ML070430020

Radiological Accident Analysis

ML070430020
Person / Time
Site:	Kewaunee
Issue date:	03/08/2007
From:	Robert Kuntz NRC/NRR/ADRO/DORL/LPLIII-1
To:	Christian D Dominion, Dominion Energy Kewaunee
Kuntz, Robert , NRR/DLPM, 415-3733
Shared Package
ML070430017	List: ML070430020
References
TAC MC9715
Download: ML070430020 (29)

Text

March 8, 2007Mr. David A. ChristianSenior Vice President and Chief Nuclear Officer Innsbrook Technical Center 5000 Dominion Boulevard Glen Allen, VA 23060-6711

SUBJECT:

KEWAUNEE POWER STATION - ISSUANCE OF AMENDMENT RE:

RADIOLOGICAL ACCIDENT ANALYSIS AND ASSOCIATED TECHNICALSPECIFICATIONS CHANGE (TAC NO. MC9715)

Dear Mr. Christian:

The U.S. Nuclear Regulatory Commission has issued the enclosed Amendment No. 190 toFacility Operating License No. DPR-43 for the Kewaunee Power Station. This amendment revises the Technical Specifications in response to your application dated January 30, 2006, as supplemented by letter dated January 23, 2007.The amendment revises radialogical accident analyses and associated technical specifications.A copy of the NRC staff's Safety Evaluation is also enclosed. The Notice of Issuance will beincluded in the Commission's next regular biweekly Federal Register notice.Sincerely, /RA/Robert F. Kuntz, Project ManagerPlant Licensing Branch III-1 Division of Operating Reactor Licensing Office of Nuclear Reactor RegulationDocket No. 50-305

Enclosures:

1. Amendment No. 190 to License No. DPR-43

2. Safety Evaluationcc w/encls: See next page Mr. David A. ChristianMarch 8, 2007Senior Vice President and Chief Nuclear Officer Innsbrook Technical Center 5000 Dominion Boulevard Glen Allen, VA 23060-6711

SUBJECT:

KEWAUNEE POWER STATION - ISSUANCE OF AMENDMENT RE:

RADIOLOGICAL ACCIDENT ANALYSIS AND ASSOCIATED TECHNICALSPECIFICATIONS CHANGE (TAC NO. MC9715)

Dear Mr. Christian:

The U.S. Nuclear Regulatory Commission has issued the enclosed Amendment No. 190 toFacility Operating License No. DPR-43 for the Kewaunee Power Station. This amendment revises the Technical Specifications in response to your application dated January 30, 2006, as supplemented by letter dated January 23, 2007.The amendment revises radialogical accident analyses and associated technical specifications.A copy of the NRC staff's Safety Evaluation is also enclosed. The Notice of Issuance will beincluded in the Commission's next regular biweekly Federal Register notice.Sincerely, /RA/Robert F. Kuntz, Project ManagerPlant Licensing Branch III-1 Division of Operating Reactor Licensing Office of Nuclear Reactor RegulationDocket No. 50-305

Enclosures:

1. Amendment No. 190 to License No. DPR-43

2. Safety Evaluationcc w/encls: See next pageDISTRIBUTION:PUBLICLPL3-1 r/fRidsNrrDorlLpleRidsNrrPMRKuntz RidsNrrLATHarrisRidsOGCRpRidsAcrsAcnwMailCenterRidsNrrDirsltsb GHill, OISRidsRgn3MailCenterRidsNrrDorlDprADAMS ACCESSION NOs.:PKG: ML070430017 Amd: ML070430020 TS:ML070680320*See SE**NLO with commentsOFFICENRR/LPL3-1/PMNRR/LPL3-1/LANRR/AADB/BCNRR/SCVB/BCOGCNRR/LPL3-1/BCNAMERKuntzTHarrisMKotzalasRDennig*TCampbell**LRaghavanDATE 3/07/07 3/06/07 2 /22/07 7/21/07 2/28/07 3/08/07OFFICIAL RECORD COPY Kewaunee Power Station cc:

Resident Inspectors OfficeU.S. Nuclear Regulatory Commission N490 Hwy 42 Kewaunee, WI 54216-9510Regional Administrator, Region IIIU.S. Nuclear Regulatory Commission Suite 210 2443 Warrenville Road Lisle, IL 60532-4351Ms. Leslie N. HartzDominion Energy Kewaunee, Inc.

Kewaunee Power Station N 490 Highway 42 Kewaunee, WI 54216Mr. Chris L. FunderburkDirector, Nuclear Licensing and Operations Support Innsbrook Technical Center 5000 Dominion Boulevard Glen Allen, VA 23060-6711Mr. Thomas L. BreeneDominon Energy Kewaunee, Inc.

Kewaunee Power Station N490 Highway 42 Kewaunee, WI 54216Ms. Lillian M. Cuoco, Esq.Senior Counsel Dominion Resources Services, Inc.

Millstone Power Station Building 475, 5th Floor Rope Ferry Road Waterford, CT 06385 DOMINION ENERGY KEWAUNEE, INC.DOCKET NO. 50-305KEWAUNEE POWER STATIONAMENDMENT TO FACILITY OPERATING LICENSE Amendment No. 190 License No. DPR-431.The U.S. Nuclear Regulatory Commission (the Commission) has found that:A.The application for amendment by Dominion Energy Kewaunee, Inc. dated January 30, 2006 as supplemented by letter dated January 23, 2007, complies with the standards and requirements of the Atomic Energy Act of 1954, as amended (the Act), and the Commission's rules and regulations set forth in 10 CFR Chapter

I;B.The facility will operate in conformity with the application, the provisions of the Act,and the rules and regulations of the Commission;C.There is reasonable assurance (i) that the activities authorized by this amendmentcan be conducted without endangering the health and safety of the public, and (ii) that such activities will be conducted in compliance with the Commission's regulations;D.The issuance of this amendment will not be inimical to the common defense andsecurity or to the health and safety of the public; andE.The issuance of this amendment is in accordance with 10 CFR Part 51 of theCommission's regulations and all applicable requirements have been satisfied. 2.Accordingly, the license is amended by changes to the Technical Specifications asindicated in the attachment to this license amendment, and paragraph 2.C.(2) of Facility Operating License No. DPR-43 is hereby amended to read as follows: (2)Technical SpecificationsThe Technical Specifications contained in Appendix A, as revised throughAmendment No. 190, are hereby incorporated in the license. The licensee shall operate the facility in accordance with the Technical Specifications. 3.This license amendment is effective as of its date of issuance and shall be implementedwithin 60 days of the date of issuance. FOR THE NUCLEAR REGULATORY COMMISSION/RA/Lakshminaras Raghavan, ChiefPlant Licensing Branch III-1 Division of Operating Reactor Licensing Office of Nuclear Reactor Regulation

