ML063400359

Issuance of Amendments 210 and 202 Full-scope Implementation of an Alternative Accident Source Term (TAC Nos. MC5495 and MC5496)

ML063400359
Person / Time
Site:	San Onofre
Issue date:	12/29/2006
From:	Kalyanam N K NRC/NRR/ADRO/DORL/LPLIV
To:	Rosenblum R M Southern California Edison Co
Kalynanam N, NRR/DORL/LP4, 415-1480
Shared Package
ML063400318	List: ML063400359
References
TAC MC5495, TAC MC5496
Download: ML063400359 (51)

Text

December 29, 2006Mr. Richard M. RosenblumSenior Vice President and Chief Nuclear Officer Southern California Edison Company San Onofre Nuclear Generating Station P.O. Box 128 San Clemente, CA 92674-0128

SUBJECT:

SAN ONOFRE NUCLEAR GENERATING STATION, UNITS 2 AND 3 -ISSUANCE OF AMENDMENTS RE: FULL-SCOPE IMPLEMENTATION OF AN ALTERNATIVE SOURCE TERM (TAC NOS. MC5495 AND MC5496)

Dear Mr. Rosenblum:

The Commission has issued the enclosed Amendment No. 210 to Facility Operating LicenseNo. NPF-10 and Amendment No. 202 to Facility Operating License No. NPF-15 for San Onofre Nuclear Generating Station, Units 2 and 3 (SONGS 2 and 3), respectively. The amendments consist of changes to the Updated Final Safety Analysis Report in response to your applicationdated December 27, 2004, as supplemented by letters dated October 27, 2005, March 10, and October 6, 2006.The amendments revise the SONGS 2 and 3 accident source term (AST) used in the design-basis radiological consequence analyses. These license amendments are inaccordance with the requirements of Section 50.67 of Title 10 of the Code of FederalRegulations, which addresses the use of an AST at operating reactors, and relevant guidanceof Regulatory Guide 1.183, "Alternative Radiological Source Terms for Evaluating Design Basis Accidents at Nuclear Power Reactors." These license amendments represent full-scope implementation of the AST described in Regulatory Guide 1.183.

R. Rosenblum-2-A copy of our related Safety Evaluation is also enclosed. The Notice of Issuance will beincluded in the Commission's next biweekly Federal Register notice. Sincerely,/RA/N. Kalyanam, Project ManagerPlant Licensing Branch IV Division of Operating Reactor Licensing Office of Nuclear Reactor RegulationDocket Nos. 50-361 and 50-362

Enclosures:

1. Amendment No. 210 to NPF-102. Amendment No. 202 to NPF-15

3. Safety Evaluationcc w/encls:See next page

ML063400359OFFICENRR/LPL4/PMNRR/LPL4/LANRR/DCI/CSGB*NRR/DSS/SPWB*NAMENKalyanamLFeizollahiAHiserJNakoski DATE12/28/0612/12/064/20/0612/1/06 OFFICENRR/DRA/AADB*OGC-NLONRR/LPL4/BC NAMEMKotzalasBPooleDTerao:JND for DT DATE11/29/0612/28/0612/29/06 March 2006San Onofre Nuclear Generating Station Units 2 and 3 cc:Mr. Daniel P. Breig Southern California Edison Company San Onofre Nuclear Generating Station P.O. Box 128 San Clemente, CA 92674-0128Mr. Douglas K. Porter, EsquireSouthern California Edison Company 2244 Walnut Grove Avenue Rosemead, CA 91770Mr. David Spath, ChiefDivision of Drinking Water and Environmental Management P.O. Box 942732 Sacramento, CA 94234-7320Chairman, Board of SupervisorsCounty of San Diego 1600 Pacific Highway, Room 335 San Diego, CA 92101Mark L. ParsonsDeputy City Attorney City of Riverside 3900 Main Street Riverside, CA 92522Mr. Gary L. Nolff Assistant Director - Resources City of Riverside 3900 Main Street Riverside, CA 92522Regional Administrator, Region IVU.S. Nuclear Regulatory Commission 611 Ryan Plaza Drive, Suite 400 Arlington, TX 76011-8064Mr. Michael R. OlsonSan Diego Gas & Electric Company 8315 Century Park Ct. CP21G San Diego, CA 92123-1548Director, Radiologic Health BranchState Department of Health Services P.O. Box 997414, MS 7610 Sacramento, CA 95899-7414Resident Inspector/San Onofre NPS c/o U.S. Nuclear Regulatory Commission Post Office Box 4329 San Clemente, CA 92674Mayor City of San Clemente 100 Avenida Presidio San Clemente, CA 92672Mr. James T. Reilly Southern California Edison Company San Onofre Nuclear Generating Station P.O. Box 128 San Clemente, CA 92674-0128Mr. James D. Boyd, CommissionerCalifornia Energy Commission 1516 Ninth Street (MS 31)

Sacramento, CA 95814Mr. Ray Waldo, Vice PresidentSouthern California Edison Company San Onofre Nuclear Generating Station P.O. Box 128 San Clemente, CA 92764-0128Mr. Brian KatzSouthern California Edison Company San Onofre Nuclear Generating Station P.O. Box 128 San Clemente, CA 92764-0128Mr. Steve HsuDepartment of Health Services Radiologic Health Branch MS 7610, P.O. Box 997414 Sacramento, CA 95899 March 2006San Onofre Nuclear Generating Station-2-Units 2 and 3 cc:Mr. A. Edward Scherer Southern California Edison Company San Onofre Nuclear Generating Station P.O. Box 128 San Clemente, CA 92674-0128 SOUTHERN CALIFORNIA EDISON COMPANYSAN DIEGO GAS AND ELECTRIC COMPANYTHE CITY OF RIVERSIDE, CALIFORNIATHE CITY OF ANAHEIM, CALIFORNIADOCKET NO.50-361SAN ONOFRE NUCLEAR GENERATING STATION, UNIT 2AMENDMENT TO FACILITY OPERATING LICENSEAmendment No. 210License No. NPF-101.The Nuclear Regulatory Commission (the Commission) has found that: A.The application for amendment by Southern California Edison Company, et al.(SCE or the licensee), dated December 27, 2004, as supplemented by letters dated October 27, 2005, March 10, and October 6, 2006, complies with the standards and requirements of the Atomic Energy Act of 1954, as amended (the Act), and the Commission's regulations set forth in 10 CFR Chapter I;B.The facility will operate in conformity with the application, the provisions of theAct, and the rules and regulations of the Commission;C.There is reasonable assurance (i) that the activities authorized by thisamendment can 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, by Amendment No. 210, the license is amended to approve changes to theUpdated Final Safety Analysis Report (UFSAR), as set forth in the application for amendment by SCE dated December 27, 2004, as supplemented by letters dated October 27, 2005, March 10, and October 6, 2006. SCE shall update the UFSAR to reflect the revised licensing basis authorized by this amendment in accordance with 10 CFR 50.71(e).3.This license amendment is effective as of the date of its issuance and shall beimplemented within 180 days of issuance. Implementation of the amendment is the incorporation of the UFSAR changes to the description of the facility as described in the licensee's application dated December 27, 2004, as supplemented by letters dated October 27, 2005, March 10, and October 6, 2006, and evaluated in the staff's Safety Evaluation dated December 29, 2006.FOR THE NUCLEAR REGULATORY COMMISSION/RA/David Terao, ChiefPlant Licensing Branch IV Division of Operating Reactor Licensing Office of Nuclear Reactor RegulationDate of Issuance: December 29, 2006 SOUTHERN CALIFORNIA EDISON COMPANYSAN DIEGO GAS AND ELECTRIC COMPANYTHE CITY OF RIVERSIDE, CALIFORNIATHE CITY OF ANAHEIM, CALIFORNIADOCKET NO.50-362SAN ONOFRE NUCLEAR GENERATING STATION, UNIT 3AMENDMENT TO FACILITY OPERATING LICENSEAmendment No. 202License No. NPF-151.The Nuclear Regulatory Commission (the Commission) has found that:A.The application for amendment by Southern California Edison Company, et al.(SCE or the licensee), dated December 27, 2004, as supplemented by letters dated October 27, 2005, March 10, and October 6, 2006, complies with the standards and requirements of the Atomic Energy Act of 1954, as amended (the Act), and the Commission's regulations set forth in 10 CFR Chapter I;B.The facility will operate in conformity with the application, the provisions of theAct, and the rules and regulations of the Commission;C.There is reasonable assurance (i) that the activities authorized by thisamendment can 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, by Amendment No. 202, the license is amended to approve changes to theUpdated Final Safety Analysis Report (UFSAR), as set forth in the application for amendment by SCE dated December 27, 2004, as supplemented by letters dated October 27, 2005, March 10, and October 6, 2006. SCE shall update the UFSAR to reflect the revised licensing basis authorized by this amendment in accordance with 10 CFR 50.71(e).3.This license amendment is effective as of the date of its issuance and shall beimplemented within 180 days of issuance. Implementation of the amendment is the incorporation of the UFSAR changes to the description of the facility as described in the licensee's application dated December 27, 2004, as supplemented by letters dated October 27, 2005, March 10, and October 6, 2006, and evaluated in the staff's Safety Evaluation dated December 29, 2006.FOR THE NUCLEAR REGULATORY COMMISSION/RA/David Terao, ChiefPlant Licensing Branch IV Division of Operating Reactor Licensing Office of Nuclear Reactor RegulationDate of Issuance: December 29, 2006 SAFETY EVALUATION BY THE OFFICE OF NUCLEAR REACTOR REGULATIONRELATED TO AMENDMENT NO. 210 TO FACILITY OPERATING LICENSE NO. NPF-10AND AMENDMENT NO. 202 TO FACILITY OPERATING LICENSE NO. NPF-15FOR IMPLEMENTATION OF ALTERNATIVE SOURCE TERMSOUTHERN CALIFORNIA EDISON COMPANYSAN DIEGO GAS AND ELECTRIC COMPANYTHE CITY OF RIVERSIDE, CALIFORNIATHE CITY OF ANAHEIM, CALIFORNIASAN ONOFRE NUCLEAR GENERATING STATION, UNITS 2 AND 3DOCKET NOS. 50-361 AND 50-36

