ML040290120

License Amendment Request 201 to the Kewaunee Nuclear Power Plant Technical Specifications, Equipment Hatch and Control Room Post Accident Recirculation System.

ML040290120
Person / Time
Site:	Kewaunee
Issue date:	01/16/2004
From:	Coutu T Nuclear Management Co
To:	Document Control Desk, Office of Nuclear Reactor Regulation
References
NRC-04-006
Download: ML040290120 (61)

Text

- - 4 Committed to NulearExcellence Kewaunee Nuclear Power Plant Operated by Nuclear Management Company, LLC NRC-04-006 10 CFR 50.90 January 16, 2004 U.S. Nuclear Regulatory Commission ATTN: Document Control Desk Washington, DC 20555 Kewaunee Nuclear Power Plant DOCKET 50-305 LICENSE No. DPR-43 License Amendment Request 201 To The Kewaunee Nuclear Power Plant Technical Specifications, Equipment Hatch And Control Room Post Accident Recirculation System."

References:

1) Letter from John G. Lamb (NRC) to Thomas Coutu (NMC),

Kewaunee Nuclear Power Plant - Issuance of Amendment Regarding Implementation of Alternate Source Term (TAC NO.

MB4596), dated March 17, 2003 (Adams Accession NO.

ML030210062)

2) Letter from Thomas Coutu (NMC) to Document Control Desk (NRC), License Amendment Request 195, Application For Stretch Power Uprate For Kewaunee Nuclear Power Plant," dated May 22, 2003 (Adams Accession NO. ML031500724)

3) Letter from Edward Weinkam (NMC) to Document Control Desk (NRC), "Generic Letter 2003-01: Control Room Habitability -

Response To Commitments," dated November 25, 2003 (Adams Accession NO. ML033300162)

In accordance with 10 CFR 50.90, Nuclear Management Company, LLC, (NMC) submits an application for amendment to Facility Operating License No. DPR-43, "Kewaunee Nuclear Power Plant (KNPP)."

N490 Highway 42

• Kewaunee, Wisconsin 54216-9511 Ao03 Telephone: 920.388.2560 Aoo3

Docket 50-305 NRC-04-006 January 16,2004 Page 2 This license amendment application would revise KNPP Technical Specifications (TS) 3.8.a.1, "Containment Closure," TS 4.4, Containment Tests," and TS 3.12, "Control Room Post-Accident Recirculation System." TS 3.8.a.1 is being modified to allow the equipment hatch to be open during REFUELING OPERATIONS and/or during movement of irradiated fuel assemblies within containment. TS 4.4.g is being added to require verification of the ability to close the equipment hatch periodically during refueling operations. TS 3.12 is modified to include requirements for operability of the Control Room Post-Accident Recirculation System (CRPARS) during fuel handling operations in which the fuel that is being moved has been irradiated less than 30 days ago. Appropriate TS Bases changes are included to reflect the proposed changes.

The Plant Operation Review Committee and the Offsite Safety Review Committee have reviewed this amendment application. Attachments 1 through 6 provide the description of proposed license changes and assessment, existing marked-up TS pages, revised TS pages, proposed TS Bases changes (provided for information only), and summary of regulatory commitments made in this submittal. During a meeting with the NRC on November 6, 2003, the NRC requested NMC address questions sent to San Onofre Nuclear Generating Station (Adams Accession NO. ML033160331). Attachment 7 is NMC's response to those questions.

NMC requests approval of the proposed license amendment by August 27, 2004 with a 60-day implementation period to allow sufficient time for planning prior to KNPP's Refueling Outage scheduled for October 2004. Additionally NMC is currently examining rescheduling the performance of the control room envelope unfiltered inleakge measurement test (reference 3). NMC is reviewing the possibility of performing this test in July of 2004 or earlier.

This submittal is not risk informed and only makes those commitments contained in attachment 6. In development of this submittal technical specification task force travelers (TSTF's) 023, 051, 068, 287, and 441 were reviewed and incorporated as appropriate.

The proposed change involves a change to the Technical Specifications (TS) that would allow the equipment hatch to remain open during irradiated fuel movement within the containment and adds supportive control room post-accident recirculation system TS.

Allowing the equipment hatch to be open during REFUELING OPERATIONS and movement of irradiated fuel assemblies inside containment does not alter any plant equipment or operating practices in such a manner that the probability of an accident is increased. The proposed change does not involve the addition or modification of any plant equipment nor alter the design, configuration, or method of operation of the plant beyond the standard functional capabilities of the equipment. Having the equipment hatch open does not create the possibility of a new accident. Analysis demonstrates that the resultant doses associated with a fuel handling accident are well within the

Docket 50-305 NRC-04-006 January 16, 2004 Page 3 appropriate acceptance limits. This change removes a defense-in-depth barrier that the analysis did not credit but provides additional restrictions on fission product release.

Administrative provisions that facilitate closing the equipment hatch following an evacuation of the containment further reduces the offsite doses in the event of a fuel handling accident (FHA) and provides additional margin to the calculated offsite doses.

It has been determined that this amendment application does not involve a significant hazard consideration as determined per 10 CFR 50.92. Pursuant to 10 CFR 51.22(b),

no environmental assessment need be prepared in connection with the issuance of this amendment.

In accordance with 10 CFR 50.91, a copy of this application, with attachments, is being provided to the designated State Official. If you should have any questions regarding this submittal, please contact Mr. Gerald Riste at (920) 388-8424.

Summary of Commitments This letter makes the following new commitments:

1) Written procedures will be developed describing compensatory measures to be taken in the event of an intentional or unintentional entry into a condition where both trains of Control Room Post-Accident Recirculation System are inoperable.

2) Written procedures will be developed describing measures to be taken to close the containment equipment hatch inthe event of adverse weather conditions (Tomado Watch).

3) Administrative controls consisting of written procedures will be established that would require: 1) appropriate personnel are aware of the open status of the containment during movement of recently irradiated fuel or REFUELING OPERATIONS, 2) specified individuals and equipment are designated and readily available to close the equipment hatch following an evacuation that would occur in the event of a fuel handling accident, and 3) any obstructions (e.g.,

cables and hoses) that would prevent closure of an open equipment hatch can be quickly removed, and 4) Procedures to verify closure of the equipment hatch within 90 minutes to be completed prior to the start of refueling activities where the equipment hatch will be open.

4) NMC will inform the State and County Emergency Governments of this accident scenario.

5) Accident scenario(s) will be developed to train the emergency response organization on this scenario.

6) Training will be developed to train those designated to close the equipment hatch.

Docket 50-305 NRC-04-006 January 16, 2004 Page 4 I declare under penalty of perjury that the foregoing is true and correct.

Executed on January 16, 2004.

Thomas Coutu Site Vice-President, Kewaunee Plant Nuclear Management Company, LLC GOR Attachments:

1) Safety Analysis

2) TS Markup Pages

3) TS Revised Pages

4) TS Basis Markup Pages

5) TS Basis Revised Pages

6) NMC Commitments

7) NMC Responses to request for additional information sent to San Onofre Nuclear Generating Station cc- Administrator, Region IlIl, USNRC Project Manager, Kewaunee Nuclear Power Plant, USNRC Senior Resident Inspector, Kewaunee Nuclear Power Plant, USNRC Electric Division, PSCW

ATTACHMENT 1 NUCLEAR MANAGEMENT COMPANY, LLC KEWAUNEE NUCLEAR PLANT DOCKET 50-305 January 16, 2004 Letter from Thomas Coutu (NMC)

To Document Control Desk (NRC)

License Amendment Request 201 Description of the Proposed Change Safety Analysis Significant Hazards Determination Environmental Considerations 19 Pages to Follow

Docket 50-305 NRC-04-006 January 16, 2004 , page 1 EVALUATION

1.0 INTRODUCTION

During a typical refueling outage, one of the factors used in scheduling the outage is the availability of the equipment hatch. Current technical specifications require the equipment hatch to be closed during refueling operations[U, any operation involving movement of reactor vessel internal components (those that could affect the reactivity of the core) within the containment when the vessel head is unbolted or removed. Because of the limitations on space within the containment the requirement to have the equipment hatch closed during fuel movement results in some activities in containment hinging on the scheduling of movement of equipment in and out of the equipment hatch, finding storage locations for equipment, rearranging equipment to make room for maintenance, and removing trash and replaced components. If the containment equipment hatch could remain open during movement of recently irradiated fuel, the extra time would allow equipment to be moved into and out of the containment other than during critical path time and in a more schedule efficient manner.

The proposed changes permit the optimization of outages to achieve an overall risk reduction while also reducing the outage time and cost. A significant contributor to this risk reduction is the ability to postpone operations early in the outage that, from a practical standpoint to achieve short outage duration, must be performed soon after shutdown when there is no TS requirement for a closed containment. The proposed change will allow some of these operations to be accomplished later, when the reactor vessel is open and the core is covered by the TS required level of water at which time the risk of a severe core damage accident is low.

In the event of a fuel handling accident (FHA) inside containment, an open equipment hatch will be the most limiting containment opening with respect to establishment of containment closure.

In order to minimize the impact on the health and safety of the public, equipment hatch closure, as well as closure of the personnel air lock and other penetrations, will be completed within the accident release timeframe assumed in the offsite dose analyses (2 hours2.314815e-5 days <br />5.555556e-4 hours <br />3.306878e-6 weeks <br />7.61e-7 months <br />).

The condition of the open equipment hatch will be monitored during irradiated fuel movement inside containment to assure closure of the equipment hatch door following containment evacuation. The assurance that the open equipment hatch will remain capable of closure will be administratively controlled in site procedures, similar to the controls applied to the personnel air lock and other containment penetrations. For example, any items passing through the equipment hatch that could obstruct closure of the door will have either quick disconnect capability or will be readily removable.

The savings gained from leaving the equipment hatch open during a refueling outage would be due to greater efficiency in the scheduling of refueling activities and could result in significant cost savings over the life of the plant.

The detailed description of what constitutes equipment hatch closure is already located in the Bases, and in greater detail. Therefore, the LCO phrase, "and held in place by four bolts is not included since this level of detail does not meet the inclusion requirements of 10 CFR 50.36.

Docket 50-305 NRC-04-006 January 16, 2004 , page 2 To allow for the equipment hatch to be open during refueling operations, NMC has analyzed the consequences from a Fuel Handling accident with no credit taken for the containment structure.

This analysis, performed for KNPP LAR 195, showed acceptable results without the containment structure.

2.0 DESCRIPTION

OF PROPOSED AMENDMENT This application is a request to amend Operating License DPR-43 for the Kewaunee Nuclear Power Plant (KNPP).

This amendment application would revise Technical Specifications (TS) 3.8.a.1, "Containment Closure," and TS 4.4, "Containment Tests" to allow the equipment hatch to be open during REFUELING OPERATIONS and/or during movement of irradiated fuel assemblies within containment. Additionally, the Bases are revised to clarify the administrative controls associated with the allowance to maintain the equipment hatch open and other appropriate TS Bases changes are included to reflect the proposed changes. The proposed changes will permit the optimization of outages to achieve an overall risk reduction while also reducing outage time and cost. A significant contributor to this risk reduction is the ability to postpone operations early in the outage that, from a practical standpoint to achieve a short outage time, must be performed soon after shutdown when there is no TS requirement for a closed containment. The proposed changes will allow some of these operations to be accomplished later, when the reactor vessel is open and covered by 23 feet of water at which time the risk of a severe core damage accident is very low.

Additionally TS 3.12, "Control Room Post Accident Recirculation System," is modified to add limiting conditions for operation (LCO) for the system during refueling operations and reformat the operating LCO's to the general format associated with the majority of KNPP TS. TS section 4.17, "Control Room Post Accident Recirculation System," was reviewed to determine if changes should be made to the surveillance requirements for the system; no changes were necessary because these TS requirements adequately cover the additional LCO requirements for refueling operations.

Specifically, TS 3.8.a.1.a, is being revised in four ways: 1) to state that the equipment hatch shall be capable of being closed within 90 minutes, 2) the 30-minute requirement to close the personnel air lock is being changed to 90 minutes, 3) the requirement to have at least o'ne door in each personnel air lock and the equipment hatch closed during the reactor vessel head lift is being removed, and 4) the allowance for penetration flow path(s) providing direct access from the containment atmosphere to the outside atmosphere to be unisolated under administrative controls is being added. Footnote number one is being revised to add the equipment hatch to the administrative controls description.

TS 3.12.a, is being revised to: 1) reformat to follow the standard convention generally found throughout KNPP's TS, 2) add a completion time to shutdown the reactor when the allowed LCO's cannot be met, and 3) add a LCO with a 24-hour allowed outage time for the condition where both trains of Control Room Post Accident Recirculation are out-of-service.

TS 3.12.b, is being added to require the Post Accident Recirculation trains to be operable during refueling operations with applicable LCO's and allowed outage times.

Docket 50-305 NRC-04-006 January 16, 2004 , page 3 TS 4.4.g, is being added to require verification of capability to close the equipment hatch when open during refueling operations.

3.0 BACKGROUND

The equipment hatch, which is part of the containment pressure boundary, provides a means for moving large equipment and components into and out of containment. Technical Specification 3.8.a.1, "Containment Closure," requires that the equipment hatch be closed during REFUELING OPERATIONS. This requirement ensures that a release of fission products within the containment will be restricted from escaping to the environment.

As described in Section 5.2.1, uReactor Containment Vessel Design," of the Updated Safety Analysis Report (USAR), the equipment hatch is 21 feet! 2] in diameter, supported entirely by the Reactor Containment Vessel and is not connected either directly or indirectly to any other structure. The equipment hatch was fabricated from welded steel and furnished with a double-gasketed flange and bolted dished door. Provision is made to pressure-test the space between the double gaskets of its flange. A moveable missile shield, a part of the shield building, is provided on the outside of the reactor building to protect the equipment hatch. During shutdown conditions, administrative controls ensure that an appropriate missile barrier is in place during the threat of severe weather that could result in the generation of tornado driven missiles.

The equipment hatch is removed by loosening 12 swing bolts until the bolts can be swung free.

Two chains are then used to remove the equipment hatch and place it in a storage position.

The first chain operates a mechanism actuator) that moves the hatch and the beam away from the hatch flange while the second chain operates trolleys to place the equipment hatch in the stored position(3 1(see attachment 7 for diagram).

The current radiological consequence analysis for the postulated design-basis FHA is based on the accident source term allowed by 10 CFR §50.67 and is described in KNPP USAR Section 14.2.1 and approved by the NRC 4]. The NMC evaluated the radiological consequences of a postulated FHA in the containment with no credit taken for containment isolation implementing the alternate source term (AST). Since the assumptions and parameters used for a FHA inside containment are identical to those for a FHA in the auxiliary building, the resulting radiological consequences are the same regardless of the location of the accident. NMC concluded that the radiological consequences resulting from the postulated FHA in the containment with no credit taken for containment isolation are within the dose acceptance criteria specified in SRP 15.0.1, "Radiological Consequence Analyses Using Alternative Source Terms," and GDC 19. The NMC reached this conclusion as a result of: (1) using the guidance provided in Appendix B to RG 1.183, "Assumptions for Evaluating the Radiological Consequences of a Fuel Handling Accident," (2) taking no credit for containment isolation, (3) taking no credit for removal of fission products by the spent fuel pool ventilation system in the auxiliary building, (4) using an overall decontamination factor of 200 for iodine in elemental and particulate forms in the spent fuel pool water with minimum water depth of 23 feet consistent with the guidelines provided in RG 1.183, (5) releasing all fission products within 2 hours2.314815e-5 days <br />5.555556e-4 hours <br />3.306878e-6 weeks <br />7.61e-7 months <br /> using an exponential release model with higher release in the initial period, (6) assuming all fuel rods in one fuel assembly with an axial power peaking factor of 1.7 are damaged to the extent that the entire gap activity inventory of the damaged fuel rods is released to the surrounding water, and (7) using a fission product decay

Docket 50-305 NRC-04-006 January 16, 2004 , page 4 period of 100 hours0.00116 days <br />0.0278 hours <br />1.653439e-4 weeks <br />3.805e-5 months <br /> (assumed time period from the reactor shutdown to the first fuel movement).

