 50% of the technical specification leak rate. leakage rate (see Section 4.1.3 of the Leakage from subatmospheric containments is assumed to terminate when the License Amendment Request).containment is brought to and maintained at a subatmospheric condition as defined by technical specifications.

NLS2008014 Attachment 1 Page 67 of 75 Appendix A Regulatory Guide 1.183 Comparison Regulatory Position No. Regulatory Position Statement CNS Analysis A-3.8 If the primary containment is routinely purged during power operations, releases Not Applicable.

The Primary via the purge system prior to containment isolation should be analyzed and the Containment is not routinely purged resulting doses summed with the postulated doses from other release paths. The during power operations.

purge release evaluation should assume that 100% of the radionuclide inventory in the reactor coolant system liquid is released to the containment at the initiation of the LOCA. This inventory should be based on the technical specification reactor coolant system equilibrium activity.

Iodine spikes need not be considered.

If the purge system is not isolated before the onset of the gap release phase, the release fractions associated with the gap release and early in-vessel phases should be considered as applicable.

__________________

A-4. 1 Leakage from the primary containment should be considered to be collected, Conforms.

The LOCA AST analysis processed by engineered safety feature (ESF) filters, if any, and released to the credits an elevated release from the environment via the secondary containment exhaust system during periods in Secondary Containment, consistent with which the secondary containment has a negative pressure as defined in technical the current design basis LOCA specifications.

Credit for an elevated release should be assumed only if the calculation.

At a height of 325 feet, the point of physical release is more than two and one-half times the height of any Elevated Release Point (ERP) is adjacent structure.

approximately twice the height of the tallest building at CNS (the Reactor Building).

However, since the Reactor Building's closest point is over 350 feet away from the ERP, it is not considered to be "adjacent" for purposes of this RG___________criterion.

A-4.2 Leakage from the primary containment is assumed to be released directly to the Conforms.

The LOCA AST analysis environment as a ground-level release during any period in which the secondary assumes a 5 minute ground release from containment does not have a negative pressure as defined in technical the Reactor Building following specifications.

initiation of the event (see Section___________

______________________________________________________4.1.3).

NLS2008O014 Attachment 1 Page 68 of 75 Appendix A Regulatory Guide 1.183 Comparison Regulatory Position No. Regulatory Position Statement CNS Analysis A-4.3 The effect of high wind speeds on the ability of the secondary containment to Not Applicable.

The CNS current maintain a negative pressure should be evaluated on an individual case basis. licensing basis is that Secondary The wind speed to be assumed is the 1 -hour average value that is exceeded only Containment remains at negative 5% of the total number of hours in the data set. Ambient temperatures used in pressure under neutral wind conditions.

these assessments should be the 1 -hour average value that is exceeded only 5%or 95% of the total numbers of hours in the data set, whichever is conservative for the intended use (e.g., if high temperatures are limiting, use those exceeded___________only 5%).A-4.4 Credit for dilution in the secondary containment may be allowed when adequate Conforms.

Credit is not taken for means to cause mixing can be demonstrated.

Otherwise, the leakage from the dilution or mixing in Secondary primary containment should be assumed to be transported directly to exhaust Containment (See Enclosure 1, Sections systems without mixing. Credit for mixing, if found to be appropriate, should 2.3.2 and 4).generally be limited to 50%. This evaluation should consider the magnitude of the containment leakage in relation to contiguous building volume or exhaust rate, the location of exhaust plenums relative to projected release locations, the recirculation ventilation systems, and internal walls and floors that impede stream flow between the release and the exhaust.A-4.5 Primary containment leakage that bypasses the secondary containment should Conforms.

CNS does not currently have be evaluated at the bypass leak rate incorporated in the technical specifications.

a Secondary Containment Bypass If the bypass leakage is through water, e.g., via a filled piping run that is Leakage technical specification.

maintained full, credit for retention of iodine and aerosols may be considered on However, NPPD has determined that a case-by-case basis. Similarly, deposition of aerosol radioactivity in gas-filled Secondary Containment bypass is lines may be considered on a case-by-case basis. limited to the MS lines and the MS inboard drain line. The proposed MS Pathway aggregate leakage limit (new proposed CNS Technical Specification SR 3.6.1.3.12) would apply to leakage from the MSIVs and the MS inboard drain line.

NLS2008014 Attachment 1 Page 69 of 75 Appendix A Regulatory Guide 1.183 Comparison Regulatory Position No. Regulatory Position Statement CNS Analysis A-4.6 Reduction in the amount of radioactive material released from the secondary Conforms.

See Enclosure 1, Sections containment because of ESF filter systems may be taken into account provided 2.3.2, 2.7, and 4.that these systems meet the guidance of Regulatory Guide 1.52 and Generic Letter 99-02.A-5 ESF systems that recirculate sump water outside of the primary containment are Conforms.

ESF leakage was assumed in assumed to leak during their intended operation.

This release source includes the LOCA AST analysis and dose from leakage through valve packing glands, pump shaft seals, flanged connections, this pathway was combined with dose and other similar components.

This release source may also include leakage from the other release pathways.through valves isolating interfacing systems. The radiological consequences from the postulated leakage should be analyzed and combined with consequences postulated for other fission product release paths to determine the total calculated radiological consequences from the LOCA. The following assumptions are acceptable for evaluating the consequences of leakage from ESF components outside the primary containment for BWRs and PWRs.A-5.1 With the exception of noble gases, all the fission products released from the Conforms.

See Enclosure 1, Section fuel to the containment (as defined in Tables 1 and 2 of this guide) should be 2.3.assumed to instantaneously and homogeneously mix in the primary containment sump water (in PWRs) or suppression pool (in BWRs) at the time of release from the core. In lieu of this deterministic approach, suitably conservative mechanistic models for the transport of airborne activity in containment to the sump water may be used. Note that many of the parameters that make spray and deposition models conservative with regard to containment airborne leakage are nonconservative with regard to the buildup of sump activity.

NLS2008014 Attachment 1 Page 70 of 75 Appendix A Regulatory Guide 1.183 Comparison Regulatory Position No. Regulatory Position Statement CNS Analysis A-5.2 The leakage should be taken as two times the sum of the simultaneous leakage Conforms.

See Enclosure 1, Section from all components in the ESF recirculation systems above which the technical 2.6.specifications, or licensee commitments to item III.D. 1.1 of NUREG-0737, would require declaring such systems inoperable.

The leakage should be assumed to start at the earliest time the recirculation flow occurs in these systems and end at the latest time the releases from these systems are terminated.

Consideration should also be given to design leakage through valves isolating ESF recirculation systems from tanks vented to atmosphere, e.g., emergency core cooling system (ECCS) pump miniflow return to the refueling water storage tank.A-5.3 With the exception of iodine, all radioactive materials in the recirculating liquid Conforms.

See Enclosure 1, Section should be assumed to be retained in the liquid phase. 2.6.A-5.4 If the temperature of the leakage exceeds 212°F, the fraction of total iodine in Not Applicable.

Suppression Pool the liquid that becomes airborne should be assumed equal to the fraction of the water remains below 212'F (see leakage that flashes to vapor. This flash fraction, FF, should be determined Enclosure 1, Section 2.6).using a constant enthalpy, h, process, based on the maximum time-dependent temperature of the sump water circulating outside the containment:

F=N -_N Where: hn is the enthalpy of liquid at system design temperature and pressure;hf2 is the enthalpy of liquid at saturation conditions (14.7 psia, 212'F); and hfg is the heat of vaporization at 212'F.A-5.5 If the temperature of the leakage is less than 212'F or the calculated flash Conforms.

See Enclosure 1, Section fraction is less than 10%, the amount of iodine that becomes airborne should be 2.6.assumed to be 10% of the total iodine activity in the leaked fluid, unless a smaller amount can be justified based on the actual sump pH history and area ventilation rates.

NLS2008014 Attachment 1 Page 71 of 75 Appendix A Regulatory Guide 1.183 Comparison Regulatory Position No. Regulatory Position Statement CNS Analysis A-5.6 The radioiodine that is postulated to be available for release to the environment Conforms.

See Enclosure 1, Section is assumed to be 97% elemental and 3% organic. Reduction in release activity 2.5.by dilution or holdup within buildings, or by ESF ventilation filtration systems, may be credited where applicable.

Filter systems used in these applications should be evaluated against the guidance of Regulatory Guide 1.52 and Generic Letter 99-02.A-6.1 For the purpose of this analysis, the activity available for release via MSIV Conforms.

See Enclosure 1, Section leakage should be assumed to be that activity determined to be in the drywell 2.2. No credit for iodine partitioning in for evaluating containment leakage (see Regulatory Position 3). No credit the reactor vessel was assumed.should be assumed for activity reduction by the steam separators or by iodine partitioning in the reactor vessel.A-6.2 All the MSIVs should be assumed to leak at the maximum leak rate above Conforms.

See License Amendment which the technical specifications would require declaring the MSIVs Request Section 4.1.4. Leakage rates inoperable.

The leakage should be assumed to continue for the duration of the assumed in the analysis are those for accident.

Postulated leakage may be reduced after the first 24 hours2.777778e-4 days 0.00667 hours 3.968254e-5 weeks 9.132e-6 months , if which new TS limits are proposed in supported by site-specific analyses, to a value not less than 50% of the this License Amendment Request.maximum leak rate.A-6.3 Reduction of the amount of released radioactivity by deposition and plateout on Conforms.

See Enclosure 1, Section steam system piping upstream of the outboard MSIVs may be credited, but the 2.4.amount of reduction in concentration allowed will be evaluated on an individual case basis. Generally, the model should be based on the assumption of well-mixed volumes, but other models such as slug flow may be used if justified.

A-6.4 In the absence of collection and treatment of releases by ESFs such as the MSIV Conforms.

See Enclosure 1, Section leakage control system, or as described in paragraph 6.5 below, the MSIV 2.5.leakage should be assumed to be released to the environment as an unprocessed, ground-level release. Holdup and dilution in the turbine building should not be assumed.

NLS2008014 Attachment 1 Page 72 of 75 Appendix A Regulatory Guide 1.183 Comparison Regulatory Position No. Regulatory Position Statement CNS Analysis A-6.5 A reduction in MSIV releases that is due to holdup and deposition in main Conforms.

See Section 4.1.2 of this steam piping downstream of the MSIVs and in the main condenser, including License Amendment Request.the treatment of air ejector effluent by offgas systems, may be credited if the components and piping systems used in the release path are capable of performing their safety function during and following a safe shutdown earthquake (SSE). The amount of reduction allowed will be evaluated on an individual case basis. References A-9 and A- 10 provide guidance on acceptable models.A-7 The radiological consequences from post-LOCA primary containment purging Not Applicable.

CNS does not rely on as a combustible gas or pressure control measure should be analyzed.

If the Primary Containment purging for post-installed containment purging capabilities are maintained for purposes of severe LOCA combustible gas control.accident management and are not credited in any design basis analysis, radiological consequences need not be evaluated.

If the primary containment purging is required within 30 days of the LOCA, the results of this analysis should be combined with consequences postulated for other fission product release paths to determine the total calculated radiological consequences from the LOCA. Reduction in the amount of radioactive material released via ESF filter systems may be taken into account provided that these systems meet the guidance in Regulatory Guide 1.52 and Generic Letter 99-02.Footnote 6 For example, a proposed modification to change the timing of a containment Conforms.

No facility modifications are isolation valve from 2.5 seconds to 5.0 seconds might be acceptable without any proposed with this License Amendment dose calculations.

However, a proposed modification that would delay Request. Future modifications will containment spray actuation could involve recalculation of DBA LOCA doses, consider the effects upon the inputs and re-assessment of the containment pressure and temperature transient, assumptions of the LOCA AST.recalculation of sump pH, re-assessment of the emergency diesel generator loading sequence, integrated doses to equipment in the containment, and more.

NLS2008014 Attachment 1 Page 73 of 75 Appendix A Regulatory Guide 1.183 Comparison Regulatory Position No. Regulatory Position Statement CNS Analysis Footnote 7 In performing screenings and evaluations pursuant to 10 CFR 50.59, it may be Conforms.

Approval of this License necessary to compare dose results expressed in terms of whole body and thyroid Amendment Request will result in the with new results expressed in terms of TEDE. In these cases, the previous necessary implementation within the thyroid dose should be multiplied by 0.03 and the product added to the whole CNS 10 CFR 50.59 program.body dose. The result is then compared to the TEDE result in the screenings and evaluations.

This change in dose methodology is not considered a change in the method of evaluation if the licensee was previously authorized to use an AST and the TEDE criteria under 10 CFR 50.67.Footnote 8 The uncertainty factor used in determining the core inventory should be that Conforms.

See Enclosure 1, Section value provided in Appendix K to 10 CFR Part 50, typically 1.02. 2.2.Footnote 9 Note that for some radionuclides, such as Cs-137, equilibrium will not be Conforms.

See Enclosure 1, Section reached prior to fuel offload. Thus, the maximum inventory at the end of life 2.2.should be used.Footnote 10 The release fractions listed here have been determined to be acceptable for use Conforms.

The design of the current with currently approved LWR fuel with a peak burnup up to 62,000 core will remain within the Footnote 10 MWD/MTU. The data in this section may not be applicable to cores containing constraints.

NPPD is monitoring mixed oxide (MOX) fuel. Industry/NRC progress in eliminating this restriction for future core designs.Footnote 12 In lieu of treating the release in a linear ramp manner, the activity for each Not Applicable.