Attachment:

Changes to the Facility Operating License and Technical Specifications Date of Issuance: March 8, 2007 ATTACHMENT TO LICENSE AMENDMENT NO. 190FACILITY OPERATING LICENSE NO. DPR-43DOCKET NO. 50-305Replace the following page of the Facility Operating License No. DPR-43 with the attachedrevised page. The changed area is identified by a marginal line.REMOVEINSERTPage 3Page 3Replace the following pages of the Appendix A Technical Specifications with the attachedrevised pages. The revised pages are identified by amendment number and contain marginal lines indicating the areas of change. REMOVE INSERTTS 3.1-7TS 3.1-7TS 3.6-4TS 3.6-4 TS 6.20-1TS 6.20-1 C.This license shall be deemed to contain and is subject to the conditions specified inthe following Commission regulations in 10 CFR, Chapter 1: (1) Part 20, Section 30.34 of Part 30, Section 40.41 of Part 40, Section 50.54 and 50.59 of Part 50, and Section 70.32 of Part 70, (2) is subject to all applicable provisions of the Act and to the rules, regulations, and orders of the Commission now or hereafter in effect, and (3) is subject to the additional conditions specified or incorporated below:(1) Maximum Power LevelThe licensee is authorized to operate the facility at steady-state reactor corepower levels not in excess of 1772 megawatts (thermal).(2) Technical SpecificationsThe Technical Specifications contained in Appendix A, as revised throughAmendment No. 190, are hereby incorporated in the license. The licensee shall operate the facility In accordance with the Technical Specifications.(3) Fire ProtectionThe licensee shall implement and maintain in effect all provisions of the approved Fire Protection Program as described in the licensee's Fire Plan, and as referenced in the Updated Safety Analysis Report, and as approved in the Safety Evaluation Reports, dated November 25, 1977, and December 12, 1978 (and supplement dated February 13, 1981) subject to the following provision:The licensee may make changes to the approved Fire Protection Programwithout prior approval of the Commission, only if those changes would not adversely affect the ability to achieve and maintain safe shutdown in the event of a fire.(4) Physical ProtectionThe licensee shall fully implement and maintain in effect all provisions of theCommission-approved physical security, training and qualification, and safeguards contingency plans including amendments made pursuant to provisions of the Miscellaneous Amendments and Search Requirements revisions to 10 CFR 73.55 (51 FR 27817 and 27822) and to the authority of 10 CFR 50.90 and 10 CFR 50.54(p). The combined set of plans, which contain Safeguards Information protected under 10 CFR 73.21, is entitled: "Nuclear Management Company Kewaunee Nuclear Power Plant Physical Security Plan (Revision 0)" submitted by letter dated October 18, as supplemented by letter dated October 21, 2004, July 26, 2005, and May 15, 2006.(5) Fuel Burn-upThe maximum rod average Burn-up for any rod shall be limited to 60 GWD/MTUuntil completion of an NRC environmental assessment supporting an increased limit. Amendment No. 190 Revised by letter dated February 8, 2007 3

SAFETY EVALUATION BY THE OFFICE OF NUCLEAR REACTOR REGULATIONRELATING TO AMENDMENT NO. 190 TO FACILITY OPERATING LICENSE NO. DPR-43DOMINION ENERGY KEWAUNEE, INC.KEWAUNEE POWER STATIONDOCKET NO. 50-30

51.0INTRODUCTION

By letter dated January 30, 2006 (Agencywide Documents Access and Management System (ADAMS) Accession No. ML060540217), as supplemented by letter dated January 23, 2007 (ADAMS Accession No. ML070240543) Dominion Energy Kewaunee (DEK, the licensee) requested an amendment to facility Operating License No. DPR-43 for the Kewaunee Power Station (KPS). The licensee proposed to modify the currently approved radiological accident analyses and associated technical specifications (TSs). This proposed amendment incorporates TS changes to compensate for the higher control room emergency zone (CREZ) unfiltered inleakage measured during the American Society for Testing and Materials (ASTM)

E741 (tracer gas) leakage test conducted in December 2004. Results from the ASTM testing of the KPS control room envelope showed the CREZ unfilteredinleakage to be greater than that assumed in the approved radiological accident analyses. The revised CREZ unfiltered inleakage was determined to be a facility change, which caused an increase in the dose consequences of the approved radiological accident analyses. Currently, the KPS CREZ is operable but non-conforming. The resolution of this condition is to incorporate the increase in assumed CREZ unfiltered inleakage into the radiological accident analyses.The supplemental letter contained clarifying information, did not change the initial no significanthazards consideration determination, and did not expand the scope of the original FederalRegister notice.The licensee proposed changes to the TSs for certain plant parameters to compensate for thehigher measured CREZ unfiltered inleakage. This safety evaluation (SE) addresses the Nuclear Regulatory Commission (NRC) staff's review of the licensee's revised radiological accident analyses. In this license amendment request, the licensee proposed to change:1.TS 3.1.c.2.A, "Maximum Coolant Activity," coolant activity limit that requires intermediateshutdown from 60 Ci/gram DOSE EQUIVALENT I-131 to 20 Ci/gram DOSEEQUIVALENT I-131. 2The use of 0.25 SV (25rem) TEDE is not intended to imply that this value constitutes anacceptable limit for emergency doses to the public under accident conditions. Rather, this 0.25 SV (25 rem) TEDE value has been stated in this section as a reference value, which can be used in the evaluation of proposed design basis changes with respect to potential reactor accidents of exceedingly low probability of occurrence and low risk of public exposure to radiation.2.TS 3.6.c.3.B, "Performance Requirement," Shield Building Ventilation System and theAuxiliary Building Special Ventilation system filter removal efficiency from 95 percentradioactive methyl iodide removal to 97.5 percent radioactive methyl iodide removal.3.TS 6.20, "Containment Leakage Rate Testing Program," maximum allowable leakagerate from 0.5 weight percent of the contained air per 24 hours1 days <br />0.143 weeks <br />0.0329 months <br /> at the peak test pressure (P a) of 46 psig to 0.2 weight percent.

2.0REGULATORY EVALUATION

The U.S. Nuclear Regulatory Commission (NRC) staff finds that the licensee in Section 5.2 ofits January 30, 2006, submittal, identified the applicable regulatory requirements. The regulatory requirements and guidance which the staff considered in its review of the requested action are as follows:Title 10 of the Code of Federal Regulations (10 CFR) Part 50, Appendix A, General DesignCriterion 19 (GDC-19) requires, in part, that "[a] control room shall be provided from which actions can be taken to operate the nuclear power unit safely under normal conditions and to maintain it in a safe condition under accident conditions, including loss-of-coolant accidents.