The NRC staff reviewed the NMC's methods, parameters, and assumptions used in its radiological dose consequence analyses and determined that they are consistent with the conservative guidance provided in RG 1.183. To verify NMC's radiological consequence assessments, the NRC staff performed confirmatory radiological consequence dose calculations for the postulated FHA. The radiological consequences calculated by the NRC staff are well within the dose criterion specified in GDC 19 (5 rem TEDE in the control room), and meet the dose acceptance criterion specified in the SRP 15.0.1 (6.3 rem TEDE at the EAB). Even though the NRC staff performed its confirmatory dose calculations, the NRC staff's acceptance was based on NMC's analyses.. The radiological consequences at the EAB, and the LPZ, and in the control room as calculated by NMC are all well within the dose criterion specified in GDC 19 and meet the dose acceptance criterion specified in the SRP 15.0.1. Therefore, the NRC staff concluded that the proposed AST implementation revising the current design-basis radiological consequence analysis for the postulated FHA is acceptable.

4.0 REGULATORY REQUIREMENTS AND GUIDANCE The regulatory basis for TS 3.8.a.1, "Containment Closure," is to ensure that the primary containment is capable of containing fission product radioactivity that may be released from the reactor core following a fuel handling accident inside containment. This ensures that offsite radiation exposures are maintained well within the requirements of 10 CFR 50.67.

The US Atomic Energy Commission (AEC) issued their Safety Evaluation of the Kewaunee Nuclear Power Plant on July 24, 1972 with supplements dated December 18, 1972 and May 10, 1973. In the AEC's, section 3.1, "Conformance with AEC General Design Criteria," described the conclusions the AEC reached associated with the General Design Criteria in effect at the time. The AEC stated:

The Kewaunee plant was designed and constructed to meet the intent of the AEC's General Design Criteria, as originally proposed in July 1967. Construction of the plant was about 50% complete and the Final Safety Analysis Report (Amendment No. 7) had been filed with the Commission before publication of the revised General Design Criteria in Febrary 1971 and the present version of the criteria in July 1971. As a result, we did not require the applicant to reanalyze the plant or resubmit the FSAR. However, our technical review did assess the plant against the General Design Criteria now in effect and we are satisfied that the plant design generally conforms to the intent of these criteria.

As such the appropriate 10 CFR 50 Appendix A General Design Criteria are listed below with the associated criteria KNPP is licensed to from the Final Safety Analysis (Amendment 7),

which has been updated and now titled the Updated Safety Analysis Report (USAR). Below are listed the applicable 10 CFR Part 50, Appendix A, General Design Criterion (GDC) with the associated information for the KNPP USAR described afterward.

10 CFR Part 50, Appendix A, GDC 16, "Containment Design," requires that reactor containment and associated systems shall be provided to establish an essentially leak-tight barrier against the uncontrolled release of radioactivity to the environment and to assure that the containment

Docket 50-305 NRC-04-006 January 16, 2004 , page 5 design conditions important to safety are not exceeded for as long as the postulated accident conditions require.

KNPP USAR section 5.1 states that the Containment System is designed to provide protection for the public from the consequences of a Design Basis Accident as defined in KNPP USAR Section 14.3E5]

10 CFR Part 50, Appendix A, GDC 56, "Primary Containment Isolation," describes the isolation provisions that must be provided for lines that connect directly to the containment atmosphere and which penetrate primary reactor containment unless it can be demonstrated that the isolation provisions for a specific class of lines are acceptable on some other defined basis.

KNPP USAR section 5.3 states that penetrations that require closure for the containment function shall be protected by redundant valving and associated apparatus (GDC 53)16].

Isolation valves are provided as required for fluid system lines penetrating containment to assure that: a) Leakage through all fluid line penetrations not serving accident-consequence-limiting systems is minimized by a double-barrier. The double-barriers take the form of closed pipe systems, both inside and outside the Reactor Containment Vessel, and various types of isolation valves. The double barrier arrangement provides two reliable low leakage barriers between the Reactor Coolant System or containment atmosphere and the environment. The failure of any one barrier will not prevent suitable isolation; b) Fluid line penetrations normally serving accident-consequence-limiting systems can be isolated by manual action if the automatic system should malfunction; and c) No single credible failure or malfunction (expected fault condition) occurring in any active system component can result in loss-of-isolation or intolerable leakage.

An isolation actuation system is provided to close those automatically operated containment isolation valves in fluid line penetrations used during normal operation but not required for Engineered Safety Features functions. The automatic closure is initiated by a Safety Injection Signal or by manual initiation.

10 CFR Part 50, Appendix A, GDC 61, "Fuel Storage and Handling and Radioactivity Control,"

requires that the fuel storage and handling, radioactive waste, and other systems which may contain radioactivity shall be designed to assure adequate safety under normal and postulated accident conditions.

10 CFR Part 50, Appendix A, GDC 64, "Monitoring Radioactivity Releases," requires monitoring the reactor containment atmosphere, spaces containing components for recirculation of loss-of-coolant accident fluids, effluent discharge paths, and the plant environs for radioactivity that may be released from normal operations, including anticipated operational occurrences, and from postulated accidents.

KNPP USAR section 11.2, "Radiation Protection," contains the design criteria for the KNPP including, monitoring radioactivity releases, monitoring fuel and waste storage areas, and protection against radioactivity release from spent fuel and waste storage areas.m It states that monitoring radioactive releases means shall be provided for monitoring the containment atmosphere and the facility effluent discharge paths for radioactivity released from normal operations, from anticipated transients, and from accident conditions. An environmental monitoring program shall be maintained to confirm that radioactivity releases to the environs of

Docket 50-305 NRC-04-006 January 16, 2004 , page 6 the plant have not been excessive. For monitoring fuel and waste storage areas monitoring and alarm instrumentation shall be provided for fuel and waste storage and associated handling areas for conditions that might result in loss of capability to remove decay heat and to detect excessive radiation levels. And for and protection against radioactivity release from spent fuel and waste storage areas, provisions shall be made in the design of fuel and waste storage facilities such that no undue risk to the health and safety of the public could result from an accidental release of radioactivity.

U.S. NRC Regulatory Guide 1.183, "Alternative Radiological Source Terms for Evaluating Design Basis Accidents at Nuclear Power Plants,' is NRC guidance that describes a method acceptable to the NRC staff for licensee evaluation of the potential radiological consequences of a fuel handling accident.

NUREG-0800, "U. S. NRC Standard Review Plan," Section 15.0.1, provides guidance to the NRC staff for the review and evaluation of system design features and plant procedures provided for the mitigation of the radiological consequences of postulated fuel handling accidents using alternative source terms.

The parameters of concern and the acceptance criteria applied are based on the requirements of 10 CFR 50.67 as modified by Regulatory Guide 1.183 with respect to the calculated radiological consequences of a fuel handling accident and GDC 61 with respect to appropriate containment, confinement, and filtering systems as described in the KNPP USAR.

5.0 TECHNICAL ANALYSIS

The proposed changes would allow the equipment hatch to be open under administrative controls during reactor vessel head lift, REFUELING OPERATIONS and/or during movement of recently irradiated fuel assemblies within containment, provided that it is capable of being closed. Allowing the equipment hatch to be open during reactor vessel head lift, REFUELING OPERATIONS or movement of recent irradiated fuel raises the concern that radioactive materials could potentially be released through the open hatch and vented to the outside environment should accidents that involve fission product releases occur. Postulated accidents that could result in a release of radioactive material through the open hatch include a fuel handling accident that results in breaching of the fuel rod cladding, and a loss of residual heat removal (RHR) cooling event that leads to core boiling and uncovery. To provide the basis for justifying the proposed change, the concern with the potential radiological consequences of the two accidents that could result in a release of radioactive material through the open equipment hatch are discussed below.

Fuel Handling Accident During movement of recently irradiated fuel assemblies within containment, the most severe radiological consequences are anticipated to result from a fuel handling accident. The fuel handling accident is a postulated event that involves damage to irradiated fuel. Fuel handling accidents include dropping a single irradiated fuel assembly, or a handling tool or heavy object, onto other irradiated fuel assemblies causing all the rods in one fuel assemble to burst releasing 100% of the rod gap activity of iodines and noble gases.

Docket 50-305 NRC-04-006 January 16, 2004 , page 7 All activity released from the fuel pool is assumed to be released to the atmosphere in 2 hours2.314815e-5 days <br />5.555556e-4 hours <br />3.306878e-6 weeks <br />7.61e-7 months <br />, using an exponential release model with higher releases in the initial periods since this is conservative for the Control Room doses. No credit is taken for operation of the Spent Fuel Pool Ventilation System in the auxiliary building. No credit is taken for isolation of containment for the FHA in containment. Since the assumptions and parameters for a FHA inside containment are identical to those for a FHA in the auxiliary building, the radiological consequences are the same regardless of the location of the accident.

It is assumed that all of the fuel rods in the equivalent of one fuel assembly are damaged to the extent that all their gap activity is released. The assembly inventory is based on the assumption that the subject fuel assembly has been operated at 1.7 times the core average power. The decay time used in the analysis is 100 hours0.00116 days <br />0.0278 hours <br />1.653439e-4 weeks <br />3.805e-5 months <br />. Thus, the analysis is bounded by the Technical Specifications limit of 14818] hours decay time prior to fuel movement. The calculated offsite and control room operator doses are within the acceptance criteria of 10 CFR 50.67, as modified by Regulatory Guide 1.183 Table 6, and General Design Criteria (GDC) 19.

KNPP license amendment 150 changed the time the reactor was to be sub-critical before movement of irradiated fuel assemblies from 100 hours0.00116 days <br />0.0278 hours <br />1.653439e-4 weeks <br />3.805e-5 months <br /> to 148 hours0.00171 days <br />0.0411 hours <br />2.44709e-4 weeks <br />5.6314e-5 months <br />. This change was not due to limitations imposed by the FHA but due to limitations imposed by the heat removal capabilities of the spent fuel pool with its increased assembly handling capability. The NRC safety evaluation1 91 confirmed that the limiting case was a full core offload 148 hours0.00171 days <br />0.0411 hours <br />2.44709e-4 weeks <br />5.6314e-5 months <br /> after shutdown. Under these conditions the bulk temperature of the spent fuel pool remained less then 150 0F.

During refueling operations, the potential for containment pressurization as a result of a fuel handling accident is not likely. Therefore, the majority of the radioactive material releases from the accident would be held up inside containment with only a minimal amount of radioactive material released through the open equipment hatch. However, the combined dose consequences of this potential release with the releases through other unisolated penetration flow paths and the open personnel airlock doors, will be bounded by the current licensing basis fuel handling accident analysis"'01. The current design basis fuel handling analysis does not credit the containment building barriers or the Spent Fuel Pool Sweep System. It is assumed that all gap activity is released from the damaged rods and all the gaseous effluent escaping from the refueling pool is released directly to the environment within two hours. In addition, no credit is taken for mixing of the gaseous effluents with the surrounding building atmosphere and removal of any iodine by the atmosphere filtration system filters except by the Control Room Post Accident Recirculation system filters.

NMC has received approval from the NRC to use the Accident Source Term methodology (10 CFR 50.67) for analysis of off-site and control room post accident dose consequences"i. For the FHA the offsite-dose limit is 6.3 rem TEDE112 1, as stated in NRC Regulatory Guide 1.183.

This is -25 percent of the limit stated in 10 CFR 50.67. The limit for the Control Room dose is 5.0 rem TEDE per 10 CFR 50.67, this is also the limit stated on GDC 19. As stated in KNPP LAR 195113], the resulting offsite dose consequences were calculated to be 0.70 rem TEDE and 0.11 rem TEDE at the Site Boundary and Low Population Zone respectively, while the Control Room dose consequence is 1.00 rem TEDE. These results are well within the 10 CFR 50.67 limits. Since the total amount of radioactive material available for immediate release into the

Docket 50-305 NRC-04-006 January 16, 2004 , page 8 water during a postulated fuel handling accident will be the same, the potential dose consequences from a simultaneous release of the gaseous effluents through the unisolated penetration flow paths, the open personnel airlock doors and the open equipment hatch will not differ greatly from the previous analysis that assumes radioactivity to be released only through the open personnel airlock doors 14 1. Therefore, allowing the equipment hatch to be open during reactor vessel head lift, upper internal removal, REFUELING OPERATIONS or movement of recently irradiated fuel would not invalidate the conclusion that the potential dose consequences from a fuel handling accident will be well within the 10 CFR 50.67 guideline limits.

NMC is adding a TS restriction on operation if two control room post accident recirculation trains are out of service due to control room boundary failure. This restriction is contained in NUREG 1431, "Standard Technical Specifications Westinghouse Plants," revision 2, section 3.7.10.

TSTF 287 added this restriction to NUREG 1431, which stated:

Requiring the plant to enter LCO 3.0.3 (standard shutdown sequence) when the ventilation envelope is not intact is excessive and, in the case of the FBACS OR FSPVS, is not appropriate. Modeling these specifications on the Shield Building specification (NUREG-1431, LCO 3.6.19) for a Dual or Ice Condenser containment would provide consistency within the NUREG. NUREG-1 431 Specification 3.6.19 allows 24 hours2.777778e-4 days <br />0.00667 hours <br />3.968254e-5 weeks <br />9.132e-6 months <br /> to restore the envelope to Operable status before requiring an orderly shutdown from operating conditions (MODE 3 in 6 hours6.944444e-5 days <br />0.00167 hours <br />9.920635e-6 weeks <br />2.283e-6 months <br />, MODE 5 in 36 hours4.166667e-4 days <br />0.01 hours <br />5.952381e-5 weeks <br />1.3698e-5 months <br />). This would allow for routine repairs. The proposed change is acceptable because of the low probability of a DBA occurring during the 24 hour2.777778e-4 days <br />0.00667 hours <br />3.968254e-5 weeks <br />9.132e-6 months <br /> AOT. Furthermore, (modeling an allowance on the CIV allowance to intermittently open penetrations that are otherwise required to be closed), an LCO Note is added to allow intermittent opening (e.g. as for entering and exiting) without entering the Actions.

Therefore NMC requests to incorporate the changes associated with TSTF 287 into KNPP TS as the justification contained in TSTF 287 is applicable to KNPP.

Although analysis for the FHA at KNPP shows acceptable results without the containment structure, a restriction is placed on refueling operation such that the equipment hatch and the personnel air locks shall be capable of being closed within 90 minutes of indication of a FHA.

This restriction provides additional reduction in the dose consequences by ensuring the containment structure is intact prior to the time the analysis assumes the release continues (2 hours2.314815e-5 days <br />5.555556e-4 hours <br />3.306878e-6 weeks <br />7.61e-7 months <br />). Therefore, although not required for acceptable results, this restriction provides additional defense-in-depth to limit the consequences of the FHA.

Loss of RHR Cooling The release of radioactive material is anticipated to be insignificant as a result of core boil-off due to a loss of RHR cooling, if the event does not continue for an extended period of time resulting in core uncovery and subsequent core damage. If core boil-off continues, the compartments in the vicinity of the core could be pressurized and thereby provide a driving force for the containment atmosphere to be released via the open hatch flow path to the outside atmosphere. However, the radiological consequences of this release of radioactive materials due to core boil-off, with no consideration for core uncovery and core damage, is expected to be significantly less than the radiological consequences arising from a postulated fuel handling

Docket 50-305 NRC-04-006 January 16, 2004 , page 9 accident because the total coolant activity (corresponding to a 1 % fuel defect) is less than the total gap activities in the damaged rods at the earliest time fuel offloading may be commenced (analysis assumes 100 hours0.00116 days <br />0.0278 hours <br />1.653439e-4 weeks <br />3.805e-5 months <br /> after shutdown).