The LOCA release is phase can be modeled as being released instantaneously at the start of that modeled in a linear fashion.release phase, i.e., in step increases.

Footnote 13 The prior practice of basing inhalation exposure on only radioiodine and not Conforms.

See Section 4.1.7 of this including radioiodine in external exposure calculations is not consistent with License Amendment Request for a the definition of TEDE and the characteristics of the revised source term. discussion of Control Room shine as a contributor to total dose.Footnote 14 With regard to the EAB TEDE, the maximum two-hour value is the basis for Conforms.

Future changes made screening and evaluation under 10 CFR 50.59. Changes to doses outside of the pursuant to 10 CFR 50.59 will be in the two-hour window are only considered in the context of their impact on the context of the maximum 2-hour dose maximum two-hour EAB TEDE. consequences.

NLS2008014 Attachment 1 Page 74 of 75 Appendix A Regulatory Guide 1.183 Comparison Regulatory Position No. Regulatory Position Statement CNS Analysis Footnote 15 The iodine protection factor (IPF) methodology of Reference 22 may not be Conforms.

LOCA AST calculation was adequately conservative for all DBAs and control room arrangements since it developed using RADTRAD Version models a steady-state control room condition.

Since many analysis parameters 3.03.change over the duration of the event, the IPF methodology should only be used with caution. The NRC computer codes HABIT and RADTRAD incorporate suitable methodologies.

Footnote 16 This occupancy is modeled in the X/Q values determined in Reference 22 and Conforms.

The required occupancy should not be credited twice. The ARCON96 Code does not incorporate these assumptions are incorporated into the occupancy assumptions, making it necessary to apply this correction in the dose dose calculation.

See Enclosure 1, calculations.

Section 3.Footnote 18 Note that for some parameters, the technical specification value may be Not Applicable.

For the SGT System, adjusted for analysis purposes by factors provided in other regulatory guidance.

Technical Specification flows and For example, ESF filter efficiencies are based on the guidance in Regulatory efficiencies were conservatively Guide 1.52 and in Generic Letter 99-02 rather than the surveillance test criteria adjusted in accordance with RG 1.52, in the technical specifications.

Generally, these adjustments address potential Rev. 3, Table 1, and further reduced 1%changes in the parameter between scheduled surveillance tests. to account for filter bypass (See Enclosure 1, Section 2.3.2). For the Control Room Emergency Filtration System, a similar approach was taken, with further reductions taken in filtration efficiency for elemental and organic material to provide greater margin (See Enclosure 1, Section 2.7).

NLS2008014 Attachment 1 Page 75 of 75 Appendix A Regulatory Guide 1.183 Comparison Regulatory Position No. Regulatory Position Statement CNS Analysis Footnote 19 The ARCON96 computer code contains processing options that may yield X/Q Conforms.

The X/Qs used in the LOCA values that are not sufficiently conservative for use in accident consequence AST analysis were developed using assessments or may be incompatible with release point and ventilation intake ARCON96 and were previously configurations at particular sites. The applicability of these options and reviewed by the NRC pursuant to associated input parameters should be evaluated on a case-by-case basis. The License Amendment 187.assumptions made in the examples in the ARCON96 documentation are illustrative only and do not imply NRC staff acceptance of the methods or data used in the example.

NLS2008014 Attachment 2 Page 1 of 12 ATTACHMENT 2 Response to Standard NRC Questions Regarding Crediting the Standby Liquid Control System for Loss of Coolant Accident (LOCA)Alternative Source Term (AST) Suppression Pool pH Control This attachment provides responses to the "generic" questions developed by the NRC Staff addressing the SLC system which have been transmitted as Requests for Additional Information to other BWR licensees requesting application of the AST in their LOCA analyses.

The specific questions repeated below were taken from the NRC's April 26, 2007, Request for Additional Information on the Hatch AST license amendment application.

Question No. 1 Please identify whether the SLC system is classified as a safety-related system as defined in 10 CFR 50.2, "Definitions," and whether the system satisfies the regulatory requirements for such systems. If the SLC system is not classified as safety-related, please provide the information requested in Items 1.1 to 1.5 below to show that the SLC system is comparable to a system classified as safety-related.

If any item is answered in the negative, please explain why the SLC system should be found acceptableforpH control agent injection.

Response to Question No. 1 The SLC system at CNS is classified as Non-Essential (non-safety related).

A simplified diagram of the SLC system is shown in Figure B-1 below. The responses to questions

1.1 through

1.5 follow.

NLS2008014 Attachment 2 Page 2 of 12 WrTNMOVUMOMMMM4TO STURAMTA4K kWRSN.awP usmoN*rl ACCtLJIALLATO 40UA?8 QLI Figure B-1 Standby Liquid Control System (Simplified Diagram)

NLS2008014 Attachment 2 Page 3 of 12 Question No. 1.1 Is the SLC system provided with standby AC power supplemented by the emergency diesel generators?

Response to Question No. 1.1 Power and control for each of the two SLC system pumps are fed from separate essential Motor Control Centers (MCCs) which are in turn fed from divisionally separate essential buses with standby power supplied by the emergency diesel generators.

The SLC Squib valves and their associated continuity meters are powered by these same essential MCCs, providing divisionally separate essential power to these valves. SLC System pressure and level indication is powered by a non-essential power panel that is powered via a switchable power source to either of two essential buses, each of which can be supplied by a diesel generator.

The SLC system tank heaters, suction line heat tracing, and associated RTDs, temperature elements, and temperature indicating controllers are provided with backup power from the Division I bus only. However, as discussed in CNS USAR Section 111-9.3, the SLC System does not rely on this heating equipment to meet its safety objective.

The CNS Technical Specifications require verification that the sodium pentaborate solution and the SLC suction piping is within established temperature limits every 24 hours2.777778e-4 days 0.00667 hours 3.968254e-5 weeks 9.132e-6 months .Therefore, it is concluded that the SLC system is provided with standby AC power supplemented by the emergency diesel generators.

Question No. 1.2 Is the SLC system seismically qualified in accordance with Regulatory Guide (RG) 1.29,"Seismic Design Classification" and Appendix A to 10 CFR Part 100 (or equivalent used for original licensing)?

Response to Question No. 1.2 CNS was designed and constructed prior to the adoption of Appendix A to 10 CFR 100 or Safety Guide 29. However, in the CNS Safety Evaluation Report, the Atomic Energy Commission concluded that CNS met the intent of both 10 CFR 100 Appendix A and Safety Guide 29. As described in the CNS USAR, the SLC system equipment and piping required for injection of sodium pentaborate has been designed or qualified to CNS Seismic Class I requirements.

Question No. 1.3 Is the SLC system incorporated into the plant's ASME Code inservice inspection and inservice testing programs based upon the plant's code of record (10 CFR 50.55a)?

NLS2008014 Attachment 2 Page 4 of 12 Response to Question No. 1.3 The applicable components (Class 1 components) of the SLC system are inspected and tested in accordance with the CNS ASME Boiler and Pressure Vessel Code Inservice Inspection (ISI) and Inservice Testing (IST) programs, as required by 10 CFR 50.55(a).

The Class 1 boundary is from the SLC outboard injection check valves to the reactor vessel. The ISI scope includes the SLC piping and components within the Class I boundary, and the IST program scope includes the SLC injection check valves. CNS includes as part of the its Augmented IST program, testing of the SLC pumps, the SLC pump discharge check valves, the SLC pump discharge relief valves, and the SLC explosive (Squibb) valves.Question No. 1.4 Is the SLC system incorporated into the plant's Maintenance Rule program consistent with 10 CFR 50.65?Response to Question No. 1.4 The SLC system is in the CNS Maintenance Rule program consistent with 10 CFR 50.65.Question No. 1.5 Does the SLC system meet 10 CFR 50.49 and Appendix A to 10 CFR Part 50 (GDC-4, or equivalent used for original licensing)?

Response to Question No. 1.5 The new Alternative Source Term (AST) based post-LOCA mission for the SLC system has been evaluated for environmental qualification of electric equipment important to safety based on the requirements contained in 10 CFR 50.49. The SLC system electrical equipment can be categorized into (1) equipment required to operate that is exposed to a harsh environment during its mission time; (2) equipment required to operate that is not exposed to a harsh environment during its mission time, and (3) equipment that is not exposed to a harsh environment, i.e., Control Room equipment (mild environment).

Equipment in category (1) has been evaluated as either qualified or identical to qualified equipment.

Equipment in category (2) and (3) are exempt from 10 CFR 50.49 qualification.

Necessary programmatic updates to the EQ documentation for category (1) equipment will be made as part of the implementation of this License Amendment Request (LAR).Question No. 2 Please describe proposed changes to plant procedures that implement SLC sodium pentaborate injection as a pH control additive.

In addition, please address Items 2.1 to 2.5 below in your response.

If any item is answered in the negative, please explain why the SLC system should be found acceptablefor pH control additive injection.

NLS2008014 Attachment 2 Page 5 of 12 Question No. 2.1 Are the SLC injection steps part of a safety-related plant procedure?

Response to Question No. 2.1 The steps necessary for the Control Room operator to manually initiate SLC injection will be contained in a station Emergency Procedure used for LOCA mitigation.

This is a "safety related" procedure in that it falls under the auspices of the CNS Quality Assurance program. Future changes to the procedure would be made under the provisions of 1 OCFR50.59.

A controlled copy of this procedure is located in the Control Room.As discussed in Section 4.1.1 of this LAR, initiation of the SLC system for control of Suppression Pool pH will be based on indication of a LOCA with core damage (high drywell pressure with high drywell radiation levels). The Emergency Procedure that will initiate SLC injection is the same one that currently directs the configuration of the Main Steam Pathway under the same indications of LOCA with core damage. The necessary procedures will be in place as part of the requested 30-day implementation of the LOCA AST license amendment.

Question No. 2.2 Are the entry conditions for the SLC injection procedure steps symptoms of imminent or actual core damage?Response to Question No. 2.2 The parameters that will be used to direct activation of the SLC system are high drywell pressure combined with high drywell radiation as indicated on the Containment High-Range Radiation Monitors.

High drywell radiation is a symptom of actual core damage. High drywell pressure is an indication of a LOCA.Question No. 2.3 Does the instrumentation cited in the procedure entry conditions meet the quality requirements for a Type E variable as defined in Tables ] and 2 of RG 1.97 on accident monitoring instrumentation?

Response to Question No. 2.3 The high drywell pressure instruments are required to be operable by TS 3.3.6.1. This instrumentation meets the quality requirements for a Type A and B variable as defined in Regulatory Guide (RG) 1.97, Revision 2, Table 1 for BWRs. The drywell high radiation instruments are required to be operable by TS 3.3.3.1. The drywell high range radiation monitors NLS2008014 Attachment 2 Page 6 of 12 meet the quality requirements for a Type C and E variable as defined in RG 1.97, Revision 2 Table 1 for BWRs.Question No. 2.4 Have plant personnel received initial and periodic refresher training in the SLC injection procedure?

Response to Question No. 2.4 As an implementing action for this LAR, licensed personnel and Shift Technical Engineers will be initially trained and receive periodic refresher training on these procedures as part of the Licensed Operator Requalification Training program. Initial training on the procedure revisions discussed in 2.1 above will be conducted as part of the requested 30-day implementation of the LOCA AST license amendment.

Question No. 2.5 Have other plant procedures, (e.g., Emergency Response Guidelines) that callfor termination of SLC as a reactivity control measure been appropriately revised to prevent blocking of SLC injection as a pH control measure? (For example, the override before Step RC/Q-1, "If while executing the following steps: .... it has been determined that the reactor will remain shutdown under all conditions without boron, terminate boron injection and....')Response to Question No. 2.5 As an implementing action for this LAR, NPPD will review the EOPs, SAGs, and supporting abnormal operating procedures and revise as necessary to ensure that there are no procedural steps that would inhibit or suspend the injection of the entire contents of the SLC system sodium pentaborate solution storage tank under the LOCA conditions assumed in the LOCA Analysis.Question No. 3 Please show that the SLC system has suitable redundancy in components and features to assure that for onsite or offsite electric power operation its safety function of injecting sodium pentaborate for the purpose of suppression pool pH control can be accomplished assuming a single failure. For this purpose, the check valve is considered an active device since the check valve must open to inject sodium pentaborate.

If the SLC system cannot be considered redundant with respect to its active components, the licensee should implement one of the three options described below, providing the information specified for that option for staff review.3.1 Option ] Show acceptable quality and reliability of the non-redundant active components and/or compensatory actions in the event offailure of the nonredundant active components.

If you choose this option, please provide the following information to justify the lack of NLS2008014 Attachment 2 Page 7 of 12 redundancy of active components in the SLC system: 3.1.1 Identify the non-redundant active components in the SLC system and provide their make, manufacturer, and model number.3.1.2 Provide the design-basis conditions for the component and the environmental and seismic conditions under which the component may be required to operate during a design-basis accident.

Environmental conditions include design-basis pressure, temperature, relative humidity and radiation fields.3.1.3 Indicate whether the component was purchased in accordance with Appendix B to 10 CFR Part 50. If the component was not purchased in accordance with Appendix B, provide information on the quality standards under which it was purchased.

3.1.4 Provide

the performance history of the component both at the licensee's facility and in industry databases such as Equipment Performance and Information and Exchange System and Nuclear Plant Reliability Data System.3.1.5 Provide a description of the component's inspection and testing program, including standards, frequency, and acceptance criteria.3.1.6 Indicate potential compensating actions that could be taken within an acceptable time period to address the failure of the component.