Adequate radiation protection shall be provided to permit access and occupancy of the control room under accident conditions without personnel receiving radiation exposures in excess of 5 rem whole body, or its equivalent to any part of the body, for the duration of the accident."

Title 10 of the Code of Federal Regulations (10 CFR) 50.36 "Technical specifications," requiresthat "an applicant for a license authorizing operation of a production or utilization facility include proposed technical specifications" in its license application. Title 10 of the Code of Federal Regulations (10 CFR) 50.67 "Accident source term" establishesanalyzed dose limits for acceptable adoption of the accident source term. Title 10 of the Codeof Federal Regulations (10 CFR) 50.67(b)(2) states that "[t]he NRC may issue the amendmentonly if the applicant's analysis demonstrates with reasonable assurance that:(i) An individual located at any point on the boundary of the exclusion area for any 2-hourperiod following the onset of the postulated fission product release, would not receive a radiation dose in excess of 0.25 Sv (25 rem) 22 total effective dose equivalent (TEDE). (ii) An individual located at any point on the outer boundary of the low population zone, who isexposed to the radioactive cloud resulting from the postulated fission product release (during the entire period of its passage), would not receive a radiation dose in excess of 0.25 Sv (25 rem) TEDE. (iii) Adequate radiation protection is provided to permit access to and occupancy of the controlroom under accident conditions without personnel receiving radiation exposures in excess of 0.05 Sv (5 rem) TEDE for the duration of the accident."The NRC staff also considered the guidance contained in Regulatory Guide (RG) 1.183,"Alternate Radiological Source Terms for Evaluating Design Basis Accidents at Nuclear PowerReactors," RG 1.196, "Control Room Habitability at Light-Water Nuclear Power Reactors" as well as NUREG-0800 "Standard Review Plan for the Review of Safety Analysis Reports for Nuclear Power Plants" (SRP) Section 15.0.1"Radiological Consequence Analyses UsingAlternative Source Terms" in reviewing the licensee's amendment request.

3.0TECHNICAL EVALUATION

The NRC staff reviewed the licensee's analysis methods, assumptions, and inputs usingdocketed information provided by the licensee. These radiological accident analyses of record for the KPS license were previously docketed inAmendment No. 166, issued March 17, 2003 (ADAMS Accession No. ML030210062), which implemented an alternate source term; and Amendment No. 172, issued February 27, 2004 (ADAMS Accession No. ML040430633), which implemented a stretch power uprate to 1772 mega-watt thermal (MWt). These previously approved radiological accident analyses used the analytical methods and assumptions outlined in RG 1.183. The revised radiological accident analyses for design-basis accidents (DBA) incorporatechanges for the control room isolation parameters based on air flow measurements of ASTM E741 (tracer gas) testing conducted in response to Generic Letter 2003-01. The control room envelope unfiltered inleakage was measured by tracer gas testing onDecember 14 and 15, 2004. The leak test measured 409 +/- 29 cubic feet per minute (cfm) for system train A, and 447 +/- 51 cfm for train B. In the revised radiological accident analyses, DEKadjusted the control room unfiltered inleakage rate during normal mode heating ventilation and air conditioning (HVAC) operation to range between 1620 to 2750 cubic feet per minute (cfm).

DEK specified two unfiltered inleakage for control room emergency ventilation. For events that model isolation actuated by a safety injection (SI) set point, CREZ unfiltered inleakage was assumed to be at least 800 cfm. For events actuated by high radiation monitor in the control room air supply duct, the CREZ unfiltered inleakage was assumed to be at least 1500 cfm.

Isolation actuated by the control room air supply duct radiation monitor does not close allcontrol room isolation dampers. The unfiltered inleakage rate of 1500 cfm compensates for the dampers that remain open. DEK also increased the assumed control room damper closure time to 20 seconds from the previous value of 10 seconds. The increased damper closure time bounds actual measured closure times.DEK compensated for the higher control room unfiltered inleakage in the radiological accidentanalysis by proposing modifications to TSs limits. The three proposed TS changes reduce the calculated fission product released to the environment, thus allowing for higher control room unfiltered inleakage. In this amendment request, DEK submitted revised radiological analyses of the DBAs. The dose acceptance criteria for the DBAs and the revised licensee-calculated radiological consequence are listed in Table 1. The CREZ unfiltered inleakage parameters and assumptions used by the licensee and acceptable to the NRC staff are listed in Table 2. 3.1 Main Steamline Break (MSLB) AccidentIn the revised radiological analysis for MSLB accident, DEK changed the assumed CREZunfiltered inleakage from 200 to 1000 cfm, which bounds the ASTM E741 test results. This is the only assumption that has changed from the previous radiological accident analysis of MSLB.The licensee assumed that the faulted steam generator (SG) boils dry within 2 minutes. Theentire liquid inventory of the faulted SG is steamed off and all the iodine initially in the SG is released to the outside environment. The primary-to-secondary SG tube leakage rate is assumed to be at the TS limit of 150 gallons per day (gpd) per SG. The 150 gpd leakage for the faulted SG, along with its noble gas and iodine, is assumed released directly to the outsideatmosphere. In the intact SG, the 150 gpd primary-to-secondary leakage mixes with the bulk SG secondary coolant water. Transferred noble gases are released without holdup, and iodine is released to the outside environment at the steaming rate of the intact SG, with credit for partitioning when the SG tubes are covered with water.DEK analyzed the MSLB for two iodine spiking cases. The pre-accident iodine spiking caseassumed that a reactor transient has occurred prior to the MSLB, and has raised the reactor coolant system (RCS) iodine concentration to 60 micro curies per gram (Ci/gm) of doseequivalent (DE) Iodine 131 (I-131). The accident-initiated iodine spiking case assumed that the reactor trip associated with the MSLB creates an increase in the iodine release rate from the fuel to the coolant to a value 500 times greater than the release rate corresponding to a maximum equilibrium RCS concentration of 1.0 Ci/gm DE I-131. The accident-initiated spikeduration is 4 hours0.167 days <br />0.0238 weeks <br />0.00548 months <br />. The secondary coolant activity in both cases is assumed to be the TS limit of 0.1 Ci/gm DE I-131. No fuel damage is projected for the MSLB.The low steamline pressure SI set point will be reached shortly after the onset of an MSLB. The SI signal causes the control room HVAC to switch from normal operation mode to the accident mode of operation. DEK conservatively assumed that the control room HVAC does not fully enter the accident mode of operation until 5 minutes after the event begins. The NRC staff reviewed the licensee's analysis of the MSLB radiological consequences, andfinds that they remain consistent with the guidance provided in RG 1.183. The licensee's calculated radiological consequences at the Exclusion Area Boundary (EAB), Low PopulationZone (LPZ) and in the KPS control room are within the dose criteria specified in 10 CFR 50.67 and GDC-19, and are within the acceptance criteria given in SRP 15.0.1 for the MSLB. The NRC staff finds the results of the licensee's calculations (Table 1), and the major parameters and assumptions used by the licensee (Table 3) acceptable. 3.2 Locked Rotor AccidentIn the revised locked rotor accident analysis, DEK changed the assumption for the fraction offailed fuel rods from 100 percent down to 50 percent, which is less conservative. DEK based the 50 percent assumption on the reload safety analysis limit. DEK also increased the length of time assumed for control room HVAC to enter accident mode of operation from 10 to 45 minutes. DEK also conservatively revised the CREZ unfiltered inleakage to 1500 cfm based on tracer gas test results, and because the control room air supply duct radiation monitor actuates the HVAC accident mode of operation. These are the only assumptions that have changed from the previous radiological accident analysis of the locked rotor accident.The licensee's analysis assumes that a reactor transient has occurred prior to the locked rotoraccident and that the transient has raised the RCS iodine activity concentration to 60 Ci/gmDE I-131, which bounds the proposed TS 3.1.c.2.A limit of 20 Ci/gm DE I-131. The noble gasand alkali metal activity concentration in the primary coolant is based on a fuel defect level of 1.0 percent. The iodine activity concentration in the secondary coolant is assumed to be 0.1 Ci/gm DE I-131, and the alkali metal activity concentration is assumed to be 10 percent of theprimary coolant concentration. Accident-induced activity is assumed to be released to the environment as a result of primary-to-secondary leakage through the SG tubes and steaming from the secondary side, released through either the atmospheric relief valves or safety valves.