A review of calculations performed for the outage risk assessment revealed that the time to core boil would be greater than 5 hours5.787037e-5 days <br />0.00139 hours <br />8.267196e-6 weeks <br />1.9025e-6 months <br /> should a loss of RHR cooling event occur at the beginning of fuel off loading, based on the normal water level maintained in the refueling pool (i.e., > 23 ft above the top of the reactor vessel flange). Technical Specification 3.1.a.2.B.2 requires that corrective actions be taken immediately to restore the RHR cooling as soon as possible if RHR loop requirements are not met (by having one RHR loop operable). In addition, operators are required to close all containment penetrations providing direct access from the containment atmosphere to the outside environment within 4 hours4.62963e-5 days <br />0.00111 hours <br />6.613757e-6 weeks <br />1.522e-6 months <br />. If an operator takes actions to restore the RHR cooling capability or uses an alternative method of core cooling within the five hour time interval, the scenario involving core boiling and subsequent containment pressurization would not be present. With all penetrations closed within the specified time period, the potential for the coolant to boil and subsequently cause the release of radioactive gas to the containment atmosphere, if RHR cooling was not restored, would not be of concern.

During reactor vessel head lift, loss of core cooling is highly unlikely. KNPP TS require two trains of RHR to be operable during this evolution (TS 3.1.a.2.B), unless there is at least 23 feet of water above the reactor vessel flange. With 2 23 feet of water in the refueling canal one of the two trains of RHR may be out-of-service for maintenance. By utilization of two trains of RHR or one train of RHR and 2 23 feet of water in the refueling canal, no single active failure will cause a loss of core cooling. To prevent a loss of suction to the RHR pumps removal of the reactor vessel head is performed with the water level in the reactor vessel well above the level required for minimum net positive suction head (NPSH) for the RHR pumps. Operating procedures for draining of the reactor coolant system list associated reactor vessel levels.

Table 1 contains excerpts associated with the RHR system and reactor vessel head lift the Reactor Coolant System draining procedure.['5 1 TABLE 1 REACTOR VESSEL LEVELS (1%= 6.5 INCHES WATER COLUMN) 1 Vessel Level (%) Description 9.0% Inlet to RHR System from Hotleg 13.0% Limit RHR flow to < 1300 gpm 16.0% 36" below Reactor Vessel Flange - reduced inventory condition 20.6% 6" below the reactor vessel flange 21.5% Reactor Vessel Flange To prevent a loss of suction to the RHR pumps, the reactor vessel head is lifted when the water level in the reactor vessel is above 20.6% (6" below the flange). This provides 11.6% (75.4")

between the RHR System inlet from the hotleg and the reactor vessel water level when the reactor vessel head is lifted. It also provides 7.6% (49.4" of water) between the level at which the reactor vessel head is lifted and the level at which RHR system flow is limited to prevent cavitation from vortexing.

Docket 50-305 NRC-04-006 January 16, 2004 , page 10 Administrative Controls NUMARC 93-01[16] "Industry Guideline for Monitoring the Effectiveness of Maintenance at Nuclear Power Plants," Section 11.3.6.5, provides the following guidance:

".... for plants which obtain license amendments to utilize shutdown safety administrative controls in lieu of Technical Specification requirements on primary or secondary containment operability and ventilation system operability during fuel handling or core alterations, the following guidelines should be included in the assessment of systems removed from service:

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 proposed changes do not affect the OPERABILITY requirements for any ventilation system or radiation monitors, nor does it affect their availability. The Control Room Post-Accident Recirculation System will be required to be OPERABLE by TS 3.12, "Control Room Post-Accident Recirculation System (CRPAR)," as well as the containment purge and vent system and associated atmosphere radioactivity monitors (TS 3.8.a.8). The only affected containment penetration that provides direct access to the outside atmosphere is the equipment hatch.

Existing TS requirements on other penetrations that provide direct access are not affected.

Containment ventilation at KNPP is accomplished via the Containment Purge and Vent System.

Although this system is not credited in any of the dose analyses, there are associated TS OPERABILITY requirements for this system. These operability requirements ensure the Containment Purge and Vent systems ability for automatic containment isolation on high radiation in containment. The Containment Purge and Vent System operates to supply outside air into the containment for ventilation, climate control, and dilution as needed for prolonged containment access following a shutdown and during refueling. The Containment Post LOCA Hydrogen Control System is used during power operation to equalize internal and external pressures and may be used to reduce the concentration of noble gases within the containment prior to and during personnel access. Each penetration associated with the Containment Purge and Vent System and the Post LOCA Hydrogen Control System, used during operation, is equipped with two valves in series, one inside containment and one outside containment to perform its containment isolation function.

Docket 50-305 NRC-04-006 January 16, 2004 , page 11 The function of the Purge and Ventilation System, as described in the KNPP USAR section 5.4, is to provide fresh, tempered air for comfort during maintenance and refueling operations and to purge contaminated air from the Reactor Containment Vessel whenever required for access.

The Containment Vent Supply unit will furnish tempering of makeup air under shutdown conditions The Purge and Vent System is not used when the reactor is above the hot shutdown operating condition. The Purge and Vent Valves RBV-1, RBV-2, RBV-3 and RBV-4, see USAR Figure 5.4-1, are sealed closed when above hot shutdown because of NRC concerns over the ability of these valves to close under Design Basis LOCA conditions. '71 For entry at or below hot shutdown, the Reactor Containment Vessel may be vented using the Purge and Vent system to reduce the concentration of radioactive gases and airborne particulates. The health and safety of the public is assured by the quick closure of the purge, and vent valves in the event of high radiation in the containment system vent.

The Purge and Ventilation System is sized to provide a reduction of the radioactivity in the Reactor Containment Vessel air following normal full-power operation to the level defined by 10 CFR 20 for a 40 hour4.62963e-4 days <br />0.0111 hours <br />6.613757e-5 weeks <br />1.522e-5 months <br /> occupational work week, within 2-6 hours after reactor shutdown. Purging of the Reactor Containment Vessel will normally be accomplished within two hours following the beginning of purge. Provision is made in the design of the Purge and Ventilation System for 11/2 air changes per hour during refueling and maintenance operations.

The Containment Purge and Vent Subsystem consists of a fresh air supply and exhaust and filtration. Isolation valves are installed both inside and outside the Containment where ducts from this system penetrate Containment. Air can be exhausted from the Containment by either of two modes, exhaust or purge. The Containment System Vent provides for the discharge of air at an elevation near the top of the Shield Building within the influence of the building wake effect, to improve the dispersion of gaseous releases.

The Exhaust Mode provides a high rate of air circulation through the Containment Vent Filter Assembly that includes a pre-filter and a bank of HEPA (high efficiency particulate activity) filters and exhausts out the Reactor Bldg Discharge Vent to the atmosphere. This mode is used when the radiation level in Containment is below the limits defined in 10CFR20. The Containment air radioactive level is constantly monitored by means of Radiation Monitors R-1 1, R-12, or R-21.

High radiation, as detected by one of these monitors, generates a Containment Vent Isolation signal, shutting down and isolating the Containment Purge and Vent system.

The Purge mode of operating the Purge and Vent Subsystem is through a purge exhaust fan and purge exhaust filter assembly with additional prefilter, HEPA and charcoal filters. A smaller amount of Containment air is exhausted via the prefilter, HEPA filter, charcoal filter, purge exhaust fan, through the Containment Vent Filter Assembly, and then to the suction of the ventilation exhaust fan where it is mixed with fresh outside air. Any contamination that gets through the filters is significantly diluted before finally being discharged to atmosphere via the Reactor Building Discharge Vent. In either mode the same quantity of air is exhausted to the atmosphere. In the purge mode, only a fraction of the air is taken from Containment.

Both the exhaust and purge mode are controlled remotely from the Control Room. If Containment air activity increases above the setpoint of the Containment Atmosphere Radiation

Docket 50-305 NRC-04-006 January 16, 2004 , page 12 Monitors R-1 1, R-12, and/or the Containment Vent Radiation Monitor R-21, a Containment Vent Isolation signal closes the containment ventilation isolation dampers.

During REFUELING OPERATIONS, TS 3.8.a.6 requires that direct communications be maintained between the control room and personnel at the refueling station. Therefore, if a fuel handling accident were to occur inside containment, the control room would be immediately informed, and action would be promptly initiated in accordance with off-normal procedures to mitigate the consequences.

If open, the equipment hatch will be maintained in an isolable condition, and the TS and Bases provides the requirements for closure of the equipment hatch. Administrative controls consisting of written procedures will be established prior to the implementation of the proposed change.

These procedural controls would require:

1. Appropriate personnel are aware of the open status of the containment during movement of irradiated fuel or REFUELING OPERATIONS.

2. Specified individuals are designated and readily available to close the equipment hatch following an evacuation that would occur in the event of a fuel handling accident.

3. Any obstructions (e.g., cables and hoses) that would prevent rapid closure of an open equipment hatch can be quickly removed.

4. Necessary hardware, tools, and equipment required for closure are available.

These administrative controls provide protection equivalent to that afforded by the administrative controls used to establish containment closure for a containment personnel air lock. Outage shift/containment supervision is responsible for coordinating the equipment hatch closure activities. Personnel are designated for each shift during which REFUELING OPERATIONS and/or movement of irradiated fuel (with the equipment hatch open) will take place. While these personnel will have normal outage related duties, these duties will not interfere with their availability to respond to the closure of the equipment hatch. Personnel responsible for closure of the equipment hatch will receive training associated with the equipment hatch operation.

An assessment of the radiological consequences, as described above for the proposed changes, concludes that site boundary doses remain well within the 10 CFR 50.67 limits and control room doses meet GDC 19 criteria without taking credit for closure of the equipment hatch. The administrative controls provide reasonable assurance that containment closure as a defense-in-depth measure can be reestablished quickly to limit releases much lower than assumed in the dose calculation.

Risk Significance Based on the results of conservative dose calculations provided in this submittal, the risk to the health and safety of the public as a result of a fuel handling accident inside the containment with the equipment hatch open is minimal. Actual fuel handling accidents which have occurred in the past have resulted in minimal or no releases, which shows that the assumptions and methodology utilized in the radiological dose calculations are very conservative. Radioactive decay is a natural phenomenon. It has a reliability of 100 percent in reducing the radiological

Docket 50-305 NRC-04-006 January 16, 2004 , page 13 release from fuel bundles. In addition, the water level that covers the fuel bundles is another natural method that provides an adequate barrier to a significant radiological release. The requirement for at least 148 hours0.00171 days <br />0.0411 hours <br />2.44709e-4 weeks <br />5.6314e-5 months <br /> of decay prior to fuel movement (100 hours0.00116 days <br />0.0278 hours <br />1.653439e-4 weeks <br />3.805e-5 months <br /> is assumed in the accident analysis, 148 hours0.00171 days <br />0.0411 hours <br />2.44709e-4 weeks <br />5.6314e-5 months <br /> is based on heat removal capacity of the spent fuel pool) is maintained in TS 3.8, Refueling Operations, item 3.8.a.3 and the requirement for water level is maintained in TS item 3.8.a.1 0. In addition, the requirements for isolable air locks, an isolable equipment hatch, isolable penetrations, and containment radiation monitors are maintained in TS item 3.8.a.1. The Containment Purge and Exhaust System will be available in accordance with TS 3.8.a.8 and the aforementioned NUMARC 93-01 guidelines to further reduce radiological release. Therefore, the risk to the health and safety of the public as a result of allowing the equipment hatch to be open during fuel movement is minimal.

6.0 REGULATORY ANALYSIS

The method of analysis used for evaluating the potential radiological consequences of the postulated fuel handling accident is in compliance with Regulatory Guide 1.83 and the guidance in NUREG-0800, Section 15.0.1. The analysis presented in Section 14.3 of the KNPP USAR, demonstrating the adequacy of the system design features and plant procedures provided for the mitigation of the radiological consequences of postulated fuel handling accidents, assumes no credit is taken for iodine removal by the atmosphere filtration system filters. All radioactivity released to the containment is assumed to be released to the environment at ground level over a two hour period.

The technical analysis performed by NMC demonstrates that the consequent doses at the site boundary and low population zone boundaries are well within the limits of 10 CFR 50.67.

Therefore, the proposed license amendment is in compliance with GDC 16, 56, 61, and 64 (as discussed) as well as Regulatory Guide 1.183, and the criteria contained in NUREG-0800, Section 15.0.1.

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.

7.0 DIFFERENCES FROM IMPROVED STANDARD TECHNICAL SPECIFICATION (NUREG 1431, REV 2)

NMC submits this license amendment application with the following differences from Improved Standard Technical Specifications (ISTS), NUREG 1431, Revision 2:

1. CRPARS Operable in Modes 3 & 4 [5&6]

2. End State with CRPARS OOS (Mode 5)

3. Toxic Chemical Mode of Operation

4. Two CRPARS Trains OOS (non Boundary) enter 3.0.3

Docket 50-305 NRC-04-006 January 16, 2004 , page 14

5. Positive Pressure in Control Room CRPARS operable in modes 3, 4, 5, or 6 KNPP TS require the control room post-accident recirculation system to be operable when the reactor is critical (ISTS Modes 1 and 2). Other dose consequences analyses that credit the control room post-accident recirculation system (CRPARS) are the fuel handling accident (FHA), rupture of a gas decay tank (GDTR), and rupture of the volume control tank (VCTR). A FHA, GDTR, or a VCTR can happen at any plant mode of operation depending on the conditions present.

A FHA can occur during Refueling Operations or anytime an irradiated fuel assembly is moved or a load is transported over the storage location for an irradiated fuel assembly (spent fuel pool). To ensure dose consequences are well below 10 CFR 50.67 and Regulatory Guide 1.183 guidance, restrictions are placed when critical and non-critical. When critical, restrictions are placed on containment penetrations and CRPARS operation. When shutdown, restrictions are placed on containment penetrations, minimum decay time for movement of irradiated fuel (100 hours0.00116 days <br />0.0278 hours <br />1.653439e-4 weeks <br />3.805e-5 months <br />), CRPARS operation, and minimum decay time where no restriction, are applied (30 days). Therefore, any time a fuel handling accident of consequence can occur, restrictions are placed on appropriate plant operations to ensure the dose consequence requirements of 10 CFR 50.67 and the guidelines of Regulatory Guide 1.183 are met, so additional mode restraints are not required Accident analysis assumes the operation of the CRPARS in the analysis for a GDTR and VCTR. The analysis was run with 0 cfm, 200 cfm, and 400 cfm unfiltered inleakage into the control room. The limiting analysis occurred when the unfiltered inleakage was assumed to be 0 cfm. This is due to the analysis assuming that the radioactive plume entered the control room ventilation system, activates the radiation monitor and isolated the control room. The operators then remain in the control room for 30 days with the radioactivity that was present at the instant of the control room isolation. If the control room did not isolate, the unfiltered inleakage would be 2500 cfm, which would reduce the dose consequences below the results obtained in the analysis. Listed below are the results obtained from the analysis with differeing control room unfiltered inleakage.

Dose Conseque ces (rem)

GDTR Calculated VCTR Calculated TEDE Dose Limit DOSE (TEDE) Dose (TEDE)

Site Boundary 0.10 0.10 0.5lkb Low Population Zone 0.02 0.01 0.5 Control Room (0 cfm)') 0.80 0.40 5.0 Control Room (200 cfm) 0.08 0.04 5.0 Control Room (400 cfm) ( 0.06 0.03 5.0 Note (1) Control Room unfiltered inleakage Additionally, 10 CFR §50.36 contains four criteria for which limiting conditions for operation must be established. These criteria relate to the reactor coolant pressure boundary, the integrity of a fission product barrier (fuel clad, reactor coolant pressure boundary, or containment), or

Docket 50-305 NRC-04-006 January 16, 2004 , page 15 significant to public health and safety. The GDTR or VCTR do not contain items that meet these criteria.