An example of a compensating action might be the ability to jumper a switch in the control room to overcome its failure. In your response, please consider the availability of compensating actions and the likelihood of successful injection of the sodium pentaborate when non-redundant active components fail to perform their intended functions.

3.2 Option

2 Provide for an alternative success path for injecting chemicals into the suppression pool. Ifyou chose this option, please provide the following information:

3.2.1 Provide

a description of the alternative injection path, its capabilities for performing the pH control function, and its quality characteristics.

3.2.2 Do the components which make up the alternative path meet the same quality characteristics required of the SLC system as described in Items 1.1 to 1.5, 2, and 3 above?3.2.3 Does the alternate injection path require actions to be taken in areas outside the control room? How accessible will these areas be? What additional personnel would be required?3.3 Option 3 Show that 10 CFR 50.67 dose criteria are met even if the pH is not controlled.

If you chose this option, demonstrate through analyses that the projected accident doses will NLS2008014 Attachment 2 Page 8 of 12 continue to meet the criteria of 10 CFR 50.67 assuming that the suppression pool pH is not controlled.

The dissolution of cesium iodide (CsI) and its re-evolution from the suppression pool as elemental iodine must be evaluated by a suitably conservative methodology.

The analysis of iodine speciation should be provided for staff review. The analysis documentation should include a detailed description and justification of the analysis assumptions, inputs, methods, and results. The resulting iodine speciation should be incorporated into the dose analyses.

The calculation may take credit for the mitigating capabilities of other equipment, for example the standby gas treatment system, ifsuch equipment would be available.

A description of the dose analysis assumptions, inputs, methods, and results should be provided.

Licensees proposing this approach should recognize that this option may incur longer staff review times and will likely involve fee-billable support from NRC staff contractor support.Response to Question No. 3 The CNS SLC system can be considered redundant with respect to its active components, except with respect to its final injection path. The following information is provided in response to Option 1 of Question No. 3.Question No. 3.1.1 Identify the non-redundant active components in the SLC system and provide their make, manufacturer, and model number.Response to Question No. 3.1.1 The only non-redundant active components of the SLC system are the two check valves in series (one inboard and one outboard) located on containment penetration X-42 for the SLC injection line. The make, manufacturer, and model number for the check valves are as follows: CNS Functional Location:

CNS-3-SLC-CV-12V (outboard) and CNS-3-SLC-CV-13CV (inboard)Make: Lift Check Valve Manufacturer:

Dresser Valve and Control 3 Model Number: 1 1/2" -7440W The check valves are inherently rugged. The check valves are constructed by placing the valve into the body (see Table 1 and Figure 2 for parts nomenclature).

The cap is then screwed into the body extension, torqued to 50 ft/lbs, and then seal-welded.

The only moving part is the valve within the slot in the cap.3. Dresser was purchased by Yarway Corporation.

Yarway is now the manufacturer.

NLS2008014 Attachment 2 Page 9 of 12 Table 1 Check Valve List of Materials Part No. Part Name Material Specifications 1 Body Stainless Steel ASTM A182 Grade F316 2 Cap Stainless Steel ASTM A182 Grade F316 3 Valve* Stainless Steel AISI Type 316 4 Body Extension Stainless Steel ASTM A 182 Grade F316 Figure 2 Dresser 1-1/2" -7440W Check Valve NLS2008014 Attachment 2 Page 10 of 12 Question No. 3.1.2 Provide the design-basis conditions for the component and the environmental and seismic conditions under which the component may be required to operate during a design-basis accident.

Environmental conditions include design-basis pressure, temperature, relative humidity and radiation fields.Response to Question No. 3.1.2 The seismic design is discussed in the response to Item 1.2. The non-redundant components are 1 1/2 inch check valves as identified above. These components are purely mechanical in nature and have no "soft" organic subcomponents such as o-rings or gaskets. The check valve has only four parts: a body, a cap, a valve disc, and a body extension.

These are made of stainless steel as indicated in Table 1. Therefore, the components are not affected by the design basis accident environmental conditions of temperature, pressure, relative humidity, and radiation.

Check valves CNS-3-SLC-CV-12CV and 13CV are classified as Essential and Class I (seismic).

Question No. 3.1.3 Indicate whether the component was purchased in accordance with Appendix B to 10 CFR Part 50. If the component was not purchased in accordance with Appendix B, provide information on the quality standards under which it was purchased.

Response to Question No. 3.1.3 The SLC injection check valves were purchased as original equipment for CNS. They were designated as Essential equipment and are maintained within the requirements of the CNS 10 CFR 50 Appendix B Quality Assurance program and subject to the same requirements as other components installed in safety-related systems.Question No. 3.1.4 Provide the performance history of the component both at the licensee's facility and in industry databases such as Equipment Performance and Information and Exchange System and Nuclear Plant Reliability Data System..Response to Question No. 3.1.4 CNS has researched the performance history of the check valves which are not functionally redundant within the SLC system using CNS maintenance and surveillance records, EPIX (Equipment Performance and Information Exchange) and NPRDS (Nuclear Plant Reliability Data System). The failure mode of concern is failure of the check valve to open. The applicable performance history is detailed below:

NLS2008014 Attachment 2 Page 11 of 12 CNS Records: CNS maintenance and surveillance records were researched to determine the Dresser 7440W check valve performance history. The review indicates that there were no failures to open for any of the Dresser 7440W check valves. The only incident for the Dresser 7440W check valve was failure of a Local Leak Rate Test (LLRT) for SLC-CV-13CV (inboard)in October 1998. It should be noted that this failure does not indicate any problem with the ability of the check valve to open for SLC fluid injection.

EPIX Database (1/1/97 to 9/2005): There were only the following two EPIX listings for the Dresser 7440W check valve: " Beaver Valley- Failure was internal leakage. This failure mode is not applicable to the injection safety function of the CNS SLC check valves.* Perry 1- Failure was in the HPCS system fill water line. The waterleg pump discharge check valve did not re-open after HPCS system testing. It was attributed to general corrosion; the valve body material was carbon steel; and the expected working pressure was about 50 psig. The check valve was agitated and flow was resumed. The CNS SLC check valves are stainless steel (see Table 1) and the working pressure is in the range of 1,000 psig. Therefore, this reported failure is not applicable to the CNS SLC check valves.NPRDS Database (prior to January 1, 1997): There were 21 failures identified for Dresser with model identification numbers containing "7440." Of these failures, there were no failures to open. The listed failures were either internal leakage or failure to close.The EPIX and NPRDS reviews demonstrate the SLC system check valves are of acceptable quality and reliability.

Question No. 3.1.5 Provide a description of the component's inspection and testing program, including standards, frequency, and acceptance criteria.Response to Question No. 3.1.5 SLC check valve testing is accomplished each refueling outage during the flow test which directs demineralized water into the RPV at rated SLC pump flow. This testing consists of a functional test of the SLC system and requires initiation of one train of SLC, from the control room, using the SLC test tank as a source of demineralized water. This is performed per TS Surveillance Requirement 3.1.7.8.

NLS2008014 Attachment 2 Page 12 of 12 Question No. 3.1.6 Indicate potential compensating actions that could be taken within an acceptable time period to address the failure of the component.

An example of a compensating action might be the ability to jumper a switch in the control room to overcome its failure. In your response, please consider the availability of compensating actions and the likelihood of successful injection of the sodium pentaborate when non-redundant active components fail to perform their intended functions.

Response to Question No. 3.1.6 The SLC check valves are considered highly reliable and the lack of functional redundancy is offset by the reliability as discussed in the response to RAI Nos. 3.1.1 -3.1.5 above. Therefore, no additional compensatory actions are considered necessary to ensure injection of SLC through this flow path.

NLS2008014 Attachment 3 Page I of 6 ATTACHMENT 3 Proposed Exemption to 10 CFR 50 Appendix J

1.0 INTRODUCTION

10 CFR 50.54(o) requires that primary reactor containments be subject to the requirements of Appendix J to 10 CFR 50. Appendix J specifies the leakage rate test requirements, schedules, and acceptance criteria for tests of the leak-tight integrity of the primary reactor containment and systems and components that penetrate the containment.

In Option B of Appendix J, Paragraph III.A requires that the overall integrated leakage rate (Type A test) must not exceed the allowable leakage (La) with margin, as specified in the Technical Specifications (TS), and Paragraph III.B requires the sum of the leakage of Type B and Type C leakage rate tests to be less than the performance criterion (La) with margin as specified in the TS. The overall integrated leakage rate, as specified in the 10 CFR 50 Appendix J definitions, includes the contribution from Main Steam (MS)Pathway leakage. (MS Pathway leakage includes leakage through the four Main Steam lines and the Main Steam inboard drain line).By letter dated March 15, 2006 (Reference 1), the Nebraska Public Power District (NPPD) previously requested exemption from Paragraphs III.A and III.B of 10 CFR 50 Appendix J, Option B for Cooper Nuclear Station (CNS) for the leakage contribution from the four MS line penetrations, tested in accordance with CNS Technical Specifications Surveillance Requirement 3.6.1.3.10.

That exemption was granted by the Nuclear Regulatory Commission (NRC), by letter dated October 30, 2006 (Reference 2), and corresponding CNS License Amendment No. 226 (Reference

3) to revise CNS Technical Specifications, Section 5.5.12, "Primary Containment Leakage Rate Testing Program," to reflect exemption from Section III.A and Section llI.B of 10 CFR 50, Appendix J, Option B for the four MS line penetrations.

Concurrent with the request for license amendment, Nebraska Public Power District (NPPD) hereby requests an exemption from the requirements of 10 CFR 50, Appendix J, Option B, Paragraphs III.A and III.B for Cooper Nuclear Station (CNS) to permit exclusion of the MS Pathway leakage contribution from the overall integrated leakage rate Type A test measurement and from the sum of the leakage rates from Type B and Type C tests. The MS Pathway leakage includes the leakage from the four MS line penetrations plus the leakage from the MS inboard drain line (penetration X-8). In References 2 and 3, the NRC granted CNS an exemption from 10 CFR 50, Appendix J, Option B, Paragraphs III.A and III.B for the MS line penetrations.

This exemption request asks that CNS be exempted from including the entire MS Pathway leakage (the MS line penetrations and Main Steam inboard drain line leakage) in determining the Type A and Type B and C total leakage. This request for exemption is similar to exemptions granted from the requirements of Paragraphs III.A and III.B of Option B for the NLS2008014 Attachment 3 Page 2 of 6 Monticello Nuclear Generating Plant on December 7, 2006 (Reference 4), and for the Browns Ferry Nuclear Plant (Units 2 and 3), on March 14, 2000 (Reference 5).2.0 10 CFR 50.12 -SPECIFIC EXEMPTIONS 10 CFR 50.12 states that the Commission will not consider granting an exemption unless special circumstances are present. NPPD considers this request as meeting the special circumstance as defined in 50.12(a)(2)(ii), which states, "Application of the regulation in the particular circumstances would not serve the underlying purpose of the rule or is not necessary to achieve the underlying purpose of the rule." The applicable rule is 10 CFR 50, Appendix J, Option B, Paragraphs III.A and III.B.These state (in part): 1. Type A test The leakage rate must not exceed the allowable leakage rate (La) with margin, as specified in the Technical Specifications.

2. Type B and C Tests The tests must demonstrate that the sum of the leakage rates at accident pressure of Type B tests, and pathway leakage rates from Type C tests, is less than the performance criterion (La) with margin, as specified in the Technical Specification.

The underlying purpose of the rule is to ensure the actual radiological consequences of design basis accidents (DBAs) remain below the analyzed consequences as demonstrated through the measured containment and local leakage rate tests.3.0 REQUESTED EXEMPTION NPPD requests a permanent exemption from: 1) the requirements of 10 CFR 50, Appendix J, Option B, Paragraph III.A. to allow exclusion of the MS Pathway leakage, including the leakage from the MS inboard drain line, from the overall integrated leakage rate measured when performing a Type A test, and 2) the requirements of 10 CFR 50, Appendix J, Option B, Paragraph III.B, to allow exclusion of the MS Pathway leakage, including the MS inboard drain line leakage, from the combined leakage rate of the penetrations and valves subject to Type B and C tests.4.0 JUSTIFICATION CNS License Amendment 180 (Reference

6) authorized CNS to use 10 CFR 50, Appendix J, Option B provisions for Type A, Type B and Type C tests. MSIV leakage NLS2008014 Attachment 3 Page 3 of 6 was included in the Type A overall containment integrated leakage rate total, and added to the combined Type B and C leakage rate total.NRC Exemption (Reference

2) and CNS License Amendment 226 (Reference 3), approved an exemption from Section III.A of 10 CFR 50, Appendix J, Option B, to allow the leakage contribution from MSIV leakage to be excluded from the overall integrated leakage rate from Type A test, and an exemption from Section III.B of 10 CFR 50, Appendix J, Option B, to allow the contribution from MSIV leakage to be excluded from the sum of the leakage rates from Type B and Type C tests.Under the AST design basis accident radiological consequence analyses, MS Pathway leakage has been accounted for separately from the overall containment integrated leakage, local leakage across pressure retaining, leakage limiting boundaries, and containment isolation valve leakage. Specifically, the AST design basis accident analyses use the MS piping, the MS drain lines, and the Main Condenser as an alternate means for MS Pathway leakage treatment.