An iodine partitioning factor in the SGs of 0.01 is used to account for retention of iodine in the SG as the water turns to steam. The partitioning factor of 0.01 is also applied to the alkali metal activity release. All noble gas activity carried over to the secondary side of the SGs is assumed to be immediately released to the outside atmosphere. At 8 hours0.333 days <br />0.0476 weeks <br />0.011 months <br /> after the accident, the licensee assumed that the residual heat removal (RHR) system has removed all decay heat with no further releases to the environment after that time.

The NRC staff reviewed the licensee's methods, inputs and assumptions used in its revised radiological consequences analysis of the locked rotor accident and finds that they are consistent with the guidance given in RG 1.183. The licensee's calculated radiological consequences at the EAB, LPZ and in the KPS control room are within the dose limits specified in 10 CFR 50.67 and GDC-19, and are within the acceptance criteria given in SRP 15.0.1 for the locked rotor accident. The NRC staff finds the results of the licensee's calculations (Table 1), and the major parameters and assumptions used by the licensee (Table 4) acceptable.3.3 Control Rod Ejection AccidentIn the revised radiological analysis of the control rod ejection accident, DEK maintained thesame assumptions as applied in previous radiological analysis, except the CREZ unfiltered inleakage rate has been conservatively increased to 1000 cfm to account for tracer gas test

results. This DBA postulates the mechanical failure of a control rod drive mechanism pressure housingthat results in the ejection of a rod cluster control assemble and drive shaft. Localized damage to fuel cladding and a limited amount of fuel melting are projected. The radioactivity in the primary coolant is assumed to leak through the SG tubes into the secondary coolant. A portion of this activity is released to the outside atmosphere through the main condenser, atmospheric relief valves or safety valves. Additionally, radioactive primary coolant is discharged to the containment through the opening in the reactor vessel head where the control rod assembly was ejected. The activity in the containment is assumed to be released to the environment as a result of design-basis containment leakage evaluated at the proposed TS limit of 0.5 percent per day for the first 24 hours1 days <br />0.143 weeks <br />0.0329 months <br />. After that, the containment is assumed to leak at half that rate until the end of the 30-day period considered in the analysis. In each case, the containment and secondary coolant release pathways are considered separately with bounding source termrelease for the combined release path ways.DEK assumed that 15 percent of the fuel rods in the core suffer sufficient damage such that alltheir gap activity is released. The licensee assumed that 10 percent of the total core activity of iodine and noble gases and 12 percent of the total core activity for alkali metals are in the fuel gap, consistent with guidance provided in RG 1.183. A small fraction of the fuel in the failed rods is assumed to melt as a result of the rod ejection. The licensee estimated this melting to be limited to 0.375 percent of the core. This estimate was previously accepted in amendment No.166 (ML030210062).The licensee assumed 100 percent of noble gases and alkali metals in the failed fuel gap andmelted fuel are released to either the RCS or the containment, depending on the pathway assumed. For the containment leakage pathway, the licensee assumed that all the iodine from the gap of the failed fuel and 25 percent of the iodine released from melted fuel are released to the containment atmosphere. For the primary-to-secondary leakage release pathway, the licensee assumed that all the iodine from the gap of the failed fuel and 50 percent of the iodine released from melted fuel is released to the RCS. As discussed for the locked rotor accident (Section 3.2 of this SE), the licensee assumed aniodine partitioning factor of 0.01 in the SGs for the primary-to-secondary leakage release pathway. For the containment leakage release pathway, no credit was taken for iodine or particulate removal mechanisms. The low pressurizer pressure SI setpoint is expected to be reached within 60 seconds of theonset of the control rod ejection. The SI signal causes the control room HVAC to switch from the normal operation mode to the accident mode of operation. For this accident, DEK conservatively assumed that the control room HVAC does not fully enter the accident mode of operation until 2.5 minutes after the event begins. The NRC staff reviewed the licensee's methods, inputs and assumptions used in its revisedradiological consequences analysis of the control rod ejection accident and finds that they remain consistent with the conservative guidance given in RG 1.183. The licensee's calculated radiological consequences at the EAB, LPZ and in the KPS control room are within the dose limits specified in 10 CFR 50.67 and GDC-19 and are within the acceptance criteria given in SRP 15.0.1 for the control rod ejection accident. The NRC staff finds the results of the licensee's calculations (Table 1), and the major parameters and assumptions used by the licensee (Table 5) acceptable. To verify the licensee's dose results the NRC staff performed confirmatory calculations for the control rod ejection accident and finds the licensee's results to be reasonable. 3.4Steam Generator Tube Rupture (SGTR)In the revised radiological analysis of the SGTR, DEK changed the assumed pre-accidentiodine spike from a value of 60 Ci/gm DE I-131 to a value of 20 Ci/gm DE I-131 per TS3.1.c.2.A. DEK also increased the CREZ unfiltered inleakage from 200 to 1000 cfm based on the tracer gas test results.The SGTR is analyzed for two iodine spiking cases; a pre-existing iodine spike resulting in elevated primary coolant activity, and an iodine spike initiated by the accident. For the pre-existing iodine spike case, the RCS iodine activity concentration is assumed to be at the proposed TS 3.1.c.2.A limit for a transient, equal to 20 Ci/gm DE I-131. For the accident initiated iodine spike case, the associated reactor trip causes an increase in the iodine release rate from the fuel to the RCS to a value 500 times the rate associated with the TS equilibrium RCS activity concentration of 1.0 Ci/gm DE I-131. The duration of the accidentinitiated iodine spike is limited by the amount of iodine in the fuel gap. Based on having 8 percent of the core inventory of iodine in the fuel gap, the spike would last 4 hours0.167 days <br />0.0238 weeks <br />0.00548 months <br />. RG 1.183 allows an accident initiated spiking factor of 335 for the SGTR, and the NRC staff finds the licensee's assumed factor of 500 is conservative compared to the RG value. All other analysis inputs are consistent with the guidance in RG 1.183.The low pressurizer pressure SI setpoint is expected to be reached at around 2.9 minutes afterthe onset of the SGTR. The SI signal causes the control room HVAC to switch from the normal operation mode to the accident mode of operation. The licensee conservatively assumed that the control room HVAC does not fully enter the accident mode of operation until 5 minutes after the event begins.The NRC staff reviewed the licensee's methods, inputs and assumptions used in its revisedradiological consequences analysis of the SGTR and finds that they are consistent with the conservative guidance given in RG 1.183. The licensee's calculated radiological consequencesat the EAB, LPZ and in the KPS control room, are within the dose limits specified in 10 CFR 50.67 and GDC-19 and are within the acceptance criteria given in SRP 15.0.1 for the SGTR.