21.0INTRODUCTION

By application dated December 27, 2004 (Agencywide Documents Access and ManagementSystem (ADAMS) Accession No. ML043650403), as supplemented by letters dated October 27, 2005 (ADAMS Accession No. ML053040458), March 10 (ADAMS Accession No. ML060750661), and October 6, 2006 (ADAMS Accession No. ML062850525), Southern California Edison Company (SCE or the licensee) requested license amendments to change the Updated Final Safety Analysis Report (UFSAR) for San Onofre Nuclear Generating Station,Units 2 and 3 (SONGS 2 and 3). The proposed amendments would revise the SONGS 2 and 3 design basis to replace the existing accident radiological source term (Technical Information Document (TID)-14844, "Calculation of Distance Factors for Power and Test Reactor Sites.")

with a full implementation of the alternative source term (AST) pursuant to Title 10 of the Codeof Federal Regulations (10 CFR) Section 50.67, "Accident Source Term." The supplementsdated October 27, 2005, March 10, and October 6, 2006, provided additional information that clarified the application, did not expand the scope of the application as originally noticed, and did not change the Nuclear Regulatory Commission (NRC) staff's original proposed no significant hazards consideration determination as published in the Federal Register onFebruary 1, 2005 (70 FR 5248). The amendments revise the SONGS 2 and 3 AST used in the design-basis radiologicalconsequence analyses. These license amendments are in accordance with the requirements of 10 CFR 50.67, which addresses the use of an AST at operating reactors, and relevant guidance of Regulatory Guide (RG) 1.183, "Alternative Radiological Source Terms for Evaluating Design Basis Accidents at Nuclear Power Reactors." These license amendments represent full-scope implementation of the AST described in RG 1.183. In the 180-dayresponse to Generic Letter 2003-1, "Control Room Habitability," the licensee stated in its letter dated August 5, 2003 (ADAMS Accession No. ML032230360), that the current value of assumed unfiltered air inleakage into the control room envelope (CRE) in the design-basis radiological analyses is 10 cubic feet per minute (cfm) assumed for ingress and egress, with no other source of unfiltered inleakage into the CRE (i.e., 0 cfm of unfiltered inleakage), for a total analysis input of 10 cfm unfiltered inleakage to the CRE. The licensee subsequently stated in its letter dated September 17, 2004 (ADAMS Accession No. ML042650353), that the CRE unfiltered air inleakage testing performed at SONGS 2 and 3 in May 2004 provided results that exceeded 0 cfm assumed in the current design-basis radiological analyses and that the licensee performed an operability assessment in accordance with the guidance provided in "Operability Determinations and Degraded/Non-Confirming Condition Resolution," in an NRC letter to Nuclear Energy Institute (NEI) dated January 30, 2004 (ADAMS Accession No. ML040160868). Full-scope implementation of an AST requires, at a minimum, reanalysis of the loss-of-coolantaccident (LOCA). In its LAR, the licensee reanalyzed the LOCA, the fuel handling accident (FHA) inside containment, FHA in the fuel handling building (FHB), and the Main Steamline Break (MSLB) accident outside containment. As part of implementing the AST, the licensee also requested to depart from the nucleate boiling (DNB) statistical convolution methodology for estimating fuel failure for non-LOCA events (i.e., the MSLB accident). Based on its operability assessment, the licensee determined that the maximum amount ofCRE unfiltered inleakage that would be consistent with an OPERABLE CRE is greater than the maximum unfiltered inleakage that has been demonstrated by CRE unfiltered inleakage testing.

The licensee further determined that the CRE is nonconforming, but operable. In the time since the CRE unfiltered inleakage testing in May 2004, subsequent operability assessment was completed, and SONGS 2 and 3 have continued to operate in a nonconforming, but operable status. Accordingly, the licensee submitted this license amendment request (LAR) on December 27, 2004 to restore SONGS 2 and 3 to full qualification for meeting the control room (CR) dose acceptance criteria specified in 10 CFR 50.67 and General Design Criterion (GDC) 19 of Appendix A to 10 CFR Part 50 and to incorporate an AST into the SONGS 2 and 3 design and licensing bases. There are no physical changes to plant equipment or operation of the plant requested in this LAR and the licensee requested no changes to the SONGS 2 and 3 technical specifications (TSs) in this LAR. Thus, the license amendment fulfills the commitmentdescribed in SCE's letter dated September 17, 2004, and establishes use of an AST methodology that documents the acceptability of an assumed increase in SONGS 2 and 3 CRE unfiltered inleakage rate to a value of 1,000 cfm (including ingress and egress related inleakage). As described in the September 17, 2004, letter, this is necessary to restore the CRE to full qualification.

2.0REGULATORY EVALUATION

In this LAR, the licensee requested a full-scope implementation of the AST pursuant to10 CFR 50.67, as described in RG 1.183. 10 CFR 50.67 provides a mechanism for licensed power reactors to replace the traditional source term used in their design-basis accident (DBA) radiological consequence analyses. The NRC staff evaluated the radiological consequences of the proposed DBAs against thedose criteria specified in 10 CFR 50.67, "Accident Source Term"; these criteria are 25 rem, total effective dose equivalent (TEDE) at the exclusion area boundary (EAB) for any 2-hour period following the onset of the postulated fission product release, 25 rem TEDE at the outer boundary of the low population zone (LPZ), and 5 rem TEDE in the CR. The TEDE dose includes both noble gas and radioiodine exposure.This safety evaluation (SE) addresses the impact of the proposed changes on previouslyanalyzed DBA radiological consequences and the acceptability of the revised analysis results.