Under the original plant design basis, certain MS and MS drain line piping, as well as the Main Condenser, were not classified as seismic category I components.

However, NPPD has demonstrated these components to be seismically rugged and thus capable of performing as an MS Pathway leakage treatment system. NRC approval of the NPPD evaluation was documented in Amendment 206, dated September 1, 2004 (Reference 7).5.0 AUTHORIZED BY LAW The proposed exemption has been previously granted to other licensees.

For example, the NRC granted this exemption to the Monticello Nuclear Generating Plant (Reference 4)and the Browns Ferry Nuclear Plant, Units 2 and 3 (Reference 5). Based on this, the proposed exemption is authorized by law.6.0 NO UNDUE RISK TO PUBLIC HEALTH AND SAFETY The exemption presents no undue risk to public health and safety. MSIV and MS Pathway leakage for the CNS design basis accident analyses has been accounted for separately from the overall leakage associated with the primary containment boundary, local leakage, and containment isolation valve leakage. As such, including the MSIV and MS Pathway leakage as part of the Type A and Type B and C test results is not necessary to ensure the actual radiological consequences of design basis accidents remain below those previously evaluated and accepted.

The exemption will not result in any change to the previously evaluated consequences associated with design basis accidents.

As such, the proposed exemption presents no undue risk to public health and safety.

NLS2008014 Attachment 3 Page 4 of 6 7.0 CONSISTENT WITH COMMON DEFENSE AND SECURITY This exemption is consistent with the common defense and security of the United States.The Commission's Statement of Considerations in support of the exemption rule noted with approval in the explanation as set forth in Long Island Lighting Company (Shoreham Nuclear Power Station, Unit 1), LBP-84-45, 20 NRC 1343, 1400 (October 29, 1984) that the term "common defense and security" refers principally to the safeguarding of special nuclear material, the absence of foreign control over the applicant, the protection of Restricted Data, and the availability of special nuclear material for defense needs.Granting the requested exemption does not affect any of these matters, and thus, is consistent with common defense and security.8.0 SPECIAL CIRCUMSTANCES 10 CFR 50.12(a)(2) identifies six specific special circumstances, on which must be present for the NRC to consider granting an exemption.

The special circumstance present for this exemption is 10 CFR 50.12(a)(2)(ii), which states: "Application of the regulation in the particular circumstances would not serve the underlying purpose of the rule or is not necessary to achieve the underlying purpose of the rule." The underlying purpose of the rule is to ensure the actual radiological consequences of design basis accidents remain below those previously evaluated and accepted, as demonstrated by the actual, periodic measurement of containment leakage, local leakage and containment isolation valve leakage. Although Type A and Type B and C leakage tests are defined as a measurement of the overall primary containment leak rate, local leakage and containment isolation valve leakage, inclusion of the MSIV and MS Pathway leakage measurements in the Type A and Type B and C test leakage results would result in double counting of the MSIV leakage in assessing the actual integrated leakage of the containment and containment isolation valve leakage. As such, this exemption addresses a special circumstance in which application of the regulation requiring the inclusion of MSIV and MS Pathway leakage in the Type A and Type B and C leakage testing results is not necessary to achieve the underlying purpose of the regulation.

9.0 ENVIRONMENTAL

IMPACT The proposed exemption does not cause additional construction or operational activities to be conducted that may significantly affect the environment.

No plant configuration changes are required.The exemption does not result in an increase in adverse environmental impact previously evaluated, does not result in a change to effluents or power levels, and does not affect any NLS2008014 Attachment 3 Page 5 of 6 matter not previously reviewed by the NRC which may have a significant adverse environmental impact.The proposed exemption does not alter the land use, water uses, or impacts on water or air quality at CNS. It does not affect the ecology of the site and vicinity and does not affect the noise emitted by the station. Therefore, the exemption does not affect the analysis of the environmental impacts described in the environmental report.

10.0 REFERENCES

1. Letter from R. Edington (NPPD) to USNSRC, dated March 15, 2006, "Request for Exemption from Certain Requirements of 10 CFR 50, Appendix J and Corresponding Change to Technical Specification 5.5.12, Primary Containment Leakage Rate Testing Program, Cooper Nuclear Station, Docket No. 50-298, License No. DPR-46." 2. Letter from B. Benney (USNRC) to R. Edington (NPPD), dated October 30, 2006,"Cooper Nuclear Station -Exemption From the Requirements of Title 10 of the Code of Federal Regulations, Part 50, Appendix J (TAC No. MD0568)." 3. Letter from B. Benney (USNRC) to R. Edington (NPPD), dated October 31, 2006,"Cooper Nuclear Station -Issuance of Amendment Re: MSIV Leakage Exemption From 10 CFR Part 50, Appendix J (TAC No. MD0570)." 4. Letter from P. Tam (USNRC) to J. Conway (Nuclear Management Company, LLC), dated December 7, 2006, "Monticello Nuclear Generating Plant -Issuance of Amendment Re: Full-Scope Implementation of the Alternative Source Term Methodology (TAC No. MC8971)," [ADAMS Accession Number ML062790015].

5. Letter from W. Long (USNRC) to J. Scalice (Tennessee Valley Authority), dated March 14, 2000, "Browns Ferry Nuclear Plant, Issuance of Exemption From 10 CFR 50, Appendix J, (TAC Nos. MA6815 and MA6816)," [ADAMS Accession Number ML003691985].

6. Letter from L. J. Burkhart (USNRC) to J. Swailes (NPPD), dated March 3, 2000,"Cooper Nuclear Station -Issuance of Amendment Re: Changes to the Technical Specifications (TSs) to Implement 10 CFR 50 Appendix J, Option B, and Changes to the TS Associated With Containment Airlock Interlock Mechanism, Isolation Valve Time Testing, and Credit for Administrative Means of Securing Isolation Devices, (TAC No. MA 6877)," [Amendment 180].7. Letter from M. Honcharik (USNRC) to R. Edington (NPPD), dated September 1, 2004, "Cooper Nuclear Station -Issuance of Amendment on Loss-of Coolant NLS2008014 Attachment 3 Page 6 of 6 Accident (LOCA) Dose Methodology and Resolution of Remaining Condition 2.C.(6) Issues, (TAC No. MC 1572) -Amendment 206.

NLS2008014 Attachment 4 Page 1 of 9 ATTACHMENT 4 Proposed Technical Specifications Markup Format Cooper Nuclear Station, Docket 50-298, DPR-46 Listing of Revised Pages TS Pages 1.1-2 1.1-3 3.1-20 3.6-14 3.6-15 5.0-16 5.0-17 Definitions 1.1 1.1 Definitions CHANNEL CHECK.(continued)

CHANNEL FUNCTIONAL TEST CORE ALTERATION status derived from independent instrument channels measuring the same parameter.

A CHANNEL FUNCTIONAL TEST shall be the.injection of a simulated br actual signal into the channel as close to the sensor as practicable to verify OPERABILITY of all devices in the channel required for channel OPERABILITY.

The CHANNEL FUNCTIONAL TEST may be performed by means of any series of sequential, overlapping, or total.channel steps.CORE ALTERATION shall be the movement of any fuel, sources, or reactivity control componenls within the reactor vessel with the vessel head removed and fuel in the vessel. The following exceptions are not considered lo be CORE ALTERATIONS:

a. Movement of source range monitors, local power range monitors, intermediate range monitors, traversing incore probes, or special movable delectors (including undervessel replacement);

and b. Control rod movement, provided Ihere are no fuel assemblies in the associated core cell.Suspension of CORE ALTERATIONS shall not preclude completion of movement 6f a component to a safe position.CORE OPERATING LIMITS REPORT (COLR)DOSE EQUIVALENT 1-131 The COLR is the unit spe'cific document that provides cycle specific parameter limits for the current reload cycle. These cycle specific limits shall be determined for each reload cycle in accordance with Specification 5.6.5.Plant operation within these limits is addressed in individual Specifications.

DOSE EQUIVALENT 1-131 shall be that concenlration of 1-13 fricrocuries/gram) thal ould produce the same.:ld ose as the quantily and isotopic mixlure of 1-131, 1-132, (continued)

Cooper 1.1-2 Amendment No.-9""7 Definitions 1.1 1.1' Definitions DOSE EQUIVALENT 1-131 (continued)

LEAKAGE LEAKAGE shall be: 1.'a. Identified LEAKAGE 1. LEAKAGE into the drywell, such as that from pump seals or valve packing, that is captured and conducted to a Sump or collecting tank; or 2. LEAKAGE into the drywell almosphere from sources (hat are both specifically located and known either not to interfere with the operation of leakage detection systems or not to be pressure boundary LEAKAGE;b. Unidentified LEAKAGE All LEAKAGE into the drywell that is not identified LEAKAGE;c. Total LEAKAGE Sum of the identified and unidentified LEAKAGE;d. Pressure Boundary LEAKAGE LEAKAGE through a nonisolable fault in a Reactor Coolant System (RCS) component body, pipe wall, or vessel wall.A LOGIC SYSTEM FUNCTIONAL TEST shall be a test of all logic components required for OPERABILITY of a logic circuit, LOGIC SYSTEM FUNCTIONAL TEST (continued)

Cooper 1.1 -3 Amendment No. 2+/-r Insert 1.1-1 The DOSE EQUIVALENT 1-131 concentration is calculated as follows: DOSE EQUIVALENT 1-131 = (1-131) + 0.0060 (1-132) + 0.17 (1-133) + 0.0010 (1-134) + 0.029 (I-135). The dose conversion factors used for this calculation are those listed in Federal Guidance Report (FGR) 11, "Limiting Values of Radionuclide Intake and Air Concentration and Dose Conversion Factors for Inhalation, Submersion, and Ingestion," 1989.

SLC System 3.1.7 3.1 REACTIVITY CONTROL SYSTEMS 3.1.7 Standby Liquid Control (SLC) System LCO 3.1.7 Two SLC subsystems shall be OPERABLE.APPLICABILITY:

MODES 1, a/d21 0 F~c3 ACTIONS CONDITION REQUIRED ACTION COMPLETION TIME A. One SLC subsystem A.1 Restore SLC subsystem 7 days inoperable, to OPERABLE status.B. Two SLC subsystems B.1 Restore one SLC 8 hours9.259259e-5 days 0.00222 hours 1.322751e-5 weeks 3.044e-6 months inoperable, subsystem to OPERABLE status.C. -Reqirid Action and C.I Be in MODE 3. 12 hours1.388889e-4 days 0.00333 hours 1.984127e-5 weeks 4.566e-6 months assOfiated Completion Time not met.tCZ 5 ; , f v =ýo ý e- : DZJ; r\ fv-, o ý e.3Gc ~O Cooper 3.1-20 Amendment No. +7&=-

PCIVs 3.6.1.3 SURVEILLANCE REQUIREMENTS (continued)

SURVEILLANCE FREQUENCY SR 3.6.1.3.6 Verify the isolation time of each MSIV is In accordance

> 3 seconds and < 5 seconds. with the Inservice Testing Program SR 3.6.1.3.7 Verify each automatic PCIV actuates to the 18 months isolation position on an actual or simulated isolation signal.SR 3.6.1.3.8 Verify a representative sample of reactor 18 months instrumentation line EFCVs actuate to the isolation position on an actual or simulated instrument line break.SR 3.6.1.3.9 Remove and test the explosive squib from each 18 months on a shear isolation valve of the TIP System. STAGGERED TEST BASIS SR 3.6.1.3.10 Verify ed main stepmn line leak@ r ate I In accordance

<46 scfhen tested qr' 29psig. with the Primary Containment Leakage Rate Testing Program Ve-- CN r* i- CA +c tý\eo Ut e-*CL(:,1j le Ck rv,.VCAlae at z Cý 5L ck, at (continued).

Cooper 3.6-14 Amendment No.-2-=2"6 PCIVs 3.6.1.3 SURVEILLANCE REQUIREMENTS (continued)

SURVEILLANCE FREQUENCY SR 3.6.1.3.11 Verify each inboard 24 inch primary containment purge and vent valve is blocked to restrict the maximum valve opening angle to 600.18 months I ILA ck, i a C .C~cicA4/V(C014,'Cooper 3.6-15 Amendment No.WT4-8 Programs and Manuals 5.5 5.5 Programs and Manuals 5.5.11 Safety Function Determination Program (SFDP) (continued)

For the purpose of this program, a loss of safety function may exist when a support system is inoperable, and: 1. A required system redundant to system(s) supported by the inoperable support system is also inoperable; or 2. A required system redundant to system(s) in turn supported by the inoperable supported system is also inoperable; or 3. A required system redundant to support system(s) for the supported systems b.1 and b.2 above is also inoperable.

The SFDP identifies where a loss of safety function exists. If a loss of safety function is determined to exist by this program, the appropriate Conditions and Required Actions of the LCO in which the loss of safety function exists are required to be entered.5.5.12 Primary Containment Leakage Rate Testing Program a. A program shall establish the leakage rate testing of the containment as required by 10 CFR 50.54(o) and 10 CFR 50, Appendix J, Option B, as modified by approved exemptions.

