ML052770499

License Amendment Request for Application of the Alternative Source Term for Reevaluation of the Fuel Handling Accident Dose Consequences, Cooper Nuclear Station

ML052770499
Person / Time
Site:	Cooper
Issue date:	09/29/2005
From:	Edington R Nebraska Public Power District (NPPD)
To:	Document Control Desk, Office of Nuclear Reactor Regulation
References
NLS2005075, NUREG-1465
Download: ML052770499 (99)

Text

Nebraska Public Power District Always there when you need us 50.90 NLS2005075 September 29, 2005 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 Reevaluation of the Fuel Handling 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.

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 Fuel Handling Accident (FHA) dose consequences. This letter also requests approval of associated changes to the Cooper Nuclear Station (CNS) Technical Specifications (TS).

NPPD has completed a new design basis calculation (Enclosure 1) using the guidelines detailed in References 1 and 2. This calculation demonstrates that radiological dose consequences of an FHA using the AST are within regulatory limits. The associated TS changes would eliminate operability requirements for Secondary Containment, Secondary Containment Isolation Valves, the Standby Gas Treatment System, and Secondary Containment Isolation Instrumentation when handling irradiated fuel that has decayed for 24 hours2.777778e-4 days <br />0.00667 hours <br />3.968254e-5 weeks <br />9.132e-6 months <br /> since critical reactor operations, and when performing Core Alterations. Similar TS relaxations are proposed for the Control Room Emergency Filter System and its initiation instrumentation after a decay period of 7 days.

The proposed TS changes are based on TSTF-5 1, Revision 2, "Revise Containment Requirements During Handling Irradiated Fuel and Core Alterations" for equipment whose performance is no longer credited in the FHA. Submittal of this License Amendment Request for the FHA is a selective scope application of the AST, as provided for in Reference 2.

COOPER NUCLEAR STATION P.O. Box 98 / Brownville, NE 68321-0098 Telephone: (402) 825-3811 / Fax: (402) 825-5211 www.nppd.com

NLS2005075 Page 2 of 3 provides a description of the License Amendment Request, 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. Additionally, Appendix A to Attachment 1 describes the conformance of this License Amendment Request to Reference 2. Appendix B to Attachment 1 provides the responses to applicable issues that have been typically raised by the NRC in reviewing the FHA AST applications of other licensees. provides the proposed changes to the current CNS TS on marked up pages. provides the revised TS pages in final typed format. Attachment 4 provides the corresponding changes to the current Bases on marked up pages for your information. NRC approval is requested by September 1, 2006, with a 30-day implementation period, to support Refueling Outage 23, which is scheduled to begin in October 2006.

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 211 issued March 22, 2005, 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 criteria 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, 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).

Should you have any questions concerning this matter, please contact Paul Fleming, Licensing Manager, at (402) 825-2774.

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

Executed on

' /Z '7 5 (Dat )

Since y,

Randall K. Edington I

Vice President - Nuclear and Chief Nuclear Officer

/WV Attachments Enclosure

NLS2005075 Page 3 of 3 cc: Regional Administrator w/attachments, enclosure USNRC - Region IV Senior Project Manager w/attachments, enclosure USNRC - NRR Project Directorate IV-1 Senior Resident Inspector w/attachments, enclosure USNRC - CNS Nebraska Health and Human Services w/attachments, enclosure Department of Regulation and Licensure NPG Distribution w/o attachments, enclosure CNS Records w/attachments, enclosure

NLS2005075 Page 1 of 53 ATTACHMENT 1 License Amendment Request for Application of the Alternative Source Term for Reevaluation of the Fuel Handling Accident Dose Consequences Cooper Nuclear Station, Docket 50-298, DPR46 Revised TS Pages 3.3-57 3.6-34 3.6-40 3.3-63 3.6-36 3.7-8 3.6-32 3.6-38 3.7-9 3.6-33 3.6-39 1.0 Description 2.0 Proposed Change 2.1 Licensing Basis Change for Fuel Handling Accident 2.2 Technical Specification Changes 3.0

Background

4.0 Technical Analysis 4.1 Radiological Consequences of the Fuel Handling Accident 4.2 Atmospheric Dispersion (X/Q) 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 Appendix B Responses to Commonly Raised NRC Issues On FHA AST License Amendment Requests

NLS2005075 Page 2 of 53

1.0 DESCRIPTION

This letter is a request to amend Operating License DPR-46 for Cooper Nuclear Station (CNS).

The proposed change revises the CNS design basis accident (DBA) radiological assessment calculational methodology for the Fuel Handling Accident (FHA) through application of the Alternative Source Term (AST), in accordance with the provisions of 10 CFR 50.67, "Accident Source Term." This application represents a selective scope application of AST, as provided for in Regulatory Guide 1.183 (Reference 7.1-1). The Nebraska Public Power District (NPPD) requests Nuclear Regulatory Commission (NRC) review and approval of the AST FHA methodology for application to CNS. Approval of this License Amendment Request will supersede the previous design basis source term assumptions and radiological criteria for the FHA. Future revisions of this analysis, if any, will use the updated approved assumptions and criteria. In accordance with the AST FHA analysis results, revisions to the CNS Technical Specifications (TS) and TS Bases are proposed that reflect the revised safety analysis assumptions for a postulated FHA in the Reactor Building.

The CNS TS currently impose restrictions on plant operations when handling irradiated fuel assemblies or performing Core Alterations. These restrictions require that certain structures, systems or components be operable, and thereby assure that the radiological consequences of an FHA do not exceed those calculated in design basis analyses. The TS changes proposed in this application are based on NRC-approved TSTF-5 1, "Revise Containment Requirements During Handling Irradiated Fuel and Core Alterations" (Reference 7.1-2). The proposed TS and TS Bases changes are consistent with this guidance with regard to the relaxation of Secondary Containment, Secondary Containment Isolation Valve (SCIV), Standby Gas Treatment (SGT)

System, Secondary Containment Isolation Instrumentation, Control Room Emergency Filter System (CREFS), and CREFS Instrumentation operability for the movement of irradiated fuel after a sufficiently long decay time has passed, and during Core Alterations. A 24 hour2.777778e-4 days <br />0.00667 hours <br />3.968254e-5 weeks <br />9.132e-6 months <br /> decay period and a 7 day decay period are analyzed for the timing of these relaxations. NPPD is not requesting the TSTF-51 relaxations for AC/DC Power Sources - Shutdown and AC Distribution Systems - Shutdown in this License Amendment Request, which would otherwise be applicable to CNS.

The implementation of these changes will reduce the duration and cost of planned outages while maintaining an adequate safety margin. For example, windows of Secondary Containment inoperability of established time periods are currently built into outage schedules. During these windows, critical path fuel movements cannot be performed, which can result in extended outage durations.

NRC approval is requested by September 1, 2006, with a 30-day implementation period. This is to support utilization of this License Amendment Request for Refueling Outage 23, which is scheduled to begin in October 2006.

NLS2005075 Page 3 of 53

2.0 PROPOSED CHANGE

NPPD is proposing to modify (a) the licensing basis of the FHA as described in the CNS Updated Safety Analysis Report (USAR) and (b) Technical Specifications for Secondary Containment (TS 3.6.4.1), Secondary Containment Isolation Valves (TS 3.6.4.2), Standby Gas Treatment System (TS 3.6.4.3), Secondary Containment Isolation Instrumentation (TS 3.3.6.2),

Control Room Emergency Filter System (TS 3.7.4), and Control Room Emergency Filter System Instrumentation (TS 3.3.7.1).

2.1 Licensing Basis Change for Fuel Handling Accident NPPD is revising the licensing basis of the FHA described in Section XIV-6.4 of the CNS USAR. The proposed licensing basis change is the reevaluation of the FHA using the AST methodology and dose consequences analysis in accordance with 10 CFR 50.67, Regulatory Guide (RG) 1.183, and NUREG-1465 (Reference 7.1-3). Sections 3.0 and 4.0 of this evaluation provide the background and technical analysis in support of the licensing basis change. Approval of this License Amendment Request will result in the necessary revisions to the USAR, with revised USAR pages submitted pursuant to 10 CFR 50.71(e).

2.2 Technical Specification Changes An AST analysis has been performed for the radiological dose consequences of a postulated FHA involving irradiated fuel assemblies for two cases: 1) after a 24 hour2.777778e-4 days <br />0.00667 hours <br />3.968254e-5 weeks <br />9.132e-6 months <br /> decay time since reactor shutdown, and 2) after a 7 day decay time. The 24 hour2.777778e-4 days <br />0.00667 hours <br />3.968254e-5 weeks <br />9.132e-6 months <br /> decay time case shows that the calculated total effective dose equivalent (TEDE) values to the control room occupants, the exclusion area boundary (EAB), and low population zone (LPZ) are below the allowable TEDE limits established in 10 CFR 50.67 without reliance on the safety functions of Secondary Containment, the SGT System, SCIVs, or Secondary Containment isolation instrumentation.

The 7 day decay time case shows that in addition to the above Engineered Safety Features, there is sufficient radioactive decay that CREFS and its initiation instrumentation need not be credited in order for the FHA dose consequences to remain within these regulatory limits. These conclusions form the basis for the proposed TS changes that allow, after 24 hours2.777778e-4 days <br />0.00667 hours <br />3.968254e-5 weeks <br />9.132e-6 months <br /> since reactor shutdown, movement of irradiated fuel assemblies without the operability requirements for Secondary Containment, the SGT System, SCIVs, or Secondary Containment Isolation Instrumentation. TS changes are similarly proposed that relax CREFS and CREFS Instrumentation operability requirements after 7 days have elapsed since reactor shutdown. Due to the inherent time required to disassemble the reactor vessel and prepare to move the fuel, it is not credible that fuel movement would commence prior to the 24-hr decay time elapsing. Thus, there is no need for a dose consequence analysis for an FHA occurring prior to that time. The Technical Requirements Manual (TRM), which is incorporated by reference into the USAR, will continue to provide the administrative controls to preclude handling irradiated fuel prior to the analyzed decay period. (Refer also to the response to Question 4 of Appendix B for additional supporting information).

NLS2005075 Page 4 of 53 The operability requirements for Secondary Containment, the SGT System, SCIVs, Secondary Containment Isolation Instrumentation, CREFS, and CREFS Instrumentation are also being deleted during Core Alterations as part of this License Amendment Request. The USAR Chapter XIV events postulated to occur during Core Alterations, in addition to the postulated FHA, are:

a) Control Rod Removal Error During Refueling, b) Fuel Assembly Insertion Error During Refueling, and c) Fuel Assembly Loading Error (due to either a mislocated bundle or misoriented bundle). Except for the FHA, these other events are not analyzed as resulting in fuel cladding integrity damage. Since the FHA is the only event postulated to occur during Core Alterations that presents a challenge to a fission product barrier, the proposed TS changes omitting Core Alterations is justified per the criteria of 10 CFR 50.36(c)(2).

The Reviewer's Note in TSTF-51 requires that licensees adding the term "recently" make a commitment consistent with draft NUMARC 93-01, Revision 3, Section 11.2.6 "Safety Assessment for Removal of Equipment from Service During Shutdown Conditions", subheading "Containment - Primary (PWR)/Secondary (BWR)." The commitment in the Reviewer's Note reads:

The following guidelines are included in the assessment of systems removed from service during movement of irradiated fuel:

During fuel handling/core alterations, ventilation system and radiation monitor availability (as defined in NUMARC 91-06) should be assessed, with respect to filtration and monitoring of releases from the fuel. Following shutdown, radioactivity in the fuel decays away fairly rapidly. The basis of the Technical Specification operability amendment is the reduction in doses due to such decay. The goal of maintaining ventilation system and radiation monitor availability is to reduce doses even further below that provided by the natural decay.

A single normal or contingency method to promptly close primary or secondary containment penetrations should be developed. Such prompt methods need not completely block the penetration or be capable of resisting pressure.

The purpose of the "prompt methods" mentioned above are to enable ventilation systems to draw the release from a postulated fuel handling accident in the proper direction such that it can be treated and monitored.

When draft NUMARC 93-01, Revision 3, was issued in the final version, Section 11.2.6 was renumbered to Section 11.3.6 and the guidelines quoted above were numbered 11.3.6.5 and modified slightly as follows:

... the following guidelines should be included in the assessment of systems removed from service:

NLS2005075 Page 5 of 53 During fuel handling/core alterations, ventilation system and radiation monitor availability (as defined in NUMARC 91-06) should be assessed, with respect to filtration and monitoring of releases from the fuel. Following shutdown, radioactivity in the RCS decays fairly rapidly. The basis of the Technical Specification operability amendment is the reduction in doses due to such decay.

The goal of maintaining ventilation system and radiation monitor availability is to reduce doses even further below that provided by the natural decay, and to avoid unmonitored releases.

A single normal or contingency method to promptly close primary or secondary containment penetrations should be developed. Such prompt methods need not completely block the penetration or be capable of resisting pressure. The purpose is to enable ventilation systems to draw the release from a postulated fuel handling accident in the proper direction such that it can be treated and monitored.

The NRC has considered the final version and the draft version of NUMARC 93-01, Revision 3, to be functionally equivalent (Reference 7.2-2). NPPD commits to implement the guidelines of Section 11.3.6.5 of NUMARC 93-01, Revision 3, prior to the implementation of this license amendment. (Refer to the response to Question 2 in Appendix B for further information on the NPPD approach to conformance with this commitment).

Approval of this License Amendment Request will result in the necessary revisions to the associated TS Bases in accordance with TS 5.5.10 (Technical Specifications Bases Control Program). These TS Bases changes are provided for information in Attachment 4 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. 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 License Amendment Request addresses these requirements in proposing selectively to use an AST in evaluating the offsite and control room radiological consequences of an FHA. 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 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.

NLS2005075 Page 6 of 53

4.0 TECHNICAL ANALYSIS

4.1 Radiological Consequences of the Fuel Handling Accident NPPD has completed a calculation following both a 24 hour2.777778e-4 days <br />0.00667 hours <br />3.968254e-5 weeks <br />9.132e-6 months <br /> decay time since reactor shutdown and a 7 day decay time evaluating the potential dose consequences of the FHA. A copy of the calculation is provided in Enclosure 1. This calculation uses the AST guidelines outlined in NUREG-1465 and RG 1.183. This calculation demonstrates that after a 24 hour2.777778e-4 days <br />0.00667 hours <br />3.968254e-5 weeks <br />9.132e-6 months <br /> decay period following reactor shutdown, the radiological doses at the EAB, LPZ, and in the control room are within regulatory limits without crediting the operability of Secondary Containment, Secondary Containment Isolation Valves, the Standby Gas Treatment System, or Secondary Containment Isolation Instrumentation. After a 7 day decay time, CREFS and CREFS Instrumentation are additionally not credited to meet the regulatory dose limits. A loss-of-offsite-power (LOOP) is not assumed to occur during this accident. (Refer to the response to Question 5 of Appendix B for additional information about the LOOP assumption). Enclosure 1, Attachments A through F provide the RADTRAD files used for the 24 hour2.777778e-4 days <br />0.00667 hours <br />3.968254e-5 weeks <br />9.132e-6 months <br /> decay time case and the 7 day decay time case.

In the postulated FHA, a fuel assembly is assumed to be dropped, thereby damaging 151 fuel rods during fuel handling. To support the proposed TS changes, the calculation does not take credit for Secondary Containment isolation or filtration by the SGT System, assuming a 24 hour2.777778e-4 days <br />0.00667 hours <br />3.968254e-5 weeks <br />9.132e-6 months <br /> period has elapsed after reactor shutdown from full power prior to the FHA. The calculation does not take credit for CREFS or its initiation instrumentation after 7 days have elapsed since reactor shutdown prior to the FHA.

The entire gap activity from the damaged fuel is assumed to be released to the Reactor Cavity Pool, and then directly to the outside atmosphere from the Reactor Building refueling floor over a 2-hour period. The calculated activity in the gap of the fuel rods assumes the assembly has been operated at 102 percent of rated thermal power times a maximum radial peaking factor of 2.0.

The analysis assumes the RG 1.183 Table 3 non-loss-of-coolant accident gap fractions. In accordance with RG 1.183, the assumed iodine species released from the fuel gap to the water is 95 percent cesium iodide, 4.85 percent elemental, and 0.15 percent organic.

The radionuclides are assumed to be released from the damaged fuel rods, pass through the water in the reactor cavity, and enter the Reactor Building atmosphere instantaneously. As the released gases rise through the overlaying water, halogens are scrubbed by the water column, resulting in an effective halogen decontamination factor of 200. No decontamination of noble gases or organic iodine forms was assumed. The iodine species above the pool were assumed to be 57 percent elemental and 43 percent organic, in accordance with RG 1.183. The guidance in RG 1.183 allows an effective halogen decontamination factor of 200 when the overlaying water column is at least 23 feet. This pre-condition is met for the Reactor Cavity Pool, but not for the Spent Fuel Pool where the water depth may be a minimum of 21 feet 6 inches per TS requirements. For the case of an FHA occurring in the Spent Fuel Pool, the implied reduction in scrubbing efficiency is offset by the reduced number of fuel rods (48 vs. 151) that are projected to be damaged by a fuel assembly drop over the Spent Fuel Pool. The effective decontamination

NLS2005075 Attachment I Page 7 of 53 factor in RG 1.183 is based on an exponential function. In this function, more scrubbing occurs at the bottom of the water column than at the top of the water column. As such, a pool level of 21 feet 6 inches results in a conservative reduction of about 6.5 percent. This is less than the 68 percent reduction in the amount of damaged rods, and hence the radionuclides released from the Spent Fuel Pool are less than from the Reactor Cavity Pool. (Refer to the response to Question 1 of Appendix B for further information on an FHA occurring in the Spent Fuel Pool).

The analysis assumes a release rate based on the release of essentially 100 percent of the radionuclides in the Reactor Building to the environment over a 2-hour period. This release rate is equivalent to a release flow rate of 45,764 cfm.