Therefore, requiring the CRPARS to be operable based on a GDTR or VCTR is unnecessary as the higher the unfiltered inleakage the less the consequences and they do not meet the criteria of 10 CFR §50.36 for inclusion in KNPP's TS.

End State with CRPARS Out-Of-Service (OOS) (Mode 5)

The current end state for KNPP TS with a CRPARS out-of service is for the reactor to be subcritical. This is KNPP's licensing basis as a hot shutdown plant.

Toxic Chemical Mode of Operation In November of 1983 the NRC issued Generic Letter (GL) 83-37, "NUREG-0737 Technical Specifications." Enclosure 1, item 11, requested TS associated with control room habitability requirements (NUREG 0737, Item III.D.3.4) which request that if an analysis of postulated accidental release of toxic gases indicated the need for installing a toxic gas detection system, it should be included in the TS. KNPP's analysis showed that a toxic gas detection system was not needed and the system was not installed and no change was needed for KNPP's TS.

Therefore KNPP does not have a toxic chemical mode of operation and the TS is not applicable to KNPP.

Two CRPARS Trains OOS (non Boundary) enter 3.0.3 KNPP's current TS require that the reactor shall not be made critical unless two trains of the CRPARS are operable. Entry into KNPP's equivalent to ISTS TS 3.0.3, KNPP TS 3.0.c, Standard Shutdown Sequence," would require placing the plant in cold shutdown (RCS <

2000F). Instead, this submittal directs the operators to shutdown the reactor if this condition exist. This action is consistent with current KNPP TS.

Positive Pressure in Control Room KNPP is an isolation type control room ventilation system, this TS is not applicable.

8.0 NO SIGNIFICANT HAZARDS DETERMINATION The Nuclear Management Company, LLC, (NMC) is submitting a license amendment application that would revise KNPP Technical Specifications (TS) 3.8.a.1, "Containment Closure," TS 4.4, Containment Tests," and TS 3.12, "Control Room Post-Accident Recirculation System." TS 3.8.a.1 is being modified to allow the equipment hatch to be open during, REFUELING OPERATIONS and/or during movement of irradiated fuel assemblies within containment. TS 4.4.g is being added to require verification of the ability to close the equipment hatch periodically during refueling operations. TS 3.12 is modified to include requirements for

Docket 50-305 NRC-04-006 January 16, 2004 , page 16 operability of the Control Room Post-Accident Recirculation System (CRPARS) during refueling operations.

During movement of recently irradiated fuel assemblies within containment (refueling operations), the most severe radiological consequences result from a fuel handling accident (FHA). The acceptance limits for a FHA are contained in Standard Review Plan (SRP) 15.0.1, "Radiological Consequence Analyses Using Alternative Source Terms," and are approximately 25% of the 10CFR50.67 value of the 25 rem Total Effective Dose Equivalent (TEDE) for offsite doses (6.3 rem) and 5 rem TEDE for the control room.

The Technical Specifications have traditionally limited the consequences of a FHA inside containment by limiting the potential escape paths for fission product radioactivity released within containment. However, this philosophy has been replaced with a new approach which demonstrates that the effects of an FHA with the containment personnel air lock and containment penetrations open will still meet the SRP acceptance criteria. These analyses typically assume no holdup of radioactive material within the containment. However, the technical specification (TS) still requires the containment equipment hatch to be closed, even though that is not an assumption in the FHA dose analysis. The proposed change to the Containment Penetration TS allows the containment equipment hatch to be open during movement of recently irradiated fuel within the containment provided that it can be closed.

Additionally, TS requirements are being added to restrict Control Room Post Accident Recirculation System operation during movement of recently irradiated fuel assemblies. This restriction is added to conform to the assumption made in the accident analysis.

NMC has evaluated whether or not a significant hazards consideration is involved with the proposed changes by focusing on the three standards set forth in 10 CFR 50.92(c) 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 change would allow the containment equipment hatch to remain open during irradiated fuel movement in containment. This penetration is not an initiator of any accident. The probability of a fuel handling accident (FHA) in the containment is unaffected by the position of the equipment hatch. Adoption of this change requires analyses, approved by the NRC staff, demonstrating that the dose consequences of a FHA with the equipment hatch open are acceptable. Therefore, the proposed change 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 change does not involve the addition or modification of any plant equipment.

Also, the proposed change would not alter the design, configuration, or method of operation of

Docket 50-305 NRC-04-006 January 16, 2004 , page 17 the plant beyond the standard functional capabilities of the equipment. The proposed change involves a change to the Technical Specifications (TS) that would allow the equipment hatch to remain open during irradiated fuel movement within the containment. Having the equipment hatch open does not create the possibility of a new accident. Therefore, the proposed change does not create the possibility of a new or different kind of accident from any previously evaluated.

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

Response: No Analysis demonstrates that the resultant doses associated with a fuel handling accident are well within the appropriate acceptance limits. This change removes a defense-in-depth barrier that the analysis did not credit but provides additional restrictions on fission product release. Thus this proposed change has the potential for an increased dose at the site boundary due to a FHA; however, the analysis demonstrates that the resultant doses are well within the appropriate acceptance limits. Without the containment structure, analysis demonstrates that the dose consequences are still approximately 20% of the allowable value for the control room dose and less than 2% of the allowable value for offsite dose. Thus, the margin of safety has not been significantly reduced.. Administrative provisions that facilitate closing the equipment hatch following an evacuation of the containment further reduces the offsite doses in the event of a FHA and provides additional margin to the calculated offsite doses. Therefore, the proposed change does not involve a significant reduction in a margin of safety.

Based on the above, the proposed change 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

8.0 ENVIRONMENTAL CONSIDERATION

NMC has determined that the proposed amendment would change requirements with respect to the installation or use of a facility component located within the restricted area, as defined in 10 CFR 20, or would change an inspection or surveillance requirement. NMC has evaluated the proposed change and has determined that the change does not involve (i) a significant hazards consideration, (ii)a significant change in the types of or significant increase in the amounts of any effluents that may be released offsite, or (iii) a significant increase in individual or cumulative occupational radiation exposure. As discussed above, the proposed changes do not involve a significant hazards consideration and the analysis demonstrates that the consequences from a fuel handling accident inside containment are well within the 10 CFR 50.67 limits. The implementation of administrative controls precludes a significant increase in occupational radiation exposure. Accordingly, the proposed change meets the eligibility criteria for categorical exclusion set forth in 10 CFR 51.22(c)(9). Therefore, pursuant to 10 CFR 51.22(b), an environmental assessment of the proposed change is not required.

Docket 50-305 NRC-04-006 January 16, 2004 , page 18 9.0 PRECEDENTS There are precedents for allowing the equipment hatch to be open during REFUELING OPERATIONS and/or during movement of irradiated fuel assemblies within containment.

1. The Wolf Creek Nuclear Operating Corporation operating license for the Wolf Creek Generating Station amendment number 146 issued on July 30, 2002.1'91

2. The Arizona Public Service Company operating license for the Palo Verde Nuclear Generating Station Units 1, 2, and 3, amendment number 143.[20]

3. The Union Electric Company operating license for the Callaway Plant Unit 1 amendment 152, issued on September 9, 2002.1211

4. The Southern Nuclear Operating Company operating licenses for the Vogtle Generating Electric Plant Units 1 and 2, amendments, Nos. 115 and 93, were issued on September 11, 2000.[221

Docket 50-305 NRC-04-006 January 16, 2004 , page 19 10.0 References

['l KNPP Technical Specification 1.0.1, page TS 1.0-4

[2] KNPP USAR, TABLE H.2-1, "Leakage From Containment Vessel To Shield Building Annulus" (3) KNPP Corrective Maintenance Procedure (CMP) CMP-89A-02, Revision C.

141 Letter from John G. Lamb (NRC) to Thomas Coutu NMC, "Kewaunee Nuclear Power Plant - Issuance Of Amendment Regarding Implementation Of Alternate Source Term (TAC NO. MB4596)", dated March 17, 2003 (Adams Accession NO ML030210062) 15 KNPP USAR revision 17, section 5.1-2, page 5.1-2.

[6] KNPP USAR revision 17, section 5.3.1, page 5.3-1.

[7] KNPP USAR revision 17, section 11.2-1, page 11.2-1

[8] KNPP Technical Specification 3.8.a.3

[9] Letter from John Lamb (NRC) to Mark Reddemann (NMC), 'Kewaunee Nuclear Power Plant-Issuance of Amendment (TAC NO MA7278) (Adam Accession NO ML010240051)

[10] Letter from Tom Coutu (NMC) to Document Control Deck (NRC), 'License Amendment Request 195, Application For Stretch Power Uprate For Kewaunee Nuclear Power Plant," dated May 22, 2003.

[11] Letter from John G. Lamb (NRC) to Thomas Coutu NMC, 'Kewaunee Nuclear Power Plant - Issuance Of Amendment Regarding Implementation Of Alternate Source Term (TAC NO. MB4596)", dated March 17, 2003

[12] Regulatory Guide 1.183

[13] Letter from Thomas Coutu (NMC) to Document Control Desk (NRC), 'License Amendment Request 195, Application for Stretch Power Uprate for Kewaunee Nuclear Power Plant," dated May 22, 2003 (Adam Accession NO.ML031500724)

[14] Letter from W.O. Long (NRC) to M.L. Marchi (WPSC), Kewaunee Nuclear Power Plant, Supplemental Safety Evaluation Associated with Amendment NO. 132, (TAC NO. M99937) dated September 3, 1998.

1151 KNPP Operating Procedure N-RC-36E, 'Draining the Reactor Coolant System," Revision AC

[161 NUMARC 93-01, Revision 3, "Industry Guideline for Monitoring the Effectiveness of Maintenance at Nuclear Power Plants," July 2000.

[171 Letter from Steven A. Varga (NRC) to C.W. Giesler (WPSC), 'Completion of the Review of Venting and Purging Containment While at Full Power and Effect on LOCA (MPA B-24)," dated April 22, 1983.

18] Safety Evaluation of the Kewaunee Nuclear Power Plant, issued July 24, 1972, supplemented December 18,1972 and May 10, 1973.

[19] Letter from Jack Donohew (NRC) to Otto I. Maynard (WCNOC), 'Wolf Creek Generating Station -

Issuance of Amendment Re: Equipment Hatch Open During Refueling (TAC NO. MB2599)," dated July 30, 2002 (Adams Accession NO. ML022120428)

[20] Letter from Jack Donohew (NRC) to Gregg R. Overbeck (APSC), "Palo Verde Nuclear Generating Station, Units 1, 2, And 3 -Issuance of Amendments Re: Equipment Hatch Open During Refuel Operations (TAC NOS. MB3690, MB3691, AND MB3692) ," dated July 25, 2002 (Adams Accession NO. ML021290540)

[21] Letter from Jack Donohew (NRC) to Garry L. Randolph (UEC), "Callaway Plant, Unit 1 - Issuance Of Amendment Re: Equipment Hatch And Emergency Air Lock Open During Core Alterations Or Movement of Irradiated Fuel Assemblies Inside Containment (TAC NO. MB3605)," dated July 25, 2002 (Adams Accession NO. ML022620599)

[22] Letter from Ramin Assa (NRC) to J.B. Beasley, Jr. (SNOC), "Vogtle Electric Generating Plant, Units 1 and 2 RE: Issuance of Amendments (TAC NOs. MA8501 AND 8502)," dated September 11, 2000 (Adams Accession NO. ML003749439)

ATTACHMENT 2 NUCLEAR MANAGEMENT COMPANY, LLC KEWAUNEE NUCLEAR PLANT DOCKET 50-305 January 16, 2004 Letter from Thomas Coutu (NMC)

To Document Control Desk (NRC)

LAR 201 Strikeout Technical Specification Pages TS 3.8-1 TS 3.12-1 TS 3.12-2 TS 4.4-4 4 Pages Follow

3.8 REFUELING OPERATIONS APPLICABILITY Applies to operating limitations during REFUELING OPERATIONS.

OBJECTIVE To ensure that no incident occurs during REFUELING OPERATIONS that would affect public health and safety.

SPECIFICATION

a. During REFUELING OPERATIONS:

1. Containment Closure

a. The equipment hatch shall be closodand at least one door in each personnel air lock-shall be closed or capable of being closed (1)in 3-920minutes or less-k4 addition, at loast ono door in each personnel air lock shall be cloosed. Thisalso applie when theroactor vesoel head or-.upper internals are lifted.

b. Each line that penetrates containment and which provides a direct air path from containment atmosphere to the outside atmosphere shall have a closed isolation valves or an operable automatic isolation valve, or may be unisolated under administrative controls.

2. Radiation levels in fuel handling areas, the containment and the spent fuel storage pool, shall be monitored continuously.

3. The reactor will be subcritical for 148 hours0.00171 days <br />0.0411 hours <br />2.44709e-4 weeks <br />5.6314e-5 months <br /> prior to movement of its irradiated fuel assemblies. Core subcritical neutron flux shall be continuously monitored by at least two neutron monitors, each with continuous visual indication in the control room and one with audible indication in the containment whenever core geometry is being changed. When core geometry is not being changed at least one neutron flux monitor shall be in service.

4. At least one residual heat removal pump shall be OPERABLE.

5. When there is fuel in the reactor, a minimum boron concentration as specified in the COLR shall be maintained in the Reactor Coolant System during reactor vessel head

') Administrative controls ensure that:

• Appropriate personnel are aware that the equipment hatch or both personnel air lock doors are open,

• A specified individual(s) is designated and available to close the equipment hatch anrd air lock following a required evacuation of containment,-and

• Any obstruction(s) (e.g., cables and hoses) that could prevent closure of an open air lock/equipment hatch can be quickly removed.

• Necessary hardware, tools. and equipment required for closure are available.

LAR 201Amondmont No. xxx TS 3.8-1 DA/20Dx_4

3.12 CONTROL ROOM POST-ACCIDENT RECIRCULATION SYSTEM APPLICABILITY Applies to the OPERABILITY of the Control Room Post-Accident Recirculation System.

OBJECTIVE To specify OPERABILITY requirements for the Control Room Post-Accident Recirculation System.

SPECIFICATION

a. The reactor shall not be made critical unless the following conditions are satisfied.

1.unkIGss-bethTwQ trains of the Control Room Post-Accident Recirculation System are OPERABLE, except as provided byTS 3.12.a.2.

2. During power operation or recovery from an inadvertent trip. the following conditions of inoperability may exist during the time interval specified. If OPERABILITY is not restored within the time specified, then within 1 hour1.157407e-5 days <br />2.777778e-4 hours <br />1.653439e-6 weeks <br />3.805e-7 months <br /> action shall be initiated to:

- Achieve HOT STANDBY within the next 6 hours6.944444e-5 days <br />0.00167 hours <br />9.920635e-6 weeks <br />2.283e-6 months <br />.

- Achieve HOT SHUTDOWN within the following 6 hours6.944444e-5 days <br />0.00167 hours <br />9.920635e-6 weeks <br />2.283e-6 months <br />.

A. One train of Control Room Post-Accident Recirculation may be out of service for 7 days B. Two trains of Control Room Post-Accident Recirculation may be out of service for 24 hours2.777778e-4 days <br />0.00667 hours <br />3.968254e-5 weeks <br />9.132e-6 months <br /> if due to control room boundary failure.

Both trains of the Control Room Poest Accident Recirculation System, including filters, chall be OPERABLE or the reactor shall be chut down within 12 hourE, except that when one of the two trains of the Control Room Post Accident Recirculation System is made or found to be inoperable for any rearen, reactor operation is permissible only during the eUcceeding 7 days.

b. During REFUELING OPERATIONS, or movement of irradiated fuel assemblies that have decayed less than 30 days, or when any heavy load is carried over the spent fuel pool if irradiated fuel in the pool has decayed less than 30 days:

1. Two trains of the Control Room Post-Accident Recirculation System shall be OPERABLE, except as provided byTS 3.12.b.2 and TS 3.12.b.3.