This program shall be in accordance with the guidelines contained in Regulatory Guide 1.163, "Performance-Based Containment Leak-Test Program," dated September, 1995, as modified by the following exceptions:

1. Exemption from Appendix J to 10CFR Part 50 to allow reverse direction local leak rate testing of four containment isolation valves at Cooper Nuclear Station (TAC NO. M89769) (July 22, 1994).2. Exemption from Appendix J to I OCFR Part 50 to allow MSIV testing at 29 psig and expansion bellows testing at 5 psig between the plies (Sept. 16, 1977).3. Exception to NEI 94-01, "Industry Guideline for Implementing Performance-Based Option of 10 CFR Part 50, Appendix J," Section 9.2.3: The first Type A test performed after the December 7, 1998 Type A test shall be performed no later than December 7, 2013.4. Exemption from Section III.A of 10CF.art 50, Appendix J, Option B, to.-,eo -.% k, e an. A allow the leakage con ri u ion fror l eakage to be exclud S ur the overall integrated leakage rate from Type A tests Octer 30,4006)drmi' A C (conlinued)

Cooper 5.0-16 Amendment,2-?ý Programs and Manuals 5.5 5.5 Programs and Manuals 5.5.12 Primary Containment Leakage Rate Testing Program (continued)(f'ickjv G4ecý w% 5. Exemption from Section IIIB of 10CFR Part 50, Appendix J, Option B, to S'% 'a I (f 0rI% allow the contribution froMcileakage to be excluded from the sum of S4 e the leakage rates from Type B and Type C tests (.6_Cto~r 30/2006j o,'i v%'% b. The peak calculated containment internal pressure for the asi of j~looctcdý ok-e,,) ," coolant accident, P,, is 58.0 psig. The containment design pressure is 56.0 psig.c. The maximum allowable containment leakage rate, La, at Pa, shall be 0.635% of containment air weight per day.d. Leakage Rate acceptance criteria are: 1. Containment leakage rate acceptance criterion is 5 1.0 L,. During the first unit startup following testing in accordance with this program, the leakage rate acceptance criteria are, <0.60 L) for the Type B and C tests and 0.75 L, for Type A tests.2. Air lock testing acceptance criteria are: a. Overall air lock leakage rate is : 12 scfh when tested at _ P,.b. Overall air lock leakage rate is !5 0.23 scfh when tested at-, 3.0 psig.e. The provisions of SR 3.0.2 do not apply to the test frequencies specified in the Primary Containment Leakage Rate Testing Program.The provisions of SR 3.0.3 are applicable to the Primary Containment Leakage Rate Testing Program.5.5.13 Control Room Envelope Habitability Program A Control Room Envelope (CRE) Habitability Program shall be established and implemented to ensure that CRE habitability is maintained such that, with an OPERABLE Control Room Emergency Filter (CREF) System, CRE occupants can control the reactor safely under normal conditions and maintain it in a safe condition following a radiological event, hazardous chemical release, or a smoke challenge.

The program shall ensure that adequate radiation protection is provided to permit access and occupancy of the CRE under design basis accident (DBA) conditions without (continued)

Cooper 5.0-17 Amendment No. 2 NLS2008014 Attachment 5 Page 1 of 8 ATTACHMENT 5 Proposed Technical Specifications Final Typed Format Cooper Nuclear Station, Docket 50-298, DPR-46 Listing of Revised Pages TS Pages 1.1-2 1.1-3 3.1-20 3.6-14 3.6-15 5.0-16 5.0-17 Definitions 1.1 1.1 Definitions CHANNEL CHECK (continued)

CHANNEL FUNCTIONAL TEST CORE ALTERATION status derived from independent instrument channels measuring the same parameter.

A CHANNEL FUNCTIONAL TEST shall be the injection of a simulated or actual signal into the channel as close to the sensor as practicable to verify OPERABILITY of all devices in the channel required for channel OPERABILITY.

The CHANNEL FUNCTIONAL TEST may be performed by means of any series of sequential, overlapping, or total channel steps.CORE ALTERATION shall be the movement of any fuel, sources, or reactivity control components within the reactor vessel with the vessel head removed and fuel in the vessel. The following exceptions are not considered to be CORE ALTERATIONS:

a. Movement of source range monitors, local power rang e monitors, intermediate range monitors, traversing incore probes, or special movable detectors (including undervessel replacement);

and b. Control rod movement, provided there are no fuel assemblies in the associated core cell.Suspension of CORE ALTERATIONS shall not preclude completion of movement of a component to a safe position.The COLR is the unit specific document that provides cycle specific parameter limits for the current reload cycle. These cycle specific limits shall be determined for each reload cycle in accordance with Specification 5.6.5.Plant operation within these limits is addressed in individual Specifications.

DOSE EQUIVALENT 1-131 shall be that concentration of 1-131 (microcuries/gram) that would produce the same dose as the quantity and isotopic mixture of 1-131, 1-132, (continued)

CORE OPERATING LIMITS REPORT (COLR)DOSE EQUIVALENT 1-131 Cooper 1.1-2 Amendment Definitions 1.1 1.1 Definitions DOSE EQUIVALENT 1-131 (continued) 1-133, 1-134, and 1-135 actually present. The DOSE EQUIVALENT 1-131 concentration is calculated as follows: DOSE EQUIVALENT 1-131 = (1-131) + 0.0060 (1-132) + 0.17 (1-133) + 0.0010 (1-134) + 0.029 (1-135). The dose conversion factors used for this calculation are those listed in Federal Guidance Report (FGR) 11, "Limiting Values of Radionuclide Intake and Air Concentration and Dose Conversion Factors for Inhalation, Submersion, and Ingestion," 1989.LEAKAGE LEAKAGE shall be: a. Identified LEAKAGE 1. LEAKAGE into the drywell, such as that from pump seals or valve packing, that is captured and conducted to a sump or collecting tank; or 2. LEAKAGE into the drywell atmosphere from sources that are both specifically located and known either not to interfere with the operation of leakage detection systems or not to be pressure boundary LEAKAGE;b. Unidentified LEAKAGE All LEAKAGE into the drywell that is not identified LEAKAGE;c. Total LEAKAGE Sum of the identified and unidentified LEAKAGE;d. Pressure Boundary LEAKAGE LEAKAGE through a nonisolable fault in a Reactor Coolant System (RCS) component body, pipe wall, or vessel wall.A LOGIC SYSTEM FUNCTIONAL TEST shall be a test of all logic components required for OPERABILITY of a logic circuit, LOGIC SYSTEM FUNCTIONAL TEST (continued)

Cooper 1.1-3 Amendment SLC System 3.1.7 3.1 REACTIVITY CONTROL SYSTEMS 3.1.7 Standby Liquid Control (SLC) System LCO 3.1.7 APPLICABILITY:

Two SLC subsystems shall be OPERABLE.MODES 1, 2, and 3.ACTIONS CONDITION REQUIRED ACTION COMPLETION TIME A. One SLC subsystem A.1 Restore SLC subsystem to 7 days inoperable.

OPERABLE status.B. Two SLC subsystems B.1 Restore one SLC subsystem to 8 hours9.259259e-5 days 0.00222 hours 1.322751e-5 weeks 3.044e-6 months inoperable.

OPERABLE status.C. Required Action and C.1 Be in MODE 3. 12 hours1.388889e-4 days 0.00333 hours 1.984127e-5 weeks 4.566e-6 months associated Completion Time not met. AND C.2 Be in MODE 4. 36 hours4.166667e-4 days 0.01 hours 5.952381e-5 weeks 1.3698e-5 months Cooper 3.1-20 Amendment No.

PCIVs 3.6.1.3 SURVEILLANCE REQUIREMENTS (continued)

SURVEILLANCE FREQUENCY SR 3.6.1.3.6 Verify the isolation time of each MSIV is In accordance

> 3 seconds and < 5 seconds. with the Inservice Testing Program SR 3.6.1.3.7 Verify each automatic PCIV actuates to the 18 months isolation position on an actual or simulated isolation signal.SR 3.6.1.3.8 Verify a representative sample of reactor 18 months instrumentation line EFCVs actuate to the isolation position on an actual or simulated instrument line break.SR 3.6.1.3.9 Remove and test the explosive squib from each 18 months on a shear isolation valve of the TIP System. STAGGERED TEST BASIS SR 3.6.1.3.10 Verify leakage rate through each Main Steam line In accordance is < 106 scfh when tested at > 29 psig. with the Primary Containment Leakage Rate Testing Program (continued)

Cooper 3.6-14 Amendment No.

PCIVs 3.6.1.3 SURVEILLANCE REQUIREMENTS (continued)

SURVEILLANCE FREQUENCY SR 3.6.1.3.11 Verify each inboard 24 inch primary containment 18 months purge and vent valve is blocked to restrict the maximum valve opening angle to 600.SR 3.6.1.3.12 Verify leakage rate through the Main Steam In accordance Pathway is < 212 scfh when tested at > 29 psig. with the Primary Containment Leakage Rate Testing Program Cooper 3.6-15 Amendment No.

Programs and Manuals 5.5 5.5 Programs and Manuals 5.5.11 Safety Function Determination Program (SFDP) (continued)

For the purpose of this program, a loss of safety function may exist when a support system is inoperable, and: 1. A required system redundant to system(s) supported by the inoperable support system is also inoperable; or 2. A required system redundant to system(s) in turn supported by the inoperable supported system is also inoperable; or 3. A required system redundant to support system(s) for the supported systems b.1 and b.2 above is also inoperable.

The SFDP identifies where a loss of safety function exists. If a loss of safety function is determined to exist by this program, the appropriate Conditions and Required Actions of the LCO in which the loss of safety function exists are required to be entered.5.5.12 Primary Containment Leakage Rate Testing Program a. A program shall establish the leakage rate testing of the containment as required by 10 CFR 50.54(o) and 10 CFR 50, Appendix J, Option B, as modified by approved exemptions.

This program shall be in accordance with the guidelines contained in Regulatory Guide 1.163, "Performance-Based Containment Leak-Test Program," dated September, 1995, as modified by the following exceptions:

1. Exemption from Appendix J to 10CFR Part 50 to allow reverse direction local leak rate testing of four containment isolation valves at Cooper Nuclear Station (TAC NO. M89769) (July 22, 1994).2. Exemption from Appendix J to 10CFR Part 50 to allow MSIV testing at 29 psig and expansion bellows testing at 5 psig between the plies (Sept. 16, 1977).3. Exception to NEI 94-01, "Industry Guideline for Implementing Performance-Based Option of 10 CFR Part 50, Appendix J," Section 9.2.3: The first Type A test performed after the December 7, 1998 Type A test shall be performed no later than December 7, 2013.4. Exemption from Section III.A of 10CFR Part 50, Appendix J, Option B, to allow the leakage contribution from Main Steam Pathway (Main Steam lines and Main Steam inboard drain line) leakage to be excluded from the overall integrated leakage rate from Type A tests ((continued)

Cooper 5.0-16 Amendment No.

Programs and Manuals 5.5 5.5 Programs and Manuals 5.5.12 Primary Containment Leakage Rate Testing Program (continued)

5. Exemption from Section III.B of 10CFR Part 50, Appendix J, Option B, to allow the contribution from Main Steam Pathway (Main Steam lines and Main Steam inboard drain line) leakage to be excluded from the sum of the leakage rates from Type B and Type C tests (b. The peak calculated containment internal pressure for the design basis loss of coolant accident, Pa, is 58.0 psig. The containment design pressure is 56.0 psig.c. The maximum allowable containment leakage rate, La, at Pa, shall be 0.635% of containment air weight per day.d. Leakage Rate acceptance criteria are: 1. Containment leakage rate acceptance criterion is _ 1.0 La. During the first unit startup following testing in accordance with this program, the leakage rate acceptance criteria are, <0.60 La for the Type B and C tests and 0.75 La for Type A tests.2. Air lock testing acceptance criteria are: a. Overall air lock leakage rate is 12 scfh when tested at >_ P_b. Overall air lock leakage rate is 0.23 scfh when tested at>_ 3.0 psig.e. The provisions of SR 3.0.2 do not apply to the test frequencies specified in the Primary Containment Leakage Rate Testing Program.f. The provisions of SR 3.0.3 are applicable to the Primary Containment Leakage Rate Testing Program.5.5.13 Control Room Envelope Habitability ProQram A Control Room Envelope (CRE) Habitability Program shall be established and implemented to ensure that CRE habitability is maintained such that, with an OPERABLE Control Room Emergency Filter (CREF) System, CRE occupants can control the reactor safely under normal conditions and maintain it in a safe condition following a radiological event, hazardous chemical release, or a smoke challenge.

The program shall ensure that adequate radiation protection is provided to permit access and occupancy of the CRE under design basis accident (DBA) conditions without (continued)

Cooper 5.0-17 Amendment No.

NLS2008014 Attachment 6 Page 1 of 26 ATTACHMENT 6 Proposed Technical Specifications Bases Revisions Markup Format Cooper Nuclear Station, Docket 50-298, DPR-46 Listing of Revised Pages TS Bases Pages B 2.0-5 B 2.0-6 B 2.0-8 B 3.1-39 B 3.1-40 B 3.1-41 B 3.1-45 B 3.1-46 B 3.1-47 B 3.1-51 B 3.3-144 B 3.3-148 B 3.3-151 B 3.3-185 B 3.4-29 B 3.4-32 B 3.6-28 B 3.6-29 Reactor Core SLs B 2.1.1 BASES SAFETY LIMIT Exceeding an SL may cause fuel damage and create a potential VIOLATIONS for radioactive releases in excess of 10 CFR 100, "Reactor Site-Criteria," limits (Ref. 31. Therefore, it is required to insert all insertable control rods and res ore compliance with the SLs within 2 hours2.314815e-5 days 5.555556e-4 hours 3.306878e-6 weeks 7.61e-7 months . The 2 hour2.314815e-5 days 5.555556e-4 hours 3.306878e-6 weeks 7.61e-7 months Completion Time ensures that the operators take prompt remedial action and also ensures that the probability of an accident occurring during this period is minimal.REFERENCES

1. USAR, Appendix F, Section F-2.2.1.2. NEDE-24011-P-A, "General Electric Standard Application for Reactor Fuel," (Revision specified in the COLR).3. 10 CFR 100.(4. OCf OC)Cooper B 2.0-5 Insert B 2.1.1 -I for the Control Rod Drop and Main Steam Line Break accidents, and 10 CFR 50.67,"Accident Source Term," limits (Ref. 4) for the Fuel Handling and Loss-of-Coolant accidents.