To calculate the dose in the Control Room for the 24 hour2.777778e-4 days <br />0.00667 hours <br />3.968254e-5 weeks <br />9.132e-6 months <br /> decay time case where CREFS is credited, two air intake flow rates were used, representing the two distinct operating modes of the Control Room Air Conditioning System. The first mode was for normal air intake of 3635 cfm for a 1-minute time period. This flow rate was the sum of the normal fresh air intake flow of 3235 cfm and an assumed unfiltered inleakage of 400 cfm. The unfiltered inleakage assumption is conservative relative to the inleakage values reported in the NPPD response to Generic Letter 2003-01 (Reference 7.1-4). (Refer to the response to Question 3 in Appendix B for additional information). The second mode was for CREFS operation. CREFS initiates on high radiation detected in the Reactor Building Exhaust Plenum. NPPD has assessed the TS setpoint for the radiation monitor, and has determined that with a Reactor Building exhaust fan or SGT fan initially running, the design basis AST FHA will cause the credited Reactor Building Exhaust Plenum high radiation actuation within the 1-minute delay time assumed for CREFS initiation.

(Refer also to the response to Question 5 of Appendix B for additional information).

In calculating the dose in the Control Room for the 7 day decay time case where CREFS is not credited, the normal air intake of 3635 cfm (including 400 cfm unfiltered inleakage) was used for the duration of the accident.

The analysis used assumptions and inputs that follow the guidance in RG 1.183. The key assumptions are listed in Table 1 of this attachment. The calculated dose results are given in Table 2. The calculated doses are within the Standard Review Plan 15.0.1 radiological dose acceptance criteria for an FHA. These TEDE criteria are 6.3 rem at the EAB for the worst two hours, 6.3 rem at the LPZ for the duration of the accident and 5 rem in the Control Room for the duration of the accident.

4.2 Atmospheric Dispersion (x/O)

The Control Room intake X/Q values used in this analysis were developed previously and were submitted for NRC staff review (Reference 7.1-5) as part of the approval of the current FHA of record in License Amendment 187. The release models used for the currently licensed FHIA and the one proposed in this License Amendment Request are somewhat different. The previous FHA assumed a 90-second unfiltered ground level release via the Reactor Building vent followed by an elevated filtered release for the balance of the 30-day duration, while this analysis assumes

NLS2005075 Page 8 of 53 a 30-day unfiltered ground level release from the Reactor Building vent. In the Reference 7.1-5 submittal, X/Qs were calculated for the vent release for different values of Reactor Building vent flow rate (51333 cfln, 9500 cfln, and 1780 cfm). The three flow rates modeled a Reactor Building exhaust fan coastdown after a Group 6 Primary Containment Isolation System trip on high radiation in the Reactor Building Exhaust Plenum. For conservatism, NPPD has elected to use X/Q values calculated for the 1780 cfln flow rate. Although the X/Q values for 1780 cfm were not specifically cited in the License Amendment 187 Safety Evaluation (as they were not applicable until after the 90-second Reactor Building ground level release had terminated), they are based on the same approved meteorological data, and were developed using ARCON96.

The X/Qs used for the EAB and LPZ are ground level releases from the Reactor Building vent calculated using site specific inputs and methodology described in Regulatory Guide 1.3. These X/Q values were previously submitted to the NRC (Reference 7.1-6). These X/Q values were approved for use, in part, in License Amendment 187. Because of the previously described modeling differences between the current FHA and the one proposed in this License Amendment Request, only the EAB and LPZ XIQs for the 90-second unfiltered ground level release were cited in License Amendment 187.

Table 1 FHA Analysis Assumptions Reactor power 2429 Mwt Radial peaking factor 2.0 Fission product decay period 24 hours2.777778e-4 days <br />0.00667 hours <br />3.968254e-5 weeks <br />9.132e-6 months <br /> and 7 days Number of fuel rods damaged 151 Number of rods in core (equivalent full length) 47,857 Fuel Gap fission product inventory I-131 8 percent Kr-85 10 percent Other iodines and noble gases 5 percent Alkali metals 0.12 percent Iodine species fractions Elemental 0.9985 Organic 0.0015 Particulates none Reactor cavity water depth 23 feet Pool iodine effective decontamination factor 200

NLS2005075 Page 9 of 53 Table 1 (continued)

Chemical form of iodine above pool Elemental 57 percent Organic 43 percent Release modeling Immediate release from fuel through Reactor Cavity Pool to building 100% release from building in 2 hours2.314815e-5 days <br />5.555556e-4 hours <br />3.306878e-6 weeks <br />7.61e-7 months <br /> No credit for building holdup or filtration prior to release Control Room Envelope volume 141,860 ft3 Normal ventilation unfiltered fresh intake 3,235 cfm CREFS filtered intake 810 cfin CREFS start delay time, minutes 1 minute Unfiltered outside inleakage 400 cfm CREFS filter efficiency 89%

Control room occupancy factors 0-24 hr 1.0 24-96 hr 0.6 96-720 hr 0.4 Control room breathing rate 3.5E-4 m3/sec Offsite breathing rate 0-8 hr 3.5E-4 m3/sec 8-24 hr 1.8E-4 m3/sec 24-720 hr 2.3E-4 m3/sec Atmospheric dispersion factors, sec/M3 Ground level release from reactor building vent Period EAB LPZ Cc 0-2 hours 5.2E-4 2.9E-4 4.)

2-8 hours 2.9E-4 8-24 hours 7.3E-5 24-96 hours 2.5E-5 9.(96-720 hours 2.5E-6 7.2 ontrol Room 15E-3 24E-3 32E-3 01E-4 22E-4

NLS2005075 Page 10 of 53 Table 2 Calculated MA Radiological Consequences TEDE (rem)

EAB LPZ Control Room Calculated results (24-hr decay period) 1.459 0.815 4.507 (7 day decay period) 0.627 0.350 4.446 Dose acceptance criteria 6.3 6.3 5

5.0 REGULATORY SAFETY ANALYSIS 5.1 No Significant Hazards Consideration In accordance with 10CFR50.92, 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) has evaluated whether or not a significant hazards consideration is involved with the proposed amendment by focusing on the three standards set forth in 10 CFR 50.92, "Issuance of amendment," as discussed below:

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

Response: No The proposed amendment involves implementation of the Alternative Source Term (AST) for the fuel handling accident (FHA) at Cooper Nuclear Station (CNS). There are no physical design modifications to the plant associated with the proposed amendment.

The FHA AST calculation does not impact the initiators of an FHA in any way.

The changes also do not impact the initiators for any other design basis accident (DBA) or events. Therefore, because DBA initiators are not being altered by adoption of the AST analyses the probability of an accident previously evaluated is not affected.

With respect to consequences, the only previously evaluated accident that could be affected is the FHA. The AST is an input to calculations used to evaluate the consequences of the accident, and does not, in and of itself, affect the plant response or the actual pathways to the environment utilized by the radiation/activity released by the

NLS2005075 Page 11 of 53 fuel. It does, however, better represent the physical characteristics of the release, so that appropriate mitigation techniques may be applied. For the FHA, the AST analyses demonstrate acceptable doses that are within regulatory limits after 24 hours2.777778e-4 days <br />0.00667 hours <br />3.968254e-5 weeks <br />9.132e-6 months <br /> of radioactive decay since reactor shutdown, without credit for Secondary Containment, the Standby Gas Treatment System, Secondary Containment Isolation Valves, or Secondary Containment Isolation Instrumentation, and that the Control Room Emergency Filter System (CREFS) and CREFS Instrumentation need not be credited after a 7 day period of decay. Therefore, the consequences of an accident previously evaluated are not significantly increased.

Based on the above conclusions, this proposed amendment does 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 The proposed amendment does not involve a physical alteration of the plant. No new or different types of equipment will be installed and there are no physical modifications to existing equipment associated with the proposed changes. The proposed changes to the control of Engineered Safety Features during handling of irradiated fuel do not create new 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 the margin of safety?

Response: No The proposed amendment is associated with the implementation of a new licensing basis for the CNS FHA. Approval of this change from the original source term to an AST derived in accordance with the guidance of Regulatory Guide (RG) 1.183 is being requested. The results of the FHA analysis, revised in support of the proposed license amendment, are subject to revised acceptance criteria. The AST FHA analysis has been performed using conservative methodologies, as specified in RG 1. 183. Safety margins have been evaluated and analytical conservatism has been utilized to ensure that the analysis adequately bounds the postulated limiting event scenario. The dose consequences of the limiting FHA remain within the acceptance criteria presented in 10 CFR 50.67, the Standard Review Plan, and RG 1. 183.

NLS2005075 Page 12 of 53 The proposed changes continue to ensure that the doses at the Exclusion Area Boundary (EAB) and Low Population Zone (LPZ) boundary, as well as the Control Room, are within the corresponding regulatory limits. For the FHA, RG 1.183 conservatively sets the EAB and LPZ limits below the 10 CFR 50.67 limit, and sets the Control Room limit consistent with 10 CFR 50.67.

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.

5.2 Applicable Regulatory Requirements/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 (GDCs), 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 Draft GDC is applicable to Secondary Containment when the Primary Containment is open.

This License Amendment Request does not alter CNS commitments to conformance with this Draft GDC, although Secondary Containment will no longer be credited for the mitigation of an FHA.

Draft GDC 69 - Protection Against Radioactivity Release From Spent Fuel and Waste Storage "Containment of fuel and waste storage shall be provided if accidents could lead to release of undue amounts of radioactivity to the public environs."

The consequences of an FHA, as presented in this License Amendment Request demonstrate that undue amounts of radioactivity are not released to the public.

NLS2005075 Page 13 of 53 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:

"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 NRC acceptance of this AST License Amendment Request, the 5 rem whole body (or equivalent) criteria for the FHA will be replaced with the 5 rem TEDE limits of 10 CFR 50.67 (as provided for in GDC 19).

1971 GDC 20 - Protection System Functions NPPD is committed to the provisions of 1971 GDC 20 as it applies to the design of the safety-related actuation instrumentation of the Reactor Building Ventilation Radiation Monitoring System:

"The protection system shall be designed (1) to initiate automatically the operation of appropriate systems including the reactivity control systems, to assure that specified acceptable fuel design limits are not exceeded as a result of anticipated operational occurrences and (2) to sense accident conditions and to initiate the operation of systems and components important to safety."

The deletion of the TS operability requirement for the Reactor Building Ventilation Exhaust Plenum - High function of TS 3.3.6.2 during Core Alterations and movement of non-recently irradiated fuel does not alter NPPD commitments to GDC 20.

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 approval of this AST License Amendment Request will result in a revision to the evaluation and consequences of a design basis FHA previously reported in the Updated Safety Analysis Report.

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.

NLS2005075 Attachment I Page 14 of 53 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).

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

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 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.

TSTF-51, Rev. 2, "Revise Containment Requirements During Handling Irradiated Fuel and Core Alterations," Excel Services Corporation.

3.

NUREG-1465, "Accident Source Terms for Light-Water Nuclear Power Plants,"

L. Soffer et al., February 1995.

4.

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).

5.

Letter from J. Swailes (NPPD) to USNRC, dated September 14, 2001, "Proposed License Amendment Related to the Design Basis Accident Radiological Assessment Calculational Methodology - Supplemental Information" (NLS2001085).

NLS2005075 Page 15 of 53

6.

Letter from J. Swailes (NPPD) to USNRC, dated February 28, 2001, "Proposed License Amendment Related to the Design Basis Accident Radiological Assessment Calculational Methodology" (NLS200101 1).

7.2 Precedent There are numerous NRC-approved Boiling Water Reactor (BWR) precedents for use of AST for the FHA and adoption of TSTF-5 1. The following examples are approved selective scope FHA AST submittals that were used in the development of this License Amendment Request.

1.

Letter from G. Vissing (NRC) to M. Kansler (Entergy Nuclear Operations), dated September 12, 2002, "James A. Fitzpatrick Nuclear Power Plant - Amendment Re: Technical Specification Change to the Requirements for Handling Irradiated Fuel (TAC No. MB5328)."

2.

Letter from J. Boska (NRC) to M. Kansler (Entergy Nuclear Operations), dated April 28, 2005, "Pilgrim Nuclear Power Station - Issuance of Amendment Re:

Alternative Source Term for the Fuel Handling Accident Dose Consequences (TAC No. MC2705).

NLS2005075 Page 16 of 53 Appendix A Regulatory Guide 1.183 Comparison The Appendix provides a comparison table of the Cooper Nuclear Station FHA AST calculation (Enclosure 1) to the criteria of RG 1.183 that pertain to BWR guidance for AST analysis development and License Amendment Request content.

RG CNS Section RG Position Analysis 1.1.1 The proposed uses of an AST and the associated proposed facility Conforms-Adequate safety margins are modifications and changes to procedures should be evaluated to determine maintained, as discussed in the No whether the proposed changes are consistent with the principle that sufficient Significant Hazards Consideration.

safety margins are maintained, including a margin to account for analysis Future changes will be evaluated under uncertainties. The safety margins are products of specific values and limits the provisions of 10 CFR 50.59.

contained in the 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.

1.1.2 The proposed uses of an AST and the associated proposed facility Conforms-Adoption of the AST for the modifications and changes to procedures should be evaluated to determine FHA makes possible the TS relaxations of whether the proposed changes are consistent with the principle that adequate Secondary Containment, SGT System, defense in depth is maintained to compensate for uncertainties in accident SCIV, and Secondary Containment progression and analysis data. Consistency with the defense-in-depth Isolation instrumentation operability after philosophy is maintained if system redundancy, independence, and diversity a 24 hour2.777778e-4 days <br />0.00667 hours <br />3.968254e-5 weeks <br />9.132e-6 months <br /> decay period, and CREFS and are preserved commensurate with the expected frequency, consequences of CREFS Instrumentation after a 7 day

NLS2005075 Page 17 of 53 RG CNS Section RG Position Analysis challenges to the system, and uncertainties. In all cases, compliance with the decay period, as well as during Core General Design Criteria in Appendix A to 10 CFR Part 50 is essential.

Alterations. These TS changes are based Modifications proposed for the facility generally should not create a need for on TSTF-5 1, which was approved for use compensatory programmatic activities, such as reliance on manual operator by the NRC. Compliance with the GDCs actions.

are maintained (both the 1967 draft GDCs and the 1971 GDCs to which NPPD has committed). No reliance is placed on compensatory programmatic actions (including manual operator actions) to maintain adequate defense-in-depth.

1.1.2 Proposed modifications that seek to downgrade or remove required Not applicable-Changes to the TS which engineered safeguards equipment should be evaluated to be sure that the remove the operability requirements for modification does not invalidate assumptions made in facility PRAs and Secondary Containment, the SGT System, does not adversely impact the facility's severe accident management SCIV and Secondary Containment program.

Isolation Instruments during movement of irradiated fuel assemblies 24 hours2.777778e-4 days <br />0.00667 hours <br />3.968254e-5 weeks <br />9.132e-6 months <br /> after shutdown, and CREFS and CREFS instrumentation after a 7 day decay period, have negligible affect on the assumptions made in the PRA or the severe accident management program.

The affected systems have no affect on Core Damage Frequency. In addition, Large Early Release Frequency is not affected based on the assumed damage to 0.3% (151/47857) fuel rods in an FHA and the CNS definition of a large release as greater than 10% Cesium Iodine fraction.

NLS2005075 Page 18 of 53 RG CNS Section RG Position Analysis 1.1.3 The design basis accident source term is a fundamental assumption upon Conforms-See RG Section 1.3 which a significant portion of the facility design is based. Additionally, many discussions.

aspects of facility operation derive from the design analyses that incorporated 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 This approach would create two tiers of analyses, those based on the Conforms-The CNS FHA AST License previous source term and those based on an AST. The radiological Amendment Request is a selective scope acceptance criteria would also be different with some analyses based on AST application. Approval of this whole body and thyroid criteria and some based on TEDE criteria. Full License Amendment Request will implementation of the AST revises the plant licensing basis to specify the supersede the previous design basis AST in place of the previous accident source term and establishes the TEDE source term assumptions and radiological dose as the new acceptance criteria. Selective implementation of the AST criteria for the FHA. Future revisions of also revises the plant licensing basis and may establish the TEDE dose as the this analysis, if any, will use the updated new acceptance criteria. Selective implementation differs from full approved assumptions and criteria.

implementation only in the scope of the change. In either case, the facility design bases should clearly indicate that 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 Radiological analyses generally should be based on assumptions and inputs Conforms-The FHA analysis is a that are consistent with corresponding data used in other design basis safety selective scope application of the AST. It analyses, radiological and nonradiological, unless these data would result in relies on assumptions and inputs that do nonconservative results or otherwise conflict with the guidance in this guide.

not create a conflict with, or render non-conservative, other design basis safety analyses.

NLS2005075 Page 19 of 53 RG CNS Section RG Position Analysis 1.1.4 Although the AST provided in this guide was based on a limited spectrum of Not Applicable-No changes are proposed severe accidents, the particular characteristics have been tailored specifically in this License Amendment Request to for DBA analysis use. The AST is not representative of the wide spectrum of Emergency Preparedness requirements.

possible events that make up the planning basis of emergency preparedness.

Therefore, the AST is insufficient by itselfas a basis for requesting relief from the emergency preparedness requirements of 10 CFR 50.47 and Appendix E to 10 CFR Part 50.

1.2.1 Full implementation is a modification of the facility design basis that Not Applicable-This License addresses all characteristics of the AST, that is, composition and magnitude Amendment Request is a selective scope of the radioactive material, its chemical and physical form, and the timing of implementation of the AST to the FHA.

its release. Full implementation revises the plant licensing basis to specify the AST in place of the previous accident source term and establishes the TEDE dose as the new acceptance criteria. This applies not only to the analyses performed in the application (which may only include a subset of the plant analyses), but also to all future design basis analyses. At a minimum for full implementations, the DBA LOCA must be re-analyzed using the guidance in Appendix A of this guide. Additional guidance on analysis is provided in Regulatory Position 1.3 of this guide. Since the AST and TEDE criteria would become part of the facility design basis, new applications of the AST would not require prior NRC approval unless stipulated by 10 CFR 50.59, "Changes, Tests, and Experiments," or unless the new application involved a change to a technical specification. However, a change from an approved AST to a different AST that is not approved for use at that facility would require a license amendment under 10 CFR 50.67.

1.2.2 Selective implementation is a modification of the facility design basis that Conforms-This License Amendment (1) is based on one or more of the characteristics of the AST or (2) entails re-Request is a selective scope application of evaluation of a limited subset of the design basis radiological analyses. The the AST to change the design basis NRC staff will allow licensees flexibility in technically justified selective radiological dose consequence analysis of implementations provided a clear, logical, and consistent design basis is the FHA.

NLS2005075 Page 20 of 53 RG I

CNS Section RG Position

[

Anal sis 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 adjust the closure time on isolation valves not included in the original approval.