The regulatory requirements for which the NRC staff based its acceptance are the accident dose criteria in 10 CFR 50.67, as supplemented in Regulatory Position 4.4 of RG 1.183, GDC 19, and Standard Review Plan (SRP) Section 15.0.1. Except where the licensee has proposed a suitable alternative, the NRC staff used 10 CFR 50.67, as well as the regulatory guidance in the following documents below in its radiological consequence dose analyses:*RG 1.23, "Onsite Meteorological Programs"

• RG 1.145, "Atmospheric Dispersion Models for Potential Accident Consequence Assessments at Nuclear Power Plants"*RG 1.183, "Alternative Radiological Source Terms for Evaluating Design BasisAccidents at Nuclear Power Reactors"*RG 1.194, "Atmospheric Relative Concentrations for Control Room RadiologicalHabitability Assessments at Nuclear Power Plants"RG 1.197, "Demonstrating Control Room Envelope Integrity at Nuclear PowerReactors"NUREG-0800, "Standard Review Plan," Section 6.4, "Control Room HabitabilitySystems"*SRP Section 2.3.4, "Short-Term Diffusion Estimates for Accidental Atmospheric Releases"*SRP Section 6.4, "Control Room Habitability Systems" (with regard to controlroom meteorology)*SRP Section 15.0.1, "Radiological Consequence Analyses Using AlternativeSource Term"The NRC staff also considered relevant information in the SONGS 2 and 3 UFSAR and TSs.

3.0TECHNICAL EVALUATION

3.1Atmospheric Dispersion EstimatesThe licensee calculated new CR atmospheric dispersion factors (/Q values) for use in reanalyzing the bounding UFSAR Chapter 15 DBAs. The resulting SONGS 2 and 3 CR /Qvalues represent a change from those currently presented in Table 15B-4 of the SONGS 2 and 3 UFSAR. The licensee used the existing exclusion area boundary (EAB) and LPZ /Qvalues listed in Table 15B-4 of the SONGS 2 and 3 UFSAR to perform the offsite dose assessments for the bounding UFSAR Chapter 15 DBAs.3.1.1Meteorological DataIn its December 27, 2004, submittal, the licensee presented new CR /Q values that weregenerated using site meteorological data collected during the period 1993-2002. The licensee previously used these data to generate FHA CR /Q values associated with SONGS 2 LicenseAmendment No. 193 and SONGS 3 License Amendment No. 184. The licensee previously provided a copy of these data in its letter to the NRC staff dated October 6, 2004 (ADAMS Accession No. ML042870468).Wind speed and wind direction were measured at 10 and 40 meters above ground level andatmospheric stability classification was based on temperature difference measurements between these two levels. The combined data recovery of wind speed, wind direction, and stability (delta-temperature) exceeded the RG 1.23 goal of 90 percent during this 10-year period, although the upper level wind data recovery for 1994 and 1995 averaged around 80 percent. Section 2.3.3.1 of the SONGS 2 and 3 UFSAR states that the onsite meteorological measurements system is consistent with the recommendations of RG 1.23.The NRC staff previously performed a basic review of a subset of the 1993-2002 sitemeteorological data as discussed in the SE associated with SONGS 2 and 3 License Amendments Nos. 193 and 184. The NRC staff performed a more comprehensive review of these data for this LAR using the methodology described in NUREG-0917, "Nuclear Regulatory Commission Staff Computer Programs for Use with Meteorological Data." Further review was performed using computer spreadsheets. As expected, the NRC staff's examination of the data revealed generally stable and neutral atmospheric conditions at night and unstable and neutral conditions during the day. Wind speed, wind direction, and stability class frequency distributions were reasonably similar from year to year and generally consistent with the 1973-1976 and 1979-1983 data presented in the SONGS 2 and 3 UFSAR, with the exception that the average lower and upper level wind speeds in 1999 were approximately 1.8 times higher than the lower and upper level wind speeds averaged over the remaining 9-year period (1993-1998 and 2000-2002). In its request for additional information (RAI) letter dated October 7, 2005, the NRC staff askedthe licensee to explain the abnormal 1999 wind speed data and their impact on the resulting ARCON96 dispersion analyses. In its submittal dated October 27, 2005, the licensee stated that the 1999 wind speed data were in error and revised its accident analyses (including the ARCON96 CR dispersion analyses) to reflect the correct meteorological data. The licensee submitted an electronic copy of the corrected meteorological data in its RAI response letter dated March 10, 2006. The NRC staff has reviewed the corrected 1999 wind speed data andhas concluded that the revised 1999 wind speed frequency distributions more closely resemble the wind speed frequency distributions for the remaining 9-year period (1993-1998 and 2000-2002).3.1.2CR Atmospheric Dispersion FactorsSONGS 2 and 3 share a combined CR, with one normal mode air intake (located near thenorthwest corner of the CRE) and two emergency mode air intakes (located near the northwest and southwest corners of the CRE, respectively). During the CR normal mode of operation, unfiltered outside air is introduced into the CRthrough the normal mode air intake. During the CR emergency mode of operation, the CR is isolated and filtered outside air is introduced into the CR through the emergency mode air intakes in order to pressurize the CR. The emergency mode of operation can be actuated either manually or automatically following a CR isolation signal (CRIS). The CRIS may be generated automatically by a safety injection actuation signal or by the detection of high radioactivity concentrations in the CR outside air flow. The emergency mode of operation is facilitated by two 100 percent redundant subsystems,each with its own air intake. The licensee assumed the failure of one emergency train and modeled single emergency train operation throughout the duration of each event. The licensee stated that this single failure results in the largest CR doses. The licensee also assumed that unfiltered inleakage occurs at the beginning of each accident scenario and continues throughout the duration of each event.The licensee calculated new /Q values to evaluate the impact of SONGS 2 and 3 main plantvent, containment shell, containment equipment hatch, main steam safety valve (MSSV),

atmospheric dump valve (ADV), steamline break outside containment (SLB-OC), auxiliary feedwater (AFW) turbine steam discharge, refueling water storage tank (RWST) vent, and FHB releases on the SONGS 2 and 3 CR. Each of these nine release scenarios was evaluated for both SONGS 2 and 3 and for each of the three CR air intakes (the one normal mode air intake and the two emergency mode air intakes). For each potential release scenario, the licensee used the maximum /Q values calculated for each of these three air intakes to model air intake(including filtered air intake and unfiltered inleakage) into the CR. The licensee provided a plant layout showing the location of potential radiological release points with respect to the CR outside air intakes in its RAI response letter dated March 10, 2006.In its RAI letter dated October 7, 2005, the NRC staff asked the licensee to confirm that thereare no potential unfiltered inleakage pathways during both normal and emergency modes that could result in /Q values that are higher than the three CR air intake /Q values. In its RAIresponse dated March 10, 2006, the licensee stated that only the west side of the CRE is exposed to radioactive plumes released to the outside environment and all three CR air intakes are located on the west side of the CRE. The licensee stated that adjacent areas and structures to the north, south, and east of the CRE, and the adjacent areas and structures above and below the CRE, do not contain activity release points. These adjacent areas and locations can become contaminated only with air introduced via intake of infiltration of radioactive material contained in the radioactive plumes released to the outside environment which are then recirculated and diluted throughout these regions. Consequently, the licenseestated that there should be no potential unfiltered inleakage pathways during both normal and emergency modes that could result in /Q values that are higher than the three CR air intake/Q values.The licensee used guidance provided in RG 1.194 to generate the new CR atmosphericdispersion factors. The licensee calculated these new CR /Q values using the ARCON96computer code (NUREG/CR-6331, Revision 1, "Atmospheric Relative Concentrations in Building Wakes"). RG 1.194 states that ARCON96 is an acceptable methodology for assessing