RCS Pressure SL B 2.1.2 B 2.0 SAFETY LIMITS (SLs)B 2.1.2 Reactor Coolant System (RCS) Pressure SL BASES BACKGROUND The SL on reactor steam dome pressure protects the RCS against overpressurization.

In the event of fuel cladding failure, fission products are released into the reactor coolant. The RCS then serves as the primary barrier in preventing the release of fission products into the atmosphere.

Establishing an upper limit on reactor steam dome pressure ensures continued RCS integrity.

According to the USAR, Appendix F (Ref. 1), the reactor coolant pressure boundary (RCPB)shall be capable of accommodating without rupture, and with only limited allowance for energy absorption through plastic deformation, the static and dynamic loads imposed on any boundary component as a result of any inadvertent and sudden release of energy to the coolant.As a design reference, this sudden release shall be taken as that which would result from a sudden reactivity insertion such as rod ejection (unless prevented by positive mechanical means), rod dropout, or cold water addition.During normal operation and abnormal operational transients, RCS pressure is limited from exceeding the design pressure by more than 10%, in accordance with Section III of the ASME Code (Ref. 2). To ensure system integrity, all RCS components are hydrostatically tested at 125% of design pressure, in accordance with ASME Code requirements, prior to initial operation when there is no fuel in the core.Any further hydrostatic testing with fuel in the core may be done under LCO 3.10.1, "Inservice Leak and Hydrostatic Testing Operation." Following inception of unit operation, RCS components shall be pressure tested in accordance with the requirements of ASME Code,Section XI (Ref. 3).Overpressurization of the RCS could result in a breach of the RCPB, reducing the number of protective barriers designed to prevent radioactive releases from exceeding the limits specified in 1-e-R--,",-Rcactcr Sitc Critor3" (Ref. 4). If this occurred in conjunction with a fuel cladding failure, fission products could enter the containment uatmosphere.(continued)

Cooper B 2.0-6 Cooper B 2.0-6 2k1~81O~~

RCS Pressure SL B 2.1.2 BASES (continued)

SAFETY LIMIT Exceeding the RCS pressure SL may cause immediate RCS VIOLATIONS failure and create a potential for radioactive releases in excess of 10 ~ ~ ~ rt Cm 1, R SLc;cilimits (Ref. 4). Therefore, it is C F "z 5-0 required to insert all insertable control rods and restore compliance with-,the SL within 2 hours2.314815e-5 days 5.555556e-4 hours 3.306878e-6 weeks 7.61e-7 months . The 2 hour2.314815e-5 days 5.555556e-4 hours 3.306878e-6 weeks 7.61e-7 months Completion Time ensures that the'C ,ct .(L- Coo L'6e, operators take prompt remedial action and also assures that the iC e r ) probability of an accident occurring during this period is minimal.REFERENCES

1. USAR, Appendix F, Section F-2.6.1.2. ASME, Boiler and Pressure Vessel Code,Section III, Article NB-7000.3. ASME, Boiler and Pressure Vessel Code,Section XI, Article IW-5000.4. 5. ASME, Boiler and Pressure Vessel Code,Section III, 1965 Edition, Addenda winter of 1966.6. ASME, USAS, Nuclear Power Piping Code, Section B31.1, 1967 Edition.7. ASME, Boiler and Pressure Vessel Code,Section III, 1983 Edition.Cooper B 2.0-8 SLC System B 3.1.7 B 3.1 REACTIVITY CONTROL SYSTEMS B 3.1.7 Standby Liquid Control (SLC) System BASES BACKGROUND The SLC System is designed to provide the capability of bringing the reactor, at any time in a fuel cycle, from full power and minimum control rod inventory (which is at the peak of the xenon transient) to a subcritical condition with the reactor in the most reactive, xenon free state without taking credit for control rod movement.

The SLC System satisfies the requirements of 10 CFR 50.62 (Ref. 1) on anticipated transient without scram.SLC System consists of a boron solution storage tank, two positive displacement pumps, two explosive valves that are provided in parallel for redundancy, and associated piping and valves used to transfer borated water from the storage tank to the reactor pressure vessel (RPV). The borated solution is discharged near the bottom of the core shroud, where it then mixes with the cooling water rising through the core. A smaller tank containing demineralized water is provided for testing purposes.APPLICABLE SAFETY ANALYSES The SLC System is manually initiated from the main control room, as directed by the emergency operating procedures, if the operator believes the reactor cannot be shut down, or kept shut down, with the control rods. The SLC System is used in the event that enough control rods cannot be inserted to accomplish shutdown and cooldown in the normal manner. The SLC System injects borated water into the reactor core to add negative reactivity to:compensate for all of the various reactivity effects that could occur during plant operations.

To meet this objective, it is necessary to inject, using both SLC pumps, a quantity of boron, which produces a concentration of 660 ppm of natural boron, in the reactor coolant at 68°F. To allow for potential leakage and imperfect mixing in the reactor system, an amount of boron equal to 25% of the amount cited above is added (Ref. 2). The volume versus concentration limits in Figure 3.1.7-1 and the temperature versus concentration limits in Figure 3.1.7-2 are calculated such that the required concentration is achieved accounting for dilution in the RPV with normal water level and including (continued)

Cooper B 3.1-39 Insert B 3.1.7-1 The SLC System is also used to maintain the pH of the water in the Suppression Chamber at or above 7.0 following a design basis Loss-of-Coolant Accident (LOCA) with indication of fuel damage. Maintaining the pH of the water above 7.0 following a LOCA ensures that iodine will be retained in the Suppression Chamber water as assumed in the LOCA Alternative Source Term analysis (Ref. 4). The SLC System is manually initiated from the main control room upon detection of symptoms that a LOCA with fuel damage is occurring.

SLC System B 3.1.7 BASES APPLICABLE the water volume in the residual heat removal shutdown SAFETY ANALYSES cooling piping and in the recirculation loop piping. This (continued) quantity of borated solution is the amount that is above the pump suction shutoff level in the boron solution storage tank. No credit is taken for the portion of the tank volume that cannot be injected.The SLC System satisfies Criterion 4 of /10 CFR 50.36 (c)(2)(ii) (Ref. 3).LCO The OPERABILITY of the SLC System provides backup capability for reactivity control independent of .normal reactivity control provisions provided by the control rods. T OPERABILITY of the SLC System is based on the conditions of the borated solution in the storage tank and the availability of a flow path to the RPV, including the OPERABILITY of the pumps and valves. Two SLC subsystems are required to be OPERABLE; each contains an OPERABLE pump, an explosive valve, and associated piping, valves, and instruments and controls to ensure an OPERABLE flow path.APPLICABILITY In MODES 1 and 2, shutdown capability is required.

In MODES 3 and 4, control rods are not able to be withdrawn since the reactor mode switch is in shutdown and a control a -r i-, rod block is applied. This provides adequate controls to-ensure that the reactor remains subcritical.

In MODE 5, only a single control rod can be withdrawn from a core cell containing fuel assemblies.

Demonstration of adequate SDM (LCO 3.1.1, "SHUTDOWN MARGIN (SDM)") ensures that the reactor will not become critical.

Therefore,.-the SLC System is not required to be OPERABLE when-only.:a:.single control rod can be withdrawn.

ACTIONS A.I If one SLC subsystem is inoperable, the inoperable subsystem must be restored to OPERABLE status within 7 days. In this condition, the remaining OPERABLE subsystem is adequate to perform the original licensing basis shutdown function.However, the overall capability is reduced since the (continued)

Cooper B 3.1-40 Insert B 3.1.7-2 The Alternative Source Term LOCA analysis methodology (Ref. 4) credits the use of the SLC System for injecting sodium pentaborate solution into the RPV following a LOCA with fuel damage to maintain the pH of the water in the Suppression Chamber above 7.0. By maintaining the pH of the water above 7.0 following a LOCA with fuel damage, the majority of the iodine released from a damaged core will be retained in solution in the water and not released as elemental iodine in gaseous form. This will ensure that the radiological consequences from the LOCA will remain within the limits of 10 CFR 50.67 (Ref. 5). Credit for the SLC System in the radiological analyses is based on operation of one SLC pump, initiated 6 hours6.944444e-5 days 0.00167 hours 9.920635e-6 weeks 2.283e-6 months after start of the LOCA, with injection completed within 8 hours9.259259e-5 days 0.00222 hours 1.322751e-5 weeks 3.044e-6 months . (Ref. 2)Insert B 3.1.7-3 Additionally, an OPERABLE SLC System allows the injection of the sodium pentaborate solution into the RPV following a LOCA with core damage in order to maintain the pH of the water in the Suppression Chamber at or above 7.0.Insert B 3.1.7-4 The SLC System must be OPERABLE in Modes 1, 2, and 3.Insert B 3.1.7-5 In MODES 1,2, and 3, the SLC System must be OPERABLE in order to inject the boron solution into the RPV following a LOCA with fuel damage for purposes of chemistry control of the water, i.e., to maintain the pH of the combined RCS and ECCS water inventory at or above 7.0. Maintaining the pH of this water at or-above 7.0 ensures that the radiological consequences of a LOCA remain within the limits of 10 CFR 50.67 (Ref. 5).

SLC System B 3.1.7 BASES ACTIONS A.1 (continued) remaining OPERABLE subsystem cannot meet the requirements of Reference

1. The 7 day Completion Time is based on the availability of an OPERABLE subsystem capable of performing the original licensing basis SLC System function and the low probability of a Design Basis Accident (DBA) or severe transient occurring concurrent with the failure of the Control Rod Drive (CRD) System to shut down the plant.B.1 If both. SLC subsystems are inoperable, at least one subsystem must be restored to OPERABLE status within 8 hours9.259259e-5 days 0.00222 hours 1.322751e-5 weeks 3.044e-6 months . The allowed Completion Time of 8 hours9.259259e-5 days 0.00222 hours 1.322751e-5 weeks 3.044e-6 months is considered acceptable given the low probability of a DBA or transient occurring concurrent with the failure of the control rods to shut down the reactor.C.1 \ y o _5 If any Required Action and associate Completion Time is not met, the plant must be brought to a ODE in which the LCO does not apply. To achieve this s atus, the plant must be brought to MODE 3 within 12 hours1.388889e-4 days 0.00333 hours 1.984127e-5 weeks 4.566e-6 months . The allowed Completion STimeof-12 h~o,, i,'rrasonable, based on operating experience, to reach MOOgE-a-, from full power conditions in an orderly manner and witou challenging plant systems.SURVEILLANCE SR 3.1.7.1, SR. 3.1.7.2, and SR 3.1.7.3 REQUIREMENTS SR 3.1.7.1 through SR 3.1.7.3 are 24 hour2.777778e-4 days 0.00667 hours 3.968254e-5 weeks 9.132e-6 months Surveillances verifying certain characteristics of the SLC System (e.g., the volume and temperature of the borated solution in the storage tank), thereby ensuring SLC System OPERABILITY without disturbing normal plant operation.

These Surveillances ensure that the proper borated solution volume and temperature, including the temperature of the pump suction piping, are maintained.

Maintaining a minimum specified borated solution temperature is important in ensuring that the boron remains in solution and does not precipitate out in the storage tank or in the pump suction (continued)

Cooper B 3.1-41

_ SLC System B 3.1.7 BASES (continued)

REFERENCES

1. 10 CFR 50.62.2. USAR, Section 111-9.3. 10 CFR 50.36(c)(2)(ii).

koac-4 -ixJiec e &?\CO#v-S IerorucIeor X'3S I, rUaC- AC Cooper B 3.1-45 Cooperjle B6 3.-5Rv-i SDV Vent and Drain Valves B 3.1.8 B 3.1 REACTIVITY CONTROL SYSTEMS B 3.1.8 Scram Discharge Volume (SDV) Vent and Drain Valves BASES BACKGROUND The SDV vent and drain valves are normally open and discharge any accumulated water in the SDV to ensure that sufficient volume is available at all times to allow a complete scram. During a scram, the SDV vent and drain valves close to contain reactor water. The SDV is a volume of header piping.that connects to each hydraulic control unit (HCU) and drains into an instrument volume. There are two SDVs (headers) -and twoý:jnstrument

,volumes, each receiving approximately-one half of the control rod drive (CRD) discharges.

The two instrument volumes are connected to separate drain lines with two valves in series for a total of four drain valves. Each header is connected to a separate vent line with two valves in series for a total of four vent valves. The automatic SDV vent and drain valves are air to open, spring closing valves which fail closed on loss of air. The SDV drain valves are equipped with a manual override which can only be used to open the valve.The SDV vent valves do not have this feature. The header piping is sized to receive and contain all the water discharged by the CRDs during a scram. The design and functions of the SDV are described in Reference 1.APPLICABLE SAFETY ANALYSES The Design Basis Accident and transient analyses assume all of the control rods are capable of scramming.