1.3.1 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 Conforms-This selective scope AST FHA request is salient to: a) Control Room Habitability (GDC 19 and NUREG-0737 Item 1JI.D.3.4), b) AST (10 CFR 50.67), c) Emergency Response Facility Habitability (10 CFR 50 Appendix E Paragraph IV.E.8), and d)

Facility Siting (10 CFR 100.11). Control

NLS2005075 Page 21 of 53 RG CNS Section RG Position Analysis Appendix E to 10 CFR Part 50)

Room Habitability and compliance with

• Alternative Source Term (10 CFR 50.67) the Alternative Source Term requirements

• Environmental Reports (10 CFR Part 51) are the principal subjects of this submittal

• Facility Siting (10 CFR 100.11)5 and are discussed in Sections 3 and 4 of There may be additional applications of the accident source term identified this License Amendment Request.

in the technical specification bases and in various licensee commitments.

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

Habitability, CNS will continue to meet

• Post-Accident Access Shielding (NUREG-0737, II.B.2) the NUREG-0654 Planning Standard for

• Post-Accident Sampling Capability (NUREG-0737, II.B.3)

Emergency Facilities and Equipment as

• Accident Monitoring Instrumentation (NUREG-0737, II.F.1) described in the CNS Emergency Plan.

• Leakage Control (NUREG-0737, III.D. 1. 1)

As stated in Footnote 5 of this RG, the

• Emergency Response Facilities (NUREG-0737, m

I.A. 1.2) dose guidelines of 10 CFR 100.11 are

• Control Room Habitability (NUREG-0737, IILD.3.4) superseded by 10 CFR 50.67 for licensees that have implemented an AST.

1.3.2 Any implementation of an AST, full or selective, and any associated facility Conforms-The License Amendment modification should be supported by evaluations of all significant Request for this selective scope radiological and nonradiological impacts of the proposed actions. This application of the AST to the FHA evaluation should consider the impact of the proposed changes on the evaluated the impact of the proposed facility's compliance with the regulations and commitments listed above as change against the Current Licensing well as any other facility-specific requirements. These impacts may be due to Basis, mitigating system design basis (1) the associated facility modifications or (2) the differences in the AST requirements, and Technical characteristics. The scope and extent of the re-evaluation will necessarily be Specifications. No facility modifications a function of the specific proposed facility modifications6 and whether a full are proposed as part of this License or selective implementation is being pursued. The NRC staff does not expect Amendment Request and compliance a complete recalculation of all facility radiological analyses, but does expect with regulations and commitments are licensees to evaluate all impacts of the proposed changes and to update the maintained.

affected analyses and the design bases appropriately. An analysis is

NLS2005075 Page 22 of 53 RG CNS Section RG Position Analysis 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.

1.3.2 The NRC staff has performed an evaluation of the impact of the AST on Conforms-There are no plant three representative operating reactors (Ref. 14). This evaluation determined modifications that are planned to that radiological analysis results based on the TID-14844 source term implement the FHA AST analysis. The assumptions (Ref. 1) and the whole body and thyroid methodology generally radiological and nonradiological impacts bound the results from analyses based on the AST and TEDE methodology.

of the FHA AST have been considered Licensees may use the applicable conclusions of this evaluation in and discussed in the License Amendment addressing the impact of the AST on design basis radiological analyses.

Request, as applicable.

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

NLS2005075 Page 23 of 53 RG CNS Section RG Position Analysis ability of ductwork downstream of the dampers to withstand increased stresses.

1.3.2 For full implementation, a complete DBA LOCA analysis as described in Not Applicable-This License Appendix A of this guide should be performed, as a minimum. Other design Amendment Request is a selective scope basis analyses are updated in accordance with the guidance in this section.

application of the AST to the FHA.

1.3.2 A selective implementation of an AST and any associated facility Conforms-The radiological and modification based on the AST should evaluate all the radiological and nonradiological impacts of the FHA AST nonradiological impacts of the proposed actions as they apply to the have been considered as they apply to the particular implementation. Design basis analyses are updated in accordance implementation.

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.

1.3.3 It may be possible to demonstrate by sensitivity or scoping evaluations that Not Applicable-The FHA AST analysis existing analyses have sufficient margin and need not be recalculated. As does not rely on sensitivity or scoping used in this guide, a sensitivity analysis is an evaluation that considers how analyses.

the overall results vary as an input parameter (in this case, AST characteristics) is varied. A scoping analysis is a brief evaluation that uses

NLS2005075 Page 24 of 53 RG CNS Section RG Position Analysis 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 postaccident 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.

1.3.4 Full implementation of the AST replaces the previous accident source term Not Applicable-This License with the approved AST and the TEDE criteria for all design basis Amendment Request is a selective scope radiological analyses. The implementation may have been supported in part implementation of the AST to the CNS by sensitivity or scoping analyses that concluded many of the design basis FHA.

radiological analyses would remain bounding for the AST and the TEDE criteria and would not require updating. After the implementation is complete, there may be a subsequent need (e.g., a planned facility modification) to revise these analyses or to perform new analyses. For these recalculations, the NRC staff expects that all characteristics of the AST and the TEDE criteria incorporated into the design basis will be addressed in all affected analyses on an individual as-needed basis. Re-evaluation using the

NLS2005075 Page 25 of 53 RG CNS Section RG Position Analysis previously approved source term may not be appropriate. Since the AST and the TEDE criteria are part of the approved design basis for the facility, use of the AST and TEDE criteria in new applications at the facility do not constitute a change in analysis methodology that would require NRC approval.7 1.3.4 This guidance is also applicable to selective implementations to the extent Conforms-Subsequent updates to the that the affected analyses are within the scope of the approved FHA AST analysis after NRC approval implementation as described in the facility design basis. In these cases, the will consider the characteristics of the characteristics of the AST and TEDE criteria identified in the facility design AST and TEDE criteria in the facility basis need to be considered in updating the analyses. Use of other design basis.

characteristics of the AST or 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-The FHA is not a basis proposed plant modification associated with the AST implementation. The for the CNS Environmental Qualification EQ analyses that have assumptions or inputs affected by the plant program.

modification should be updated to address these impacts. The NRC staff is assessing the effect of increased cesium releases on EQ doses to determine whether licensee action is warranted. Until such time as this generic issue is resolved, licensees may use either the AST or the TID 14844 assumptions for performing the required EQ analyses. However, no plant modifications are required to address the impact of the difference in source term characteristics (i.e., AST vs TID14844) 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 The use of an AST changes only the regulatory assumptions regarding the Not applicable. No facility modifications analytical treatment of the design basis accidents. The AST has no direct are proposed or planned as effect on the probability of the accident. Use of an AST alone cannot implementation actions of the FHA AST increase the core damage frequency (CDF) or the large early release analysis.

NLS2005075 Page 26 of 53 RG CNS Section RG Position Analysis frequency (LERF). 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 facilitys PRA, the impact on the existing PRAs should be evaluated.

1.4 Consideration should be given to the risk impact of proposed Conforms-The TS and TS Bases changes implementations that seek to remove or downgrade the performance of which remove the operability previously required engineered safeguards equipment on the basis of the requirements for Secondary Containment, reduced postulated doses. The NRC staff may request risk information if Standby Gas Treatment, Secondary there is a reason to question adequate protection of public health and safety.

Containment Isolation Valves and Secondary Containment Isolation Instruments during movement of irradiated fuel assemblies 24 hours2.777778e-4 days <br />0.00667 hours <br />3.968254e-5 weeks <br />9.132e-6 months <br /> after shutdown have a negligible effect on risk due to Core Damage or Large Early Releases. The affected systems have no affect on Core Damage Frequency. In addition, Large Early Release Frequency is not affected based on the assumed damage to 0.3% (151/47857) fuel rods in an FHA and the CNS definition of a large release as greater than 10% Cesium Iodine fraction. Approval of this License Amendment Request will result in the necessary changes to shutdown risk modeling to reflect the proposed changes to the TS, as required.

1.4 The licensee may elect to use risk insights in support of proposed changes to Not Applicable-NPPD is not utilizing the design basis that are not addressed in currently approved NRC staff risk insights as a basis for this License

NLS2005075 Attachment I Page 27 of 53 RG CNS Section RG Position Analysis positions. For guidance, refer to Regulatory Guide 1.174, "An Approach for Amendment Request.

Using Probabilistic Risk Assessment in Risk-Informed Decisions on Plant-Specific Changes to the Licensing Basis" (Ref. 15).

1.5 According to 10 CFR 50.90, an application for an amendment must fully Conforms-The License Amendment describe the changes desired and should follow, as far as applicable, the form Request is formatted in accordance with prescribed for original applications. Regulatory Guide 1.70, "Standard accepted NRC/industry guidance. The Format and Content of Safety Analysis Reports for Nuclear Power Plants request describes the radiological and (LWR Edition)" (Ref 16), provides additional guidance. The NRC staffs nonradiological impacts of the FHA AST finding that the amendment may be approved must be based on the licensee's analysis. Consistent with several of the analyses, since it is these analyses that will become part of the design basis most recently approved BWR precedents of the facility. The amendment request should describe the licensee's (Pilgrim, Vermont Yankee, and River analyses of the radiological and nonradiological impacts of the proposed Bend), affected USAR pages are not modification in sufficient detail to support review by the NRC staff. The included in the analyses. Approval of this staff recommends that licensees submit affected FSAR pages annotated with License Amendment Request will result changes that reflect the revised analyses or submit the actual calculation in the necessary revisions to the USAR, documentation.

with revised USAR pages submitted pursuant to 10 CFR 50.7 1(e).

1.5 If the licensee has used a current approved version of an NRC-sponsored Conforms-NPPD has used RADTRAD computer code, the NRC staff review can be made more efficient if the Version 3.03 in the performance of the licensee identifies the code used and submits the inputs that the licensee used FHA AST analysis. The inputs made are in the calculations made with that code. In many cases, this will reduce the included with Enclosure 1. ARCON96 need for NRC staff confirmatory analyses. This recommendation does not was used to develop the Control Room constitute a requirement that the licensee use NRC-sponsored computer X/Q values. That calculation was codes.

submitted to the NRC as part of the License Amendment 187 review.

1.6 Requirements for updating the facility's final safety analysis report (FSAR)

Conforms-Approval of this License are in 10 CFR 50.71, "Maintenance of Records, Making of Reports." The Amendment Request will result in the regulations in 10 CFR 50.71(e) require that the FSAR be updated to include necessary revisions to the USAR, with I all changes made in the facility or procedures described in the FSAR and all revised USAR pages submitted pursuant

NLS2005075 Page 28 of 53 RG CNS Section RG Position Analysis safety evaluations performed by the licensee in support of requests for to 10 CFR 50.71 (e).

license 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 (Ref.

16) provides additional guidance. The descriptions of superseded analyses should be removed from the FSAR in the interest of maintaining a clear design basis.

2.1 The AST must be based on major accidents, hypothesized for the purposes of Conforms-This License Amendment design analyses or consideration of possible accidental events, that could Request applies the AST to the FHA.

result 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 2.2.

radioactive fission products released into containment, the types and quantities of the radioactive species released, and the chemical forms of iodine released.

2.3 The AST must not be based upon a single accident scenario but instead must Conforms-This License Amendment represent a spectrum of credible severe accident events. Risk insights may be Request considers a number of scenarios used, not to select a single risk-significant accident, but rather to establish involving an FHA during fuel movements the range of events to be considered. Relevant insights from applicable or Core Alterations. The most limiting of severe accident research on the phenomenology of fission product release these events is analyzed for radiological and transport behavior may be considered.

consequences.

2.4 The AST must have a defensible technical basis supported by sufficient Conforms-Enclosure 1 has been experimental and empirical data, be verified and validated, and be developed based on NUREG-1465 and

NLS2005075 Page 29 of 53 RG CNS Section RG Position Analysis documented in a scrutable form that facilitates public review and discourse.

this Regulatory Guide. The calculation, which utilizes RADTRAD Version 3.03 was developed in accordance with 10 CFR 50 Appendix B, Criterion m.

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 m.

3.1 The inventory of fission products in the reactor core and available for release Conforms-See Enclosure 1 Section 2.2.

to the containment should be based on the maximum full power operation of the core with, as a minimum, current licensed values for fuel enrichment, fuel 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 (Ref. 17) or ORIGEN-ARP (Ref 18). Core inventory factors (Ci/MWt) provided in TID14844 and used in some analysis computer codes were derived for low burnup, low enrichment fuel and should not be used with higher burnup and higher enrichment fuels.

3.1 For the DBA LOCA, all fuel assemblies in the core are assumed to be Conforms-See Enclosure 1 Section 2.1.

affected and the core average inventory should be used. For DBA events that do not involve the entire core, the fission product inventory of each of the 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

NLS2005075 Page 30 of 53 RG CNS Section RG Position Analysis report (COLR) or technical specifications should be applied in determining the inventory of the damaged rods.

3.1 No adjustment to the fission product inventory should be made for events Conforms-See Enclosure 1 Section 2.2.

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

3.2 The core inventory release fractions, by radionuclide groups, for the gap Not Applicable-This License release and early in-vessel damage phases for DBA LOCAs are listed in Amendment Request is a selective scope Table 1 for BWRs and Table 2 for PWRs. These fractions are applied to the implementation of the AST to the CNS equilibrium core inventory described in Regulatory Position 3.1.

FHA.

3.2 For non-LOCA events, the fractions of the core inventory assumed to be in the gap for the various radionuclides are given in Table 3. The release fractions from Table 3 are used in conjunction with the fission product inventory calculated with the maximum core radial peaking factor.

Table 3.11 Non-LOCA Fraction of Fission Product Inventory in Gap Table 3 Conforms-See Enclosure 1 Section 2.2.

Group 1-131 Kr-85 Other Noble Gases Other Halogens Alkali Metals Fraction 0.08 0.10 0.05 0.05 0.12 Table 4 tabulates the onset and duration of each sequential release phase for DBA LOCAs at PWRs and BWRs. The specified onset is the time following the initiation of the accident (i.e., time = 0). The early in-vessel phase Conforms-See Enclosure 1 Section 4.1.

NLS2005075 Attachment I Page 31 of 53 RG CNS Section RG Position Analysis 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.'2 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.

3.3 For facilities licensed with leak-before-break methodology, the onset of the Not Applicable-This License gap release phase may be assumed to be 10 minutes. A licensee may propose Amendment Request is a selective scope an alternative time for the onset of the gap release phase, based on facility-implementation of the AST to the CNS specific calculations using suitable analysis codes or on an accepted topical FHA.

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.

3.4 Table 5 lists the elements in each radionuclide group that should be Conforms-These radionuclide groups considered in design basis analyses.

were considered in the FHA AST analysis, as described in Enclosure 1 Table 5 Radionuclide Groups Section 4.3.

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 2.2.

containment in a postulated accident, 95 percent of the iodine released should be assumed to be cesium iodide (CsI), 4.85 percent elemental iodine,

NLS2005075 Page 32 of 53 RG CNS Section

~

RG Position Analysis 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.

3.6 The amount of fuel damage caused by non-LOCA design basis events should be analyzed to determine, for the case resulting in the highest radioactivity release, the fraction of the fuel that reaches or exceeds the initiation temperature of fuel melt and the fraction of fuel elements for which the fuel clad is breached. Although the NRC staff has traditionally relied upon the departure from nucleate boiling ratio (DNBR) as a fuel damage criterion, licensees may propose other methods to the NRC staff, such as those based upon enthalpy deposition, for estimating fuel damage for the purpose of establishing radioactivity releases.

The amount of fuel damage caused by a FHA is addressed in Appendix B of this guide.

Conforms-See Enclosure 1 Section 2.2.

4.1.1 The dose calculations should determine the TEDE. TEDE is the sum of the Conforms-See Enclosure 1 Section 2.6.

committed effective dose equivalent (CEDE) from inhalation and the deep dose equivalent (DDE) from external exposure. The calculation of these two 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.' 3 4.1.2 The exposure-to-CEDE factors for inhalation of radioactive material should Conforms-See Enclosure 1 Section 2.6.

be derived from the data provided in ICRP Publication 30, "Limits for Intakes of Radionuclides by Workers" (Ref. 19). Table 2.1 of Federal

NLS2005075 Page 33 of 53 RG CNS Section RG Position Analysis Guidance Report 11, "Limiting Values of Radionuclide Intake and Air Concentration and Dose Conversion Factors for Inhalation, Submersion, and Ingestion" (Ref. 20), 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 <br />0.00222 hours <br />1.322751e-5 weeks <br />3.044e-6 months <br />, the breathing rate of persons offsite should be assumed Conforms-See Enclosure 1 Section 2.6.

to be 3.5 x 1 0 4 cubic meters per second. From 8 to 24 hours2.777778e-4 days <br />0.00667 hours <br />3.968254e-5 weeks <br />9.132e-6 months <br /> following the accident, the breathing rate should be assumed to be 1.8 x 104 cubic meters per second. After that and until the end of the accident, the rate should be assumed to be 2.3 x 104 cubic meters per second.

4.1.4 The DDE should be calculated assuming submergence in semi-infinite cloud Conforms-See Enclosure 1 Section 2.6.

assumptions with appropriate credit for attenuation by body tissue. The DDE is 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 1I. 1 of Federal Guidance Report 12, "External Exposure to Radionuclides in Air, Water, and Soil" (Ref. 21), 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.

Conforms-See Enclosure 1 Section 2.5.

The maximum EAB TEDE for any two-hour period following the start of the 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

NLS2005075 Page 34 of 53 RG CNS Section RG Position Analysis 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 2.6.

boundary of the low population zone (LPZ) and should be used in determining 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 Conforms-See Enclosure 1 Section 3.

deposition on the ground.

4.2.1 The TEDE analysis should consider all sources of radiation that will cause Conforms-See Enclosure 1 Section 5.

exposure to control room personnel. The applicable sources will vary from facility to facility, but typically will include:

• 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 Conforms-See Enclosure 1 Section 5.

room dose analysis should be determined using the same source term, transport, and 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 Conforms-See Enclosure 1 Section 5.

control room,' 5 and the shielding models used to determine radiation dose

NLS2005075 Page 35 of 53 RG CNS Section RG Position Analysis rates from external sources, should be structured to provide suitably conservative estimates of the exposure to control room personnel.

4.2.4 Credit for engineered safety features that mitigate airborne radioactive Conforms-Under the design basis source material within the control room may be assumed. Such features may include term release and dispersion criteria of control room isolation or pressurization, or intake or recirculation filtration.