2. One train of Control Room Post-Accident Recirculation System may be out of service A. The opposite train is in service in the emergency mode or.

B. Movement of irradiated fuel assemblies that decayed less than 30 days and movement of any load over the spent fuel pool are suspended.

LAR 201Amcndment No. 152 TS 3.12-1 1/1 6 20 2812-01

3. Two trains of Control Room Post-Accident Recirculation System may be out of service provided movement of irradiated fuel assemblies that have decayed less than 30 days and movement of any load over the spent fuel pool are suspended.

c. During testing the system shall meet the following performance requirements:

1. The results of the in-place cold DOP and halogenated hydrocarbon tests at design flows on HEPA filter and charcoal adsorber banks shall show 2 99% DOP removal and 2 99% halogenated hydrocarbon removal.

2. The results of the laboratory carbon sample analysis from the Control Room Post-Accident Recirculation System carbon shall show 2 95% radioactive methyl iodide removal when tested in accordance with ASTM D3803-89 at conditions of 30 0C, and 95% RH.

3. Fans shall operate within +/-10% of design flow when tested.

d, The control room boundary may be opened intermittently under administrative control.

fl2~AmondmontF~

LAR No 5 TS 3.12-2 /16/200402/28/2004

b. Priorto entering INTERMEDIATE SHUTDOWN from COLD SHUTDOWN, if not performed in the previous 92 days, verify each containment isolation manual valve and blind flange that is located inside containment and required to be closed during accident conditions is closed, except for containment isolation valves that are locked sealed or otherwise secured closed or open as allowed by TS 3.6.b.2.

c. Valves and blind flanges in high radiation areas may be verified by use of administrative means.

a. During REFUELING OPERATIONS with the equipment hatch open. verify:

a. The equipment hatch closure capability prior to fuel movement and weekly thereafter,

b. Necessary hardware, tools, and equipment required for closure are available daily.

4LAR42mondmont No. 2lx0 TS 4.4-4 1/162004

ATTACHMENT 3 NUCLEAR MANAGEMENT COMPANY, LLC KEWAUNEE NUCLEAR PLANT DOCKET 50-305 January 16, 2004 Letter from Thomas Coutu (NMC)

To Document Control Desk (NRC)

LAR 201 Technical Specification Pages as Revised TS 3.8-1 TS 3.12-1 TS 3.12-2 TS 4.4-4 4 Pages Follow

3.8 REFUELING OPERATIONS APPLICABILITY Applies to operating limitations during REFUELING OPERATIONS.

OBJECTIVE To ensure that no incident occurs during REFUELING OPERATIONS that would affect public health and safety.

SPECIFICATION

a. During REFUELING OPERATIONS:

1. Containment Closure

a. The equipment hatch and at least one door in each personnel air lock shall be closed or capable of being closed (1)in 90 minutes or less. This also applies when the upper internals are lifted.

b. Each line that penetrates containment and which provides a direct air path from containment atmosphere to the outside atmosphere shall have a closed isolation valve, or an operable automatic isolation valve, or may be unisolated under administrative controls.

2. Radiation levels in fuel handling areas, the containment and the spent fuel storage pool, shall be monitored continuously.

3. The reactor will be subcritical for 148 hours0.00171 days <br />0.0411 hours <br />2.44709e-4 weeks <br />5.6314e-5 months <br /> prior to movement of its irradiated fuel assemblies. Core subcritical neutron flux shall be continuously monitored by at least two neutron monitors, each with continuous visual indication in the control room and one with audible indication in the containment whenever core geometry is being changed. When core geometry is not being changed at least one neutron flux monitor shall be in service.

4. At least one residual heat removal pump shall be OPERABLE.

5. When there is fuel in the reactor, a minimum boron concentration as specified in the COLR shall be maintained in the Reactor Coolant System during reactor vessel head removal or while loading and unloading fuel from the reactor. The required boron

() Administrative controls ensure that:

• Appropriate personnel are aware that the equipment hatch or both personnel air lock doors are open,

• A specified individual(s) is designated and available to close the equipment hatch and air lock following a required evacuation of containment,

• Any obstruction(s) (e.g., cables and hoses) that could prevent closure of an open air lock/equipment hatch can be quickly removed.

• Necessary hardware, tools, and equipment required for closure are available.

TS 3.8-1

3.12 CONTROL ROOM POST-ACCIDENT RECIRCULATION SYSTEM APPLICABILITY Applies to the OPERABILITY of the Control Room Post-Accident Recirculation System.

OBJECTIVE To specify OPERABILITY requirements for the Control Room Post-Accident Recirculation System.

SPECIFICATION

a. The reactor shall not be made critical unless the following conditions are satisfied,

1. Two trains of the Control Room Post-Accident Recirculation System are OPERABLE, except as provided by TS 3.12.a.2.

2. During power operation or recovery from an inadvertent trip, the following conditions of inoperability may exist during the time interval specified. If OPERABILITY is not restored within the time specified, then within 1 hour1.157407e-5 days <br />2.777778e-4 hours <br />1.653439e-6 weeks <br />3.805e-7 months <br /> action shall be initiated to:

- Achieve HOT STANDBY within the next 6 hours6.944444e-5 days <br />0.00167 hours <br />9.920635e-6 weeks <br />2.283e-6 months <br />.

- Achieve HOT SHUTDOWN within the following 6 hours6.944444e-5 days <br />0.00167 hours <br />9.920635e-6 weeks <br />2.283e-6 months <br />.

A. One train of Control Room Post-Accident Recirculation may be out of service for 7 days.

B. Two trains of Control Room Post-Accident Recirculation may be out of service for 24 hours2.777778e-4 days <br />0.00667 hours <br />3.968254e-5 weeks <br />9.132e-6 months <br /> if due to control room boundary failure.

b. During REFUELING OPERATIONS, or movement of irradiated fuel assemblies that have decayed less than 30 days, or when any heavy load is carried over the spent fuel pool if irradiated fuel in the pool has decayed less than 30 days:

1. Two trains of the Control Room Post-Accident Recirculation System shall be OPERABLE, except as provided by TS 3.1 2.b.2 and TS 3.1 2.b.3.

2. One train of Control Room Post-Accident Recirculation System may be out of service provided:

A. The opposite train is in service in the emergency mode or, B. Movement of irradiated fuel assemblies that decayed less than 30 days and movement of any load over the spent fuel pool are suspended.

3. Two trains of Control Room Post-Accident Recirculation System may be out of service provided movement of irradiated fuel assemblies that have decayed less than 30 days and movement of any load over the spent fuel pool are suspended.

c. During testing the system shall meet the following performance requirements:

TS 3.12-1

1. The results of the in-place cold DOP and halogenated hydrocarbon tests at design flows on HEPA filter and charcoal adsorber banks shall show 2 99% DOP removal and 2 99% halogenated hydrocarbon removal.

2. The results of the laboratory carbon sample analysis from the Control Room Post-Accident Recirculation System carbon shall show 2 95% radioactive methyl iodide removal when tested in accordance with ASTM D3803-89 at conditions of 300C, and 95% RH.

3. Fans shall operate within +/-10% of design flow when tested.

d. The control room boundary may be opened intermittently under administrative control.

TS 3.12-2

b. Prior to entering INTERMEDIATE SHUTDOWN from COLD SHUTDOWN, if not performed in the previous 92 days, verify each containment isolation manual valve and blind flange that is located inside containment and required to be closed during accident conditions is closed, except for containment isolation valves that are locked sealed or otherwise secured closed or open as allowed by TS 3.6.b.2.

c. Valves and blind flanges in high radiation areas may be verified by use of administrative means.

g. During REFUELING OPERATIONS with the equipment hatch open, verify:

a. The equipment hatch closure capability prior to fuel movement and weekly thereafter,

b. Necessary hardware, tools, and equipment required for closure are available daily.

TS 4.4-4

ATTACHMENT 4 NUCLEAR MANAGEMENT COMPANY, LLC KEWAUNEE NUCLEAR PLANT DOCKET 50-305 January 16, 2004 Letter from Thomas Coutu (NMC)

To Document Control Desk (NRC)

LAR 201 Strikeout Technical Specification Basis Pages TS B3.8-1 TS B3.8-2 TS B3.8-3 TS B3.12-1 TS B3.12-2 TS B4.4-5 6 Pages Follow

BASIS - Refueling Operations (TS 3.8)

The equipment and general procedures to be utilized during REFUELING OPERATIONS are discussed in the USAR. Detailed instructions, the above specified precautions, and the design of the fuel handling equipment incorporating built-in interlocks and safetyfeatures, provide assurance that no incident occurs during the REFUELING OPERATIONS that would result in a hazard to public health and safety.(') Whenever changes are not being made in core geometry, one flux monitor is sufficient. This permits maintenance of the instrumentation. Continuous monitoring of radiation levels (TS 3.8.a.2) and neutron flux provides immediate indication of an unsafe condition. The residual heat removal pump is used to maintain a uniform boron concentration.

Containment Closure (TS 3.8.a.1)

During movement of recently irradiated fuel assemblies (i.e. fuel which has decayed less than 30 days) within containment. a release of fission product radioactivity within containment will be restricted from escaping to the environment when the TS requirements are met. When above COLD SHUTDOWN. this is accomplished by maintaining containment OPERABLE as described in TS 3.6, "Containment." In COLD SHUTDOWN. the potential for containment pressurization as a result of an accident is not likely: therefore, requirements to isolate the containment from the outside atmosphere can be less stringent. The TS requirements are referred to as ucontainment closure" rather than "containment OPERABILITY." Containment closure means that all potential escape paths are closed or caiable of being closed. Since there is no potential for containment pressurization, the Appendix J leakage criteria and tests are not required, The containment serves to contain fission product radioactivity that may be released from the reactor core following an accident, such that offsite radiation exposures are maintained well within the requirements of 10 CFR 50.67. Additionally, the containment provides radiation shielding from the fission products that may be Present in the containment atmosphere following accident conditions.

The containment equipment hatch, which is part of the containment pressure boundary, provides a means for moving larae equipment and components into and out of containment. During movement of recently irradiated fuel assemblies within containment, the equipment hatch must be capable of being closed and held in place by at least four swing bolts. Good engineering practice dictates that the bolts required by this TS be approximately equally spaced.

The containment air locks, which are also part of the containment pressure boundary, provide a means for personnel access during operation above COLD SHUTDOWN in accordance with TS 3.6.a. "Containment System Integrity". Each air lock has a door at both ends. The doors are normally interlocked to prevent simultaneous opening when CONTAINMENT INTEGRITY is required. During periods of unit shutdown when containment closure is not requiredl the door interlock mechanism may be disabled, allowing both doors of an air lock to remain open for extended periods when frequent containment entry is necessary. During movement of recently irradiated fuel assemblies within containment, containment closure is required: therefore. the door interlock mechanism may remain disabled, but one air lock door must always remain closed or capable of being closed.

USAR Section 9.5.2 LAR 20A1Amedm9t No. 165 TS B3.8-1 01 /16/200403/1-1/2003

The requirements for containment penetration closure provides additional defense-in-depth to further ensure that a release of fission product radioactivity within containment will be restricted to within regulatory limits.

The other containment penetrations that provide direct access from containment atmosphere to outside atmosphere must be isolated on at least one side. Isolation may be achieved by an OPERABLE automatic isolation valve, or by a manual isolation valve, blind flange, or equivalent.

Equivalent isolation methods must be approved and may include use of a material that can provide a temporary, atmospheric pressure, ventilation barrier for the other containment penetrations during recently irradiated fuel movements.

During REFUELING OPERATIONS or movement of irradiated fuel assemblies within containment, the most severe radiological consequences result from a fuel handling accident involving handling recently irradiated fuel. The fuel handling accident is a postulated event that involves damage to irradiated fuel. Fuel handlinq accidents, analyzed include dropping a single irradiated fuel assembly and handling tool or a heavy object onto other irradiated fuel assemblies (i.e., all rods in one assembly are damaged releasing the gap activity of iodines and noble gases). The requirements of TS 3.8.a.1 0. refueling cavity water level, in conjunction with a minimum decay time of 100 hours0.00116 days <br />0.0278 hours <br />1.653439e-4 weeks <br />3.805e-5 months <br /> prior to irradiated fuel movement with containment closure capability or a minimum decay time of 30 days without containment closure capability, ensures that the release of fission product radioactivity. subsequent to a fuel handling accident. results in doses that are within the values specified in 10 CFR 50.67 as modified by Regulatory Guide 1.183, Alternative Radiological Source Terms for Evaluating Design Basis Accidents at Nuclear Power Reactors" (RG 1.183). The acceptance limits for offsite radiation exposure for a Fuel Handling Accident is listed in RG 1.183 as 6.3 rem TEDE, which is 25% of the 10 CFR 50.67 limits. Containment penetrations satisfy Criterion 3 of 10 CFR 50.36(c)(2)(ii).

This TS limits the consequences of a fuel handling accident involving handling recently irradiated fuel in containment by limiting the potential escape paths for fission product radioactivity released within containment. The TS requires any penetration providing direct access from the containment atmosphere to the outside atmosphere to be closed, have an OPERABLE automatic isolation, or. as in the case of the containment personnel air lock and equipment hatch, capable of being closed.

The TS is modified allowing penetration flow paths with direct access from the containment atmosphere to the outside atmosphere to be unisolated under administrative controls.

Administrative controls ensure that 1) appropriate personnel are aware of the open status of the penetration flow path during REFUELING OPERATIONS or movement of irradiated fuel assemblies within containment, and 2) specified individuals are designated and readily available to isolate the flow path in the event of a fuel handling accident.

The containment personnel air lock doors many be open during movement of irradiated fuel in the containment and during REFUELING OPERATIONS provided that one door is capable of being closed in the event of a fuel handling accident. Should a fuel handling accident occur inside containment. one personnel air lock door will be closed following an evacuation of containment The containment penetration requirements are applicable during movement of irradiated fuel assemblies that have decayed less than 30 days within containment to maintain a defense-in-depth philosophy against a potential for the limiting fuel handling accident. When above COLD SHUTDOWN, containment penetration requirements are addressed byTS 3.6. In COLD LAI201Amendment No. 165 TS B3.8-2 01/161200403!112003

SHUTDOWN or REFUELING, when movement of irradiated fuel assemblies within containment is not being conducted. the potential for a fuel handling accident does not exist. Additionally, due to radioactive decay, a fuel handling accident involving handling irradiated fuel that has decayed greater than 100 hours0.00116 days <br />0.0278 hours <br />1.653439e-4 weeks <br />3.805e-5 months <br /> results in doses that are well within the guideline values specified in 10 CFR 50.67 even without containment closure capability. Therefore, under these conditions no requirements need be placed on containment penetration status for irradiated fuel that has decayed greater than 100 hours0.00116 days <br />0.0278 hours <br />1.653439e-4 weeks <br />3.805e-5 months <br />. To provide defense-in-depth, requirements are placed on the containment penetrations until the irradiated fuel has decayed greater than 30 days. This relaxation of TS requirements when moving irradiated fuel that has decayed areater than 30 days is only used in the Spent Fuel Pool Sweep System TS.

If the containment equipment hatch, air locks, or any containment penetration that provides direct access from the containment atmosphere to the outside atmosphere is not in the required status, including the Containment Purge and Exhaust Isolation System not capable of automatic actuation when the purge and exhaust valves are open, the unit must be placed in a condition where the isolation function is not needed. This is accomplished by immediately suspending movement of recently irradiated fuel assemblies within containment. Performance of these actions shall not preclude completion of movement of a component to a safe position.

A minimum shutdown margin of greater than or equal to 5% Ak/k must be maintained in the core.

The boron concentration as specified in the COLR is sufficient to ensure an adequate margin of safety. The specification for REFUELING OPERATIONS shutdown margin is based on a dilution during refueling accident.(2 ) With an initial shutdown margin of 5% Ak/k, under the postulated accident conditions, it will take longer than 30 minutes for the reactor to go critical. This is ample time for the operator to recognize the audible high count rate signal, and isolate the reactor makeup water system. Periodic checks of refueling water boron concentration ensure that proper shutdown margin is maintained. Specification 3.8.a.6 allows the control room operator to inform the manipulator operator of any impending unsafe condition detected from the main control board indicators during fuel movement.