CR /Q values for use in DBA radiological analyses.The licensee executed ARCON96 using the 1993-2002 hourly data from the sitemeteorological tower. Wind speed and wind direction data from the tower's 10-meter and 40-meter levels were provided as input and stability class was calculated using the temperature difference between the 40-meter and 10-meter levels. The resulting /Q values are presentedin Table 2 of this SE. Details on the licensee's assessments of CR post-accident atmospheric dispersion conditions for each release scenario are provided below.1.Main Plant Vent Releases: Each of the SONGS 2 and 3 units has a main plant ventwhich is located on top of each unit's containment structure. The licensee modeled the main plant vent releases as point sources using the ARCON96 ground-level release option. The plant vent release heights are both at 53.6 meters above plant grade (APG) as compared to the normal air intake height of 1.7 meters APG and the emergency air intake heights of 4.0 meters APG. The resulting ARCON96 analysis showed that the SONGS 2 main plant vent release to the SONGS 2 emergency air intake was the bounding source-receptor combination that resulted in the highest (i.e., most conservative) /Q values.2.Containment Shell Releases: The licensee modeled the SONGS 2 and 3 containmentshell (surface) releases as area (diffuse) sources using the ARCON96 ground-level release option. The area source dimensions were the maximum vertical and horizontal dimensions of the above-grade containment shell cross-sectional area perpendicular to the line of sight from the containment shell center to the CR air intakes. The initial diffusion coefficients (plume dimensions) were determined by dividing the area sourcedimensions by a factor of six in accordance with RG 1.194. The release heights (24.5 meters APG) were set equal to the mid-height of the containment shells above grade. The resulting ARCON96 analysis showed that the SONGS 2 containment shell release to the SONGS 2 emergency air intake was the bounding source-receptor combination that resulted in conservative /Q values.3.Equipment Hatch Releases: The licensee modeled the SONGS 2 and 3 containmentequipment hatch releases as area (diffuse) sources using the ARCON96 ground-level release option. The area source dimensions were based on the 5.8-meter diameter of the hatch openings. The initial diffusion coefficients (plume dimensions) were determined by dividing the area source dimensions by a factor of six in accordance with RG 1.194. The equipment hatch release heights (2.4 meters APG) were set equal to the mid-height of the hatch openings above grade. Since the SONGS 2 and 3 containment equipment hatches are on the opposite side of their respective containment (1)In its RAI letter dated October 7, 2005, the NRC staff asked the licensee whether a stuck-openMSSV concurrent with other DBAs is excluded from the licensing basis for the MSSVs. In its letter response dated March 10, 2006, the licensee responded that the SONGS 2 and 3 licensing basis, as reflected in UFSAR Section 10.3, does not require that a stuck-open MSSV be considered concurrently with other DBAs.structures from the CR air intakes, atmospheric dispersion factors were calculatedassuming flow both around and over (through) the containment buildings and the higher values were used. The resulting ARCON96 analysis showed that (1) the SONGS 2 containment shell release over (through) the containment building to the SONGS 2 emergency air intake was the bounding source-receptor combination for the 8 to 24-hour time period and (2) the SONGS 2 containment shell release around the containment building to the SONGS 2 emergency air intake was the bounding source-receptor combination for the remaining time periods. In its analyses, the licensee used a composite of these two conservative release pathways to derive conservative equipment hatch release /Q values.4.MSSV Releases: SONGS 2 and 3 are both Combustion Engineering 2-loop pressurizedwater reactors which have two main steamlines for each unit. Each steamline has a set of nine safety valves centered around a main steamline isolation valve (MSIV). These valves may open either automatically, when the pressure in the main steamline reaches the valve setpoint, or manually by use of a valve lever. The licensee used the center of each MSIV as the MSSV release location for each steamline in determining the horizontal distances and directions to each of the three CR air intakes. Consequently the licensee modeled four MSSV release point locations (i.e., SONGS 2 MSSVs centered at MSIVs 8204 and 8205 and SONGS 3 MSSVs centered at MSIVs 8204 and 8205) to each of the three CR air intakes.The licensee modeled the MSSV releases as point sources using the ARCON96ground-level release option. The MSSV release heights are 13.2 meters APG as compared to the normal air intake height of 1.7 meters APG and the emergency air intake heights of 4.0 meters APG. The resulting ARCON96 analysis showed that the SONGS 2 MSIV 8204 MSSV release to the SONGS 2 emergency air intake was the bounding source-receptor combination.RG 1.194 allows the ground level /Q values calculated with ARCON96 (on the physicalheight of the release point) to be reduced by a factor of 5 if (1) the release point is uncapped and vertically oriented and (2) the time-dependent vertical velocity exceeds the 95 th percentile wind speed. MSSV releases satisfy both criteria, namely, (1) MSSVrelease points are uncapped and vertically oriented and (2) the calculated minimum MSSV stack exit velocity is 72 meters per second, which is considerably higher than the 10-meter 95 th percentile wind speed value of 5.5 meters per second. Consequently, thelicensee reduced the resulting ARCON96 MSSV /Q values by a factor of 5.(1)5.ADV Releases: Each of the two steamlines for each of the two SONGS units has apower-operated atmospheric ADV. Consequently, the licensee modeled four ADV release points (i.e., SONGS 2 ADVs 606 and 607 and SONGS 3 ADVs 606 and 607) to each of the three CR air intakes. (2)Table 4.4-11 of Enclosure 2 of the licensee's December 27, 2004, submittal (which wasmodified in the licensee's October 27, 2005, submittal) also presented /Q values for ADV releases withplume rise credit. None of the DBA events addressed in the licensee's current AST application use these ADV release /Q values with plume rise credit. The licensee presented these ADV release /Q valueswith plume rise credit for use in potential future applications. However, in order to credit plume rise in an ADV release dose analysis, the time period for which the ADV stack vertical flow exit velocity exceeds five times the 95 th percentile upper level wind speed would need to be determined before the plume riseadjustment factor could be applied.The licensee modeled the ADV releases as point sources using the ARCON96 ground-level release option. The locations of the ADV stacks were used to determine the distance and direction to the CR air intakes. The ADV release heights are 25.6 meters APG as compared to the normal air intake height of 1.7 meters APG and the emergency air intake heights of 4.0 meters APG. No credit was taken for a plume rise associated with the ADV release pathways. The resulting ARCON96 analysis showed that the SONGS 2 607 ADV release to the SONGS 2 emergency air intake was the bounding source-receptor combination(2) that resulted in conservative /Q values.6.Steamline Break Outside Containment (SLB-OC) Releases: The locations of thepostulated break for each of the two steamlines for each of the two SONGS units are assumed to occur downstream of the MSIVs and the releases are assumed to occur through blowout panels mounted on the roof of the MSIV/MFIV enclosure structures located directly above each main steamline. Consequently, the licensee modeled four SLB-OC release points (i.e., SONGS 2 north and south MSIV/MFIV enclosure roof blowout panels and SONGS 3 north and south MSIV/MFIV enclosure roof blowout panels) to each of the three CR air intakes. The licensee modeled the SLB-OC releases as area (diffuse) sources using theARCON96 ground-level release option. The area source dimensions were based on the width of the area formed by the three blowout panels mounted on the roof of the MSIV/MFIV enclosure structure perpendicular to the line of sight from the MSIVs to the respective CR intake. The initial diffusion coefficients (plume dimensions) were determined by dividing the source dimensions by a factor of six in accordance with RG 1.194. The SLB-OC release heights (10.2 meters AGL) were set equal to the height of the blowout panels and no credit was taken for a plume rise. The resulting ARCON96 analysis showed that the SONGS 2 south MSIV/MFIV enclosure structure roof blowout panel release to the SONGS 2 emergency air intake was the bounding source-receptor combination that resulted in conservative /Q values.7.Auxiliary Feedwater (AFW) Turbine Exhaust Releases: There is one AFW turbine foreach of the two SONGS units. Consequently, the licensee modeled two AFW turbine release points (i.e., SONGS 2 turbine stack and SONGS 3 turbine stack) to each of the three CR air intakes.The licensee modeled the AFW turbine exhaust releases as point sources using theARCON96 vent level release option which is not in accordance with RG 1.194.

However, the licensee assumed both the vertical exit velocity and stack flow were zero.