Appendix B to this Regulatory Guide, the Refer to Section 6.5.1, "ESF Atmospheric Cleanup System," of the SRP Reactor Building Exhaust Plenum (Ref. 3) and Regulatory Guide 1.52, "Design, Testing, and Maintenance radiation monitor will detect high Criteria for Postaccident Engineered-Safety-Feature Atmosphere Cleanup radiation in the plenum and start CREFS System Air Filtration and Adsorption Units of Light-Water-Cooled Nuclear within the assumed 1-minute initiation Power Plants" (Ref. 25), for guidance. The control room design is often time.

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-No credit is taken for the use equipment or prophylactic drugs. Deviations may be considered on a case-of personal protective equipment or by-case basis.

prophylactic drugs.

4.2.6 The dose receptor for these analyses is the hypothetical maximum exposed Conforms-See Enclosure 1 Section 2.6.

individual who is present in the control room for 100% of the time during the first 24 hours2.777778e-4 days <br />0.00667 hours <br />3.968254e-5 weeks <br />9.132e-6 months <br /> after the event, 60% of the time between 1 and 4 days, and 40% of the time from 4 days to 30 days.' 6 For the duration of the event, the breathing rate of this individual should be assumed to be 3.5 x 1 0 4 cubic meters per second.

NLS2005075 Page 36 of 53 RG CNS Section RG Position Analysis 4.2.7 Control room doses should be calculated using dose conversion factors Conforms-See Enclosure 1, Section 2.6.

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, DDEfinie, where the control room is modeled as a hemisphere that has a volume, V, in cubic feet, equivalent to that of the control room (Ref. 22).

DDEfinite = DDEV 43 338 Equation 1 1173 4.3 The guidance provided in Regulatory Positions 4.1 and 4.2 should be used, Not Applicable-This is a selective scope as applicable, in re-assessing the radiological analyses identified in application of the AST to the FHA. The Regulatory Position 1.3.1, such as those in NUREG-0737 (Ref. 2). Design NUREG-0737 analyses are based on non-envelope source terms provided in NUREG-0737 should be updated for AST LOCA results. Therefore a consistency with the AST. In general, radiation exposures to plant personnel conversion to AST methodology is not identified in Regulatory Position 1.3.1 should be expressed in terms of appropriate.

TEDE. Integrated radiation exposure of plant equipment should be determined using the guidance of Appendix I of this guide.

4.4 The radiological criteria for the EAB, the outer boundary of the LPZ, and for Conforms-See Enclosure 1, Section 5.

the control room are in 10 CFR 50.67. These criteria are stated for evaluating 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.

4.4 The acceptance criteria for the various NUREG-0737 (Ref. 2) items Not Applicable-This is a selective scope generally reference General Design Criteria 19 (GDC 19) from Appendix A application of the AST to the FHA. The

NLS2005075 Page 37 of 53 RG CNS Section RG Position Analysis to 10 CFR Part 50 or specify criteria derived from GDC 19. These criteria NUREG-0737 analyses are based on non-are generally specified in terms of whole body dose, or its equivalent to any TEDE LOCA results. Therefore a change body organ. For facilities applying for, or having received, approval for the from GDC 19 to TEDE methodology is use of an AST, the applicable criteria should be updated for consistency with not appropriate.

the TEDE criterion in 10 CFR 50.67(b)(2)(iii).

5.1.1 The evaluations required by 10 CFR 50.67 are re-analyses of the design basis Conforms-Enclosure 1 was prepared and safety analyses and evaluations required by 10 CFR 50.34; they are accepted by NPPD under a 10 CFR 50 considered to be a significant input to the evaluations required by 10 CFR Appendix B Quality Assurance program.

50.92 or 10 CFR 50.59. These analyses should be prepared, reviewed, and maintained in accordance with quality assurance programs that comply with Appendix B, "Quality Assurance Criteria for Nuclear Power Plants and Fuel Reprocessing Plants," to 10 CFR Part 50.

5.1.1 These design basis analyses were structured to provide a conservative set of Not Applicable-NPPD is not proposing assumptions to test the performance of one or more aspects of the facility deviations to conformance with this 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.

5.1.2 Credit may be taken for accident mitigation features that are classified as Conforms-The accident mitigation safety-related, are required to be operable by technical specifications, are features that are credited in the FHA AST powered by emergency power sources, and are either automatically actuated analysis are: 1) 23 foot water level over or, in limited cases, have actuation requirements explicitly addressed in fuel in the reactor (TS-controlled), and 2) emergency operating procedures. The single active component failure that CREFS and CREFS initiation

NLS2005075 Page 38 of 53 RG CNS Section RG Position Analysis results in the most limiting radiological consequences should be assumed.

instrumentation (TS-controlled). A loss-Assumptions regarding the occurrence and timing of a loss of offsite power of-offsite power is not assumed for should be selected with the objective of maximizing the postulated accidents occurring during Modes 4 or 5 radiological consequences.

per the CNS TS Bases for AC Power Sources.

5.1.3 The numeric values that are chosen as inputs to the analyses required by 10 Conforms-The numeric values chosen CFR 50.67 should be selected with the objective of determining a provide a conservative dose result. There conservative postulated dose. In some instances, a particular parameter may are no parameters in the FHA AST be conservative in one portion of an analysis but be nonconservative in analysis that have conflicting credit another portion of the same analysis. For example, assuming minimum within the analysis. Parameters that are containment system spray flow is usually conservative for estimating iodine controlled by TS either use the TS values, scrubbing, but in many cases may be nonconservative when determining or are otherwise bounded by them (e.g.

sump pH. Sensitivity analyses may be needed to determine the appropriate CREFS filter efficiency relaxation). The value to use. As a conservative alternative, the limiting value applicable to parameters are not based on surveillance each portion of the analysis may be used in the evaluation of that portion. A testing results.

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.' 8 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.

5.1.4 The NRC staff considers the implementation of an AST to be a significant Conforms-The FHA analysis change to the design basis of the facility that is voluntarily initiated by the assumptions and methods are compatible licensee. In order to issue a license amendment authorizing the use of an with the AST and the TEDE criteria.

AST and the TEDE dose criteria, the NRC staff must make a current finding

NLS2005075 Page 39 of 53 RG CNS Section RG Position Analysis of 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 The appendices to this regulatory guide provide accident-specific Conforms-See RG 1.183 Appendix B assumptions that are acceptable to the staff for performing analyses that are Sections of this appendix.

required by 10 CFR 50.67. The DBAs addressed in these attachments were 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.

5.2 The NRC staff has determined that the analysis assumptions in the Conforms-See RG 1.183 Appendix B appendices to this guide provide an integrated approach to performing the Sections of this appendix.

individual analyses and generally expects licensees to address each assumption 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

NLS2005075 Page 40 of 53 RG CNS Section RG Position Analysis 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 The NRC is committed to using probabilistic risk analysis (PRA) insights in Conforms-PRA was not used as a basis its regulatory activities and will consider licensee proposals for changes in for acceptability of this FHA AST analysis assumptions based upon risk insights. The staff will not approve License Amendment Request.

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 Atmospheric dispersion values (X/Q) for the EAB, the LPZ, and the control Conforms-X/Qs that are used were room that were approved by the staff during initial facility licensing or in previously submitted for NRC review (see subsequent licensing proceedings may be used in performing the radiological License Amendment Request Section analyses identified by this guide. Methodologies that have been used for 4.2).

determining X/Q values are documented in Regulatory Guides 1.3 and 1.4, 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" (Refs. 6, 7, 22, and 28).

5.3 References 22 and 28 should be used if the FSAR X/Q values are to be Not Applicable-The X/Q values were revised or if values are to be determined for new release points or receptor previously submitted to the NRC and are distances. Fumigation should be considered where applicable for the EAB based on approved meteorological data and LPZ. For the EAB, the assumed fumigation period should be timed to be (see License Amendment Request Section included in the worst 2-hour exposure period. The NRC computer code 4.2).

PAVAN (Ref. 29) implements Regulatory Guide 1.145 (Ref. 28) and its use

NLS2005075 Page 41 of 53 RG CNS Section RG Position Analysis is acceptable to the NRC staff. The methodology of the NRC computer code ARCON9619 (Ref. 26) 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" (Ref 30). All changes in X/Q analysis methodology should be reviewed by the NRC staff.

6.0 The assumptions in Appendix Ito this guide are acceptable to the NRC staff Not Applicable-10 CFR 50.49 for performing radiological assessments associated with equipment compliance is not affected by the Fuel qualification. The assumptions in Appendix I will supersede Regulatory Handling Accident.

Positions 2.c(1) and 2.c(2) and Appendix D of Revision 1 of Regulatory Guide 1.89, "Environmental Qualification of Certain Electric Equipment Important to Safety for Nuclear Power Plants" (Ref. 11), for operating reactors that have amended their licensing basis to use an alternative source term. Except as stated in Appendix I, all other assumptions, methods, and provisions of Revision 1 of Regulatory Guide 1.89 remain effective.

The NRC staff is assessing the effect of increased cesium releases on EQ doses to determine whether licensee action is warranted. Until such time as this generic issue is resolved, licensees may use either the AST or the TID14844 assumptions for performing the required EQ analyses. However, no plant modifications are required to address the impact of the difference in source term characteristics (i.e., AST vs TID14844) on EQ doses pending the outcome of the evaluation of the generic issue.

B-1 Acceptable assumptions regarding core inventory and the release of Conforms-See discussions in RG Section radionuclides from the fuel are provided in Regulatory Position 3 of this 3.

guide.

NLS2005075 Page 42 of 53 RG CNS Section RG Position Analysis B-l.1 The number of fuel rods damaged during the accident should be based on a Conforms-The FHA is based on GE14 conservative analysis that considers the most limiting case. This analysis fuel which assumes 151 fuel rods are should consider parameters such as the weight of the dropped heavy load or damaged per Amendment 22 of NEDE-the weight of a dropped fuel assembly (plus any attached handling grapples),

24011-P-A (GESTAR II).

the height of the drop, and the compression, torsion, and shear stresses on the irradiated fuel rods. Damage to adjacent fuel assemblies, if applicable (e.g.,

events over the reactor vessel), should be considered.

B-1.2 The fission product release from the breached fuel is based on Regulatory Conforms-See Enclosure 1 Section 2.2.

Position 3.2 of this guide and the estimate of the number of fuel rods breached. All the gap activity in the damaged rods is assumed to be instantaneously released. Radionuclides that should be considered include xenons, kryptons, halogens, cesiums, and rubidiums.

B-1.3 The chemical form of radioiodine released from the fuel to the spent fuel Conforms-See Enclosure 1 Section 2.2.

pool should be assumed to be 95% cesium iodide (CsI), 4.85 percent elemental iodine, and 0.15 percent organic iodide. The CsI released from the fuel is assumed to completely dissociate in the pool water. Because of the low pH of the pool water, the iodine re-evolves as elemental iodine. This is assumed to occur instantaneously. The NRC staff will consider, on a case-by-case basis, justifiable mechanistic treatment of the iodine release from the pool.

B-2 If the depth of water above the damaged fuel is 23 feet or greater, the Conforms-See Enclosure 1 Section 2.2.

decontamination factors for the elemental and organic species are 500 and 1, Section 4.1 of this License Amendment respectively, giving an overall effective decontamination factor of 200 (i.e.,

Request discusses the case of an FHA 99.5% of the total iodine released from the damaged rods is retained by the occurring in the Spent Fuel Pool.

water). This difference in decontamination factors for elemental (99.85%)

and organic iodine (0.15%) species results in the iodine above the water being composed of 57% elemental and 43% organic species. If the depth of water is not 23 feet, the decontamination factor will have to be determined on a case-by-case method (Ref. B-1).

NLS2005075 Page 43 of 53 RG CNS Section RG Position Analysis B-3 The retention of noble gases in the water in the fuel pool or reactor cavity is Conforms-See Enclosure 1 Section 4.1.

negligible (i.e., decontamination factor of 1). Particulate radionuclides are assumed to be retained by the water in the fuel pool or reactor cavity (i.e.,

infinite decontamination factor).

B-4 For fuel handling accidents postulated to occur within the fuel building, the Not Applicable-CNS does not have a following assumptions are acceptable to the NRC staff.

separate fuel building.

B-5.1 If the containment is isolated2 during fuel handling operations, no Not Applicable-NPPD is not proposing radiological consequences need to be analyzed.

an isolated Secondary Containment during fuel handling operations.

B-5.2 If the containment is open during fuel handling operations, but designed to Not Applicable-NPPD is not taking any automatically isolate in the event of a fuel handling accident, the release credit for Secondary Containment duration should be based on delays in radiation detection and completion of isolation after an FHA.

containment isolation. If it can be shown that containment isolation occurs before radioactivity is released to the environmentl no radiological consequences need to be analyzed.

B-5.3 If the containment is open during fuel handling operations (e.g., personnel air Conforms-Radioactive material that lock or equipment hatch is open),3 the radioactive material that escapes from escapes the Reactor Cavity Pool is the reactor cavity pool to the containment is released to the environment over released to the environment over a 2-hour a 2-hour time period.

period.

B-5.4 A reduction in the amount of radioactive material released from the Not Applicable-The SGT System is not containment by ESF filter systems may be taken into account provided that credited in the release from the Secondary these systems meet the guidance of Regulatory Guide 1.52 and Generic Containment to the environment.

Letter 99-02 (Refs. B-2 and B-3). Delays in radiation detection, actuation of the ESF filtration system, or diversion of ventilation flow to the ESF filtration system should be determined and accounted for in the radioactivity release analyses.'

5.5 Credit for dilution or mixing of the activity released from the reactor cavity Not Applicable-No credit is taken for by natural or forced convection inside the containment may be considered on mixing or dilution.

a case-by-case basis. Such credit is generally limited to 50% of the

NLS2005075 Page 44 of 53 RG

_CNS Section RG Position Analysis containment free volume. This evaluation should consider the magnitude of the contairnment volume and exhaust rate, the potential for bypass to the environment, the location of exhaust plenums relative to the surface of the reactor cavity, recirculation ventilation systems, and internal walls and floors that impede stream flow between the surface of the reactor cavity and the exhaust plenums.

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 proposed with this License Amendment any dose calculations. However, a proposed modification that would delay Request. Future modifications will containment spray actuation could involve recalculation of DBA LOCA consider the effects upon the inputs and doses, re-assessment of the containment pressure and temperature transient, assumptions of the FHA AST.

recalculation of sump pH, re-assessment of the emergency diesel generator loading sequence, integrated doses to equipment in the containment, and more.

Footnote 7 In performing screenings and evaluations pursuant to 10 CFR 50.59, it may Conforms-Approval of this License be necessary to compare dose results expressed in terms of whole body and Amendment Request will result in the thyroid with new results expressed in terms of TEDE. In these cases, the necessary implementation within the CNS previous thyroid dose should be multiplied by 0.03 and the product added to 10 CFR 50.59 program.

the whole 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 2.2.

value provided in Appendix K to 10 CFR Part 50, typically 1.02.

Footnote 9 Note that for some radionuclides, such as Cs-137, equilibrium will not be Not applicable-Particulates are not reached prior to fuel offload. Thus, the maximum inventory at the end of life applicable to the FHA, as they are 100%

should be used.

retained in the Reactor Cavity Pool.

NLS2005075 Page 45 of 53 RG CNS Section RG Position Analysis Footnote The release fractions listed here have been determined to be acceptable for Conforms-The design of the current core 10 use with currently approved LWR fuel with a peak burnup up to 62,000 will remain within the Footnote 10 MWD/MTU. The data in this section may not be applicable to cores constraints. NPPD is monitoring containing mixed oxide (MOX) fuel.

Industry/NRC progress in eliminating this footnote for future core design.

Footnote The release fractions listed here have been determined to be acceptable for Conforms-Using the CNS modeling 11 use with currently approved LWR fuel with a peak burnup up to 62,000 code, NPPD has determined that the MWD/MTU provided that the maximum linear heat generation rate does not current core will remain within the exceed 6.3 kw/ft peak rod average power for bumups exceeding 54 constraints of Footnote 11. However, the GWD/MTU. As an alternative, fission gas release calculations performed fuel vendor has been contacted to provide using NRC-approved methodologies may be considered on a case-by-case confirmation with the licensing basis. To be acceptable, these calculations must use a projected power code/methodology. The results are history that will bound the limiting projected plant-specific power history for expected first quarter 2006, and the specific fuel load. For the BWR rod drop accident the PWR rod ejection supplemental information will be accident, the gap fractions are assumed to be 10% for iodines and noble provided to the NRC, as necessary.

gases.

NPPD is monitoring Industry/NRC progress in eliminating this footnote for future core design.

Footnote In lieu of treating the release in a linear ramp manner, the activity for each Conforms-See Enclosure 1, Section 2.3.

12 phase can be modeled as being released instantaneously at the start of that release phase, i.e., in step increases.

Footnote The prior practice of basing inhalation exposure on only radioiodine and not Conforms-RADTRAD Version 3.03 was 13 including radioiodine in external exposure calculations is not consistent with used which calculates offsite exposure the definition of TEDE and the characteristics of the revised source term.

TEDE for the AST.

Footnote With regard to the EAB TEDE, the maximum two-hour value is the basis for Conforms-Future changes made pursuant 14 screening and evaluation under 10 CFR 50.59. Changes to doses outside of to 10 CFR 50.59 will be in the context of the two-hour window are only considered in the context of their impact on the maximum 2-hour dose consequences.

the maximum two-hour EAB TEDE.

NLS2005075 Page 46 of 53 RG CNS Section RG Position Analysis Footnote The iodine protection factor (IPF) methodology of Reference 22 may not be Conforms-FHA AST calculation was 15 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 3.03.

parameters change over the duration of the event, the IPF methodology should only be used with caution. The NRC computer codes HABIT (Ref.

23) and RADTRAD (Ref. 24) incorporate suitable methodologies.

Footnote This occupancy is modeled in the X/Q values determined in Reference 22 Conforms-The required occupancy 16 and should not be credited twice. The ARCON96 Code (Ref. 26) does not assumptions are incorporated into the incorporate these occupancy assumptions, making it necessary to apply this dose calculation. See Enclosure 1, correction in the dose calculations.

Section 2.6.