Interlocks are utilized during REFUELING OPERATIONS to ensure safe handling. Only one assembly at a time can be handled. The fuel handling hoist is dead weight tested prior to use to assure proper crane operation. Itwill not be possible to lift or carry heavy objects overthe spentfuel pool when fuel is stored therein through interlocks and administrative procedures. Placement of additional spent fuel racks will be controlled by detailed procedures to prevent traverse directly above spent fuel.

The one hundred forty-eight hour decay time following plant shutdown is consistent with the spent fuel pool cooling analysis and also bounds the assumption used in the dose calculation for the fuel handling accident. The requirement for the spent fuel pool sweep system, including charcoal adsorbers, to be operating when spent fuel movement is being made provides added assurance that the off-site doses will be within acceptable limits in the event of a fuel handling accident. The spent fuel pool sweep system is designed to sweep the atmosphere above the refueling pool and release to the Auxiliary Building vent during fuel handling operations. Normally, the charcoal adsorbers are bypassed but for purification operation, the bypass dampers are closed routing the air flow through the charcoal adsorbers. If the dampers do not close tightly, bypass leakage could exist to negate the usefulness of the charcoal adsorber. If the spent fuel pool sweep system is found not to be operating, fuel handling within the Auxiliary Building will be terminated until the system can be

( 2)USAR Section 14.1 LAR 201Amedment No. 165 TS B3.8-3 1L1/62004O3I4412003

BASIS - Control Room Post-Accident Recirculation System (TS 3.12)

The Control Room Post-Accident Recirculation System LEBPABS)is designed to filter the Control Room atmosphere during Control Room isolation conditions. The Control Room Post-Accident Recirculation System is designed to automatically start upon SIS or high radiation signal. The ORPARS consists of two independent, redundant trains that recirculate and filter the control room air. Each train consists of a prefilter. a high efficiency particulate air (HEPA) filter, an activated charcoal adsorber section. for removal of aaseous activity (principally iodines), and a fan. Ductwork.

valves or dampers, and instrumentation also form part of the system.

If the system ir found to be inoperable, there ir no immediate thrmatto the Ceotrol Room and rcacter operation may continue for a limited period of time while repairs are being made. If the syStem canRnt be repaired within 7 days, the reactor is placed in HOT STANDBY uRVI the repainr are mrade.

The CRPARS satisfies Criterion 3 of 10 CFR 50.36(c)(2)(ii) for design basis accidents and fuel handling accidents.

Two independent and redundant CRPARS trains are required to be OPERABLE to ensure that at least one is available assuming a single failure disables the other train. Total system failure could result in exceeding a dose of 5 rem to the control room operator in the event of a large radioactive release. The CRPARS is considered OPERABLE when the individual components necessary to limit operator exposure are OPERABLE in both trains.

A CRPARS train is OPERABLE when the associated:

a. Fan is OPERABLE.

b. HEPA filters and charcoal adsorbers 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 In addition, the control room boundary must be maintained, including the integrity of the walls, floors, ceilings, ductwork, and access doors. Allowing the control room boundary to be opened intermittently under administrative controls modifies the TS. For entry and exit through doors. the administrative control of the opening is performed by the person(s) entering or exiting the area.

For other openings, these controls consist of stationing a dedicated individual at the opening who is in continuous communication with the control room. This individual will have a method to rapidly close the opening when a need for control room isolation is indicated.

When CRITICAL and during REFUELING OPERATION, CRPARS must be OPERABLE to control operator exposure during and following a DBA. In all Modes, the CRPARS is required to cope with a release from the rupture of a gas decay tank (GDT) or the Volume Control Tank (VCT). Although required to cope with a release from the gas decay tank or the Volume Control Tank the requirement does not meet the criteria found in 10 CFR 50.36 for inclusion in Technical Specifications, therefore the requirement for operability of the CRPARS when in all modes due to the presents of radioactive gases in the GDT or VCT is not included in the Technical Specifications. During movement of irradiated fuel assemblies that have decayed less than 30 days, the CRPARS must be OPERABLE to cope with the release from a fuel handling accident.

When one CRPARS train is inoperable, action must be taken to restore OPERABLE status within 7 days. In this Condition. the remaining OPERABLE CRPARS train is adequate to perform LAR 201Amondmont No. 162 TS B3.12-14 Pg/o1/20 /2o24

the control room protection function. However, the overall reliability is reduced because a single failure in the OPERABLE CRPARS train could result in loss of CRPARS function. The 7 day Completion Time is based on the low probability of a DBA occurring during this time period, and ability of the remaining train to provide the required capability.

If the control room boundary is inoperable when critical, the CRPARS trains cannot perform their intended functions. Actions must be taken to restore an OPERABLE control room boundary within 24 hours2.777778e-4 days <br />0.00667 hours <br />3.968254e-5 weeks <br />9.132e-6 months <br />. During the period that the control room boundary is inoperable, appropriate compensatory measures (consistent with the intent of GDC 19) should be utilized to protect control room operators from potential hazards such as radioactive contamination, toxic chemicals, smoke. temperature and relative humidity. and physical security. Preplanned measures should be available to address these concerns for intentional and unintentional entry into the condition. The 24 hour2.777778e-4 days <br />0.00667 hours <br />3.968254e-5 weeks <br />9.132e-6 months <br /> Completion Time is reasonable based on the low probability of a DBA occurring during this time period, and the use of compensatory measures. The 24 hour2.777778e-4 days <br />0.00667 hours <br />3.968254e-5 weeks <br />9.132e-6 months <br /> Completion time is a typically reasonable time to diagnose, plan and possibly repair, and test most problems with the control room boundary.

When critical, if the inoperable CRPARS train or control room boundar cannot be restored to OPERABLE status within the required Completion Time. the unit must be placed in a MODE that minimizes accident risk. To achieve this status, the unit must be placed in at least HOT SHUTDOWN within 12 hours1.388889e-4 days <br />0.00333 hours <br />1.984127e-5 weeks <br />4.566e-6 months <br />. The allowed Completion Time is reasonable, based on operating experience, to reach the required unit conditions from full power conditions in an orderly manner and without challenging unit systems.

During REFUELING OPERATIONS. if the inoperable CRPARS train cannot be restored to OPERABLE status within the required Completion Time, action must be taken to immediately place the OPERABLE CRPARS train in the emergency mode. This action ensures that the remaining train is OPERABLE. that no failures preventing automatic actuation will occur, and that any active failure would be readily detected. An alternative is to immediately suspend activities that could result in a release of radioactivity that might require isolation of the control room. This places the unit in a condition that minimizes risk. This does not preclude the movement of fuel to a safe position.

TS 3.12.b imposes restrictions on movement of heavy loads over the spent fuel pool. NUREG-0612, "Control of Heavy Loads at Nuclear Power Plants." dated July 1980, provides guidelines and recommendations to assure safe handling of heaw loads by prohibiting. to the extent practicable, heavy load travel over stored spent fuel assemblies, fuel in the reactor core. safety-related equipment, and equipment needed for decay heat removal. The NUREG defines a heav load as any load carried in a given area during the operation of the plant that weighs more than the combined weight of a single spent fuel assembly and its associated handling tool. This restriction stems from these guidelines.

During REFUELING OPERATIONS, with two CRPARS trains inoperable, action must be taken immediately to suspend activities that could result in a release of radioactivity that might enter the control room. This pfaces the unit in a condition that minimizes accident risk. This does not preclude the movement of fuel to a safe position.

If both CRPARS trains are inoperable when critical for reasons other than an inoperable control room boundary, the CRPARS may not be capable of performing the intended function and the unit is in a condition outside the accident analyses. Therefore, a plant shutdown as stated in TS 3.1 2.a.2 must be initiated.

LAS 209Amoi oto.

TS B3.12-24 01/16/200402128/0Oo

TS 4.4.f.3.B requires verification that each containment isolation manual valve and blind flange located inside containment and not locked, sealed, or otherwise secured and required to be closed during accident conditions, is closed. The TS helps to ensure that post-accident leakage of radioactive fluids or gases outside of the containment boundary is within design limits. For containment isolation valves inside containment, the frequencyof "priorto entering INTERMEDIATE SHUTDOWN from COLD SHUTDOWN if not performed within the previous 92 days" is appropriate since these containment isolation valves are operated under administrative controls and the probability of their misalignment is low. The TS specifies that containment isolation valves that are open under administrative controls are not required to meet the TS during the time they are open.

This TS does not apply to valves that are locked, sealed, or otherwise secured in the closed position, since these were verified to be in the correct position upon locking, sealing, or securing.

TS 4.4.f.3.C modifies TS 4.4.f.3 for valves and blind flanges located in high radiation areas and allows these devices to be verified closed by use of administrative means. Allowing verification by administrative means is considered acceptable, since access to these areas is typically restricted when above COLD SHUTDOWN for ALARA reasons. Therefore, the probability of misalignment of these containment isolation valves, once they have been verified to be in the proper position, is small.

Equipment Hatch closure capability (TS 4.4.a)

This surveillance demonstrates that the equipment hatch is capable of being closed within the required time prior to movement of irradiated fuel. Repeating this closure capability demonstration weekly provides additional assurance that if called on to be closed in an emergency closure will occurexpeditiously. The surveillance interval was selected to be commensurate with the normal duration of time to complete the fuel handling operations Additionally, this surveillance demonstrates that the necessary hardware. tools. and equipment are available to install the equipment hatch. The equipment hatch is provided with a set of hardware.

tools, and equipment for moving the hatch from its storage location and installing it in the opening.

The required set of hardware. tools. and equipment shall be inspected to ensure that they can perform the required functions. The surveillance is performed daily during REFUELING OPERATIONS when the equipment hatch is open. . The daily surveillance is adequate considering that the hardware, tools and equipment are dedicated to the equipment hatch and not used for any other function LAR 201 TS B4.4-5 40 25 120020i11612D04

ATTACHMENT 5 NUCLEAR MANAGEMENT COMPANY, LLC KEWAUNEE NUCLEAR PLANT DOCKET 50-305 January 16, 2004' -

Letter from Thomas Coutu (NMC)

To Document Control Desk (NRC)

LAR 201 Technical Specification Basis Pages as Revised TS B3.8-1 TS B3.8-2 TS B3.8-3 TS B3.12-1 TS B3.12-2 TS B4.4-5 6 Pages Follow

BASIS - Refuelinq Operations (TS 3.8)

The equipment and general procedures to be utilized during REFUELING OPERATIONS are discussed in the USAR. Detailed instructions, the above specified precautions, and the design of the fuel handling equipment incorporating built-in interlocks and safety features, provide assurance that no incident occurs during the REFUELING OPERATIONS that would result in a hazard to public health and safety.(') Whenever changes are not being made in core geometry, one flux monitor is sufficient. This permits maintenance of the instrumentation. Continuous monitoring of radiation levels (TS 3.8.a.2) and neutron flux provides immediate indication of an unsafe condition. The residual heat removal pump is used to maintain a uniform boron concentration.

Containment Closure (TS 3.8.a.1)

During movement of recently irradiated fuel assemblies (i.e. fuel which has decayed less than 30 days) within containment, a release of fission product radioactivity within containment will be restricted from escaping to the environment when the TS requirements are met. When above COLD SHUTDOWN, this is accomplished by maintaining containment OPERABLE as described in TS 3.6, "Containment." In COLD SHUTDOWN, the potential for containment pressurization as a result of an accident is not likely; therefore, requirements to isolate the containment from the outside atmosphere can be less stringent. The TS requirements are referred to as "containment closure" rather than "containment OPERABILITY." Containment closure means that all potential escape paths are closed or capable of being closed. Since there is no potential for containment pressurization, the Appendix J leakage criteria and tests are not required.

The containment serves to contain fission product radioactivity that may be released from the reactor core following an accident, such that offsite radiation exposures are maintained well within the requirements of 10 CFR 50.67. Additionally, the containment provides radiation shielding from the fission products that may be present in the containment atmosphere following accident conditions.

The containment equipment hatch, which is part of the containment pressure boundary, provides a means for moving large equipment and components into and out of containment. During movement of recently irradiated fuel assemblies within containment, the equipment hatch must be capable of being closed and held in place by at least four swing bolts. Good engineering practice dictates that the bolts required by this TS be approximately equally spaced.

The containment air locks, which are also part of the containment pressure boundary, provide a means for personnel access during operation above COLD SHUTDOWN in accordance with TS 3.6.a, "Containment System Integrity". Each air lock has a door at both ends. The doors are normally interlocked to prevent simultaneous opening when CONTAINMENT INTEGRITY is required. During periods of unit shutdown when containment closure is not required, the door interlock mechanism may be disabled, allowing both doors of an air lock to remain open for extended periods when frequent containment entry is necessary. During movement of recently irradiated fuel assemblies within containment, containment closure is required; therefore, the door interlock mechanism may remain disabled, but one air lock door must always remain closed or capable of being closed.

(1)USAR Section 9.5.2 TS B3.8-1

The requirements for containment penetration closure provides additional defense-in-depth to further ensure that a release of fission product radioactivity within containment will be restricted to within regulatory limits.

The other containment penetrations that provide direct access from containment atmosphere to outside atmosphere must be isolated on at least one side. Isolation may be achieved by an OPERABLE automatic isolation valve, or by a manual isolation valve, blind flange, or equivalent.

Equivalent isolation methods must be approved and may include use of a material that can provide a temporary, atmospheric pressure, ventilation barrier for the other containment penetrations during recently irradiated fuel movements.

During REFUELING OPERATIONS or movement of irradiated fuel assemblies within containment, the most severe radiological consequences result from a fuel handling accident involving handling recently irradiated fuel. The fuel handling accident is a postulated event that involves damage to irradiated fuel. Fuel handling accidents, analyzed include dropping a single irradiated fuel assembly and handling tool or a heavy object onto other irradiated fuel assemblies (i.e., all rods in one assembly are damaged releasing the gap activity of iodines and noble gases). The requirements of TS 3.8.a.1 0, refueling cavity water level, in conjunction with a minimum decay time of 100 hours0.00116 days <br />0.0278 hours <br />1.653439e-4 weeks <br />3.805e-5 months <br /> prior to irradiated fuel movement with containment closure capability or a minimum decay time of 30 days without containment closure capability, ensures that the release of fission product radioactivity, subsequent to a fuel handling accident, results in doses that are within the values specified in 10 CFR 50.67 as modified by Regulatory Guide 1.183, Alternative Radiological Source Terms for Evaluating Design Basis Accidents at Nuclear Power Reactors" (RG 1.183). The acceptance limits for offsite radiation exposure for a Fuel Handling Accident is listed in RG 1.183 as 6.3 rem TEDE, which is 25% of the 10 CFR 50.67 limits. Containment penetrations satisfy Criterion 3 of 10 CFR 50.36(c)(2)(ii).

This TS limits the consequences of a fuel handling accident involving handling recently irradiated fuel in containment by limiting the potential escape paths for fission product radioactivity released within containment. The TS requires any penetration providing direct access from the containment atmosphere to the outside atmosphere to be closed, have an OPERABLE automatic isolation, or, as in the case of the containment personnel air lock and equipment hatch, capable of being closed.

The TS is modified allowing penetration flow paths with direct access from the containment atmosphere to the outside atmosphere to be unisolated under administrative controls.

Administrative controls ensure that 1) appropriate personnel are aware of the open status of the penetration flow path during REFUELING OPERATIONS or movement of irradiated fuel assemblies within containment, and 2) specified individuals are designated and readily available to isolate the flow path in the event of a fuel handling accident.

The containment personnel air lock doors many be open during movement of irradiated fuel in the containment and during REFUELING OPERATIONS provided that one door is capable of being closed in the event of a fuel handling accident. Should a fuel handling accident occur inside containment, one personnel air lock door will be closed following an evacuation of containment.