The NRC staff finds the results of the licensee's calculations (Table 1), and the major parameters and assumptions used by the licensee (Table 6) acceptable. 3.5Large-Break Loss-of-Coolant Accident (LBLOCA)For the LBLOCA analysis, the current radioactive methyl iodide removal percentage TS limit forthe shield building ventilation system, and the Auxiliary Building Special Ventilation Systemcarbon filters is 95 percent. The licensee reviewed historical data of radiological accidentanalysis (RAA) sensitivity cases, to ensure that although more limiting, the proposedconservative change to 97.5 percent is reasonable and continues to provide adequateoperating margin. The NRC staff finds that the revised limits bound plant charcoal filter testresults and provide sufficient operating margin and are therefore acceptable.For the LBLOCA analysis, the containment leakage rate is reduced to 0.2 weight percent perday of the contained air per 24 hours1 days <br />0.143 weeks <br />0.0329 months <br /> at a peak test pressure of 46 psig from the current analysis value of 0.5 weight percent per day. The licensee reviewed historical data of RAA sensitivity cases, to ensure that although more limiting, the proposed conservative change is reasonable and continues to provide adequate margin to actual measurements of containment leakage rates. The NRC staff finds that the revised containment leak rate limit bounds the plant measured containment leak rate test result, and provides sufficient operating margin and is therefore acceptable.In its revised radiological analysis of LBLOCA, DEK changed the assumed shield building andauxiliary building filter efficiencies, the containment leakage rate, and the CREZ unfiltered inleakage flow rate. The revised assumptions increased shield building and auxiliary building filter efficiencies from 90 percent to 95 percent for removal of both elemental and organiciodine. The containment leak rates are revised from 0.5 to 0.2 weight percent per day for the first 24 hours1 days <br />0.143 weeks <br />0.0329 months <br />, per proposed TS 6.20, and from 0.25 to 0.1 weight percent per day for greater than 24 hours1 days <br />0.143 weeks <br />0.0329 months <br />. The CREZ unfiltered inleakage flow rate is revised from 200 to 800 cfm based on tracer gas test results. These are the only assumptions that have changed from the previous radiological accident analysis of LBLOCA.In the licensee's analysis of the LBLOCA radiological consequences, activity from the damagedcore is released into the containment. Three pathways for release to the environment are considered in the analysis:(1) design-basis containment leakage,(2) leakage from engineering safety feature (ESF) systems outside containment, and (3) emergency core cooling system (ECCS) recirculation back-leakage to the refueling water storage tank (RWST).The calculated radiological consequences of these three release pathways are added togetherto determine the total LBLOCA radiological consequences.3.5.1Containment Leakage Pathways The containment is assumed to leak at the proposed TS 6.20 design-basis leak rate of 0.2percent per day for the first 24 hours1 days <br />0.143 weeks <br />0.0329 months <br /> of the accident, and then to leak at half that rate for the remainder of the 30-day analysis period. The licensee assumed that during the first 10 minutes of the accident, 90 percent of the activity leaking from the containment is discharged directly to the environment. The remaining 10 percent enters the auxiliary building where it is released through filters. After 10 minutes, only 1 percent of the activity leaking from the containment is assumed to go directly to the environment, 10 percent continues to go to the auxiliary building, and 89 percent is assumed to go into the shield building. The air discharged from the shield building is filtered. Additionally, once the shield building is brought to sub-atmospheric pressure at 30 minutes into the accident, iodine and particulate can be removed by recirculation through filters. A shield building participation fraction of 0.5 is assumed. The shield building filter efficiency for elemental and organic iodine is revised to 95 percent,which is bound by the proposed level in TS 3.6.c.3.B. The licensee assumed removal of iodine through sedimentation for particulate and the containment spray for elemental and particulate forms of iodine. The KPS containment spray system is an ESF system and is designed to provide containment cooling and fission product removal in the containment following a LBLOCA. One train of spray was assumed to operate. Switch over to recirculation spray is not credited and all spray removal is terminated when the RWST drains down at 0.91 hours3.792 days <br />0.542 weeks <br />0.125 months <br /> from the start of the accident. In determining the core spray iodine removal rates, the licensee assumed a reduction in assumed spray flow relative to that assumed in the analysis supporting Amendment No. 166 (ML030210062). This reduction is intended to bound potential pump degradation. The NRC staff finds this change acceptable. The licensee assumed a sedimentation coefficient of 0.1 hr-1 for particulate after the core spray system is terminated. The licensee used the models and guidance provided in RG 1.183 and SRP 6.5.2,"Containment Spray as a Fission Product Cleanup System," to determine the removal rates for iodine. 3.5.2Post-LOCA ESF Leakage Pathway During the recirculation phase of long-term core cooling, radioactive water from thecontainment sump is sent to ECCS equipment located outside the containment. These components may leak into the auxiliary building. Although ECCS recirculation does not occur until 0.91 hours3.792 days <br />0.542 weeks <br />0.125 months <br /> after the accident begins, the licensee conservatively assumed leakage occurs immediately upon the onset of the LBLOCA. The licensee conservatively assumed the leakage to the auxiliary building is 12 gallons per hour. The licensee assumed that 10 percent of the activity in the leaked fluid becomes airborne when the sump temperature is above 212 F. Once the sump temperature drops below 212 F at 3 hours0.125 days <br />0.0179 weeks <br />0.00411 months <br /> from the start of the event, theairborne activity fraction is reduced to 1 percent of the activity in the leaked fluid. The assumed time that the sump temperature falls below 212 F is selected to bound the results of thecontainment response analyses performed for the TS changes. The NRC staff finds these assumptions are consistent with RG 1.183 and are acceptable. The licensee also assumed that half of the airborne iodine activity in the auxiliary building is removed by plateout on surfaces. This assumption was previously approved in Amendment No. 166 (ML030210062).3.5.3ECCS Back-Leakage to the RWST RHR back-leakage to the RWST is assumed to be at a rate of 3 gallons per minute (gpm) forthe first 24 hours1 days <br />0.143 weeks <br />0.0329 months <br />, and 1.5 gpm for the remainder of the accident. It is assumed that 1 percent of the iodine becomes airborne, even when the sump temperature is above 212 F since anyincoming water would be cooled by the water remaining in the RWST. 3.5.4Control Room Ventilation System Modeling For the LBLOCA, the low pressurizer pressure SI setpoint will be reached shortly after the startof the event. The SI signal causes the control room HVAC to switch from the normal operation mode to the accident mode of operation. The licensee conservatively assumed that the control room HVAC does not fully enter the accident mode of operation until 2 minutes after the event begins. 3.5.5LBLOCA Conclusion The NRC staff reviewed the licensee's methods, inputs and assumptions used in its revisedradiological consequences analysis of the LBLOCA for TS changes, and finds that they are consistent with the conservative guidance given in RG 1.183. The licensee's calculated radiological consequences at the EAB, LPZ and in the KPS control room are within the dose limits specified in 10 CFR 50.67 and GDC-19, and are within the acceptance criteria given in SRP 15.0.1 for the LBLOCA. The NRC staff finds the results of the licensee's calculations (Table 1), and the major parameters and assumptions used by the licensee (Table 7) acceptable. To verify the licensee's dose results, the NRC staff performed confirmatory radiological consequence analyses of the LBLOCA and finds the licensee's results to be reasonable. 3.6Fuel-Handling Accident (FHA)In the revised radiological analysis, DEK conservatively changed the CREZ unfiltered inleakageflow rate from 200 to 1500 cfm. This is the only assumption that has changed from the previous radiological accident analysis of FHA.The licensee's analysis of the FHA was performed with assumptions selected so that the resultsare bounding for an accident that occurs either in the containment or in the auxiliary building.