With both exit velocity and stack flow set to zero, the ARCON96 vent level release option generates the same results as the ground-level release option recommended byRG 1.194. The AFW turbine stack release heights are 8.8 meters APG, as compared to the normal air intake height of 1.7 meters APG and the emergency air intake heights of

4.0 meters

APG. The resulting ARCON96 analysis showed that the SONGS 2 AFW turbine exhaust release to the SONGS 2 emergency air intake was the bounding source-receptor combination that resulted in conservative /Q values.8.RWST Releases: There are two RWSTs for each of the two SONGS units. Consequently, the licensee modeled four RWST release points (i.e., SONGS 2 RWSTs T005 and T006 and SONGS 3 RWSTs T005 and T006) to each of the three CR air intakes. The RWST releases are assumed to occur through the roof vent on each RWST. The licensee modeled the RWST releases as point sources using the ARCON96 ventrelease option which is not in accordance with RG 1.194. However, the licensee assumed both the vertical exit velocity and stack flow were zero. With both exit velocity and stack flow set to zero, the ARCON96 vent level release option generates the same results as the ground-level release option recommended by RG 1.194. The RWST vent release heights are approximately 12.7 meters above plant grade as compared to the normal air intake height of 1.7 meters APG and the emergency air intake heights of

4.0 meters

APG. The resulting ARCON96 analysis showed that SONGS 2 RWST T006 vent release to the SONGS 2 emergency air intake was the bounding source-receptor combination for the 0-2 hour time period whereas the SONGS 2 RWST T005 vent release to the SONGS 2 emergency air intake was the bounding source-receptor combination for the remaining time periods. In its analyses, the licensee used a composite of these two conservative release pathways to derive conservative RWST

/Q values.9.FHB Releases: There is one FHB for each of the two SONGS units. Consequently, thelicensee modeled two FHB release points (i.e., SONGS 2 FHB and SONGS 3 FHB) to each of the three CR air intakes. The FHB releases are assumed to occur through the closest and largest cask hatch in each FHB.The licensee modeled the FHB releases as point sources using the ARCON96 ventlevel release option which is not in accordance with RG 1.194. However, the licensee assumed both the vertical exit velocity and stack flow were zero. With both exit velocity and stack flow set to zero, the ARCON96 vent level release option generates the same results as the ground-level release option recommended by RG 1.194. The FHB cask hatch release heights are 10.2 meters APG as compared to the normal air intake height of 1.7 meters APG and the emergency air intake heights of 4.0 meters APG. The resulting ARCON96 analysis showed that the SONGS 2 FHB release to the SONGS 2 emergency air intake was the bounding source-receptor combination that resulted in conservative /Q values.The NRC staff evaluated the applicability of the ARCON96 model and concluded that there areno unusual siting, building arrangements, release characterization, source-receptor configuration, meteorological regimes, or terrain conditions that preclude use of the ARCON96 model for the SONGS 2 and 3 site. The NRC staff qualitatively reviewed the inputs to the ARCON96 calculations and found them generally consistent with site configuration drawingsand site practice. The NRC staff made an independent evaluation of the resulting bounding atmospheric dispersion estimates by running the ARCON96 computer code and found that the licensee's results were similar to or more conservative than the NRC staff's results. On the basis of this review, the NRC staff concludes that the /Q values for SONGS 2 and 3 DBAreleases to the SONGS 2 and 3 CR as presented in Table 2 are acceptable for use in the DBA CR dose assessments performed in support of this license amendment request.3.1.3EAB and LPZ Atmospheric Dispersion FactorsThe licensee evaluated offsite doses using the EAB and LPZ five percentile /Q valuespresented in the SONGS 2 and 3 UFSAR Appendix 15B Table 15B-4. These /Q values, whichare presented in Table 3 of this SE, were generally based on the 5 percent overall site /Qmethodology (excluding the effects of plume meander) described in RG 1.145. Onsite meteorological data from the period 1973 through 1976 were used to derive these /Q values. Further details on the calculation of the licensee's EAB and LPZ /Q values can be found inSONGS 2 and 3 UFSAR Section 2.3.4.1.The NRC staff reviewed the licensee's use of existing SONGS 2 and 3 EAB and LPZ /Qvalues and has found them to be appropriate for the applications in which they are being used.

On the basis of the review discussed above, the NRC staff concludes that the EAB and LPZ

/Q values presented in Table 3 are acceptable for use in the design-basis offsite doseassessments performed in support of this license amendment request.3.2Radiological Consequences of DBAs To support the proposed implementation of an AST, the licensee analyzed the radiological doseconsequences of the following three DBAs:*Large break LOCA

• MSLB outside containment

• FHA in containment and in FHBIn accordance with the guidance provided in RG 1.183, the licensee reanalyzed the LOCA(required as a minimum for full implementation of the AST) and in addition, the FHA inside containment, the FHA in the FHB, and the MSLB accident outside containment. The remainder of the DBA dose analyses applicable to SONGS 2 and 3 were not reanalyzed by the licensee because no other changes were proposed to the SONGS 2 and 3 licensing or design basis thatwould require analysis of the radiological consequences of those remaining accidents. The licensee's submittal and its supplements reported the results of the radiological consequence analyses for the above DBAs to show compliance with 10 CFR 50.67, or fractions thereof, as defined in SRP Section 15.0.1, for doses offsite and in the CR. Therefore, this LAR is a full implementation of the AST. The NRC staff performed confirmatory calculations to evaluate the licensee's dose analysismodeling, assumptions and results. The NRC staff used the licensee's analysis assumptions and inputs in its analyses, which were performed using version 3.03 of the computer codedescribed in NUREG/CR-6604, "RADTRAD: A Simplified Model for RAD ionuclide Transport and Removal A nd Dose Estimation," and its supplements. 3.2.1Loss-of-Coolant Accident (LOCA)The current licensing basis radiological consequence analysis for the postulated LOCA isprovided in SONGS 2 and 3 UFSAR Section 15.6.3.3, "Loss-of-Coolant Accident," and is based on the traditional accident source term described in TID-14844. To demonstrate that the engineered safety features (ESFs) designed to mitigate the radiological consequences at SONGS 2 and 3 will remain adequate after implementing the AST as requested in this LAR, the licensee reanalyzed the offsite and CR radiological consequences of the postulated LOCA. The licensee submitted the results of its DBA calculations for offsite and CR doses andprovided the major assumptions and parameters used in its dose calculations. As documented in its submittal, as supplemented, the licensee has determined that after implementation of the AST, the existing ESF systems at SONGS 2 and 3 will continue to provide reasonable assurance that the radiological consequences of the postulated LOCA at the EAB, in the LPZ, and in the CR will meet the acceptable radiation dose criteria specified in 10 CFR 50.67(b)(2).

As part of the implementation of the AST, the TEDE dose reference values of 10 CFR 50.67(b)(2) replace the previous whole-body and thyroid dose guidelines of 10 CFR 100.11 and also supplements the dose requirements in GDC 19. SCE's analyses assumed that the inventory of fission products in the reactor core and availablefor release into the containment atmosphere is based on the maximum power level of 3,507 MWt, which is 1.02 times the current licensed thermal power level of 3,438 MWt in order to account for the emergency core cooling system (ECCS) evaluation uncertainty. The licensee developed the core inventory of fission products using the SAS2H and ORIGEN-S modules of the SCALE code package developed by the NRC. As discussed in RG 1.183, the NRC staff finds the use of isotope generation and depletion computer codes such as ORIGEN acceptable for developing the core inventory of fission products.The NRC staff has reviewed the licensee's analyses for the following four potential fissionproduct release pathways:(1)primary containment leakage (2)leakage from emergency core cooling systems (ECCSs) outside containment.