The containment penetration requirements are applicable during movement of irradiated fuel assemblies that have decayed less than 30 days within containment to maintain a defense-in-depth philosophy against a potential for the limiting fuel handling accident. When above COLD SHUTDOWN, containment penetration requirements are addressed by TS 3.6. In COLD TS B3.8-2

SHUTDOWN or REFUELING, when movement of irradiated fuel assemblies within containment is not being conducted, the potential for a fuel handling accident does not exist. Additionally, due to radioactive decay, a fuel handling accident involving handling irradiated fuel that has decayed greater than 100 hours0.00116 days <br />0.0278 hours <br />1.653439e-4 weeks <br />3.805e-5 months <br /> results in doses that are well within the guideline values specified in 10 CFR 50.67 even without containment closure capability. Therefore, under these conditions no requirements need be placed on containment penetration status for irradiated fuel that has decayed greater than 100 hours0.00116 days <br />0.0278 hours <br />1.653439e-4 weeks <br />3.805e-5 months <br />. To provide defense-in-depth, requirements are placed on the containment penetrations until the irradiated fuel has decayed greater than 30 days. This relaxation of TS requirements when moving irradiated fuel that has decayed greater than 30 days is only used in the Spent Fuel Pool Sweep System TS.

If the containment equipment hatch, air locks, or any containment penetration that provides direct access from the containment atmosphere to the outside atmosphere is not in the required status, including the Containment Purge and Exhaust Isolation System not capable of automatic actuation when the purge and exhaust valves are open, the unit must be placed in a condition where the isolation function is not needed. This is accomplished by immediately suspending movement of recently irradiated fuel assemblies within containment. Performance of these actions shall not preclude completion of movement of a component to a safe position.

A minimum shutdown margin of greater than or equal to 5% Ak/k must be maintained in the core.

The boron concentration as specified in the COLR is sufficient to ensure an adequate margin of safety. The specification for REFUELING OPERATIONS shutdown margin is based on a dilution during refueling accident.( 2 ) With an initial shutdown margin of 5% Ak/k, under the postulated accident conditions, it will take longer than 30 minutes for the reactor to go critical. This is ample time for the operator to recognize the audible high count rate signal, and isolate the reactor makeup water system. Periodic checks of refueling water boron concentration ensure that proper shutdown margin is maintained. Specification 3.8.a.6 allows the control room operator to inform the manipulator operator of any impending unsafe condition detected from the main control board indicators during fuel movement.

Interlocks are utilized during REFUELING OPERATIONS to ensure safe handling. Only one assembly at a time can be handled. The fuel handling hoist is dead weight tested prior to use to assure proper crane operation. It will not be possible to lift or carry heavy objects over the spent fuel pool when fuel is stored therein through interlocks and administrative procedures. Placement of additional spent fuel racks will be controlled by detailed procedures to prevent traverse directly above spent fuel.

The one hundred forty-eight hour decay time following plant shutdown is consistent with the spent fuel pool cooling analysis and also bounds the assumption used in the dose calculation for the fuel handling accident. The requirement for the spent fuel pool sweep system, including charcoal adsorbers, to be operating when spent fuel movement is being made provides added assurance that the off-site doses will be within acceptable limits in the event of a fuel handling accident. The spent fuel pool sweep system is designed to sweep the atmosphere above the refueling pool and release to the Auxiliary Building vent during fuel handling operations. Normally, the charcoal adsorbers are bypassed but for purification operation, the bypass dampers are closed routing the air flow through the charcoal adsorbers. If the dampers do not close tightly, bypass leakage could exist to negate the usefulness of the charcoal adsorber. If the spent fuel pool sweep system is found not to be operating, fuel handling within the Auxiliary Building will be terminated until the system can be (2) USAR Section 14.1 TS B3.8-3

Completion Time is based on the low probability of a DBA occurring during this time period, and ability of the remaining train to provide the required capability.

If the control room boundary is inoperable when critical, the CRPARS trains cannot perform their intended functions. Actions must be taken to restore an OPERABLE control room boundary within 24 hours2.777778e-4 days <br />0.00667 hours <br />3.968254e-5 weeks <br />9.132e-6 months <br />. During the period that the control room boundary is inoperable, appropriate compensatory measures (consistent with the intent of GDC 19) should be utilized to protect control room operators from potential hazards such as radioactive contamination, toxic chemicals, smoke, temperature and relative humidity, and physical security. Preplanned measures should be available to address these concerns for intentional and unintentional entry into the condition. The 24 hour2.777778e-4 days <br />0.00667 hours <br />3.968254e-5 weeks <br />9.132e-6 months <br /> Completion Time is reasonable based on the low probability of a DBA occurring during this time period, and the use of compensatory measures. The 24 hour2.777778e-4 days <br />0.00667 hours <br />3.968254e-5 weeks <br />9.132e-6 months <br /> Completion time is a typically reasonable time to diagnose, plan and possibly repair, and test most problems with the control room boundary.

When critical, if the inoperable CRPARS train or control room boundary cannot be restored to OPERABLE status within the required Completion Time, the unit must be placed in a MODE that minimizes accident risk. To achieve this status, the unit must be placed in at least HOT SHUTDOWN within 12 hours1.388889e-4 days <br />0.00333 hours <br />1.984127e-5 weeks <br />4.566e-6 months <br />. The allowed Completion Time is reasonable, based on operating experience, to reach the required unit conditions from full power conditions in an orderly manner and without challenging unit systems.

During REFUELING OPERATIONS, if the inoperable CRPARS train cannot be restored to OPERABLE status within the required Completion Time, action must be taken to immediately place the OPERABLE CRPARS train in the emergency mode. This action ensures that the remaining train is OPERABLE, that no failures preventing automatic actuation will occur, and that any active failure would be readily detected. An alternative is to immediately suspend activities that could result in a release of radioactivity that might require isolation of the control room. This places the unit in a condition that minimizes risk. This does not preclude the movement of fuel to a safe position.

TS 3.12.b imposes restrictions on movement of heavy loads over the spent fuel pool. NUREG-0612, "Control of Heavy Loads at Nuclear Power Plants," dated July 1980, provides guidelines and recommendations to assure safe handling of heavy loads by prohibiting, to the extent practicable, heavy load travel over stored spent fuel assemblies, fuel in the reactor core, safety-related equipment, and equipment needed for decay heat removal. The NUREG defines a heavy load as any load carried in a given area during the operation of the plant that weighs more than the combined weight of a single spent fuel assembly and its associated handling tool. This restriction stems from these guidelines.

During REFUELING OPERATIONS, with two CRPARS trains inoperable, action must be taken immediately to suspend activities that could result in a release of radioactivity that might enter the control room. This places the unit in a condition that minimizes accident risk. This does not preclude the movement of fuel to a safe position.

If both CRPARS trains are inoperable when critical for reasons other than an inoperable control room boundary, the CRPARS may not be capable of performing the intended function and the unit is in a condition outside the accident analyses. Therefore, a plant shutdown as stated in TS 3.12.a.2 must be initiated.

TS B3.12-2

BASIS - Control Room Post-Accident Recirculation Svstem (TS 3.12)

The Control Room Post-Accident Recirculation System (CRPARS) is designed to filter the Control Room atmosphere during Control Room isolation conditions. The Control Room Post-Accident Recirculation System is designed to automatically start upon SIS or high radiation signal. The CRPARS consists of two independent, redundant trains that recirculate and filter the control room air. Each train consists of a prefilter, a high efficiency particulate air (HEPA) filter, an activated charcoal adsorber section, for removal of gaseous activity (principally iodines), and a fan. Ductwork, valves or dampers, and instrumentation also form part of the system.

The CRPARS satisfies Criterion 3 of 10 CFR 50.36(c)(2)(ii) for design basis accidents and fuel handling accidents.

Two independent and redundant CRPARS trains are required to be OPERABLE to ensure that at least one is available assuming a single failure disables the other train. Total system failure could result in exceeding a dose of 5 rem to the control room operator in the event of a large radioactive release. The CRPARS is considered OPERABLE when the individual components necessary to limit operator exposure are OPERABLE in both trains.

A CRPARS train is OPERABLE when the associated:

a. Fan is OPERABLE,

b. HEPA filters and charcoal adsorbers 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. Allowing the control room boundary to be opened intermittently under administrative controls modifies the TS. For entry and exit through doors, the administrative control of the opening is performed by the person(s) entering or exiting the area.

For other openings, these controls consist of stationing a dedicated individual at the opening who is in continuous communication with the control room. This individual will have a method to rapidly close the opening when a need for control room isolation is indicated.

When CRITICAL and during REFUELING OPERATION, CRPARS must be OPERABLE to control operator exposure during and following a DBA. In all Modes, the CRPARS is required to cope with a release from the rupture of a gas decay tank (GDT) or the Volume Control Tank (VCT). Although required to cope with a release from the gas decay tank or the Volume Control Tank the requirement does not meet the criteria found in 10 CFR 50.36 for inclusion in Technical Specifications, therefore the requirement for operability of the CRPARS when in all modes due to the presents of radioactive gases in the GDT or VCT is not included in the Technical Specifications. During movement of irradiated fuel assemblies that have decayed less than 30 days, the CRPARS must be OPERABLE to cope with the release from a fuel handling accident.

When one CRPARS train is inoperable, action must be taken to restore OPERABLE status within 7 days. In this Condition, the remaining OPERABLE CRPARS train is adequate to perform the control room protection function. However, the overall reliability is reduced because a single failure in the OPERABLE CRPARS train could result in loss of CRPARS function. The 7 day TS B3.12-1

TS 4.4.f.3.B requires verification that each containment isolation manual valve and blind flange located inside containment and not locked, sealed, or otherwise secured and required to be closed during accident conditions, is closed. The TS helps to ensure that post-accident leakage of radioactive fluids or gases outside of the containment boundary is within design limits. For containment isolation valves inside containment, the frequency of "prior to entering INTERMEDIATE SHUTDOWN from COLD SHUTDOWN if not performed within the previous 92 days" is appropriate since these containment isolation valves are operated under administrative controls and the probability of their misalignment is low. The TS specifies that containment isolation valves that are open under administrative controls are not required to meet the TS during the time they are open.

This TS does not apply to valves that are locked, sealed, or otherwise secured in the closed position, since these were verified to be in the correct position upon locking, sealing, or securing.

TS 4.4.f.3.C modifies TS 4.4.f.3 for valves and blind flanges located in high radiation areas and allows these devices to be verified closed by use of administrative means. Allowing verification by administrative means is considered acceptable, since access to these areas is typically restricted when above COLD SHUTDOWN for ALARA reasons. Therefore, the probability of misalignment of these containment isolation valves, once they have been verified to be in the proper position, is small.

Equipment Hatch closure capability (TS 4.4.q)

This surveillance demonstrates that the equipment hatch is capable of being closed within the required time prior to movement of irradiated fuel. Repeating this closure capability demonstration weekly provides additional assurance that if called on to be closed in an emergency closure will occur expeditiously. The surveillance interval was selected to be commensurate with the normal duration of time to complete the fuel handling operations Additionally, this surveillance demonstrates that the necessary hardware, tools, and equipment are available to install the equipment hatch. The equipment hatch is provided with a set of hardware, tools, and equipment for moving the hatch from its storage location and installing it in the opening.

The required set of hardware, tools, and equipment shall be inspected to ensure that they can perform the required functions. The surveillance is performed daily during REFUELING OPERATIONS when the equipment hatch is open. . The daily surveillance is adequate considering that the hardware, tools and equipment are dedicated to the equipment hatch and not used for any other function.

TS B4.4-5

ATTACHMENT 6 NUCLEAR MANAGEMENT COMPANY, LLC KEWAUNEE NUCLEAR PLANT DOCKET 50-305 January 16, 2004 Letter from Thomas Coutu (NMC)

To Document Control Desk (NRC)

• LAR 201 Commitments 1 Page Follows

Docket 50-305 NRC-04-006 January 16, 2004 , page 1 LIST OF COMMITMENTS The following table identifies those actions committed to by the Nuclear Management Company (NMC) in this document. Any other statements in this submittal are provided for information purposes and are not considered to be commitments. Please direct questions regarding these commitments to Mr. Gerald Riste at KNPP, (920) 388-8424.

COMMITMENT Due Date/Event Written procedures will be developed describing compensatory measures to be taken in the event of an intentional or unintentional entry into a condition where Pror to Implementation.

both trains of Control Room Post-Accident Recirculation System are inoperable.

Written procedures will be developed describing measures to be taken to close the containment equipment hatch in the Prior to implementation.

event of adverse weather conditions (Tornado Watch)

Administrative controls consisting of written procedures will be established that would require: 1) appropriate personnel are aware of the open status of the containment during movement of recently irradiated fuel or REFUELING OPERATIONS, 2) specified individuals and equipment are designated and readily available to close the equipment hatch following an evacuation that would occur in - Prior to implementation.

the event of a fuel handling accident, and

3) any obstructions (e.g., cables and hoses) that would prevent closure of an open equipment hatch can be quickly removed,4) Procedures to verify closure of the equipment hatch within 90 minutes to be completed prior to the start of refueling activities where the equipment hatch will be open NMC will inform the State and County Emergency Governments of this accident Prior to Implementation.

scenario Accident scenario(s) will be developed to train the emergency response organization Within 6 months after approval.

on this scenario.

Training will be developed to train those designated to close the equipment hatch. Prior to Implementation.

ATTACHMENT 7 NUCLEAR MANAGEMENT COMPANY, LLC KEWAUNEE NUCLEAR PLANT DOCKET 50-305 January 16, 2004 Letter from Thomas Coutu (NMC)

To Document Control Desk (NRC)

LAR 201 REQUEST FOR ADDITIONAL INFORMATION REGARDING PROPOSED AMENDMENT TO TECHNICAL SPECIFICATIONS FUEL HANDLING ACCIDENT ANALYSES SAN ONOFRE NUCLEAR GENERATING STATION UNITS 2 AND 3 DOCKET NO 50-361 AND 50-362 NMC Responses 10 Pages Follow

Docket 50-305 NRC-04-006 January 16, 2004 , page 1

1. What design bases parameters assumptions or methodologies (other than those provided in the August 4, 2003 submittal) were changed in the radiological design basis accident analyses as a result of the proposed change? If there are many changes it would be helpful to compare and contrast them in a table. Also, please provide justification for any changes.

NMC Response NRC approved the use of the Alternate Source Term (AST) methodology for the Kewaunee Nuclear Power Plant (KNPP) in a safety evaluation (SE) enclosed in a letter dated March 17, 2003 (Adams Accession NO. ML030210062), "Kewaunee Nuclear Power Plant - Issuance of Amendment Regarding Implementation of Alternate Source Term (TAC NO. MB4596)."

This letter referenced calculations performed by Westinghouse to support the AST License Amendment Request (LAR). The specific one used in the LAR for the fuel handling accident (FHA) was CN-CRA-00-56, Revision 0, "Kewaunee alternate source term (AST) FHA Dose." In August 2002, this calculation was revised (revision1) to address revised source terms in support of NMC's Stretch Power Uprate (SPU) for the KNPP and to document sensitivity to the control room unfiltered inleakage rate. NMC had revision 0 performed at a power level of 1650 MWt increased to 1683 MWt to cover uncertainties. The reactor core fission product inventory and the reactor coolant fission product inventory were then increased by 10% to allow for future KNPP power uprating.

In revision 1 of the KNPP FHA analysis the core power for the analysis was 1772 MWt increased to 1782.6 MWt to cover uncertaincies. Also a sensitivity case was performed with 400 cfm control room unfiltered inleakage. The following table lists the results of these analysis.