Activity released from the damaged assembly is assumed to be released to the environment through either the containment purge system or the spent fuel pool ventilation system, without credit for filtration or isolation of the containment, containment purge system, or spent fuel pool ventilation system. The decay time used, 100 hours4.167 days <br />0.595 weeks <br />0.137 months <br />, is the minimum decay time required by TS before movement of fuel. The licensee assumed that all the fuel rods in the equivalent of one fuel assembly are damaged, and all the gap activity in the rods is released to the pool. A pool iodine effective decontamination factor of 200 is assumed. All fuel gap noble gas activity is assumed to be released from the pool. All activity released from the pool is assumed to be released to the outside environment within 2 hours0.0833 days <br />0.0119 weeks <br />0.00274 months <br />.DEK assumed that the control room HVAC system is in normal operation mode at the onset ofthe FHA. A high-radiation signal for the control room air supply duct is generated as a result of the activity release to the atmosphere, and control room HVAC enters accident mode of operation within 25 minutes.The NRC staff reviewed the licensee's methods, inputs and assumptions, and finds that theyare consistent with the conservative guidance given in RG 1.183. The licensee's calculated radiological consequences at the EAB, LPZ and in the KPS control room, are within the dose limits specified in 10 CFR 50.67 and GDC-19 and are within the acceptance criteria given in SRP 15.0.1 for the FHA. The NRC staff finds the results of the licensee's calculations (Table 1), and the major parameters and assumptions used by the licensee (Table 8) acceptable. To verify the licensee's dose results, the NRC staff performed a confirmatory radiological consequence analysis of the FHA and finds the results to be reasonable.

3.7Waste Gas Decay Tank (GDT) Rupture and Volume Control Tank (VCT) RuptureThe KPS licensing basis includes analyses of the radiological consequences of the waste GDTrupture, and the VCT rupture. The radiological analyses for these two accidents were previously found acceptable by the NRC staff in its SE approving Amendment No. 166 to the KPS license. The only change to the assumptions in the existing analyses is the increase in CREZ unfiltered inleakage. The NRC staff documented in its SE approving Amendment No.

172, issued February 27, 2004 (ADAMS Accession No. ML040430633) that "[b]ecause of the short duration of the radiation release, minimizing the assumed control room unfiltered inleakage maximizes the calculated control room dose. This is due to less dilution of the radioactivity in the control room. The NRC staff finds these assumptions to be acceptable based on plant operation and the operation of the radiation monitoring and control room HVAC systems." Therefore, the assumed increase in the unfiltered inleakage for these two accidentsis bounded by the previous analyses. Consequently, these two accidents were not re-analyzed. 3.8Technical Evaluation ConclusionThe NRC staff reviewed the assumptions, inputs, and methods used by DEK to reevaluate theradiological consequences of DBAs. The NRC staff finds that DEK used analysis methods andassumptions consistent with the conservative regulatory requirements and guidance identified in Section 2. DEK's analysis demonstrated that the radiological consequences of DBAs wouldremain within applicable regulatory limits. The NRC staff has performed confirmatory calculations on selected accidents and finds, with reasonable assurance, that the revised radiological accident analyses, assuming higher CREZ unfiltered inleakage and amended TS, (TS 3.1.c.2.A, TS3.6.c.3.B, and TS 6.20) complies with the regulatory requirements and is therefore acceptable.

4.0STATE CONSULTATION

In accordance with the Commission's regulations, the Wisconsin State official was notified ofthe proposed issuance of the amendment. The State official had no comments.

5.0ENVIRONMENTAL CONSIDERATION

This amendment changes a requirement with respect to installation or use of a facilitycomponent located within the restricted area as defined in 10 CFR Part 20. The staff has determined that the amendment involves no significant increase in the amounts, and no significant change in the types, of any effluent that may be released offsite and that there is no significant increase in individual or cumulative occupational radiation exposure. TheCommission has previously issued a proposed finding that this amendment involves no significant hazards consideration and there has been no public comment on such finding (71 FR 13172). Accordingly, this amendment meets the eligibility criteria for categorical exclusion set forth in 10 CFR 51.22(c)(9). Pursuant to 10 CFR 51.22(b) no environmental impact statement or environmental assessment need be prepared in connection with the issuance of this amendment.