Footnote Note that for some parameters, the technical specification value may be Not Applicable-Although a lower 18 adjusted for analysis purposes by factors provided in other regulatory CREFS filter efficiency is assumed than guidance. For example, ESF filter efficiencies are based on the guidance in is required by the TS Ventilation Filter Regulatory Guide 1.52 (Ref. 25) and in Generic Letter 99-02 (Ref. 27) rather Testing Program, NPPD is not proposing than the surveillance test criteria in the technical specifications. Generally, changes to the TS efficiency.

these adjustments address potential changes in the parameter between scheduled surveillance tests.

Footnote The ARCON96 computer code contains processing options that may yield Conforms-The X/Qs used in the FHA 19 X/Q values that are not sufficiently conservative for use in accident AST analysis were developed using consequence assessments or may be incompatible with release point and ARCON96 and were previously reviewed ventilation intake configurations at particular sites. The applicability of these by the NRC pursuant to License options and associated input parameters should be evaluated on a case-by-Amendment 187.

case basis. The 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.

Footnote These analyses should consider the time for the radioactivity concentration to Conforms-With ventilation flow B-1 reach levels corresponding to the monitor setpoint, instrument line sampling established at the start of the accident, the time, detector response time, diversion damper alignment time, and filter Reactor Building Exhaust Plenum system actuation, as applicable.

radiation monitor will provide the

NLS2005075 Page 47 of 53 RG CNS Section RG Position Analysis necessary CREFS initiation. Refer to Section 4.1 and Appendix B Question 5 of this License Amendment Request.

Footnote Containment isolation does not imply containment integrity as defined by Not Applicable-NPPD is not proposing B-2 technical specifications for non-shutdown modes. The term isolation is used an isolated Secondary Containment here collectively to encompass both containment integrity and containment during fuel handling operations.

closure, typically in place during shutdown periods. To be credited in the analysis, the appropriate form of isolation should be addressed in technical specifications.

Footnote The staff will generally require that technical specifications allowing such Conforms-With the acceptance of TSTF-B-3 operations include administrative controls to close the airlock, hatch, or open 51, the NRC has accepted that TS penetrations within 30 minutes. Such administrative controls will generally administrative controls are not needed require that a dedicated individual be present, with necessary equipment provided there is a commitment made to available, to restore containment closure should a fuel handling accident conform with draft NUMARC 93-01, occur. Radiological analyses should generally not credit this manual Revision 3, Section 11.2.6.

isolation.

NLS2005075 Page 48 of 53 Responses to Commonly Raised NRC Issues On FMA AST License Amendment Requests A review has been performed of previous issues raised by the NRC as evidenced by the Requests for Additional Information (RAIs) that have been asked for BWR licensee FHA AST License Amendment Requests. This Appendix provides the NPPD responses to these RAI issues.

Question 1:

There is less than 23 feet of water above the Spent Fuel Pool. Justify why an FHA associated with fuel in the reactor is more limiting than an FHA occurring over the Spent Fuel Pool.

Response

CNS Technical Specifications require a Spent Fuel Pool water level to be > 21 ft 6 inches over the top of irradiated fuel assemblies seated in the spent fuel storage pool racks. NPPD has determined that the difference in kinetic energy between an FHA in Spent Fuel Pool versus the Reactor Cavity Pool results in damage to 48 fuel rods versus 151. This is a 68 percent reduction in damaged rods.

The effective Decontamination Factor (DF) in RG 1.183 is based on an exponential function. In this function, more scrubbing occurs at the bottom of the water column than at the top of the water column. Therefore, assuming a linear change in DF as a function of water level would be conservative. As such, a pool level of 21 feet 6 inches compared with 23 feet would result in about a 6.5 percent reduction in DF.

In summary, the implied reduction in scrubbing efficiency is more than offset by the reduced number of fuel rods that are projected to be damaged by a fuel assembly drop over the spent fuel pool. Accordingly, the dose consequences of an FHA occurring over the Spent Fuel Pool will be bounded by the FHA occurring in the Reactor Cavity Pool. Note-This approach is similar to the one accepted by the NRC for the James A. Fitzpatrick Nuclear Power Station FHA AST application (see Page 4 to the Safety Evaluation of Reference 7.2-1).

Question 2:

Describe the approach that will be used to conform to the commitment to implement the provisions of NUMARC 93-01 Section 11.3.6.5 when Secondary Containment and SGTS are inoperable during the handling of non-recently irradiated fuel. How much of an open area to the environment would be permitted? Describe the ventilation systems that would be used to draw the release from the postulated FHA (i.e., are they Engineered Safety Features, do they have carbon adsorber filters and high-efficiency particulate air filters, and are they tested in accordance with Regulatory Guide 1.52 or other standards?). Justify how these means will ensure that the NUMARC guidance "[these prompt methods] enable the ventilation systems to draw from the postulated FHA such that it can be treated and monitored" is met. Will there be a test to determine that

NLS2005075 Page 49 of 53 all air flow was into the Secondary Containment in the event that partial closure is allowed?

Response

The approach that will be taken by NPPD to conform to the NUMARC guidance will be to develop Secondary Containment breach control guidance that will be a prerequisite to moving irradiated fuel when Secondary Containment, the SCIVs, the SGT System, or Secondary Containment Isolation Instrumentation are not operable. This breach control guidance will include:

Ensuring that Secondary Containment access doors and hatches are closed (except for opening and closing for normal personnel travel). Fouling of an access door/hatch is not allowed unless an individual is on station to restore the closed configuration if an FHA occurs.

Ensuring that there is normally one available train of SGT, such that if an FHA occurs, the SGT train can be started manually. If both SGT trains would be out of service (such as for work on a component in the common suction line), pre-positioned means would be on hand with personnel stationed to ensure system integrity could be restored and one train placed in operation.

Ensuring there is at least the capability of closing one SCIV manually for each Secondary Containment penetration.

Ensuring there are no breaches of piping systems that penetrate both sides of Secondary Containment unless there are pre-positioned means to close the breach if an FHA occurs.

Ensuring there is no maintenance in progress that compromises the structural integrity of Secondary Containment (e.g., removal of blowout panels, or core drills).

The strategy outlined above is designed to assure that the Secondary Containment function can be restored without the allowance of partial breaches. Therefore, no SGT System tests above the applicable TS surveillance testing requirements are needed.

Question 3:

Discuss the Control Room unfiltered bypass leakage assumption used in the FHA AST calculation compared with leakage results of inleakage testing performed per Generic Letter 2003-01.

Response

There are two issues related to the Generic Letter (GL) 2003-01 inleakage testing results that factor into the FHA AST calculation: 1) the assumed inleakage in the FHA calculation must bound the testing results, and 2) the receptor location for the X/Qs used in the FHA calculation must remain appropriately conservative relative to the tested inleakage.

NLS2005075 Page 50 of 53 Regarding the first issue, the unfiltered inleakage assumed in the AST FHA calculation is 400 cfm. This bounds the testing results of 64 cfm (CREFS pressurization mode) and 223 cfm (CREFS isolation mode) reported to the NRC in the response to GL 2003-01.

Regarding the second issue, the FHA calculation assumes a single receptor point at the Control Building HVAC intake. As part of the GL 2003-01 testing, NPPD has not attempted to determine the actual locations of the inleakage, or to quantify the amount of inleakage that ingress at points that may be closer to the Reactor Building vent than the Control Building intake. There are potential points of inleakage that are both greater than and less than the distance between the Reactor Building vent and the Control Building intake. NPPD believes that the Control Room X/Q (which assumes a ground level ingress at the Control Building HVAC intake) already provides inherent conservatism. Furthermore, from a geometric standpoint, it would be difficult to envision the magnitudes of wind direction shifts needed over the 2-hour release period that would maximize both the radionuclide concentration at the Control Building intake and potential points of closer inleakage. Additionally, inleakage points would typically be from adjacent buildings which would make plume pathways resulting in higher atmospheric dispersion factors improbable. Moreover, the 400 cfm unfiltered inleakage assumed in Enclosure 1 to this License Amendment Request provides a great deal of margin to the actual inleakage. For these reasons, assuming ingress at the Control Room HVAC intake provides conservative Control Room X/Qs and dose consequences that are appropriately bounding for the GL 2003-01 testing results.

Question 4:

The only dose analyses provided for the FHA were those in which the dose was associated with fuel which was not "recently" irradiated. If it is ever intended to handle fuel which is "recently" irradiated, then an analysis needs to be provided that demonstrates acceptable dose results, both offsite and in the control room, in the event of an FHA. Are there any Technical Specifications or administrative controls that apply to when fuel movement is allowed post shutdown?

Response: assumes a 24 hour2.777778e-4 days <br />0.00667 hours <br />3.968254e-5 weeks <br />9.132e-6 months <br /> decay time since reactor shutdown prior to the FHA. Similarly, under the CNS Custom Technical Specifications 24 hours2.777778e-4 days <br />0.00667 hours <br />3.968254e-5 weeks <br />9.132e-6 months <br /> were required to elapse after shutdown before irradiated fuel movements could commence. As part of the conversion to the Improved Standard Technical Specifications (License Amendment 178), this TS was relocated to the Technical Requirements Manual (a document incorporated by reference into the USAR).

The basis for this relocation was:

The 24 hour2.777778e-4 days <br />0.00667 hours <br />3.968254e-5 weeks <br />9.132e-6 months <br /> decay time following subcriticality will always be met for a refueling outage because of the operations required prior to moving irradiated fuel in the reactor vessel (e.g., containment entry, removal of drywell head, removal of vessel head, removal of vessel internals).

NLS2005075 Page 51 of 53 Therefore, the requirement is not necessary to be in Technical Specifications and its relocation will not adversely affect the public health and safety.

As a result of License Amendment 187, the TRM was changed from the previously relocated 24 hour2.777778e-4 days <br />0.00667 hours <br />3.968254e-5 weeks <br />9.132e-6 months <br /> decay time to the new 67 hour7.75463e-4 days <br />0.0186 hours <br />1.107804e-4 weeks <br />2.54935e-5 months <br /> decay period.

Approval of this License Amendment Request will result in restoring the original 24 hour2.777778e-4 days <br />0.00667 hours <br />3.968254e-5 weeks <br />9.132e-6 months <br /> value. In summary, the TRM will continue to provide the appropriate administrative controls to ensure that the assumed FHA decay time will be met.

A dose calculation for an FHA occurring prior to the 24 hour2.777778e-4 days <br />0.00667 hours <br />3.968254e-5 weeks <br />9.132e-6 months <br /> decay time is unnecessary.

Question 5:

Response

If a fan is not operating within the reactor building (or shuts off due to Loss-of Offsite Power or single failure), would this result in the release occurring over a period longer than 2 hours2.314815e-5 days <br />5.555556e-4 hours <br />3.306878e-6 weeks <br />7.61e-7 months <br />, and would it result in a higher control room operator dose than the analysis provided? If the release occurred over 2 hours2.314815e-5 days <br />5.555556e-4 hours <br />3.306878e-6 weeks <br />7.61e-7 months <br /> without the reactor building fan operating, would it result in a larger dose?

Regulatory Guide 1.183 requires the design basis assumption of a 2-hour release regardless of whether there is forced ventilation or not. Holding all other variables constant (including the assumed CREFS actuation time within I minute), a release occurring over a longer period of time will result in lower Control Room doses. NPPD has assessed the ability of the Reactor Building Exhaust Plenum Radiation Monitor to provide the necessary CREFS initiation signal for the AST FHA. For personnel comfort considerations on the Refueling Bridge during fuel handling operations, the Reactor Building Heating and Ventilating System will typically be in service, which includes at least one exhaust fan. A LOOP is not assumed for accidents occurring during Modes 4 or 5 per the CNS TS Bases for AC Power Sources.' However, if a single failure occurs that disables the operating Reactor Building exhaust fan, a coastdown air flow will result that is sufficient (even for the 7 day decay time case) for the Radiation Monitor to provide the credited high radiation CREFS actuation within the assumed one minute time period after the design basis FHA. Since this setpoint assessment is predicated on having ventilation flow in the exhaust plenum at the start of the FHA, NPPD commits to the following prior to conducting fuel handling operations when less than a 7 day decay time has elapsed: 1) a Reactor Building exhaust fan or SGT System fan is in operation, or

2) CREFS is in operation.

Based on the above discussion, the Control Room dose consequences of are bounding over release periods that exceed 2 hours2.314815e-5 days <br />5.555556e-4 hours <br />3.306878e-6 weeks <br />7.61e-7 months <br />.

1.

Nonetheless, a LOOP occurring concurrent with the FHA would deenergize the Group 6 PCIS logic, and cause a CREFS initiation signal to occur.

NLS2005075 Page 52 of 53 Question 6: A TS change is proposed that will not require Secondary Containment operability during the movement of fuel assemblies that have not been "recently" irradiated.

The FHA analysis assumes the release to the control room intake and the environment is through the Reactor Building vent. Justify that that release point is an appropriately conservative assumption given that the Secondary Containment may be inoperable.

Response

As discussed in Question 2, the NPPD approach to Secondary Containment breach control during fuel handling operations ensures that breaches can be sealed following an FHA, and furthermore, that a train of SGT can be placed in service to provide a filtered, monitored, and elevated release. Thus, the RG 1.183 assumption of a 2-hour release to the environment is only plausible if there is Reactor Building exhaust fan flow via the Reactor Building vent. As discussed in Question 5, ventilation exhaust flow will be in service during fuel handling operations to ensure CREFS initiation.

If there were no fan flow, there would not be a significant driving force from the refuel floor to the environment for the FHA release. A release from the Reactor Building railroad airlock doors was also considered under these conditions, but physical interlocks and security considerations prevent both Reactor Building railroad airlock doors from being open concurrently. Therefore, without forced ventilation flow, it would be expected that any FHA radioactive release would occur at a much slower rate than with forced ventilation flow via the Reactor Building vent. Also, as discussed in Question 5, fuel handling operations involving irradiated fuel with less than a 7 day decay time during periods without Reactor Building exhaust fan flow or without an Operable Secondary Containment with a SGT system in operation would require CREFS to be in operation, further mitigating the dose consequences of an FHA under these conditions. Accordingly, consistent with regulatory precedent (References 7.2-1 and 7.2-2) a ground level release from the Reactor Building vent when Secondary Containment and the SGT System are inoperable, or when a Reactor Building exhaust fan is not in operation, provides an appropriately conservative release point based on the information provided above.

Question 7:

TS changes are proposed that eliminate the operability requirements for Secondary Containment and Standby Gas Treatment System operability during the movement of fuel assemblies that have not been "recently" irradiated. How will the intent of GDC 61 be met for controlling containment releases via confinement or filtering, and GDC 64 in monitoring releases?

Response

As discussed in Section 5.2, the construction of CNS predated the 1971 issuance of 10 CFR 50 Appendix A. NPPD has made no commitments to meet 1971 GDCs 61 and 64, and there are no 1967 GDCs that provide analogous requirements for assuring monitored releases following an FHA. Notwithstanding

NLS2005075 Page 53 of 53 this, with a Reactor Building exhaust fan or SGT System fan in operation while moving recently irradiated fuel (see commitment in Question 5), a monitored release is assured. Otherwise, the NPPD Secondary Containment breach control strategy (see Question 2) is designed to ensure that the Secondary Containment function, including monitored and filtered release via the SGT System, can be restored following an FEHA.

NLS2005075 ATTACHMENT 2 Proposed Technical Specifications Markup Format Cooper Nuclear Station, Docket 50-298, DPR-46 Listing of Revised Pages TS Pages 3.3-57 3.3-63 3.6-32 3.6-33 3.6-34 3.6-36 3.6-38 3.6-39 3.6-40 3.7-8 3.7-9

Secondary Containment Isolation Instrumentation 3.3.6.2 Table 3.3.6.2-1 (page 1 of 1)

Secondary Containment Isolation Instrumentation APPLICABLE MODES OR REQUIRED OTHER CHANNELS SPECIFIED PER SURVEILLANCE ALLOWABLE FUNCTION CONDITIONS TRIP SYSTEM REQUIREMENTS VALUE

1.

Reactor Vessel Water 1,2,3, 4

SR 3.3.6.2.1

>-42 inches Level - Low Low (Level 2)

(a)

SR 3.3.6.2.2 SR 3.3.6.2.3 SR 3.3.6.2.4

2.

Drywell Pressure - High 1,2,3 4

SR 3.3.6.2.2

< 1.84 psig SR 3.3.6.2.3 SR 3.3.6.2.4

3.

Reactor Building Ventilation 1,2,3, 4

SR 3.3.6.2.1

<49 mR/lhr Exhaust Plenum (a),(b)

SR 3.3.6.2.2 Radiation - High SR 3.3.6.2.3 SR 3.3.6.2.4 (a)

During operations with a potential for draining the reactor vessel.

(b)

During CORE ALTERATIONS and durin movement of recently irradiated fuel assemblies In secondary containment.

Amendment 209 3.3-57 2M 019

CREF System Instrumentation 3.3.7.1 Table 3.3.7.1-1 (page 1 of 1)

Control Room Emergency Filter System Instrumentation APPLICABLE MODES OR REQUIRED OTHER CHANNELS SPECIFIED PER TRIP SURVEILLANCE ALLOWABLE FUNCTION CONDITIONS SYSTEM REQUIREMENTS VALUE

1. Reactor Vessel Water 1,2,3, 4

SR 3.3.7.1.1

> - 42 inches Level - Low Low (Level 2)

(a)

SR 3.3.7.1.2 SR 3.3.7.1.3 SR 3.3.7.1.4

2. Drywell Pressure - High 1,2,3 4

SR 3.3.7.1.2

< 1.84 psig SR 3.3.7.1.3 SR 3.3.7.1.4

3. Reactor Building Ventilation 1,2,3, 4

SR 3.3.7.1.1

<49 mR/hr Exhaust Plenum (a),(b)

SR 3.3.7.1.2 Radiation - High SR 3.3.7.1.3 SR 3.3.7.1.4 (a)

During operations with a potential for draining the reactor vessel.

(b)

During CORE ALTERATION nd duri.g movement of recentl irradiated fuel assemblies in the secondary containment.

Amendment 33M 3.3-63 VI vie 05

Secondary Containment 3.6.4.1 3.6 CONTAINMENT SYSTEMS 3.6.4.1 Secondary Containment LCO 3.6.4.1 APPLICABILITY:

The secondary containment shall be OPERABLE.

MODES 1, 2, and 3, During movement of recently irradiated fuel assemblies in the secondary containment, Bui egeGREALTERATIONS, During operations with a potential for draining the reactor vessel (OPDRVs).

ACTIONS CONDITION REQUIRED ACTION COMPLETION TIME A.