KNPP Fuel Handling Accident Dose Results CN-CRA-00-56, Revision 1 Site Boundary Low Population Control Room Control Room Dose Zone Dose Dose (200 cfm Dose (400 cfm unfiltered unfiltered inleakage) inleakage)

TEDE (rem) 0.7 0.11 1.0 1.5 Limits (rem) 6.3 6.3 5.0 5.0

2. Based upon a preliminary review of the fuel handling accident for the proposed change the reviewer is unable to match the calculated doses. It would be helpful if the licensee would provide their design bases fuel handling calculation. If the calculation is provided, answers to questions provided in this request for additional information may reference the calculation.

NMC Response

Docket 50-305 NRC-04-006 January 16, 2004 , page 2 See response to question 1.

3. What types of hoses and cables will be allowed to pass through the open equipment hatch? What provisions will be made for the designated individual to separate these to close the air lock door while reducing hazards from these hoses and cables?

NMC Response NMC does not plan to run any hoses or cables through the open equipment hatch. If hoses or cables are required to be run through the open equipment hatch, isolation devices (e.g. isolation valve(s)) will be provided to promptly isolate the hose or cable and allow for separation of the hose or cable.

4. A value of 1000 cfm is assumed for the value for unfiltered inleakage into the control room. Because this value is not based upon a measurement, sufficient justification should be provided to explain why this number is appropriate.

Provide sufficient details regarding your control room, design, maintenance and assessments to justify the use of this number and your plans to verify this number.

NMC Response KNPP accident analysis assumes a 200 cfm unfiltered inleakage into the KNPP control room, which is an isolation control room versus a pressurized control room. Unfiltered inleakage (UFI) can enter the control room boundary through the following paths:

• Leakage through Dampers

• Leakage through Doors

• Leakage through penetrations

• Leakage through the wall KNPP UFI through dampers has been measured. This leakage is from measurements taken in support of the KNPP Control Room Habitability analysis done in 1989. In this analysis the KNPP Control room ventilation system was placed in the emergency mode and inleakage through dampers was measured.

This analysis is where the 200 cfm UFI is derived. In the same analysis 10 cfm was assumed to enter the control room through doors To preclude UFI into the control room from penetrations and through the wall NMC performs routine inspections of these penetrations and walls to ensure their structural integrity.

5. The proposed technical specification specifies that a "designated" crew is available to close the Containment Structure Equipment Hatch Shield Doors rather than a "dedicated" crew who would have no other duties. Specify what

Docket 50-305 NRC-04-006 January 16, 2004 , page 3 other duties the designated crew will have and where they will be stationed relative to the air lock doors.

NMC Response The designated crew will be individuals stationed in containment who have been trained in closure of the equipment hatch on receipt of instructions to close either by the containment evacuation alarm or verbally.

6. Provide a detailed account of the timing and flow rates, and filtration of the control room HVAC as it responds to the accident a schematic would be helpful.

NMC Response See KNPP USAR section 9.6.4, "Control Room Air Conditioning System," and USAR figure 9.6-4, "Control Room Air Conditioning System Flow Diagram," for a description of the Control Room Air Condition System operation in normal and emergency operating modes.

One analysis was performed to bound both the FHA in containment and the FHA in the auxiliary building. The volume of the buildings was not used as input to the analysis since all that matters is the release. All activity released from the fuel pool is assumed to be released to the atmosphere in two hours, using an exponential release model with higher releases in the initial periods since this is conservative for the control room doses.

The FHA has certain input assumptions. Normal Control Room HVAC unfiltered air inflow is 2500 cfm, conservatively increased by 10% to 2750 cfm for the analysis to maximize the activity entering the control room. During emergency operation, HVAC filtered recirculation air flow is 2500 +/-10% cfm, conservatively assumed to be at the low end value of 2250 cfm for the analysis to minimize cleanup of activity in the control room. Emergency HVAC unfiltered inleakage air flow is 200 cfm. Additionally, a sensitivity case with 400 cfm of unfiltered inleakage was performed. There is no filtered air inflow to the control room in either HVAC mode. Control room filter efficiency is 90% for elemental and organic Iodine. (No particulate iodine is released from the pool.) The run is extended to 30 days to calculate the accumulation of dose to the operators in the control room.

7. Please provide engineering drawings of the proposed change. A photograph of the equipment hatch would also be helpful in the review of this proposed change.

Describe the steps taken to insure that the proposed flashing will not interfere with closure of the shield doors. What is the acceptable design clearance between the flashing on the shield doors and the containment?

NMC Response Not applicable to KNPP as KNPP is not using flashings to provide the barrier, see attached drawing of the KNPP Equipment Hatch for hatch details.

Docket 50-305 NRC-04-006 January 16, 2004 , page 4

8. Provide the criterion used to decide if the Equipment Hatch Shield Doors are capable of being closed within 90 minutes.

NMC Response During normal operation of the containment equipment hatch closure time is estimated to be approximately 15 minutes. NMC has contracted to a vendor to provide transfer rails that can be disassembled quickly to allow the equipment hatch to be closed. NMC will demonstrate that the equipment hatch can be closed within 90 minutes prior to commencing movement of irradiated fuel in containment with the equipment hatch open then will demonstrate this capability every weekly. Additionally, NMC will verify the equipment necessary to close the equipment hatch is readily available daily.

9. Provide the Low Population Zone and Beta doses consistent with the information provided in current Updated FSAR.

NMC Response See response to question 1.

10. What criteria will be used to determine if closure of the containment is necessary in the event that environmental conditions could impact fuel handling? Has the impact of wind on fuel handling been evaluated (for example. reduced pool visibility due to pool surface disruption)? What steps would be taken in the event of severe weather to minimize the impact of flying debris?

NMC Response In the event where weather conditions exist which could form a tornado (Tornado Watch) the equipment hatch will be closed and secured with four bolts.

The design of the KNPP containment and internal components precludes disruption of the refueling cavity pool surface. The equipment hatch is located below the elevation of the refueling cavity pool with no direct path to the pool surface. Any wind blowing into containment from an open equipment hatch would have to be redirected towards the top of containment where the free air space would dissipate the force of the wind.

11. There appears to be inconsistencies between value used in the Updated FSAR and the values provided in Table 1 of the submittal without justification provided for the changes. Please verify the parameter provided in Table 1 against those in the analysis utilized to justify this amendment request and provide a justification for the changes in values from those previously accepted.

Docket 50-305 NRC-04-006 January 16, 2004 , page 5 Parameter Table 1 UFSAR Value UFSAR Location Inflow & Elemental & No credit taken for Page 6.4-4, section Recirculation Filter Organic Iodine 95% charcoal absorbers C, Carbon Efficiencies Particulate 99% Adsorbers

___________I____ ________________ ________________ A lso. Table 158-5 NMC Response Not applicable to KNPP as this is a specific question associated with the SONGs submittal and their UFSAR.

12. Criterion 64 of 10 CFR 50 Appendix A states that means shall be provided for monitoring the reactor containment atmosphere, spaces containing components for recirculation of loss-of-coolant accident fluids, effluent discharge paths, and the plant environs for radioactivity that may be released from normal operations, including anticipated operational occurrences and from postulated accidents. The proposed change should consider how Criterion 64 will be met in the event of a FHA with the equipment hatch open. Moreover, this information should be included as part of the Bases discussion; Provide the bases for meeting Criterion 64 for the proposed change. Please confirm that your emergency planning dose assessment methodology includes the ability to assess this accident. For example, does your methodology include the capability to determine the source term, release rate out of containment, meteorology and consider feedback via field monitoring health physics survey teams? Have you evaluated the need for any special radiological monitoring or survey equipment (ie.; in-plant equipment or field team survey equipment) to evaluate the radiological conditions of this accident scenario? Will your emergency response personnel be trained to deal with this accident scenario?

NMC Response To ensure Criterion 64 is meet during refueling operations the operations staff performs two checklists, a pre-refueling checklist and a refueling daily checklist.

These checklists verify six radiation monitors are operating. They are

1. R-2, 'Containment Area Monitor."

2. R-5, "Fuel Handling Area Monitor" (Spent Fuel Pool area)

3. R-1 1, "Containment Particulate Monitor."

4. R-12, "Containment Gas Monitor."

5. R-21, "Containment Vent Monitor."

6. R-23, "Control Room Vent Monitor."

The area radiation monitor R-2 is located on the containment refueling floor and the suction for the process radiation monitors R-11 and R-12 is located on the

Docket 50-305 NRC-04-006 January 16, 2004 , page 6 refueling floor. If leakage were to occur from a spent fuel assembly that resulted in excessive radioactive levels escaping the reactor cavity pool then those excessive levels would be monitored by one of these radiation monitors or by one of the other radiation monitors should one of these fail. As the excessive radioactivity must escape the refueling cavity pool first on the refueling floor, which is monitored for radioactivity, any excessive radioactivity that could leave containment would be previously detected. Additionally, a continuous air monitor is placed by the open equipment hatch to identify releases through the hatch.

Therefore, criterion 64 is met.

KNPP Emergency Plan Implementing Procedures (EPIPs) contain emergency action levels (EALs) for activation of the emergency response organization (ERO) based on accident conditions associated with a FHA inside containment.

Based on the severity of the accident an Alert or Site Emergency classification can be entered. If during refueling operations a fuel damage accident occurs with release of radioactivity to containment, as determined by containment radiation monitors R-1 1 or R-1 2 (R-1 1 is a particulate radioactivity monitor while R-12 is a gaseous radioactivity monitor) and a confirming report, an Alert emergency classification is entered. If a spent fuel assembly or other large object was dropped then a Site Emergency emergency classification is entered.

Either of these emergency classifications will activate the entire ERO, including the environmental monitoring teams.

Additionally an active radiation monitor is required to be on the refueling bridge during refueling operations, which along with the constant communications provides additional assurance that any accident would be expeditiously communicated to the control room and to those individuals tasked with closure of the equipment hatch.

KNPP emergency planning dose assessment methodology includes the ability to assess this accident. KNPP dose assessment uses the RASCAL 3.0.3 to assess the consequences of this type of accident. This program contains options that include an underwater release that has inputs for the assumed amount of fuel damaged to determine the source term and inputs on the assumed release rate out of containment. Feedback from the field surveys can be inputted into the program to re-assess the consequences of the accident.

Accident scenario(s) will be developed to train the emergency response organization on this scenario.

13. The proposed change states:

'With the proposed TS 3.9.3 changes; the crew tasked with closing the containment shield doors as a means of providing for containment closure will be performing this activity from outside containment." Since containment is unlikely to become pressurized during an in-containment fuel handling accident during refueling, there is no motive force for airborne radioactivity to be propelled

Docket 50-305 NRC-04-006 January 16, 2004 , page 7 through the opening. As a result, the dose to the crew is anticipated to be minimal."

Provide justification for the statement that "there is no motive force for airborne radioactivity" considering the motive force that may be caused by 11 in containment heat sources, 2) the pressure from external sources such as wind, or interfaces with pressurized buildings, or 3) heating of the containment by the sun.

NMC Response KNPP has two separate structures preventing leakage to the environment, the containment vessel and the shield building. The containment vessel is a freestanding steel structure around which is a five-foot air space with an approximate two and one half foot reinforced concrete structure (shield building).

During refueling operations with the equipment hatch open the pressure between the outside environment and the containment will essentially equalize. Although some momentary differences may occur due to wind gusts, the pressures should be essentially equal.

The other building that the containment interfaces with is the auxiliary building through the personnel air locks. If the air locks are closed then there is an equalization process that must be preformed to open the air locks. This process equalizes the pressure from the inside of the air lock to containment. The amount of airflow that could flow into containment from this process is insignificant. If the air locks are open then the pressures would equalize and, excluding momentary wind gusts, the pressures would be equal, therefore no motive force.

Air coolers hold the temperature in containment essentially constant. Following a FHA these air coolers remain in service keeping the containment air temperature constant preventing any motive force due to heating of the air.

14. 10 CFR 50.36 states that:

"'A technical specification limiting condition for operation of a nuclear reactor must be established for each item meeting one or more of the following criteria:

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

The proposed analysis utilizes an initial condition of 72 hours8.333333e-4 days <br />0.02 hours <br />1.190476e-4 weeks <br />2.7396e-5 months <br /> of fuel decay for the fuel handling accident. The proposed technical specification does not provide a limiting condition of operation for this initial condition. Please justify why this decay time does not meet criterion 2 of 10 CFR 50.36 or modify the technical specification to include the decay time.

Docket 50-305 NRC-04-006 January 16, 2004 , page 8 NMC Response Not applicable for KNPP as KNPP technical specifications contain time limits for movement of spent fuel assemblies.

EP Considerations

15. Will your Emergency Plan be updated to include an accident release through the equipment hatch? Will your Emergency Operating Procedures be updated to address the specific details needed to respond to this accident scenario?

NMC Response No. Current KNPP EPIPs contain adequate information to handle an accident release through the containment equipment hatch. Procedures will be updated to include closure of the equipment hatch during adverse weather conditions and on indications of a FHA.

16. Will you inform the State Emergency Response personnel about this accident scenario?

NMC Response Yes, the State Emergency Response Personnel be informed of this accident scenario. Additionally, the county emergency governments will be informed of this accident scenario.

Control Room Atmospheric Dispersion Factors

17. The control room (CR) radiological analysis supporting this license amendment request is based on a fuel handling accident inside containment with the containment open to the outside environment. All airborne radioactivity reaching the containment atmosphere is assumed to be exhausted within two hours to the outside environment via the open containment equipment hatch shield doors.

This analysis uses a CR atmospheric dispersion factor (X/Q value) of 3.1 E-3 sec/M 3 which is presented in Section 2.3.4.2 of the San Onofre 2&3 UFSAR.

UFSAR Section 2.3.4.2 states that the CR X/Q value of 3.1 E-3 sec/M 3 is based on the Murphy & Campe diffuse source-point receptor algorithm. This algorithm is applicable when activity is assumed to leak from many points on the surface of the containment in conjunction with a single point receptor (i.e., CR air intake);

that is, the activity is assumed to be homogeneously distributed throughout the containment and the release rate is assumed to be reasonably constant over the surface of the building. This is not the situation in this accident scenario where the release is assumed to occur through the open containment equipment hatch shield doors. As such, please justify the use of the Murphy & Campe diffuse source-point receptor algorithm in this analysis.

Docket 50-305 NRC-04-006 January 16, 2004 , page 9 NMC Response Kewaunee also uses the Murphy & Campe diffuse source-point receptor algorithm in the analysis for FHA which resulted in a X/Q value of 2.93 E-3 sec/M 3 . Sensitivity analysis show that the X/Q value would differ by a factor of approximately three if the point source - point receptor were used. This value is used in the analysis in the timing of the isolation of the control room and in the source term for the unfiltered inleakage into the control room. If this term were used the dose consequences to the operators may be increased by a factor of three. Where this to occur the dose would increase from 1.0 rem to 3.0 rem TEDE for 200 cfm unfiltered inleakage and from 1.5 rem to 4.5 rem TEDE for 400 cfm unfiltered inleakage.

Additionally, the equipment hatch is located approximately 1800 around the containment structure from the control room intakes. Elimination from the equipment hatch would travel around the containment structure causing additional dispersion due to the wake effect of the containment structure.

18. If the Murphy & Campe diffuse source-point receptor algorithm is to be used in this analysis, UFSAR Section 2.3.4.2 states that a value of 180 ft (54.9 m) was assumed for the distances between the containment surface and receptor location (i.e., CR air intake). UFSAR Figure 6.4-3 shows the location of the two emergency CR air intakes with respect to the Unit 2 and Unit 3 containment structures and seems to indicate that the distance S between the closest containment surface and each air intake is more like 90 ft (27.4 m) rather than 180 1t (54.9 m). Please justify the continued use of 180 ft (54.9 m) for the value of s in this analysis.

NMC Response

-Not applicable to KNPP as this is a site-specific question concerning the layout of the San Onofre site. KNPP containment has the diameter of 120 feet with the distance from containment to the control room inlet as 75 feet.

Docket 50-305 NRC-04-006 January 16, 2004 , page 10 KNPP Equipment Hatch I