The acceptance criteria for the SDV vent and drain valves are that they operate automatically to: a. Close during*scram to .limit the amount:of reactor coolant discharged so that adequate core cooling is maintained and offsite doses remain within the limits of 10 CFR 100 (Ref. 2ý; and b. Open on scram reset to maintain the SDV vent and drain path open so that there is sufficient volume to accept the reactor coolant discharged during a scram.Isolation of the SDV can also be accomplished by manual closure of the SDV valves. Additionally, the discharge (continued)

Cooper B 3.1-46 Insert B 3.1.8-1 for the Control Rod Drop and Main Steam Line Break accidents, and 10 CFR 50.67,"Accident Source Term,"' (Ref. 5) for the Fuel Handling and Loss-of-Coolant accidents.

SDV Vent and Drain Valves B 3.1.8 BASES APPLICABLE SAFETY ANALYSIS (continued) of reactor coolant to the SDV can be terminated by scram reset or closure of the HCU manual isolation valves. For a bounding leakage case, the offsite doses are yw-l-within the-imits of 10 CFR 100 (Ref. 2)7 and adequate core cooling is maintained (Ref. 3). The SDV vent and drain valves allow continuous drainage of the SDV during normal plant operation to ensure that the SDV has sufficient capacity to contain the reactor coolant discharge during a full core scram. To automatically ensure this capacity, a reactor scram (LCO 3.3.1.1, "Reactor Protection System (RPS)Instrumentation")

is initiated if the SDV water level in the instrument volume exceeds a specified setpoint.

The setpoint is chosen-so that all control-rods are inserted before the SDV has insufficientivolume to accept a full scram.SDV vent and drain valves satisfy Criterion 3 of 10 CFR 50.36 (c)(2)(ii) (Ref. 4).LCO The OPERABILITY of all SDV vent and drain valves ensures that the SDV vent and drain valves will close during a scram to contain reactor water discharged to the SDV piping.Since the vent and drain lines are provided with two valves in series, the single failure of one valve in the open position will not impair the isolation function of the system. Additionally, thevalves are required to open on scram reset to ensure that a path is available for the SDV piping to drain freely at other times.APPLICABILITY In MODES 1 and 2,scram may be-required;.therefore, the SDV vent and drain valves must be OPERABLE.

In MODES 3 and 4, control rods are not able to be withdrawn since the reactor mode switch is in shutdown and a control rod block is applied. Also, during MODE 5, only a single control rod can be withdrawn from a core cell containing fuel assemblies.

Therefore, the SDV vent and drain valves are not required to be OPERABLE in these MODES since the reactor is subcritical and only one rod may be withdrawn and subject to scram.(continued)

Cooper B 3.1-47 CpB 1tine Insert B 3.1.8-2 for the Control Rod Drop and Main Steam Line Break accidents, and 10 CFR 50.67,"Accident Source Term,"' (Ref. 5) for the Fuel Handling and Loss-of-Coolant accidents.

SDV Vent and Drain Valves B 3.1.8 BASES REFERENCES

1. USAR, Section 111-5.2. 10 CFR 100.3. NUREG-0803, "Generic Safety Evaluation Report Regarding Integrity of BWR Scram System Piping," August 1981.4. 10 CFR 50.36(c)(2)(ii).

5. -' -7, A"',c-8A er.er-L Scur)e.."terr M.B 3.1-51 June Cooper Primary Containment Isolation Instrumentation B 3.3.6.1 BASES APPLICABLE In general, the individual Functions are required to be SAFETY ANALYSES OPERABLE in MODES 1, 2, and 3 consistent with the LCO, and Applicability for LCO 3.6.1.1, "Primary Containment." APPLICABILITY Functions that have different Applicabilities are discussed (continued) below in the individual Functions discussion.

The specific Applicable Safety Analyses, LCO, and Applicability discussions are listed below on a Function by Function basis.Main Steam Line Isolation l.a. Reactor Vessel Water Level -Low Low Low (Level 1)Low reactor pressure vessel (RPV) water level indicates that the capability to cool the fuel may be threatened.

Should RPV water level decrease too far, fuel damage could result.Therefore, isolation of the MSIVs and other interfaces with the reactor vessel occurs to prevent offsite dose limits from being exceeded.

The Reactor Vessel Water Level--Low Low Low (Level 1) Function is one of the many Functions assumed to be OPERABLE and capable of providing isolation signals.The Reactor Vessel Water Level-Low Low Low (Level 1)Function associated with isolation is assumed in the analysis of the recirculation line break (Ref. 4). The isolation of the MSLs on Level I supports actions to ensure that offsite dose limits are not exceeded for a DBA.Reactor vessel water level signals are initiated from four level transmitters that sense the difference between the pressure due to a constant column of water (reference leg)and the pressure due to the actual water level (variable leg) in the vessel. Four channels of Reactor Vessel Water Level-Low Low Low (Level 1) Function are available and are required to be OPERABLE to ensure that no single instrument failure can preclude the isolation function.The Reactor Vessel Water Level--Low Low Low (Level 1)Allowable Value is chosen to be the same as the ECCS Level 1 Allowable Value (LCO 3.3.5.1) to ensure that the MSLs isolate on a potential loss of coolant accident (LOCA) to prevent offsite doses from exceeding 10 .OR C.EJ-limits.

This Function isolates the MSIVs and MSL drains.(continued)

Cooper B 3.3-144 Primary Containment Isolation Instrumentation B 3.3.6.1 BASES APPLICABLE SAFETY ANALYSES, LCO, and APPLICABILITY (continued)

Primary Containment Isolation 2.a. Reactor Vessel Water Level -Low (Level 3)Low RPV water level indicates that the capability to cool the fuel may be threatened.

The valves whose penetrations communicate with the primary containment are isolated to limit the release of fission products.The isolation of the primary containment on Level 3 supports actions to ensure that offsite dose limits of .-F4R-.-8e are not exceeded.

The SReactorVesselWaterLevelLow(Level 3)Function associated with isolation is implicitly assumed in the USAR analysis as these leakage paths are assumed to be isolated post LOCA.Reactor Vessel Water Level -Low (Level 3) signals are initiated from four vessel level instrument switches that sense the difference between the pressure due to a constant column of water (reference leg) and the pressure due to the actual water level (variable leg) in the vessel. Four channels of Reactor Vessel Water Level -Low (Level 3) Function are available and are required to be OPERABLE to ensure that no single instrument failure can preclude the isolation function.The Reactor Vessel Water Level -Low (Level 3) Allowable Value was chosen to be the same as the RPS Level 3 scram Allowable Value (LCO 3.3.1.1), since isolation of these valves is not critical to orderly plant shutdown.This Function isolates the Group 2, 3, and 6 valves listed in Reference 1.2.b. Drywell Pressure -High High drywell pressure can indicate a break in the RCPB inside the primary containment.

The isolation of some of the primary containment isolation valves on high drywell pressure supports actions to ensure that offsite dose limits of 1-Q--C-R-+-

are not exceeded.

The Drywell ,c- 1 ressure -High Function, associated with isolation of the primary containment, is implicitly assumed in the USAR accident analysis as these leakage paths are assumed to be isolated post LOCA.Cooper B 3.3-148 CpB32L42*2-,-

Primary Containment Isolation Instrumentation B 3.3.6.1 BASES APPLICABLE SAFETY ANALYSES, LCO, and APPLICABILITY (continued) from the reactor vessel occurs as part of the isolation of Primary Containment to prevent offsite dose limits from being exceeded.

The Reactor Vessel Water Level -Low Low Low (Level 1) Function is one of the many Functions assumed to be OPERABLE and capable of providing isolation signals. This Function, associated with Primary Containment isolation, is assumed in the analysis of the recirculation line break (Ref. 4). The isolation of the valves of the recirculation sample line on Level I supports actions to ensure that offsite dose limits are not exceeded for a DBA.Reactor vessel water level signals are initiated from four level transmitters that sense the difference between the pressure due to a constant column of water (reference leg) and the pressure due to the actual water level (variable leg) in the vessel. Four channels of Reactor Vessel Water Level -Low Low Low (Level 1) Function are available and are required to be OPERABLE to ensure that no single instrument failure can preclude the isolation function.The Reactor Vessel Water Level -Low Low Low (Level 1) Allowable Value is chosen to be the same as the ECCS Level 1 Allowable Value (LCO 3.3.5.1) to ensure that the recirculation sample valves will isolate on a potential LOCA to prevent offsite doses from exceeding 10 CFR 4e--limits.This Function isolates the recirculation sample valves. It may be bypassed using a key-locked switch during accident conditions to obtain a sample for Post Accident Sampling System (PASS).High Pressure Coolant Iniection and Reactor Core Isolation Cooling Systems Isolation 3.a., 3.b., 4.a., 4.b. HPCI and RCIC Steam Line Flow -High and Time Delay Relays Steam Line Flow -High Functions are provided to detect a break of the RCIC or HPCI steam lines and initiate closure of the steam line isolation valves of the appropriate system. If the steam is allowed to continue flowing out of the break, the reactor will depressurize and the core can uncover. Therefore, the isolations are initiated on high flow to prevent or minimize core damage. The isolation action, along with the scram function of the RPS, ensures that the fuel peak cladding temperature remains below the limits of 10 CFR 50.46. Specific credit for these Functions is not assumed in any USAR accident analyses since the bounding analysis is performed for large breaks such as recirculation and Cooper B 3.3-151 CREF System Instrumentation B 3.3.7.1 B 3.3 INSTRUMENTATION B 3.3.7.1 Control Room Emergency Filter (CREF) System Instrumentation BASES BACKGROUND The CREF System is designed to provide a radiologically controlled environment to ensure the habitability of the control room for the safety of control room operators under all plant conditions.

The instrumentation and controls for the CREF System automatically isolate the normal ventilation intake and initiate action to pressurize the main control room and filter incoming air to minimize the infiltration of radioactive material into the control room environment.

In the event of a loss of coolant accident (LOCA) signal (Reactor Vessel Water Level -Low Low, Level 2 or Drywell Pressure -High) or Reactor Building Ventilation Exhaust Plenum Radiation

-High signal, the normal control room inlet supply damper closes and the CREF System is automatically started in the emergency bypass mode. The air drawn in from the outside passes through a high efficiency filter and a charcoal filter in sufficient volume to maintain the control room slightly pressurized with respect to the adjacent areas.The CREF System instrumentation has two trip systems. Each trip system includes the sensors, relays, and switches necessary to cause initiation of the CREF System. Each trip system receives input from each of the Functions listed above (each sensor sends a signal to both trip systems).

The Reactor Vessel Water Level -Low Low, Level 2, Drywell Pressure -High, and Reactor Building Ventilation Exhaust Plenum Radiation

-High are each arranged in a one-out-of-two taken twice logic for each trip system. The channels include electronic and electrical equipment (e.g., switches and trip relays) that compares measured input signals with pre-established setpoints.

When the setpoint is exceeded, the channel output relay actuates, which then outputs a CREF System initiation signal to the initiation logic.APPLICABLE SAFETY ANALYSES, LCO, and APPLICABILITY The ability of the CREF System to maintain the habitability of the control room is explicitly assumed for certain accidents as discussed in the USAR safety analyses (Refs. 1, 2, and 3). CREF System operation ensures that the radiation exposure of control room personnel, through the duration of any one of the postulated accidents that assume CREF System operation, does not exceed the limits set by GDC 19 of 10 CFR 50, Appendix A or 10 CFR 50.67 (Fuel Handling Accident only).Cooper B 3.3-185 RCS Specific Activity B 3.4.6 B 3.4 REACTOR COOLANT SYSTEM (RCS)B 3.4.6 RCS Specific Activity BASES BACKGROUND During circulation, the reactor coolant acquires radioactive materials due to release of fission products from fuel leaks into the reactor coolant and activation of corrosion products in the reactor coolant. These radioactive materials in the reactor coolant can plate out in the RCS, and, at times, an accumulation will break away to spike the normal level of radioactivity.

The release of coolant during a Design Basis Accident (DBA) could send radioactive materials into the environment.

Limits on the maximum allowable level of radioactivity in the reactor coolant are established to ensure that in the event of a release of any radioactive material to the environment during a DBA; radiation doses are maintained within the limits of 10 CFR 100 (Ref. 1). I: This LCO contains iodine specific activity limits. The iodine isotopic activities per gram of reactor coolant are expressed in terms of a DOSE EQUIVALENT 1-131. The allowable levels are intended to limit the 2 hour2.314815e-5 days 5.555556e-4 hours 3.306878e-6 weeks 7.61e-7 months radiation dose to an individual at the site boundary to a small fraction of the 10 CFR 100 limit.APPLICABLE SAFETY ANALYSES Analytical methods and assumptions involving radioactive material in the primary coolant are presented in the USAR (Ref. 2). The specific activity in the reactor coolant (the source term) is an initial condition for evaluation of the consequences of an accident due to a main steam line break (MSLB) outside containment.

No fuel damage is postulated in the MSLB accident, and the release of radioactive material to the environment is assumed to end when the main steam isolation valves (MSIVs) close completely.