(3)RWST release (4)post-accident sampling system leakage3.2.1.1 Containment Leakage The current SONGS 2 and 3 design-basis containment leak rate specified in the SONGS 2and 3 TS and in the UFSAR is 0.1 percent of the containment-free volume per day (percent per day) at the containment design pressure of 60 pounds per square inch gauge (psig). For the radiological consequence analysis, this rate is reduced to 0.05 percent per day after 24 hour2.777778e-4 days <br />0.00667 hours <br />3.968254e-5 weeks <br />9.132e-6 months <br />sfollowing a LOCA for the remaining duration of the accident (30 days), consistent with the guidance provided in RG 1.183. The licensee has not proposed to change the design basis containment leak rate.3.2.1.1.1 Radioactivity Removal Inside the ContainmentThe fission products in the containment atmosphere following the postulated LOCA atSONGS 2 and 3 are mitigated by (1) natural deposition of fission products in aerosol form, and (2) removal by the containment spray system (CSS). SCE's analysis assumed removal of fission products in aerosol form by natural deposition in the containment following the postulated LOCA using Powers simplified natural deposition model in the RADTRAD dose consequences computer code described in NUREG/CR-6604 and its supplements. The Powers simplified natural deposition model is described in NUREG/CR-6189, "A Simplified Model of Aerosol Removal by Natural Processes in Reactor Containments," The licensee used the 10 percentile confidence interval (90 percent probability) removal values implemented in the RADTRAD code. The Powers natural deposition model was derived by correlation of results of Monte Carlo uncertainty analyses of detailed models of aerosol behavior in the containment under accident conditions. The NRC staff finds that the use of this model in NRC computer code, RADTRAD, is acceptable, as discussed in RG 1.183. Aerosol removal rates by natural deposition developed and used by the licensee are shown in Table 4.5-3 of the licensee's March 10, 2006, submittal.The licensee's analysis also assumed removal of elemental iodine by wall deposition using themethodology provided in SRP Section 6.5.2. Inputs to this methodology include a mass transfer coefficient, the wetted surface area inside containment, and the containment building net-free volume. The licensee used the mass transfer coefficient value of 4.9 meters per hour recommended by SRP Section 6.5.2. The NRC staff finds that the use of this methodology according to the guidance in SRP Section 6.5.2 is acceptable. The elemental iodine depositionremoval rate value calculated and used by the licensee is 4.26 per hour.The CSS at SONGS 2 and 3 is an ESF system. When used in conjunction with twocontainment emergency cooling units (ECUs), each rated at 31,000 cfm, and two dome air circulator units (DACUs), each rated at 37,000 cfm, the CSS is designed to ensure that containment pressure does not exceed the design-basis value of 60 psig and also to remove fission products in the containment atmosphere following the postulated LOCA. To meet the single failure criterion, only one of the two ECUs and one of the two DACUs are assumed to be operational for mixing of air in the containment. The licensee assumes that the ECU and DACU start operation 1 minute after the start of the LOCA. The licensee has determined that 99 percent of the contaminated air in the containment unsprayed region will be replaced with air from the sprayed region within 28 minutes, which equates to approximately four change-outs of the containment unsprayed region prior to the end of the activity releases from the core at conclusion of the early in-vessel phase at 1.8 hours9.259259e-5 days <br />0.00222 hours <br />1.322751e-5 weeks <br />3.044e-6 months <br />. RG 1.183 assumes that two turnovers of the unsprayed region per hour provide adequate mixing. Therefore, the NRC staff concludes that the licensee has shown that the ECUs and DACUs provide adequate mixing between sprayed and unsprayed regions of the containment atmosphere in accordance with the guidance provided in RG 1.183. As shown in the UFSAR, the CSS consists of two redundant and independent trains. Eachtrain consists of a pump, spray headers, and associated valves, piping, and instrumentation.

Each train receives power from a separate emergency diesel generator and separate actuation signals, and they are physically separated from each other. The CSS is automatically initiated by a containment spray actuation signal that is initiated by the combination of any two high-high containment pressure signals and a safety injection actuation signal. The CSS may also beinitiated manually in the CR. The CSS has two phases of operation, an injection phase and a recirculation phase. During the injection phase, the CSS draws spray water from the RWST until it reaches a pre-set nominal low level. Following the injection phase, the CSS enters recirculation phase where spray water is drawn from the containment sump and recirculated through the CSS. The licensee stated that the sprayed volume of the containment is 83.5 percent of the total free volume of the containment. The CSS is independently capable of delivering minimum flow rate of 1,606 gallons per minute(gpm) of borated water from the RWST during the injection phase and 1,991 gpm of sump water from the containment sump during recirculation phase into the sprayed region of the containment free volume. The licensee stated that it conservatively assumed a flow of 1,600 gpm throughout the CSS operation for determining the fission product removal coefficients by spray. The spray pumps are automatically started whenever two out of four high-high containment pressure signals occur or a manual signal is given. The licensee assumed that one out of two spray pump starts taking suction initially from the RWST and initiates building spray through the spray headers until the water in the RWST reaches a pre-set low level at 20 minutes after the postulated LOCA. This first 20 minutes of containment spray, taking suction from the RWST, is called the injection phase.The licensee stated that, after the RWST reaches a preset low level at 20 minutes, the spraypump suction is transferred manually to the containment sump and the spray water from the containment sump is recirculated. The recirculation phase starts at 20 minutes and continuesfor the duration of the accident. Radioactive iodine is released to the containment atmosphere in three different forms:elemental, particulate, and organic. For iodine in elemental and particulate forms, the licensee considered two distinct mechanisms by which radioactive iodine could be removed: containment sprays and natural deposition in the containment. There is no effective mechanism for removing organic iodine from the containment atmosphere. Removal of iodine in elemental and particulate forms from the containment atmosphere by the CSS is controlled by two types of parameters: those controlling the rates of removal, called removal coefficients or lambdas (), and those determining the maximum amount that can be removed, calleddecontamination factors (DF). The evaluation of the pH control in the containment sump to retain radioiodine in the sump water is addressed in Section 3.3 of this SE.The removal rate of iodine by spray is a function of the volumetric flow of the spray solution,which is reflected in the corresponding spray removal coefficient . For SONGS 2 and 3, thereare two different volumetric flow rates: one during the injection phase (1,991 gpm) and one during the recirculation phase (1,606 gpm). Therefore, there are different removal coefficients applied for the injection phase and the recirculation phase. In addition, there are separate sets of removal coefficients for elemental iodine and particulate iodine. SCE calculated the elemental iodine spray removal coefficients in accordance with themethodology described in Section 6.5.2 of the SRP, which is acceptable to the NRC staff. The licensee calculated elemental iodine removal coefficients using a computer code that incorporates the model in SRP Section 6.5.2 and its basis document NUREG/CR-0009, "Technological Bases for Models of Spray Washout of Airborne Contaminants in ContainmentVessels." The licensee calculated elemental iodine spray removal rates for each spray ring in each of the two CSS spray headers. The calculations address the fact that each spray ring is at a different height than the others, with its own unique spray flux due to different coverage areas and number of spray nozzles per spray ring. The elemental iodine spray removal rates for each ring in a spray header were summed together, and the lowest header value at each time interval was used in the dose analysis. The licensee's calculations accounted for the changing containment sump water temperature and pH, volumes of the gaseous and liquid phases, and the initial iodine inventory in the containment sump water. The resulting elemental iodine spray removal coefficients and decontamination factors vary over time. For the duration of the CSS injection phase, SCE calculated an elemental iodine spray removalcoefficient of 1.02 per hour, with a decontamination factor of 110. For the CSS recirculation phase, the calculated removal coefficients decrease over time, and vary from 20 per hour to 3.78 per hour. Accordingly, the decontamination factors for the CSS recirculation phase alsodecrease over time, varying in value from 170 to 25. In accordance with SRP Section 6.5.2,Section III.4.c.