Secondary containment A.1 Restore secondary 4 hours4.62963e-5 days <br />0.00111 hours <br />6.613757e-6 weeks <br />1.522e-6 months <br /> inoperable in MODE 1, 2, containment to or 3.

OPERABLE status.

B.

Required Action and B.1 Be in MODE 3.

12 hours1.388889e-4 days <br />0.00333 hours <br />1.984127e-5 weeks <br />4.566e-6 months <br /> associated Completion Time of Condition A not AND met.

B.2 Be in MODE 4.

36 hours4.166667e-4 days <br />0.01 hours <br />5.952381e-5 weeks <br />1.3698e-5 months <br /> C. Secondary containment C.1

---

NOTE--------------

inoperable during LCO 3.0.3 is not movement of recently applicable.

irradiated fuel assemblies in the secondary containment, during CORE Suspend movement of Immediately ALTERATIONS, or during recently irradiated fuel OPDRVs.

assemblies in the secondary containment.

AND (continued)

Cooper 3.6-32 Amendment 4%

Secondary Containment 3.6.4.1 ACTIONS CONDITION REQUIRED ACTION COMPLETION TIME C. (continued)

C.2 Suspend CORE Immediately LTERATINES.

AND C.32 Initiate action to Immediately suspend OPDRVs.

SURVEILLANCE REQUIREMENTS SURVEILLANCE FREQUENCY SR 3.6.4.1.1 Verify secondary containment vacuum is 24 hours2.777778e-4 days <br />0.00667 hours <br />3.968254e-5 weeks <br />9.132e-6 months <br />

> 0.25 inch of vacuum water gauge.

SR 3.6.4.1.2 Verify all secondary containment equipment 31 days hatches are closed and sealed.

SR 3.6.4.1.3 Verify one secondary containment access door in 31 days each access opening is closed.

SR 3.6.4.1.4 Verify each SGT subsystem can maintain 18 months on a

> 0.25 inch of vacuum water gauge in the STAGGERED secondary containment for 1 hour1.157407e-5 days <br />2.777778e-4 hours <br />1.653439e-6 weeks <br />3.805e-7 months <br /> at a flow rate TEST BASIS

< 1780 cfm.

Cooper 3.6-33 Amendment 4--8

SCIVs 3.6.4.2 3.6 CONTAINMENT SYSTEMS 3.6.4.2 Secondary Containment Isolation Valves (SCIVs)

LCO 3.6.4.2 APPLICABILITY:

Each SCIV shall be OPERABLE.

MODES 1, 2, and 3, During movement of recently irradiated fuel assemblies in the secondary containment, During CORE ALTERATIONS, During operations with a potential for draining the reactor vessel (OPDRVs).

ACTIONS

-- - --- - --------- - -------- - ---------- - ---------- - ---- Iluk j r- 'n -------------------------- - ------- - --- -

1.

Penetration flow paths may be unisolated intermittently under administrative controls.

2.

Separate Condition entry is allowed for each penetration flow path.

3.

Enter applicable Conditions and Required Actions for systems made inoperable by SCIVs.

CONDITION REQUIRED ACTION COMPLETION TIME A.

One or more penetration A.1 Isolate the affected 8 hours9.259259e-5 days <br />0.00222 hours <br />1.322751e-5 weeks <br />3.044e-6 months <br /> flow paths with one SCIV penetration flow path by inoperable.

use of at least one closed and de-activated automatic valve, closed manual valve, or blind flange.

AND (continued)

Cooper 3.6-34 Amendment 47-8

SCIVs 3.6.4.2 ACTIONS (continued)

CONDITION REQUIRED ACTION COMPLETION TIME D. Required Action and D.1

---

NOTE------------

associated Completion LCO 3.0.3 is not Time of Condition A or B applicable.

not met during movement of recently irradiated fuel assemblies in the Suspend movement of Immediately secondary containment, recently irradiated fuel during GERE assemblies in the ALTERATIONS, or during secondary containment.

OPDRVs.

AND D.2 Suspend CORE Immediately ALTERATIONS.

AND D.82 Initiate action to suspend Immediately OPDRVs.

Cooper 3.6-36 Amendment No. -78

SGT System 3.6.4.3 3.6 CONTAINMENT SYSTEMS 3.6.4.3 Standby Gas Treatment (SGT) System LCO 3.6.4.3 APPLICABILITY:

Two SGT subsystems shall be OPERABLE.

MODES 1, 2, and 3, During movement of recently irradiated fuel assemblies in the secondary containment, 1-1 I-A I T E A XT I

^

Vou II Ifj%#a

_V I-r-L A

I "1EAI x

IVI M

1%0 During operations with a potential for draining the reactor vessel (OPDRVs).

ACTIONS CONDITION REQUIRED ACTION COMPLETION TIME A.

One SGT subsystem A.1 Restore SGT subsystem 7 days inoperable.

to OPERABLE status.

B.

Required Action and B.1 Be in MODE 3.

12 hours1.388889e-4 days <br />0.00333 hours <br />1.984127e-5 weeks <br />4.566e-6 months <br /> associated Completion Time of Condition A not AND met in MODE 1, 2, or 3.

B.2 Be in MODE 4.

36 hours4.166667e-4 days <br />0.01 hours <br />5.952381e-5 weeks <br />1.3698e-5 months <br /> C. Required Action and

---

NOTE---------

associated Completion LCO 3.0.3 is not applicable.

Time of Condition A not met during movement of recently irradiated fuel C.1 Place OPERABLE SGT Immediately assemblies in the subsystem in operation.

secondary containment; during OeGR OR ALTERATIONS, or during OPDRVs.

(continued)

Cooper 3.6-38 Amendment No. +78

SGT System 3.6.4.3 ACTIONS CONDITION REQUIRED ACTION COMPLETION TIME C. (continued)

C.2.1 Suspend movement of Immediately recently irradiated fuel assemblies in secondary containment.

AND C.2.2 Suspend CORE mmediately ALTERATIONS.

AND C.2.32 Initiate action to Immediately suspend OPDRVs.

D. Two SGT subsystems DA Enter LCO 3.0.3 Immediately inoperable in MODE 1 2, or 3.

E. Two SGT subsystems E.1

---

NOTE----------

inoperable during LCO 3.0.3 is not movement of recentlv applicable.

irradiated fuel assemblies in the secondary Suspend movement of Immediately containment-,during recently irradiated fuel REC ALT-ERATINOPS, or assemblies in during OPDRVs.

secondary containment.

AND (continued)

Cooper 3.6-39 Amendment No. -78

SGT System 3.6.4.3 ACTIONS CONDITION REQUIRED ACTION COMPLETION TIME E. (continued)

[.2 Suspend CORE

• mmediately ALTERATIONSo.

ANED E.82 Initiate action to Immediately suspend OPDRVs.

SURVEILLANCE REQUIREMENTS SURVEILLANCE FREQUENCY SR 3.6.4.3.1 Operate each SGT subsystem for > 10 continuous 31 days hours with heaters operating.

TEMPORARY NOTE-- - -

The next required performance of this SR may be delayed until the current cycle refueling outage, but no later than February 2, 2005. This temporary note expires upon startup from that refueling outage.

SR 3.6.4.3.2 Perform required SGT filter testing in accordance In accordance with the Ventilation Filter Testing Program (VFTP).

with the VFTP SR 3.6.4.3.3 Verify each SGT subsystem actuates on an actual 18 months or simulated initiation signal.

a--------TEMPORARY NOTE ----- ----- -------------- -----

The next required performance of this SR may be delayed until the current cycle refueling outage, but no later than February 2, 2005. This temporary note expires upon startup from that refueling outage.

SR 3.6.4.3.4 Verify the SGT units cross tie damper is in the 18 months correct position, and each SGT room air supply check valve and SGT dilution air shutoff valve can be opened.

Amendment M5 3.6-40 08/1 114

CREF System 3.7.4 3.7 PLANT SYSTEMS 3.7.4 Control Room Emergency Filter (CREF) System LCO 3.7.4 The CREF System shall be OPERABLE.

APPLICABILITY:

MODES 1, 2, and 3, During movement of recently irradiated fuel assemblies containment,

-\\

ant~f Aarm A Tl~l.

SO in the secondary 1JUI11 If I Vd I.

rL.

I I XI-I IA

vj_,

During operations with a potential for draining the reactor vessel (OPDRVs).

ACTIONS CONDITION REQUIRED ACTION COMPLETION A.

CREF System inoperable.

A.1 Restore CREF System to 7 days OPERABLE status.

B.

Required Action and B.1 Be in MODE 3.

12 hours1.388889e-4 days <br />0.00333 hours <br />1.984127e-5 weeks <br />4.566e-6 months <br /> associated Completion Time of Condition A not AND met in MODE 1, 2, or 3.

B.2 Be in MODE 4.

36 hours4.166667e-4 days <br />0.01 hours <br />5.952381e-5 weeks <br />1.3698e-5 months <br /> (continued)

Cooper 3.7-8 Amendment No. +76

CREF System 3.7.4 ACTIONS (continued)

CONDITION REQUIRED ACTION COMPLETION TIME C. Required Action and

---

NOTE------------

associated Completion LCO 3.0.3 is not applicable.

Time of Condition A not met during movement of recently irradiated fuel C.1 Suspend movement of recently Immediately assemblies in the irradiated fuel assemblies in secondary containment; the secondary containment.

during RE ALTERATIONS, or AND during OPDRVs.

,.2 Suspend CORE Immediately ALTERATIONS.

AND C.32 Initiate action to Immediately suspend OPDRVs.

SURVEILLANCE REQUIREMENTS SURVEILLANCE FREQUENCY SR 3.7.4.1 Operate the CREF System for > 15 minutes.

31 days (continued)

Cooper 3.7-9 Amendment No. ff8

NLS2005075 ATTACHMENT 3 Proposed Technical Specifications Final Typed Format Cooper Nuclear Station, Docket 50-298, DPR-46 Listing of Revised Pages TS Pages 3.3-57 3.3-63 3.6-32 3.6-33 3.6-34 3.6-36 3.6-38 3.6-39 3.6-40 3.7-8 3.7-9

Secondary Containment Isolation Instrumentation 3.3.6.2 Table 3.3.6.2-1 (page 1 of 1)

Secondary Containment Isolation Instrumentation APPLICABLE MODES OR REQUIRED OTHER CHANNELS SPECIFIED PER SURVEILLANCE ALLOWABLE FUNCTION CONDITIONS TRIP SYSTEM REQUIREMENTS VALUE

1.

Reactor Vessel Water 1,2,3, 4

SR 3.3.6.2.1

> -42 inches Level - Low Low (Level 2)

(a)

SR 3.3.6.2.2 SR 3.3.6.2.3 SR 3.3.6.2.4

2.

Drywell Pressure - High 1,2,3 4

SR 3.3.6.2.2

< 1.84 psig SR 3.3.6.2.3 SR 3.3.6.2.4

3.

Reactor Building Ventilation 1,2,3, 4

SR 3.3.6.2.1

<49 mR/hr Exhaust Plenum (a),(b)

SR 3.3.6.2.2 Radiation - High SR 3.3.6.2.3 SR 3.3.6.2.4 (a)

(b)

During operations with a potential for draining the reactor vessel.

During movement of recently irradiated fuel assemblies in secondary containment.

I Amendment 3.3-57

CREF System Instrumentation 3.3.7.1 Table 3.3.7.1-1 (page 1 of 1)

Control Room Emergency Filter System Instrumentation APPLICABLE MODES OR REQUIRED OTHER CHANNELS SPECIFIED PER TRIP SURVEILLANCE ALLOWABLE FUNCTION CONDITIONS SYSTEM REQUIREMENTS VALUE

1. Reactor Vessel Water 1,2,3, 4

SR 3.3.7.1.1

> -42 inches Level - Low Low (Level 2)

(a)

SR 3.3.7.1.2 SR 3.3.7.1.3 SR 3.3.7.1.4

2. Drywell Pressure - High 1,2,3 4

SR 3.3.7.1.2

< 1.84 psig SR 3.3.7.1.3 SR 3.3.7.1.4

3. Reactor Building Ventilation 1,2,3, 4

SR 3.3.7.1.1

<49 mR/hr Exhaust Plenum (a),(b)

SR 3.3.7.1.2 Radiation - High SR 3.3.7.1.3 SR 3.3.7.1.4 (a)

During operations with a potential for draining the reactor vessel.

(b)

During movement of recently irradiated fuel assemblies in the secondary containment.

I Amendment 3

3.3-63

Secondary Containment 3.6.4.1 3.6 CONTAINMENT SYSTEMS 3.6.4.1 Secondary Containment LCO 3.6.4.1 APPLICABILITY:

The secondary containment shall be OPERABLE.

MODES 1, 2, and 3, During movement of recently irradiated fuel assemblies in the secondary containment, During operations with a potential for draining the reactor vessel (OPDRVs).

ACTIONS CONDITION REQUIRED ACTION COMPLETION TIME A.

Secondary containment A.1 Restore secondary 4 hours4.62963e-5 days <br />0.00111 hours <br />6.613757e-6 weeks <br />1.522e-6 months <br /> inoperable in MODE 1, 2, containment to or 3.

OPERABLE status.

B.

Required Action and B.1 Be in MODE 3.

12 hours1.388889e-4 days <br />0.00333 hours <br />1.984127e-5 weeks <br />4.566e-6 months <br /> associated Completion Time of Condition A not AND met.

B.2 Be in MODE 4.

36 hours4.166667e-4 days <br />0.01 hours <br />5.952381e-5 weeks <br />1.3698e-5 months <br /> C. Secondary containment C.1

---

NOTE------------

inoperable during LCO 3.0.3 is not movement of recently applicable.

irradiated fuel assemblies in the secondary containment or during Suspend movement of Immediately OPDRVs.

recently irradiated fuel assemblies in the secondary containment.

AND (continued)

I Cooper 3.6-32 Amendment

Secondary Containment 3.6.4.1 ACTIONS CONDITION REQUIRED ACTION COMPLETION TIME C. (continued)

C.2 Initiate action to suspend Immediately OPDRVs.

SURVEILLANCE REQUIREMENTS SURVEILLANCE FREQUENCY SR 3.6.4.1.1 Verify secondary containment vacuum is 24 hours2.777778e-4 days <br />0.00667 hours <br />3.968254e-5 weeks <br />9.132e-6 months <br />

> 0.25 inch of vacuum water gauge.

SR 3.6.4.1.2 Verify all secondary containment equipment 31 days hatches are closed and sealed.

SR 3.6.4.1.3 Verify one secondary containment access door in 31 days each access opening is closed.

SR 3.6.4.1.4 Verify each SGT subsystem can maintain 18 months on a

> 0.25 inch of vacuum water gauge in the STAGGERED secondary containment for 1 hour1.157407e-5 days <br />2.777778e-4 hours <br />1.653439e-6 weeks <br />3.805e-7 months <br /> at a flow rate TEST BASIS

< 1780 cfm.

I Cooper 3.6-33 Amendment

SCIVs 3.6.4.2 3.6 CONTAINMENT SYSTEMS 3.6.4.2 Secondary Containment Isolation Valves (SCIVs)

LCO 3.6.4.2 APPLICABILITY:

Each SCIV shall be OPERABLE.

MODES 1, 2, and 3, During movement of recently irradiated fuel assemblies in the secondary containment, During operations with a potential for draining the reactor vessel (OPDRVs).

ACTIONS

---

----------------------------------NOTES---------

1.

Penetration flow paths may be unisolated intermittently under administrative controls.

2.

Separate Condition entry is allowed for each penetration flow path.

3.

Enter applicable Conditions and Required Actions for systems made inoperable by SCIVs.

CONDITION REQUIRED ACTION COMPLETION TIME A.

One or more penetration A.1 Isolate the affected 8 hours9.259259e-5 days <br />0.00222 hours <br />1.322751e-5 weeks <br />3.044e-6 months <br /> flow paths with one SCIV penetration flow path by inoperable.

use of at least one closed and de-activated automatic valve, closed manual valve, or blind flange.

AND (continued)

Cooper 3.6-34 Amendment

SCIVs 3.6.4.2 ACTIONS (continued)

CONDITION REQUIRED ACTION COMPLETION TIME D. Required Action and D.1

---

NOTE-------------

associated Completion LCO 3.0.3 is not Time of Condition A or B applicable.

not met during movement of recently irradiated fuel assemblies in the Suspend movement of Immediately secondary containment or recently irradiated fuel during OPDRVs.

assemblies in the secondary containment.

AND D.2 Initiate action to suspend Immediately OPDRVs.

I I

Cooper 3.6-36 Amendment No.

SGT System 3.6.4.3 3.6 CONTAINMENT SYSTEMS 3.6.4.3 Standby Gas Treatment (SGT) System LCO 3.6.4.3 APPLICABILITY:

Two SGT subsystems shall be OPERABLE.

MODES 1, 2, and 3, During movement of recently irradiated fuel assemblies in the secondary containment, During operations with a potential for draining the reactor vessel (OPDRVs).

ACTIONS CONDITION REQUIRED ACTION COMPLETION TIME A.

One SGT subsystem A.1 Restore SGT subsystem 7 days inoperable.

to OPERABLE status.

B.

Required Action and B.1 Be in MODE 3.

12 hours1.388889e-4 days <br />0.00333 hours <br />1.984127e-5 weeks <br />4.566e-6 months <br /> associated Completion Time of Condition A not AND met in MODE 1, 2, or 3.

B.2 Be in MODE 4.

36 hours4.166667e-4 days <br />0.01 hours <br />5.952381e-5 weeks <br />1.3698e-5 months <br /> C. Required Action and

---

NOTE-----------------

associated Completion LCO 3.0.3 is not applicable.

Time of Condition A not met during movement of recently irradiated fuel C.1 Place OPERABLE SGT Immediately assemblies in the subsystem in operation.

secondary containment or during OPDRVs.

OR (continued)

I Cooper 3.6-38 Amendment No.

SGT System 3.6.4.3 ACTIONS CONDITION REQUIRED ACTION COMPLETION TIME C. (continued)

C.2.1 Suspend movement of Immediately recently irradiated fuel assemblies in secondary containment.

AND C.2.2 Initiate action to Immediately suspend OPDRVs.

D. Two SGT subsystems D.1 Enter LCO 3.0.3 Immediately inoperable in MODE 1 2, or 3.

E. Two SGT subsystems E.1 NOTE----------

inoperable during LCO 3.0.3 is not movement of recently applicable.

irradiated fuel assemblies in the secondary Suspend movement of Immediately containment or recently irradiated fuel during OPDRVs.

assemblies in secondary containment.