6.0CONCLUSION

The staff has concluded, based on the considerations discussed above, that: (1) there isreasonable 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 this amendment will not be inimical to the common defense and security or to the health and safety of the public. Principal Contributor: D. Chung, B. Lee Date: March 8, 2007 Table 1Design-Basis Accident Licensee Calculated Radiological ConsequencesTEDE (rem)Design-Basis AccidentEABLPZControl RoomMSLB, Pre-existing iodine spikeDose acceptance criteria 0.03 250.01 250.70 5MSLB, Accident-initiated iodine spikeDose acceptance criteria0.062.50.022.52.60 5Locked Rotor AccidentDose acceptance criteria0.402.50.062.53.90 5Control Rod Ejection AccidentDose acceptance criteria0.406.30.096.34.54 5SGTR, Pre-existing spikingDose acceptance criteria0.50 250.10 251.90 5SGTR, Accident-initiated spikingDose acceptance criteria0.802.50.202.52.80 5LBLOCA, totalDose acceptance criteria0.52 250.09 254.95 5 FHADose acceptance criteria0.906.30.156.34.0 5 3 The CREZ UFI is increased to at least 1500 cfm for events that model control roomisolation on a control room radiation monitor, R-23, high control room duct activity monitor actuation (i.e. locked rotor and fuel handling accident).Table 2Revised Control Room ParametersREVISED ASSUMPTIONPREVIOUSASSUMPTIONNormal ventilation flow ratesUnfiltered Makeup Flow Rate1620 - 2750 cfm2250 - 2750 cfmEmergency Ventilation Flow RatesUnfiltered Inleakage Following SI Unfiltered Inleakage Following R-23 3 800 cfm 1500 cfm200 cfm 200 cfmControl Room Isolation Damper Closure Time20 seconds10 seconds Table 3Assumptions Used in Radiological Consequence AnalysisMain Steamline BreakReactor coolant activityPre-existing iodine spike case Accident-initiated iodine spike case Accident-initiated iodine appearance rate spiking factor Duration of accident-initiated iodine spike60.0 Ci/gm DE I-1311.0 Ci/gm DE I-131500 times equilibrium rate 4 hours0.167 days <br />0.0238 weeks <br />0.00548 months <br />Secondary coolant activity 0.1 Ci/gm DE I-131Primary coolant mass 1.19E+08 gmSecondary coolant initial liquid massFaulted steam generator (SG)

Intact SG161,000 lbm84,000 lbmSteam release from faulted SG161,000 lbm Time to release faulted SG initial mass2 minutes Steam release from intact SG0 - 2 hours0.0833 days <br />0.0119 weeks <br />0.00274 months <br /> 2 - 8 hours0.333 days <br />0.0476 weeks <br />0.011 months <br /> 8 - 24 hours1 days <br />0.143 weeks <br />0.0329 months <br />222,000 lbm424,000 lbm 614,000 lbmTime to cool RCS and stop faulted SG release72 hours3 days <br />0.429 weeks <br />0.0986 months <br /> Steam partition coefficientFaulted steam generator Intact steam generator 10.01Steam generator tube leak rate150 gallons per day per SG Time until begin control room emergency HVAC5 minutes Normal ventilation flow ratesUnfiltered makeup2500 cfm (+10%)Emergency ventilation system flow ratesFiltered makeup Filtered recirculation Unfiltered inleakage0 cfm2500 cfm (+10%)

1000 cfmControl room filter efficienciesElemental Organic Particulate 90%90%

99%Atmospheric dispersion factors Table 9 Table 4Assumptions Used in Radiological Consequence AnalysisLocked Rotor AccidentReactor coolant activity60.0 Ci/gm DE I-131Secondary coolant activity0.1 Ci/gm DE I-131Primary coolant mass1.19E+08 gm Secondary coolant mass 0 to 30 minutesSecondary coolant mass > 30 minutes7.89E+07 gm1.06E+08 gmSecondary coolant mass > 30 minutes1.06E+8 gm Fuel rods in core failing, No fuel melting50%

Peaking Factor Applied to Calculate Activity in Failed Fuel Rods1.7Fission product gap fractionsI-131 Kr-85 Other iodines and noble gases Alkali metals0.080.10 0.05 0.12Iodine chemical form in release97% elemental, 3% organic Primary-to-secondary SG tube leak rate150 gallons per day per SG Steam release from secondary0 - 2 hours0.0833 days <br />0.0119 weeks <br />0.00274 months <br /> 2 - 8 hours0.333 days <br />0.0476 weeks <br />0.011 months <br />210,000 lbm455,000 lbmSteam partition coefficient0.01 Time to cool RCS and stop steam release8 hours0.333 days <br />0.0476 weeks <br />0.011 months <br /> Time until begin control room emergency HVAC45 minutes Normal ventilation flow ratesUnfiltered makeup2500 cfm (+10%)Emergency ventilation system flow ratesFiltered makeup Filtered recirculation Unfiltered inleakage0 cfm2500 cfm (+10%)

1500 cfmControl room filter efficienciesElemental Organic Particulate 90%90%

99%Atmospheric dispersion factorsTable 9 Table 5Assumptions Used in Radiological Consequence AnalysisControl Rod Ejection AccidentReactor power1782.6 MWtReactor coolant activity60.0 Ci/gm DE I-131Secondary coolant activity0.1 Ci/gm DE I-131Primary coolant mass1.19E+08 gm Secondary coolant mass7.89E+07 gm Radial peaking factor1.7 Fuel rods in core failing15%

Fission product gap fractionsIodines and noble gases Alkali metals0.100.12Fuel rods in core melting0.375%

Fission product activity released from melted fuelNoble gases and alkali metals Iodines 100%25% for containment leakage path 50% for SG steaming pathSG steaming release pathwayPrimary-to-secondary SG tube leak rate150 gallons per day per SGSteam release from secondary0 - 200 seconds 200 - 1800 seconds

> 1800 seconds800 lbm/sec100 lbm/sec

0 lbm/secSteam partition coefficient0.01 Iodine chemical form in steam release97% elemental, 3% organic Containment leakage pathway Containment net free volume1.32E+06 ft 3Shield building volume3.74E+05 ft 3Shield building participation fraction0.5 Containment leak rate0 - 24 hours1 days <br />0.143 weeks <br />0.0329 months <br />