(1), the elemental iodine spray removal coefficient is limited to a maximum value of 20 per hour. Therefore, although the licensee's calculation results exceeded that value early in the recirculation phase, the licensee used the limit value of 20 per hour in its dose calculations for those time periods. Because the licensee's calculated elemental iodine removalcoefficients were based on the calculated time-dependent airborne aerosol mass, the reduction in removal rate by a factor of 10, when a DF of 50 is achieved, is not required, which is in accordance with RG 1.183, Appendix A, position 3.3. Table 4.5-5 of the licensee's March 10, 2006, submittal lists the licensee calculated elemental iodine spray removal rates per time period. The licensee calculated the reduction in containment particulate iodine and aerosols by theCSS using the spray removal model for aerosols that was developed by Powers, et al., in NUREG/CR-5966, "A Simplified Model of Aerosol Removal by Containment Sprays," which is referred to as the Powers spray model in this SE. Appendix A to RG 1.183 identifies the Powers spray model as acceptable to the NRC staff. The licensee used plant-specific input to the spray aerosol removal model with regard to the CSS spray water flux and the fall height of the spray droplets. The plant-specific spray water flux and fall height values are within the applicability range of values for which the Powers spray model is valid. For conservatism, to meet the single failure criterion, the licensee's analysis assumed that only one of the two CSS headers is in operation at a flow rate of 1,600 gpm, which is rounded down from the minimum flow rate during the injection phase for one train. SCE used the Powers spray model 10th percentile correlation, which minimizes the aerosol removal and is appropriately conservative. The licensee's calculated CSS spray aerosol removal coefficient values decreaseover time, and vary from 5.15 per hour to 0.5 per hour. Table 4.5-4 of the licensee's March 10, 2006, submittal lists the licensee calculated aerosol spray removal rates per time period. SCE combined the natural deposition and containment spray removal rates discussed aboveinto an overall effective removal rate per time period for elemental iodine and aerosols for the unsprayed and sprayed regions of the containment. For example, for elemental iodine there isno separate spray removal rate and natural deposition rate for any one time period in the sprayed region of containment. The CSS is assumed to operate for 4 hours4.62963e-5 days <br />0.00111 hours <br />6.613757e-6 weeks <br />1.522e-6 months <br /> after the onset of the LOCA. After 4 hours4.62963e-5 days <br />0.00111 hours <br />6.613757e-6 weeks <br />1.522e-6 months <br />, only natural deposition is modeled in the sprayed region of containment. Tables 4.5-6 and 4.5-7 of the licensee's March 10, 2006, submittal list the licensee calculated effective iodine and aerosol removal rates per time period. Use of models for the various mechanisms for iodine removal, when more than one is usedsimultaneously for the same iodine species in a dose analysis, should consider the effect of one model on the others. Because each model used by the licensee does not account for removal through the other model, the use of both the referenced natural deposition models and spray removal models in the same (sprayed) region of containment for the same time period is recognized as potentially nonconservative. Although both natural deposition and spray removal are acting on the overall in-containment aerosol and elemental iodine source term, the total effect from both removal mechanisms is not the same as would be found by simply adding the removal coefficients for each model for a given time period together. SCE addressed this issue in its March 10, 2006, response to RAI #9. However, the modeled removal of particulate and aerosol iodine by natural deposition is not significant from a radiological consequence analysisperspective. The NRC staff did not find it necessary to have the licensee recalculate the doses using a modified natural deposition model to account for containment spray (or a modified containment spray model to account for natural deposition) because the adjustment to the overall LOCA doses would be negligible.The NRC staff has reviewed the information provided by the licensee and compared the valuesused to the guidance in RG 1.183, Appendix A, and has determined that the licensee's assumptions for the containment leakage source term and transport are consistent with the guidance in RG 1.183. The NRC staff also performed an independent calculation of the dose consequences of this LOCA pathway using the licensee's assumptions for input to the RADTRAD computer code. The NRC staff's calculation confirmed the licensee's dose results. 3.2.1.2 Post-LOCA Leakage from ESF Outside Containment As described by the licensee, during the initial phases of the postulated LOCA, both safetyinjection system (SIS) and CSS draw borated water from the RWST. As early as 20.2 minutes into the accident, the recirculation mode of SIS and CSS starts for a two-train recirculation mode. For the recirculation mode, the spray pump suction is transferred manually to the containment sump and the spray water from the containment sump is recirculated. However, the licensee conservatively assumed the ESF leakage begins at 20 minutes after the start of the postulated LOCA event, representing the earliest possible recirculation start time with the ESF operating at maximum design capacity. The ESF leakage is assumed to continue for the duration of the 30-day LOCA event.This recirculation flow causes contaminated sump water to be circulated through piping andcomponents outside of the containment where a small amount of system leakage could provide a path for the release of radionuclides to the environment. Consistent with the guidance provided in RG 1.183, the licensee conservatively assumed that all of the radioiodines released from the reactor coolant system (RCS) are instantaneously moved to the containment sump water. The licensee stated that the maximum expected leakage rate from all ESF components in therecirculation systems is 5,950 cubic centimeters per hour (cc/hr). SONGS 2 and 3 TS Section 5.5.2.8 requires establishment of a program which provides controls to minimize leakage from those portions of systems outside containment that could contain highly radioactive fluids during a serious transient or accident to a level as low as practicable. It further requires the program to include integrated leak test requirements for each system at refueling cycle intervals or less. In its radiological consequence analysis, the licensee doubled the maximum expected leakage rate of 5,950 cc/hr to 11,900 cc/hr consistent with the guidance provided in RG 1.183. The licensee's analysis assumed that 10 percent of the iodine in the ESF leaked fluid becomes airborne in accordance with guidance in Appendix A to RG 1.183. SCE determined that there are three pathways for radioactivity release from ESF systemsoutside containment during a postulated LOCA. These release pathways are (1) leakage from ECCSs (2) leakage to the RWST during ESF recirculation, and (3) leakage from the post-accident sampling system (PASS). 3.2.1.2.1 Post-LOCA ECCS LeakageAs described by the licensee, the ECCS automatically delivers cooling water to the reactor corein the event of a LOCA. The ESF recirculation system circulates containment sump liquid foruse by the CSS and the ECCS, i.e., high-pressure safety injection (HPSI) and low-pressure safety injection (LPSI). The containment sump liquid is circulated outside of the containment tothe ESF pumps. In accordance with RG 1.183, Appendix A, Section 5, ESF systems that circulate sump water outside of the primary containment are assumed to leak during their intended operation. This release includes leakage through valve packing glands, pump shaft seals, flanged connections, and other similar components. Consistent with RG 1.183 guidance, the post-LOCA ESF leakage is assumed to start at theearliest time that recirculation flow begins in the ESF recirculation system. The maximum assumed leakage rate from the components in the HPSI, LPSI and CSS during recirculation is a total 5,950 cc/hr. The licensee assumed two times the maximum assumed leakage rate (i.e.,