AND (continued)

I I

I Cooper 3.6-39 Amendment No.

SGT System 3.6.4.3 ACTIONS CONDITION REQUIRED ACTION COMPLETION TIME E. (continued)

E.2 Initiate action to Immediately suspend OPDRVs.

SURVEILLANCE REQUIREMENTS SURVEILLANCE FREQUENCY SR 3.6.4.3.1 Operate each SGT subsystem for > 10 continuous 31 days hours with heaters operating.

TEMPORARY NOTE-- --------------

The next required performance of this SR may be delayed until the current cycle refueling outage, but no later than February 2, 2005. This temporary note expires upon startup from that refueling outage.

SR 3.6.4.3.2 Perform required SGT filter testing in accordance In accordance with the Ventilation Filter Testing Program (VFTP).

with the VFTP SR 3.6.4.3.3 Verify each SGT subsystem actuates on an actual 18 months or simulated initiation signal.

---

a--------------TEMPORARY NOTE --- ----- ------------------- ---

The next required performance of this SR may be delayed until the current cycle refueling outage, but no later than February 2, 2005. This temporary note expires upon startup from that refueling outage.

SR 3.6.4.3.4 Verify the SGT units cross tie damper is in the 18 months correct position, and each SGT room air supply check valve and SGT dilution air shutoff valve can be opened.

I Amendment 3.6-40

CREF System 3.7.4 3.7 PLANT SYSTEMS 3.7.4 Control Room Emergency Filter (CREF) System LCO 3.7.4 The CREF System shall be OPERABLE.

APPLICABILITY: MODES 1, 2, and 3, During movement of recently irradiated fuel assemblies in the secondary containment, During operations with a potential for draining the reactor vessel (OPDRVs).

I ACTIONS CONDITION REQUIRED ACTION COMPLETION A.

CREF System inoperable.

A.1 Restore CREF System to 7 days OPERABLE status.

B. Required Action and B.1 Be in MODE 3.

12 hours1.388889e-4 days <br />0.00333 hours <br />1.984127e-5 weeks <br />4.566e-6 months <br /> associated Completion Time of Condition A not AND met in MODE 1, 2, or 3.

B.2 Be in MODE 4.

36 hours4.166667e-4 days <br />0.01 hours <br />5.952381e-5 weeks <br />1.3698e-5 months <br /> (continued)

Cooper 3.7-8 Amendment No.

CREF System 3.7.4 ACTIONS (continued)

CONDITION REQUIRED ACTION COMPLETION TIME C. Required Action and

---

NOTE-----------

associated Completion LCO 3.0.3 is not applicable.

Time of Condition A not met during movement of recently irradiated fuel C.1 Suspend movement of recently Immediately assemblies in the irradiated fuel assemblies in secondary containment or the secondary containment.

during OPDRVs.

AND C.2 Initiate action to Immediately suspend OPDRVs.

SURVEILLANCE REQUIREMENTS SURVEILLANCE FREQUENCY SR 3.7.4.1 Operate the CREF System for > 15 minutes.

31 days (continued)

I Cooper 3.7-9 Amendment No.

NLS2005075 ATTACHMENT 4 Proposed Technical Specifications Bases Revisions Markup Format Cooper Nuclear Station, Docket 50-298, DPR-46 Listing of Revised Pages TS Bases Pages B 3.3-172 B 3.3-189 B 3.6-67 B 3.6-68 B 3.6-69 B 3.6-70 B 3.6.72 B 3.6-73 B 3.6-74 B 3.6-76 B 3.6-80 B 3.6-81 B 3.6-82 B 3.6-83 B 3.7-18 B 3.7-19 B 3.7-20

Secondary Containment Isolation Instrumentation B 3.3.6.2 BASES APPLICABILITY SAFETY ANALYSES, LCO, AND APPLICABILITY

3.

Reactor Building Ventilation Exhaust Plenum Radiation-High (continued) instrument failure can preclude the isolation function.

The Allowable Values are chosen to promptly detect gross failure of the fuel cladding.

The Reactor Building Ventilation Exhaust Plenum Radiation -

High Functions are required to be OPERABLE in MODES 1, 2, and 3 where considerable energy exists; thus, there is a probability of pipe breaks resulting in significant releases of radioactive steam and gas. In MODES 4 and 5, the probability and consequences of these events are low due to the RCS pressure and temperature limitations of these MODES; thus, these Functions are not required. In addition, the Functions are also required to be OPERABLE during CORE ALTERATIONS, OPDRVs, and movement of recently irradiated fuel assemblies in the secondary containment, because the capability of detecting radiation releases due to fuel failures (due to fuel uncovery or dropped fuel assemblies) must be provided to ensure that offsite dose limits are not exceeded. Due to radioactive decay. this Function is only required to isolate secondary containment during fuel handling accidents involving handling recently irradiated fuel (i.e.. fuel that has occupied Dart of a critical reactor core within the previous 24 hours2.777778e-4 days <br />0.00667 hours <br />3.968254e-5 weeks <br />9.132e-6 months <br />).

ACTIONS A Note has been provided to modify the ACTIONS related to secondary containment isolation instrumentation channels. Section 1.3, Completion Times, specifies that once a Condition has been entered, subsequent divisions, subsystems, components, or variables expressed in the Condition, discovered to be inoperable or not within limits, will not result in separate entry into the Condition. Section 1.3 also specifies that Required Actions of the Condition continue to apply for each additional failure, with Completion Times based on initial entry into the Condition.

However, the Required Actions for inoperable secondary containment isolation instrumentation channels provide appropriate compensatory measures for separate inoperable channels. As such, a Note has been provided that allows separate Condition entry for each inoperable secondary containment isolation instrumentation channel.

Cooper B 3.3-1 72 Re.v.

_ n

CREF System Instrumentation B 3.3.7.1 BASES APPLICABLE SAFETY ANALYSES, LCO, and APPLICABILITY

3. Reactor Building Ventilation Exhaust Plenum Radiation - High (continued)

The Reactor Building Ventilation Exhaust Plenum Radiation -

High Function is required to be OPERABLE in MODES 1, 2, and 3 and during movement of recently irradiated fuel assemblies in the secondary containment, GORE ALTERATIONS, and operations with a potential for draining the reactor vessel (OPDRVs), to ensure control room personnel are protected during a pipe break resulting in significant releases of radioactive steam and gas, fuel handling event, or vessel draindown event. During MODES 4 and 5, when these specified conditions are not in progress (e.g., eRE-ALTERATIONSOPDRVs), the probability of a pipe break resulting in significant releases of radioactive steam and gas or fuel damage is low; thus, the Function is not required. Due to radioactive decay. this Function is only required to initiate the CREF System during fuel handling accidents involving handling recently irradiated fuel (i.e.. fuel that has occupied Dart of a critical reactor core within the Drevious 7 days).

ACTIONS A Note has been provided to modify the ACTIONS related to CREF System instrumentation channels. Section 1.3, Completion Times, specifies that once a Condition has been entered, subsequent divisions, subsystems, components, or variables expressed in the Condition, discovered to be inoperable or not within limits, will not result in separate entry into the Condition. Section 1.3 also specifies that Required Actions of the Condition continue to apply for each additional failure, with Completion Times based on initial entry into the Condition. However, the Required Actions for inoperable CREF System instrumentation channels provide appropriate compensatory measures for separate inoperable channels. As such, a Note has been provided that allows separate Condition entry for each inoperable CREF System instrumentation channel.

A.1 Because of the diversity of sensors available to provide isolation signals and the common interface with the Secondary Containment isolation Instrumentation, allowable out of service time of 12 hours1.388889e-4 days <br />0.00333 hours <br />1.984127e-5 weeks <br />4.566e-6 months <br /> for Functions 1 and 2, and 24 hours2.777778e-4 days <br />0.00667 hours <br />3.968254e-5 weeks <br />9.132e-6 months <br /> for Function 3, has been shown to be acceptable (Refs. 5, 6, and 7) to permit restoration of any inoperable channel to Cooper B 3.3-189 A 'I i94A0 'I

Secondary Containment B 3.6.4.1 B 3.6 CONTAINMENT SYSTEMS B 3.6.4.1 Secondary Containment BASES BACKGROUND The function of the secondary containment is to contain, dilute, and hold up fission products that may leak from primary containment following a Design Basis Accident (DBA) to limit fission product release to the environment. In conjunction with operation of the Standby Gas Treatment (SGT) System and closure of certain valves whose lines penetrate the secondary containment, the secondary containment is designed to reduce the activity level of the fission products released to the environment and to limit fission products that are released during certain operations that take place inside primary containment, when primary containment is not required to be OPERABLE, or that take place outside primary containment.

The secondary containment is a structure that completely encloses the primary containment and those components that may be postulated to contain primary system fluid. This structure forms a control volume that serves to hold up and dilute the fission products. It is possible for the pressure in the control volume to rise relative to the environmental pressure (e.g., due to pump and motor heat load additions). To prevent ground level exfiltration while allowing the secondary containment to be designed as a conventional structure, the secondary containment requires support systems to maintain the control volume pressure at less than the external pressure. Requirements for these systems are specified separately in LCO 3.6.4.2, "Secondary Containment Isolation Valves (SCIVs)," and LCO 3.6.4.3, "Standby Gas Treatment (SGT)

System."

APPLICABLE SAFETY ANALYSIS There are two principal accidents for which credit is taken for secondary containment OPERABILITY. These are a loss of coolant accident (LOCA) (Ref. 1) and a fuel handling accident involving handling recently irradiated fuel (i.e., fuel that has occupied Dart of acritical reactor core within the previous 24 hours2.777778e-4 days <br />0.00667 hours <br />3.968254e-5 weeks <br />9.132e-6 months <br />) inside secondary containment (Ref. 2). The secondary containment performs no active function in response to each of these limiting events; however, its leak tightness is required to ensure that fission products entrapped within the secondary containment structure following secondary containment isolation will be treated by the SGT System prior to discharge to the environment.

Cooper B 3.6-67 1 i /e4101

Secondary Containment B 3.6.4.1 BASES APPLICABLE SAFETY ANALYSES Secondary containment satisfies Criterion 3 of 10 CFR 50.36(c)(2)(ii)

(Ref. 3).

LCO An OPERABLE secondary containment provides a control volume into which fission products that leak from primary containment, or are released from the reactor coolant pressure boundary components located in secondary containment, following secondary containment isolation can be processed prior to release to the environment. For the secondary containment to be considered OPERABLE, it must have adequate leak tightness to ensure that the required vacuum can be established and maintained.

APPLICABILITY In MODES 1, 2, and 3, a LOCA could lead to a fission product release to primary containment that leaks to secondary containment. Therefore, secondary containment OPERABILITY is required during the same operating conditions that require primary containment OPERABILITY.

In MODES 4 and 5, the probability and consequences of the LOCA are reduced due to the pressure and temperature limitations in these MODES. Therefore, maintaining secondary containment OPERABLE is not required in MODE 4 or 5 to ensure a control volume, except for other situations for which significant releases of radioactive material can be postulated, such as during operations with a potential for draining the reactor vessel (OPDRVs), during GORE ALTERATIONS, or during movement of recently irradiated fuel assemblies in the secondary containment. Due to radioactive decay. secondary containment is only required to be OPERABLE during fuel handling involving handling recently irradiated fuel (i.e.. fuel that has occupied part of critical reactor core within the previous 24 hours2.777778e-4 days <br />0.00667 hours <br />3.968254e-5 weeks <br />9.132e-6 months <br />).

ACTIONS A.l If secondary containment is inoperable, it must be restored to OPERABLE status within 4 hours4.62963e-5 days <br />0.00111 hours <br />6.613757e-6 weeks <br />1.522e-6 months <br />. The 4 hour4.62963e-5 days <br />0.00111 hours <br />6.613757e-6 weeks <br />1.522e-6 months <br /> Completion Time provides a period of time to correct the problem that is commensurate with the importance of maintaining secondary Cooper B 3.6-68 X4 qn 14 In

Secondary Containment B 3.6.4.1 BASES ACTIONS A.1 (continued) containment during MODES 1, 2, and 3. This time period also ensures that the probability of an accident (requiring secondary containment OPERABILITY) occurring during periods where secondary containment is inoperable is minimal.

B.1 and B.2 If secondary containment cannot be restored to OPERABLE status within the required Completion Time, the plant must be brought to a MODE in which the LCO does not apply. To achieve this status, the plant must be brought to at least MODE 3 within 12 hours1.388889e-4 days <br />0.00333 hours <br />1.984127e-5 weeks <br />4.566e-6 months <br /> and to MODE 4 within 36 hours4.166667e-4 days <br />0.01 hours <br />5.952381e-5 weeks <br />1.3698e-5 months <br />. The allowed Completion Times are reasonable, based on operating experience, to reach the required plant conditions from full power conditions in an orderly manner and without challenging plant systems.

C.1; and C.2-,nd-C3 Movement of recently irradiated fuel assemblies in the secondary containment, CORE ALTERATIONS and OPDRVs can be postulated to cause significant fission product release to the secondary containment.

In such cases, the secondary containment is the only barrier to release of fission products to the environment. CORE ALTERATIONS and Therefore. movement of recently irradiated fuel assemblies must be immediately suspended if the secondary containment is inoperable.

Suspension of these activities shall not preclude completing an action that involves moving a component to a safe position. Also, action must be immediately initiated to suspend OPDRVs to minimize the probability of a vessel draindown and subsequent potential for fission product release. Actions must continue until OPDRVs are suspended.

LCO 3.0.3 is not applicable in MODES 4 or 5. 1 lowever, since irradiated fiel assc lnmove maen Neam can c i

Lr in MODE 1, 2, ori3, Required Action C.1 has been modified by a Note stating that LCO 3.0.3 is not applicable. If moving recentl irradiated fuel assemblies while in MODE 4 or 5, LCO 3.0.3 would not specify any action. If moving recently irradiated fuel assemblies while in MODE 1, 2, or 3, the fuel movement is independent of reactor operations. Therefore, in either case, inability to suspend Cooper B 3.6-69 Revision 0

Secondary Containment B 3.6.4.1 BASES ACTIONS C.1, and C.2. and 6.3 (continued) movement of recently irradiated fuel assemblies would not be a sufficient reason to require a reactor shutdown.

SURVEILLANCE SR 3.6.4.1.1 REQUIREMENTS This SR ensures that the secondary containment boundary is sufficiently leak tight to preclude exfiltration under expected wind conditions.

Momentary transients on installed instrumentation due to gusty wind conditions are considered acceptable and are not cause for failure to meet this SR. The 24 hour2.777778e-4 days <br />0.00667 hours <br />3.968254e-5 weeks <br />9.132e-6 months <br /> Frequency of this SR was developed based on operating experience related to secondary containment vacuum variations during the applicable MODES and the low probability of a DBA occurring between surveillances.

Furthermore, the 24 hour2.777778e-4 days <br />0.00667 hours <br />3.968254e-5 weeks <br />9.132e-6 months <br /> Frequency is considered adequate in view of other indications available in the control room, including alarms, to alert the operator to an abnormal secondary containment vacuum condition.

SR 3.6.4.1.2 and SR 3.6.4.1.3 Verifying that secondary containment equipment hatches and one access door in each access opening are closed ensures that the infiltration of outside air of such a magnitude as to prevent maintaining the desired negative pressure does not occur. Verifying that all such openings are closed provides adequate assurance that exfiltration from the secondary containment will not occur. SR 3.6.4.1.2 also requires equipment hatches to be sealed. In this application, the term "sealed" has no connotation of leak tightness. Maintaining secondary containment OPERABILITY requires verifying one door in the access opening is closed. However, each secondary containment access door is normally kept closed, except when the access opening is being used for normal transient entry and exit or when maintenance is being performed on an access. The 31 day Frequency for these SRs has been shown to be adequate, based on operating experience, and is considered adequate in view of the other indications of door and hatch status that are available to the operator.

Cooper B 3.6-70 Revision 1

SCIVs B 3.6.4.2 B 3.6 CONTAINMENT SYSTEMS B 3.6.4.2 Secondary Containment Isolation Valves (SCIVs)

BASES BACKGROUND The function of the SCIVs, in combination with other accident mitigation systems, is to limit fission product release during and following postulated Design Basis Accidents (DBAs) (Refs. 1 and 2). Secondary containment isolation within the time limits specified for those isolation valves designed to close automatically ensures that fission products that leak from primary containment following a DBA, or that are released during certain operations when primary containment is not required to be OPERABLE or take place outside primary containment, are maintained within the secondary containment boundary.

The OPERABILITY requirements for SCIVs help ensure that an adequate secondary containment boundary is maintained during and after an accident by minimizing potential paths to the environment. These isolation devices consist of either passive devices or active (automatic) devices. Manual valves, de-activated automatic valves secured in their closed position (including check valves with flow through the valve secured), and blind flanges are considered passive devices.

Automatic SCIVs close on a secondary containment isolation signal to establish a boundary for untreated radioactive material within secondary containment following a DBA or other accidents.

Other penetrations are isolated by the use of valves in the closed position or blind flanges.

APPLICABLE The SCIVs must be OPERABLE to ensure the secondary containment SAFETY ANALYSES barrier to fission product releases is established. The principal accidents for which the secondary containment boundary is required are a loss of coolant accident (Ref. 3) and a fuel handling accident involving handling recently irradiated fuel (i.e.. fuel that has occupied Dart of a critical reactor core within the previous 24 hours2.777778e-4 days <br />0.00667 hours <br />3.968254e-5 weeks <br />9.132e-6 months <br />) (Ref. 4). The secondary containment performs no active function in response to either of these limiting events, but the boundary established by SCIVs is required to ensure that leakage from the primary containment is processed by the Standby Gas Treatment System (SGT) System following secondary containment isolation, before being released to the environment.

Cooper B 3.6-72

SCIVs B 3.6.4.2 BASES APPLICABLE SAFETY ANALYSES (continued)

Maintaining SCIVs OPERABLE with isolation times within limits ensures that fission products will remain trapped inside secondary containment following secondary containment isolation so that they can be treated by the SGT System prior to discharge to the environment.

SCIVs satisfy Criterion 3 of 10 CFR 50.36(c)(2)(ii) (Ref. 5).

LCO SCIVs form a part of the secondary containment boundary. The SCIV safety function is related to control of offsite radiation releases resulting from DBAs.