> 24 hours1 days <br />0.143 weeks <br />0.0329 months <br />0.5 weight %/day0.25 weight %/day Table 5 (continued)Assumptions Used in Radiological Consequence AnalysisControl Rod Ejection AccidentContainment leak path fractions0 -10 minutesThrough shield building Through auxiliary building Direct to environment> 10 minutesThrough shield building Through auxiliary building Direct to environment0.00.1 0.90.890.1 0.01Shield building air flow0 - 10 minutesShield building to environment Shield building recirculation10 - 30 minutesShield building to environment Shield building recirculation> 30 minutesShield building to environment Shield building recirculationNot applicableNot applicable6000 cfm (+10%)0.0 cfm3100 cfm2300cfmShield building and auxiliary building filter efficienciesElemental Organic Particulate 90%90%

99%Time until begin control room emergency HVAC2.5 minutes Normal ventilation flow ratesUnfiltered makeup2500 cfm (+10%)Emergency ventilation system flow ratesFiltered makeup Filtered recirculation Unfiltered inleakage0 cfm2500 cfm (+10%)

1000 cfmControl room filter efficienciesElemental Organic Particulate 90%90%

99%Atmospheric dispersion factorsTable 9 Table 6Assumptions Used in Radiological Consequence AnalysisSteam Generator Tube RuptureReactor coolant activityPre-existing iodine spike case Accident-initiated iodine spike case Accident-initiated iodine appearance ratespiking factorDuration of accident-initiated iodine spike 20.0 Ci/gm DE I-1311.0 Ci/gm DE I-131500 times equilibrium rate4 hours0.167 days <br />0.0238 weeks <br />0.00548 months <br />Secondary coolant activity0.1 Ci/gm DE I-131Primary coolant mass1.19E+08 gm Secondary coolant initial liquid mass84,000 lbm/SG Intact steam generator tube leak rate150 gallons per day Pre-trip releases (< 173.3 seconds)Tube rupture break flow Percentage of break flow that flashes to steam Steam release to condenser16,900 lbm19.93%

1077.8 lbm/sec for each SGPost-trip releases (> 173.3 seconds)Tube rupture break flow Percentage of break flow that flashes to steam Steam release from ruptured SG, 0 - 2 hours0.0833 days <br />0.0119 weeks <br />0.00274 months <br /> Steam release from intact SG, 0 - 2 hours0.0833 days <br />0.0119 weeks <br />0.00274 months <br /> Steam release from intact SG, 2 - 8 hours0.333 days <br />0.0476 weeks <br />0.011 months <br /> Steam release from intact SG, 8 - 24 hours1 days <br />0.143 weeks <br />0.0329 months <br />138,000 lbm14.76%

86,400 lbm 233,400 lbm 488,800 lbm 662,800 lbmSteam partition coefficientRuptured steam generator, break flow Intact steam generator 10.01Time until begin control room emergency HVAC5 minutes Normal ventilation flow ratesUnfiltered makeup2500 cfm (+10%)Emergency ventilation system flow ratesFiltered makeup Filtered recirculation Unfiltered inleakage0 cfm2500 cfm (+10%)

1000 cfmControl room filter efficienciesElemental Organic Particulate 90%90%

99%Atmospheric dispersion factorsTable 9 Table 7Assumptions Used in Radiological Consequence AnalysisLarge-Break Loss-of-Coolant AccidentReactor powerSource term Containment volume1782.6 MWtBased on RG 1.183 1.32E+06 ft 3Containment leak rate0 - 24 hours1 days <br />0.143 weeks <br />0.0329 months <br />

> 24 hours1 days <br />0.143 weeks <br />0.0329 months <br />0.2 weight % per day0.1 weight % per dayShield building volumeShield building participation fraction Containment leak modeling3.74E+05 ft 30.5See Table 4Spray operationTime to initiate sprays Termination of sprays Recirculation spray Removal coefficientsElemental iodine Particulate Sedimentation (after spray termination)0.0 hours0 days <br />0 weeks <br />0 months <br />0.91 hours3.792 days <br />0.542 weeks <br />0.125 months <br /> Not credited 20 hr-14.5 hr-10.1 hr-1ECCS leakageContainment sump volume ECCS leak rate, 0 - 30 days Airborne percent iodine to auxiliary building0 - 3 hours0.125 days <br />0.0179 weeks <br />0.00411 months <br />

> 3 hours0.125 days <br />0.0179 weeks <br />0.00411 months <br />ECCS leak rate to RWST0 - 24 hours1 days <br />0.143 weeks <br />0.0329 months <br />

> 24 hours1 days <br />0.143 weeks <br />0.0329 months <br />Shield and auxiliary building filter efficienciesElemental Organic Particulate315,000 gal12 gal/hr 10%1%3 gpm1.5 gpm 95%95%

99%Time until begin control room emergency HVAC2 minutes Normal ventilation flow ratesUnfiltered makeup 2500 cfm (+10%)Emergency ventilation system flow ratesFiltered makeup Filtered recirculation Unfiltered inleakage0 cfm2500 cfm (+10%)

800 cfm Table 7 (continued)Assumptions Used in Radiological Consequence AnalysisLarge-Break Loss-of-Coolant AccidentControl room filter efficienciesElemental Organic Particulate 90%90%

99%Atmospheric dispersion factorsTable 9 Table 8Assumptions Used in Radiological Consequence AnalysisFuel-Handling AccidentReactor power1782.6 MWtRadial peaking factor1.7 Fission product decay period100 hours4.167 days <br />0.595 weeks <br />0.137 months <br /> Number of fuel assemblies damaged1 Fuel pool water depth23 ft Pool iodine effective decontamination factor200 Fraction of fuel compliant with RG 1.183, footnote 11Fuel gap fission product inventory (RG 1.183, footnote 11 compliant)I-131 Kr-85 Other iodines and noble gases0.50 8%10%

5%Fraction of fuel not compliant with RG 1.183, footnote 11Fuel gap fission product inventory (RG 1.183, footnote 11 non-compliant)I-131 Kr-85 Other iodines and noble gases0.50 12%30%

10%Duration of release2 hours0.0833 days <br />0.0119 weeks <br />0.00274 months <br /> Time until begin control room emergency HVAC1 minute Normal ventilation flow rateUnfiltered makeup2750 cfmEmergency ventilation system flow ratesFiltered makeup Filtered recirculation Unfiltered inleakage0 cfm2500 cfm (+10%)

1500 cfmControl room filter efficienciesElemental Organic Particulate 90%90%

99%Atmospheric dispersion factorsTable 9 Table 9Atmospheric Dispersion FactorsExclusion Area BoundaryTime (hr) X/Q (sec/m 3)0 - 22.232E-04Low Population ZoneTime (hr) X/Q (sec/m 3)0 - 23.977E-05 2 - 244.100E-06 24 - 482.427E-06 48 - 7204.473E-07Control RoomTime (hr) X/Q (sec/m 3)0 - 82.93E-03 8 - 241.73E-03 24 - 486.74E-04 48 - 7201.93E-04