This MSLB release forms the basis for determining offsite and control room.doses (Refs. 2 and 3). The limits on the specific activity of the primary coolant ensure that the 2 hour2.314815e-5 days 5.555556e-4 hours 3.306878e-6 weeks 7.61e-7 months thyroid and whole body doses at the site boundary, (continued)

Cooper B 3.4-29 Insert B 3.4.6-1 for the Control Rod Drop and Main Steam Line Break accidents, and 10 CFR 50.67,"Accident Source Term," (Ref. 6) for the Fuel Handling and Loss-of-Coolant accidents.

Insert B 3.4.6-2 for the Control Rod Drop and Main Steam Line Break accidents, and within the 10 CFR 50.67, "Accident Source Term," (Ref. 6) limit for the Fuel Handling and Loss-of-Coolant accidents.

RCS Specific Activity B 3.4.6 BASES ACTIONS B.1, B.2.1, B.2.2.1, and B.2.2.2 (continued)

The Completion Time of once every 4 hours4.62963e-5 days 0.00111 hours 6.613757e-6 weeks 1.522e-6 months is the time needed to take and analyze a sample. The 12 hour1.388889e-4 days 0.00333 hours 1.984127e-5 weeks 4.566e-6 months Completion Time is reasonable, based on operating experience, to isolate the main steam lines in an orderly manner and without challenging plant systems. Also, the allowed Completion Times for Required Actions B.2.2.1 and B.2.2.2 for placing the unit in MODES 3 and 4 are reasonable, based on operating experience, to achieve the required plant conditions from full power conditions in an orderly manner and without challenging plant systems.SURVEILLANCE REQUIREMENTS SR 3.4.6.1 This Surveillance is performed to ensure iodine remains within limit during normal operation.

The 7 day Frequency is adequate to trend changes in the iodine activity level.This SR is modified by a Note that requires this Surveillance to be performed only in MODE 1 because the level of fission products generated in other MODES is much less.REFERENCES

1. 10 CFR 100.11, 1973.2. USAR, Section XIV-8.1.3. USAR, Section XIV-6.5.4. 10 CFR 50, Appendix A, GDC 19.5. 10 CFR 50.36(c)(2)(ii).

10 1C9 Cooper B 3.4-32'IV-Gwý PCIVs B 3.6.1.3 BASES SURVEILLANCE REQUIREMENTS (continued)

SR 3.6.1.3.9 The TIP shear isolation valves are actuated by explosive charges. An in place functional test is not possible with this design. The explosive squib is removed and tested to provide assurance that the valves will actuate when required.

The replacement charge for the explosive squib shall be from the same manufactured batch as the one fired or from another batch that has been certified by having one of the batch successfully fired. The Frequency of 18 months on a STAGGERED TEST BASIS is considered adequate given the administrative controls on replacement charges and the frequent checks of circuit continuity (SR 3.6.1.3.4).

SR 3.6.1.3.10 The analyses in References 8 nd 9 are based on leakage that is less than the specified leakage rate. he co ined main,,steam line le-kaage ra e mustbe-< scfh w ested at _Pt (29 psigf The Frequency is.feq&FisdyPte Primary Containment Lea age Rate Testing Program.SR 3.6.1.3.11 e Verifying each inboard 24 inch primary containment purge and vent valve (PC-230 MV, PC-231 MV, PC-232 MV, and P0-233 MV) is blocked to restrict the maximum opening angle to 600 is required to ensure that the valves can close under DBA conditions within the times assumed in the analysis of References 7 and 8. If a LOCA occurs, the purge and vent valves must close to maintain containment leakage within the values assumed in the accident analysis.

At other times, pressurization concerns are not present, thus the purge valves can be fully open. The 18 month Frequency is appropriate because the blocking devices may be removed during a refueling outage.Cooper B 3.6-28 Insert B 3.6.1.3 -I A leakage rate of 150 scfh per Main Steam line at. > Pa (58 psig) was assumed in the LOCA analyses.

The equivalent leakage rate at _> Pt (29 psig) is 106 scfh. An "MSIV line" is each one of the four Main Steam lines with an inboard and an outboard Main Steam Isolation Valve (MSIV). The leakage rate to be measured is the Main Steam line"minimum path" leakage (the lesser actual pathway leakage of the two MSIVs in the Main Steam line). The leakage limit is based on the analyses of References 11 and 12.Insert B 3.6.1.3 -2 SR 3.6.1.3.12 The Main Steam Pathway is the analyzed leakage path from the four Main Steam lines and the inboard Main Steam drain line to and including the condenser.

The leakage limit imposed on the Main Steam Pathway with this surveillance requirement applies to the total (aggregate) leakage for the Main Steam Pathway. The Main Steam Pathway leakage includes the total leakage of all four Main Steam line penetrations plus the inboard Main Steam drain line penetration (penetration X-8, with Containment isolation valves MS-MOV-MO74 and MS-MOV-MO77).

The aggregate leakage of the Main Steam Pathway must be _< 212 scfh when tested at. _> 29 psig or the equivalent leakage rate of _< 300 scfh when tested at _> Pa (58 psig). The leakage rate to be measured is the Main Steam Pathway "minimum path" leakage (the lesser actual pathway leakage of the two Containment isolation valves in each penetration; i.e., the MSIVs in penetrations X-7A through X-7D, and MS-MOV-MO74 and MS-MOV-MO77 in penetration X-8). The leakage limit is based on the analyses of References 11 and 12. The Frequency is specified in the Primary Containment Leakage Rate Testing Program.

PCIVs B 3.6.1.3 BASES REFERENCES 1.2.3.4.5.6.7.8.9.10.USAR, Chapter XIV.Amendment 25 to the FSAR.NEDC 96-006, "Estimate of Steam Tunnel's HELB," dated March 30, 1996.USAR Section IV-4.9.10 CFR 50.36(c)(2)(ii).

USAR, Table V-2-2.USAR, Burns and Roe Drawing 4259, Sheets 1 and 1A, and Burns and Roe Drawing 4260, Sheets 2A and 2B (Incorporated by Reference).

USAR, Section V-2.0.USAR, Section XIV-6.3.10 CFR 50, Appendix J, Option B.11. NEDC 07-082, "Radiological Dose Analysis for a Loss of Coolant Accident (LOCA) at Cooper Nuclear Station." 12. NEDC 07-087, "AST Reduced Pressure MS Pathway Leakage Limits." Cooper B 3.6-29 ATTACHMENT 3 LIST OF REGULATORY COMMITMENTS@

O.ATTACHMENT 3 LIST OF REGULATORY COMMITMENTS@

Correspondence Number: NLS2008014 The following table identifies those actions committed to by Nebraska Public Power District (NPPD) in this document.

Any other actions discussed in the submittal represent intended or planned actions by NPPD. They are described for information only and are not regulatory commitments.

Please notify the Licensing Manager at Cooper Nuclear Station of any questions regarding this document or any associated regulatory commitments.

COMMITMENT COMMITTED DATE COMMITMENT NUMBER OR OUTAGE None 4 i 4 I 4 t 4 4 I PROCEDURE 0.42 REVISION 22 PAGE 18 OF 25 NLS2008014 Enclosure 2 ENCLOSURE 2 Alion Science and Technology Corporation Affidavit Required by 10 CFR 2.390 Alion Science and Technology AFFIDAVIT I, Peter K. Mast, state as follows: (1) I am Vice President, Innovative Technology Solutions Operations, Alion Science and Technology Corporation (ALION) and have the responsibility for reviewing the information described in paragraph (2) that is sought to be withheld, and have been authorized to apply for its withholding.

(2) The information sought to be withheld is contained in the proprietary calculation ALION-CAL-NPPD-3236-002, Revision 1.(3) In making this application for withholding of proprietary information of which it is the owner, ALION relies upon the exemption from disclosure set forth in the Freedom of Information Act ("FOIA"), 5 USC Sec. 552(b)(4), and the Trade Secrets Act, 18 USC Sec.1905, and NRC regulations 10 CFR 9.17(a)(4), and 2.390(a)(4) for "trade secrets" (Exemption 4). The materials for which exemption from disclosure is here sought is all"confidential commercial information", and some portions also qualify under the narrower definition of "trade secret", within the meanings assigned to those terms for purposes of FOIA Exemption 4 in, respectively, Critical Mass Energy Project v. Nuclear Regulatory Commission, 975F2d871 (DC Cir. 1992), and Public Citizen Health Research Group v.FDA, 704F2d1280 (DC Cir. 1983).(4) Some examples of categories of information that fit into the definition of proprietary information are: a. Information that discloses a process, method, or apparatus, including supporting data and analyses, where prevention of its use by ALION's competitors without license from ALION constitutes a competitive economic advantage over other companies;

b. Information that, if used by a competitor, would reduce his expenditure of resources or improve his competitive position in the design, preparation, manufacture, shipment, installation, assurance of quality, or licensing of a similar product or service;c. Information that reveals cost or price information, production capacities, budget levels, or commercial strategies of Alion, its customers, or its suppliers;

d. Information that reveals aspects of past, present, or future ALION customer-funded development plans and programs, resulting in potential products to ALION;e. Information that discloses patentable subject matter for which it may be desirable to obtain patent protections.

Affidavit Page 1 of 3 August 28, .2008 The information sought to be withheld is considered to be proprietary for the reasons set forth in paragraphs (4)a, and (4)b, above.(5) To address 10 CFR 2.390 (a)(4), the information sought to be withheld is being submitted to NRC in confidence.

The information is of a sort customarily held in confidence by ALIGN, and is in fact so held. The information sought to be withheld has, to the best of my knowledge and belief, consistently been held in confidence by ALIGN, no public disclosure has been made, and it is not available in public sources. All disclosures to third parties, including any required transmittals to NRC, have been made, or must be made, pursuant to regulatory provisions or proprietary agreements which provide for maintenance of the information in confidence.

Its initial designation as proprietary information, and the subsequent steps taken to prevent its unauthorized disclosure, are as set forth in paragraphs (6) and (7) following.

(6) Initial approval of proprietary treatment of a document is made by the manager of the originating division, the person most likely to be acquainted with the value and sensitivity of the information in relation to industry knowledge.

Access to such documents within ALIGN is limited on a "need to know" basis.(7 1) The procedure for approval of external release of such a document typically requires review by the manager, project manager, principal engineer or other equivalent authority, by the manager of the cognizant marketing function (or his delegate), and by Contracts, for technical content, competitive effect, and determination of the accuracy of the proprietary designation.

Disclosures outside ALIGN are limited to regulatory bodies, customers and potential customers, and their agents, suppliers, and licensees, and others with a legitimate need for the information, and then only in accordance with appropriate regulatory provisions or proprietary agreements.

(8) The information identified in paragraph (2), above, is classified as proprietary because it contains detailed methods and processes, which ALIGN has developed for the preparation of detailed safety analyses in support of the design and licensing of nuclear facilities.

The development of these methods and processes along with the interpretation and application of the analytical results was derived from extensive company experience that constitutes a maj or ALIGN asset.(9) Public disclosure of the information sought to be withheld is likely to cause substantial harm to ALIGN's competitive position and foreclose or reduce the availability of profit-making opportunities.

The information is part of ALIGN's comprehensive nuclear safety and technology base, and its commercial value extends beyond the original development cost.The value of the technology base goes beyond the extensive physical databases used and analytical methodology and includes development of the expertise to determine and apply the appropriate evaluation processes.

The precise value of the expertise to devise an evaluation process and apply the correct analytical methodology is difficult to quantify, but it clearly is substantial.

ALIGN's competitive advantage will be lost if its competitors are able to use the results of the ALIGN experience to normalize or verify their own process or if they are able to claim an equivalent understanding by demonstrating that they can arrive at the same or similar conclusions.

The value of this information to ALIGN would be lost if the information were disclosed to the public. Making such information available to competitors Affidavit Page 2 of 3 August 28, 200 8 without their having been required to undertake a similar expenditure of resources would unfairly provide competitors with a windfall, and deprive ALION of the opportunity to exercise its competitive advantage to seek an adequate return on its large investment in developing these very valuable analytical tools.I declare under penalty of perjury that the foregoing affidavit and the matters stated therein are true and correct to the best of my knowledge, information, and belief.Executed on this 28th day of August, 2008.Peter K. Mast Vice President, Alion Science and Technology Affidavit Page 3 of 3 August 28, 2008

v t e Letter Sequence Request
TAC:MD9921, (Open)
Initiation Request, Request, Request Acceptance Administration Withholding Request Acceptance Questions Answers 8 April 2009 Results Approval... Other: ML091671807, NLS2009012, Response to Nuclear Regulatory Commission Request Regarding Withholding Information from Public Disclosure, NLS2009044, Enclosure 3, Calculation, Nedc 07-082, Rev. 2, Radiological Dose Analysis for a Loss of Coolant Accident (LOCA) at Cooper Nuclear Station, (Non-Proprietary Version)
MONTHYEAR2009-04-29 [Table View]

ML082910768

Text

Subject:

References:

Dear Sir or Madam:

2.0 Proposed

4.0 Technical

5.2 Applicable

6.0 Environmental

7.0 References

7.1 General

7.2 Precedent

1.0 DESCRIPTION

2.0 PROPOSED CHANGE

3.0 BACKGROUND

4.1.1 Suppression

4.1.2 Alternate

4.4 Proposed

7.0 REFERENCES

7.1 General

4.1 Offsite

5.1 General

1.1 through

3.1.4 Provide

3.2 Option

3.2.1 Provide

1.0 INTRODUCTION

9.0 ENVIRONMENTAL

10.0 REFERENCES

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