The power operated automatic isolation valves are considered OPERABLE when their isolation times are within limits and the valves actuate on an automatic isolation signal. The valves covered by this LCO are listed in Reference 6.

The normally closed isolation valves or blind flanges are considered OPERABLE when manual valves are closed or open in accordance with appropriate administrative controls, automatic SCIVs are de-activated and secured in their closed position, and blind flanges are in place.

These passive isolation valves or devices are listed in Reference 6.

APPLICABILITY In MODES 1, 2, and 3, a DBA could lead to a fission product release to the primary containment that leaks to the secondary containment.

Therefore, the OPERABILITY of SCIVs is required.

In MODES 4 and 5, the probability and consequences of these events are reduced due to pressure and temperature limitations in these MODES. Therefore, maintaining SCIVs OPERABLE is not required in MODE 4 or 5, except for other situations under which significant radioactive releases can be postulated, such as during operations with a potential for draining the reactor vessel (OPDRVs), during CORE Cooper B, 3.6-73 November 23, 2001

SCIVs B 3.6.4.2 BASES APPLICABILITY ALTERATIEN&, or during movement of recently irradiated fuel (continued) assemblies in the secondary containment. Moving recently irradiated fuel assemblies in the secondary containment may also occur in MODES 1, 2, and 3. Due to radioactive decay. SCIVs are only required to be OPERABLE during fuel handling involving handling recently irradiated fuel (i.e.. fuel that has occupied Dart of a critical reactor core within the previous 24 hours2.777778e-4 days <br />0.00667 hours <br />3.968254e-5 weeks <br />9.132e-6 months <br />).

ACTIONS The ACTIONS are modified by three Notes. The first Note allows penetration flow paths to be unisolated intermittently under administrative controls. These controls consist of stationing a dedicated operator, who is in continuous communication with the control room, at the controls of the isolation device. In this way, the penetration can be rapidly isolated when a need for secondary containment isolation is indicated.

The second Note provides clarification that, for the purpose of this LCO, separate Condition entry is allowed for each penetration flow path. This is acceptable, since the Required Actions for each Condition provide appropriate compensatory actions for each inoperable SCIV. Complying with the Required Actions may allow for continued operation, and subsequent inoperable SCIVs are governed by subsequent Condition entry and application of associated Required Actions.

The third Note ensures appropriate remedial actions are taken, if necessary, if the affected system(s) are rendered inoperable by an inoperable SCIV.

A. 1 and A.2 In the event that there are one or more penetration flow paths with one SCIV inoperable, the affected penetration flow path(s) must be isolated.

The method of isolation must include the use of at least one isolation barrier that cannot be adversely affected by a single active failure.

Isolation barriers that meet this criterion are a closed and de-activated automatic SCIV, a closed manual valve, and a blind flange. For penetrations isolated in accordance with Required Action A. 1, the device used to isolate the penetration should be the closest available device to secondary containment. The Required Action must be completed within the 8 hour9.259259e-5 days <br />0.00222 hours <br />1.322751e-5 weeks <br />3.044e-6 months <br /> Completion Time. The specified time period is reasonable considering the time required to Cooper B 3.6-74 Revision 0

SCIVs B 3.6.4.2 BASES ACTIONS B.1 (continued)

The Condition has been modified by a Note stating that Condition B is only applicable to penetration flow paths with two isolation valves. This clarifies that only Condition A is entered if one SCIV is inoperable in multiple penetrations.

C.1 and C.2 If any Required Action and associated Completion Time cannot be met, the plant must be brought to a MODE in which the LCO does not apply.

To achieve this status, the plant must be brought to at least MODE 3 within 12 hours1.388889e-4 days <br />0.00333 hours <br />1.984127e-5 weeks <br />4.566e-6 months <br /> and to MODE 4 within 36 hours4.166667e-4 days <br />0.01 hours <br />5.952381e-5 weeks <br />1.3698e-5 months <br />. The allowed Completion Times are reasonable, based on operating experience, to reach the required plant conditions from full power conditions in an orderly manner and without challenging plant systems.

D.1. D.2. and D.3 If any Required Action and associated Completion Time are not met, the plant must be placed in a condition in which the LCO does not apply. If applicable, CORE ALTERATIONS and the movement of recently irradiated fuel assemblies in the secondary containment must be immediately suspended. Suspension of these activities shall not preclude completion of movement of a component to a safe position. Also, if applicable, actions must be immediately initiated to suspend OPDRVs in order to minimize the probability of a vessel draindown and the subsequent potential for fission product release. Actions must continue until OPDRVs are suspended.

LCO 3.0.3 is not applicable while in MODE 4 or 5. I l-wever since irradiated fuel assembly movement can occur in MODE 1, 2, or 3, Required Action D.1 has been modified by a Note stating that LCO 3.0.3 is not applicable. If moving recently irradiated fuel assemblies while in MODE 4 or 5, LCO 3.0.3 would not specify any action. If moving fuel while in MODE 1, 2, or 3, the fuel movement is independent of reactor operations. Therefore, in either case, inability to suspend movement of irradiated fuel assemblies would not be a sufficient reason to require a reactor shutdown.

Cooper B 3.6-76 Coope B 36-76Marceh 0, 2000

SGT System B 3.6.4.3 BASES BACKGROUND (continued)

g.

A centrifugal fan.

The capacity of the SGT System is sufficient to reduce and maintain the reactor building at a subatmospheric pressure of -0.25 inches water gauge (under neutral wind conditions of greater than 2 mph but less than 5 mph) with an air infiltration rate of no more than 100% of the reactor building volume per day.

The demister is provided to remove entrained water in the air, while the electric heater reduces the relative humidity of the airstream to less than 70% (Ref. 2). The prefilter removes large particulate matter, while the HEPA filter removes fine particulate matter and protects the charcoal from fouling. The charcoal adsorber removes gaseous elemental iodine and organic iodides, and the final HEPA filter collects any carbon fines exhausted from the charcoal adsorber.

The SGT System automatically starts and operates in response to actuation signals indicative of conditions or an accident that could require operation of the system. Following initiation, both charcoal filter train fans start. Upon verification that both subsystems are operating, the redundant subsystem is normally shut down.

APPLICABLE SAFETY ANALYSES The design basis for the SGT System is to mitigate the consequences of a loss of coolant accident and fuel handling accidents involving handling recently irradiated fuel (i.e.. fuel that has occupied part of a critical reactor core within the previous 24 hours2.777778e-4 days <br />0.00667 hours <br />3.968254e-5 weeks <br />9.132e-6 months <br />) (Ref. 2). For all events analyzed, the SGT System is shown to be automatically initiated to reduce, via filtration and adsorption, the radioactive material released to the environment.

The SGT System satisfies Criterion 3 of 10 CFR 50.36(c)(2)(ii) (Ref. 3).

LCO Following a DBA, a minimum of one SGT subsystem is required to maintain the secondary containment at a negative pressure with respect to the environment and to process gaseous releases. Meeting the LCO requirements for two OPERABLE subsystems ensures operation of at least one SGT subsystem in the event of a single active failure. An OPERABLE SGT Cooper B 3.6-80 Revision -O

SGT System B 3.6.4.3 BASES LCO subsystem consists of a demister, prefilter, HEPA filter, charcoal (continued) adsorber, a final HEPA filter, exhaust fan, and associated ductwork, dampers, valves and controls.

When the required decay heat removal flow through the cross tie damper is not met, only ONE SGT subsystem may be considered OPERABLE.

APPLICABILITY In MODES 1, 2, and 3, a DBA could lead to a fission product release to primary containment that leaks to secondary containment. Therefore, SGT System OPERABILITY is required during these MODES.

In MODES 4 and 5, the probability and consequences of these events are reduced due to the pressure and temperature limitations in these MODES. Therefore, maintaining the SGT System in OPERABLE status is not required in MODE 4 or 5, except for other situations under which significant releases of radioactive material can be postulated, such as during operations with a potential for draining the reactor vessel (OPDRVs), drmig CORE ALTERAT-IENS, or during movement of re entl irradiated fuel assemblies in the secondary containment. Due to radioactive decay. the SGT System is only required to be OPERABLE during fuel handling involving handling recently irradiated fuel (i.e., fuel that has occupied part of a critical reactor core within the previous 24 hours2.777778e-4 days <br />0.00667 hours <br />3.968254e-5 weeks <br />9.132e-6 months <br />).

ACTIONS A. 1 With one SGT subsystem inoperable, the inoperable subsystem must be restored to OPERABLE status in 7 days. In this Condition, the remaining OPERABLE SGT subsystem is adequate to perform the required radioactivity release control function. However, the overall system reliability is reduced because a single failure in the OPERABLE subsystem could result in the radioactivity release control function not being adequately performed. The 7 day Completion Time is based on consideration of such factors as the availability of the OPERABLE redundant SGT subsystem and the low probability of a DBA occurring during this period.

B.1 and B.2 If the SGT subsystem cannot be restored to OPERABLE status within the required Completion Time in MODE 1, 2, or 3, the plant must be brought to a MODE in which the LCO does not apply. To achieve this status, the plant must be brought to at least MODE 3 within 12 hours1.388889e-4 days <br />0.00333 hours <br />1.984127e-5 weeks <br />4.566e-6 months <br /> and to MODE 4 within Cope B_.-1

~5 Cooper B 3.6-81 12105iO3

SGT System B 3.6.4.3 BASES ACTIONS B.1 and B.2 (continued) 36 hours4.166667e-4 days <br />0.01 hours <br />5.952381e-5 weeks <br />1.3698e-5 months <br />. The allowed Completion Times are reasonable, based on operating experience, to reach the required plant conditions from full power conditions in an orderly manner and without challenging plant systems.

C.1. C.2.1. and C.2.2-,and-6.2.3 During movement of recently irradiated fuel assemblies, in the secondary containment, during CRE ALTERATIONS, or during OPDRVs, when Required Action A.1 cannot be completed within the required Completion Time, the OPERABLE SGT subsystem should immediately be placed in operation. This action ensures that the remaining subsystem is OPERABLE, that no failures that could prevent automatic actuation have occurred, and that any other failure would be readily detected.

An alternative to Required Action C.1 is to immediately suspend activities that represent a potential for releasing a significant amount of radioactive material to the secondary containment, thus placing the plant in a condition that minimizes risk. If applicable, eeRE ALTERATIONS and movement of recently irradiated fuel assemblies must immediately be suspended. Suspension of these activities must not preclude completion of movement of a component to a safe position. Also, if applicable, actions must immediately be initiated to suspend OPDRVs in order to minimize the probability of a vessel draindown and subsequent potential for fission product release. Actions must continue until OPDRVs are suspended.

LCO 3.0.3 is not applicable in MODE 4 or 5. I iowevei, since irradiated fuel assembly movement can occur in MODE 1, 2, or 3, the Required Actions of Condition C have been modified by a Note stating that LCO 3.0.3 is not applicable. If moving recently irradiated fuel assemblies while in MODE 4 or 5, LCO 3.0.3 would not specify any action. If moving recently irradiated fuel assemblies while in MODE 1, 2, or 3, the fuel movement is independent of reactor operations. Therefore, in either case, inability to suspend movement of irradiated fuel assemblies would not be a sufficient reason to require a reactor shutdown.

Cooper B 3.6-82 Revision O

SGT System B 3.6.4.3 BASES ACTIONS D.1 (continued)

If both SGTS subsystems are inoperable in MODE 1, 2, or 3, the SGT system may not be capable of supporting the required radioactivity release control function. Therefore, actions are required to enter LCO 3.0.3 immediately.

E.1. and E.2-and-E.3 When two SGT subsystems are inoperable, if applicable, CORE ALTERATIONS and movement of recently irradiated fuel assemblies in secondary containment must immediately be suspended. Suspension of these activities shall not preclude completion of movement of a component to a safe position. Also, if applicable, actions must immediately be initiated to suspend OPDRVs in order to minimize the probability of a vessel draindown and subsequent potential for fission product release. Actions must continue until OPDRVs are suspended.

LCO 3.0.3 is not applicable while in MODE 4 or 5. I lowever, sice irradiated fuel assembly movement can XT inME) OE 1, 2, er "3 r~equired Action E.1 has been modified by a Note stating that LCO 3.0.3 is not applicable. If moving recently irradiated fuel assemblies while in MODE 4 or 5, LCO 3.0.3 would not specify any action. If moving recently irradiated fuel assemblies while in MODE 1, 2, or 3, the fuel movement is independent of reactor operations. Therefore, in either case, inability to suspend movement of recently irradiated fuel assemblies would not be a sufficient reason to require a reactor shutdown.

SURVEILLANCE SR 3.6.4.3.1 REQUIREMENTS Operating each SGT subsystem, including each filter train fan, for > 10 continuous hours ensures that both subsystems are OPERABLE and that all associated controls are functioning properly. It also ensures that blockage, fan or motor failure, or excessive vibration can be detected for corrective action. Operation with the heaters on for > 10 continuous hours every 31 days eliminates moisture on the adsorbers and HEPA filters. The 31 day Frequency was developed in consideration of the known reliability of Cooper B 3.6-83 Revision 0

CREF System B 3.7.4 BASES APPLICABLE SAFETY ANALYSES The ability of the CREF System to maintain the habitability of the control room is an explicit assumption for the safety analyses presented in the USAR, Chapters X and XIV (Refs. 1 and 2, respectively). The CREF System is assumed to operate following a loss of coolant accident and a fuel handling accident involving handling recently irradiated fuel (i.e.. fuel that has occupied part of a critical reactor core within the Previous 7 days).

The CREF System satisfies Criterion 3 of 10 CFR 50.36(c)(2)(ii) (Ref. 4).

LCO The CREF System is required to be OPERABLE, since total system failure could result in exceeding a dose of 5 rem to the control room operators in the event of a DBA.

The CREF System is considered OPERABLE when the individual components necessary to control operator exposure are OPERABLE.

The system is considered OPERABLE when its associated:

a.

Fans are OPERABLE (one supply fan, the emergency booster fan and the exhaust booster fan);

b.

HEPA filter and charcoal adsorber are not excessively restricting flow and are capable of performing their filtration functions; and

c.

Ductwork, valves, and dampers are OPERABLE, and air circulation can be maintained.

In addition, the control room boundary must be maintained, including the integrity of the walls, floors, ceilings, ductwork, and access doors, such that the pressurization limit of SR 3.7.4.4 can be met. However, it is acceptable for access doors to be open for normal control room entry and exit, and not consider it to be a failure to meet the LCO.

Cooper B 3.7-18

CREF System B 3.7.4 BASES APPLICABILITY In MODES 1, 2, and 3, the CREF System must be OPERABLE to control operator exposure during and following a DBA, since the DBA could lead to a fission product release.

In MODES 4 and 5, the probability and consequences of a DBA are reduced because of the pressure and temperature limitations in these MODES. Therefore, maintaining the CREF System OPERABLE is not required in MODE 4 or 5, except for the following situations under which significant radioactive releases can be postulated:

a.

During operations with potential for draining the reactor vessel (OPDRVs);

b.

During ECRE ALTERATIONS; and eb.

During movement of recently irradiated fuel assemblies in the secondary containment. Due to radioactive decay. the CREF System is only required to be OPERABLE during fuel handling involving handling recently irradiated fuel (i.e.. fuel that has occupied Dart of a critical reactor core within the previous 7 days).

ACTIONS A.1 The inoperable CREF System must be restored to OPERABLE status within 7 days. With the unit in this condition, there is no other system to perform control room radiation protection. The 7 day Completion Time is based on the low probability of a DBA occurring during this time period.

B.1 and B.2 In MODE 1, 2, or 3, if the inoperable CREF System cannot be restored to OPERABLE status within the associated Completion Time, the unit must be placed in a MODE that minimizes risk. To achieve this status, the unit must be placed in at least MODE 3 within 12 hours1.388889e-4 days <br />0.00333 hours <br />1.984127e-5 weeks <br />4.566e-6 months <br /> and in MODE 4 within 36 hours4.166667e-4 days <br />0.01 hours <br />5.952381e-5 weeks <br />1.3698e-5 months <br />. The allowed Completion Times are reasonable, based on operating experience, to reach the required unit conditions from full power conditions in an orderly manner and without challenging unit systems.

Cooper B 3.7-19 Revision 0

CREF System B 3.7.4 BASES ACTIONS C.1-.C.2-,and C.32 The Required Actions of Condition C are modified by a Note indicating that LCO 3.0.3 does not apply. If moving recentl irradiated fuel assemblies while in MODE 1, 2, or 3, the fuel movement is independent of reactor operations. Therefore, inability to suspend movement of recently irradiated fuel assemblies is not sufficient reason to require a reactor shutdown.

During movement of recently irradiated fuel assemblies in the secondary containment, during CORE ALTERATIONS, or during OPDRVs, if the inoperable CREF System cannot be restored to OPERABLE status within the required Completion Time, activities that present a potential for releasing radioactivity that might require isolation of the control room must be immediately suspended. This places the unit in a condition that minimizes risk.

If applicable, CORE ALTERATIONS and movement of recently irradiated fuel assemblies in the secondary containment must be suspended immediately. Suspension of these activities shall not preclude completion of movement of a component to a safe position. Also, if applicable, actions must be initiated immediately to suspend OPDRVs to minimize the probability of a vessel draindown and the subsequent potential for fission product release. Actions must continue until the OPDRVs are suspended.

SURVEILLANCE REQUIREMENTS SR 3.7.4.1 This SR verifies that the CREF System in a standby mode starts on demand and continues to operate. The system should be checked periodically to ensure that it starts and functions properly. As the environmental and normal operating conditions of this system are not severe, testing the system once every month provides an adequate check on this system. Since the CREF System does not contain heaters, the system need only be operated for > 15 minutes to demonstrate the function of the system. The 31 day Frequency is based on the known reliability of the equipment.

Cooper B 3.7-20 Revision 0

ATTACHMENT 3 LIST OF REGULATORY COMMITMENTS© Correspondence Number: NLS2005075 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 NPPD commits to implement the guidelines Within 30 days of of Section 11.3.6.5 of NUMARC 93-01, NLS200507501 issuance of the Revision 3, using a strategy outlined in the 575-license amendment.

Response to Appendix B, Question 2.

NPPD commits to the following prior to conducting fuel handling operations when Within 30 days of less than a 7 day decay time has elapsed: 1) issuance of the a Reactor Building exhaust fan or SGT NLS2005075-02 license amendment.

system fan is in operation, or 2) CREFS is in operation.

PROCEDURE 0.42 REVISION 18 PAGE 20 OF 25