11,900 cc/hr) in the dose analysis, in accordance with guidance in RG 1.183. The licensee assumed that 10 percent of the iodine in the ESF leakage flashed to vapor and in available for release to the outside environment. This assumption of 10 percent is consistent with the guidance in RG 1.183, which states that if the water temperature is less than 212 degrees Fahrenheit (oF), then 10 percent of the iodine in the leakage is assumed to become airborneunless a smaller amount is justified based on actual sump pH history and ventilation rates. The licensee's containment pressure and temperature response analysis for the LOCA event shows that the temperature of the containment sump liquid has been reduced to below 212 oF whenthe ESF recirculation mode of operation begins at 20 minutes.In accordance with RG 1.183 guidance, with the exception of iodine, all radioactive materials inthe recirculating fluid are retained in the liquid phase. The licensee further assumed that 100 percent noble gases formed by the decay of the isotopes in the ESF recirculated liquid will become airborne and available for release to the outside environment. This release of noble gases goes beyond the guidance of RG 1.183 and is an additional conservatism. The licensee assumed that iodine species in the airborne release from the ESF leakage were 97 percent elemental iodine and 3 percent organic iodine, in accordance with RG 1.183. The activity released from the ESF recirculation leakage is exhausted to the environment via the main plantvent.The licensee calculated the doses for the EAB and LPZ for this release pathway and added theresults to the EAB and LPZ doses from the other three pathways to give the total offsite radiological consequences of the LOCA. The licensee calculated the doses for the CR for thisrelease pathway and added the results to the doses from the other three pathways and the direct shine dose to give the total radiological consequences of the LOCA in the CR.The NRC staff has reviewed the information provided by the licensee and compared the valuesused to the guidance in RG 1.183, Appendix A, and has determined that the licensee's assumptions for the post-LOCA ECCS leakage source term and transport are consistent with, or more conservative than, the guidance in RG 1.183. The NRC staff also performed an independent calculation of the dose consequences of this LOCA pathway using the licensee'sassumptions for input to the RADTRAD computer code. The NRC staff's calculation confirmed the licensee's dose results. 3.2.1.2.2 Post-LOCA RWST LeakageAs described by the licensee, after 20 minutes post-LOCA, contaminated fluid is circulatedoutside of the containment via the SIS and CSS pumps during the recirculation mode of operation. The recirculated water may leak past the closed valves that isolate the RWST from the ESF systems. The licensee evaluated two scenarios, dependent on whether the LOCA occurs with or without diesel generator failure. The licensee determined that the scenario that does not assume diesel generator failure leads to more severe dose consequences both offsite and in the CR. Without a diesel generator failure, two ESF leakage pathways to the RWST are likely. The first pathway is flow into the air space of the RWST through the ESF pump minimum flow (mini-flow) isolation valve leakage. The second pathway is flow into the RWST, by backleakage, through the RWST discharge check valve. ESF leakage to the RWST for potential release paths with three or more normally closed isolation valves in series was assumed to be negligible by the licensee, which the NRC staff finds to be a reasonable assumption because of three normally closed valves in series. The licensee used the guidance in RG 1.183 to perform the dose analysis of this leakagepathway. The licensee assumed that the ESF leakage to the RWST starts at the earliest time that recirculation starts, and continues for the remainder of the 30-day duration of the accident. Consistent with RG 1.183 guidance, the licensee multiplied by two the maximum allowable leak rate through the mini-flow isolation valves and the RWST discharge check valve to account for assumptions for flow into the air space of the RWST of 3 gpm and flow into the water in the RWST of 10 gpm, respectively. For the 3 gpm that enters the air space of the RWST, the licensee assumed that 10 percent ofthe iodine in the leakage flashes to vapor, in accordance with RG 1.183 guidance. The licensee addressed this in the dose analysis by modeling 10 percent of the leakage entering the RWST air space, and the remaining 90 percent entering the RWST water space. For the ESF leakage that has entered the RWST water space, including the 10 gpm from thedischarge valve leakage and 90 percent of the 3 gpm from the mini-flow leakage, the licensee applied an iodine partition coefficient to determine the amount of the iodine in the leakage thatenters the air space in the RWST and becomes available for release to the outside environment. The iodine partition coefficient is defined as the iodine concentration in the liquid divided by the iodine concentration in the gas. The licensee used a partition coefficient of 200, based on the NUREG/CR-0009 discussion of partition coefficients for borated solutions. The licensee accounted for the pH of the solution as well as air and water space mixing in the RWST due to turbulence from the mini-flow leakage that drops down into the water. No credit was taken by the licensee for mixing due to the RWST discharge valve flow or for mixing by thermal gradients in the water. In accordance with RG 1.183 guidance, with the exception of iodine, all radioactive materials inthe recirculating fluid are retained in the liquid phase. The licensee assumed that 100 percent of the noble gases formed by the decay of the isotopes in the RWST liquid will become airborne and available for release to the outside environment. This release of noble gases goes beyond the guidance in RG 1.183 and is an additional conservatism. The licensee assumed iodine species in the RWST air space were 97 percent elemental iodine and 3 percent organic iodine,in accordance with RG 1.183. The activity released from the RWST is exhausted to the environment through the RWST vent at a rate equal to the sum of the inflow to the RWST from the ESF pump mini-flow isolation valves and RWST discharge check valve. The licensee calculated the doses for the EAB and LPZ for this release pathway and added theresults to the EAB and LPZ doses from the other three pathways to give the total offsite radiological consequences of the LOCA. The licensee calculated the doses for the CR for thisrelease pathway and added the results to the doses from the other three pathways and the direct shine dose to give the total radiological consequences of the LOCA in the CR. The NRC staff has reviewed the information provided by the licensee and compared the valuesused to the guidance in RG 1.183, Appendix A, and has determined that the licensee's assumptions for the post-LOCA RWST leakage source term and transport are consistent with, or more conservative than, the guidance in RG 1.183. The NRC staff also performed an independent calculation of the dose consequences of this LOCA pathway using the licensee'sassumptions for input to the RADTRAD computer code. The NRC staff's calculation confirmed the licensee's dose results. 3.2.1.2.3 Post-LOCA PASS LeakageThe licensee stated that, at SONGS 2 and 3, the post-accident sampling system is maintainedfor severe accident management only. The PASS samples containment sump liquid, reactor coolant, and containment air. Portions of the PASS that are outside the containment provide apotential leakage pathway. The licensee's analysis assumed that the reactor coolant is the fluid in the PASS that leaks during the DBA LOCA. Reactor coolant has a greater iodine activity concentration than the containment sump liquid or containment air, therefore giving bounding dose results. Although RG 1.183 does not discuss leakage from the PASS in detail, the NRC staff is of the view that guidance given in the regulatory guide on ESF system leakage is applicable to the PASS leakage pathway analysis, because this guidance deals with reactor coolant leakage outside containment. Consistent with RG 1.183 guidance, the licensee assumed that PASS leakage starts at 30-minutes post-LOCA, which is the earliest time that PASS sampling could start according to plant procedure. The PASS leakage is assumed to continue for the remainder of the 30-day LOCA duration. In the licensee's dose analysis, the PASS leakage rate is assumed to be 700 cc/hr, which is two times the maximum expected leakage from the PASS. The licensee assumes that 10 percent of the iodine in the PASS leakage flashes to vapor and is available for release to the outside environment. This 10 percent flashing assumption is consistent with the guidance in RG 1.183, which states that if the water temperature is less than 212 "F, then10 percent of the iodine in the water leakage is assumed to become airborne. The PASS sample is cooled by the sample vessel heat exchanger to 120 oF to allow for low temperaturesample analysis, and the majority of the leakage from the PASS sample station fittings will be at low temperature. In accordance with RG 1.183 guidance, with the exception of iodine, all radioactive materials inthe recirculating fluid are retained in the liquid phase. The licensee assumed that 100 percent of the noble gases formed by the decay of the isotopes in the PASS liquid will become airborne and become available for release to the outside environment. This release of noble gases goes beyond the guidance in RG 1.183 and is an additional conservatism. The licensee assumed iodine species in the airborne release from the PASS leakage were 97 percent elemental iodineand 3 percent organic iodine, in accordance with RG 1.183. The activity released from the PASS leakage into the radwaste building is exhausted to the atmosphere via the main plant vent. The licensee calculated the doses for the EAB and LPZ for this release pathway and added theresults to the EAB and LPZ doses from the other three pathways to give the total offsite radiological consequences of the LOCA. The licensee calculated the doses for the CR for thisrelease pathway and added the results to the doses from the other three pathways and the direct shine dose to give the total radiological consequences of the LOCA in the CR. The NRC staff has reviewed the information provided by the licensee and compared the valuesused to the guidance in RG 1.183, Appendix A, and has determined that the licensee's assumptions for the post-LOCA PASS leakage source term and transport are consistent with, or more conservative than, the guidance in RG 1.183. The NRC staff also performed an independent calculation of the dose consequences of this LOCA pathway using the licensee'sassumptions for input to the RADTRAD computer code. The NRC staff's calculation confirmed the licensee's dose results. 3.2.1.3 CR Habitability for LOCA As described by the licensee, the SONGS 2 and 3 CRs are located within a common CRE. Thetechnical support center is located within the CRE. Consistent with the guidance in RG 1.183, the licensee's dose analyses consider the sources of radiation that may cause exposure to CR personnel through intake or infiltration of contaminated air and radiation shine from radioactive material in the containment, release plume, and systems and components external to the CRE. The CR emergency air cleanup system (CREACUS) is an ESF system. The CREACUSemergency mode of operation pressurizes the CR with filtered outside air and recirculates and filters the air within the CRE. The CREACUS emergency mode may be actuated either automatically following a CR isolation signal or manually. The CR isolation signal may begenerated automatically by a safety injection actuation signal (SIAS) or by detection of high radioactivity concentrations in the CR outside air intake flow. The CREACUS is automatically actuated by the SIAS within 10 seconds of the onset of the LOCA, based on high containment pressure. Because the gap release activity is not assumed to be released into the containment until 30 seconds after the onset of the LOCA per RG 1.183, the CREACUS emergency mode of operation is credited from the beginning of the LOCA dose analysis. The NRC staff finds that the CREACUS operation assumption in the dose analysis is acceptable because the CREACUS would be in operation prior to the arrival of any contaminated air at any of the CR ventilation systems outside air intakes or the CRE boundary. On June 12, 2003, the NRC staff issued Generic Letter (GL) 2003-01, "CR Habitability." ThisGL identified NRC staff concerns regarding the reliability of current surveillance testing to identify and quantify CR inleakage, and requested licensees to confirm the most limiting unfiltered inleakage into their CRE. In a letter dated August 5, 2003, the licensee submitted a "60-day" response to this GL for SONGS 2 and 3 that included a schedule for testing for the CRE unfiltered inleakage. By letter dated September 17, 2004, SCE provided the results of tracer gas testing performed during May of 2004. The AST dose analyses in the license amendment request reviewed here revised the radiological consequences analyses for both SONGS units to update the plant licensing and design basis with regard to CRE unfiltered inleakage. Based on information provided by the licensee, tracer gas testing showed that for theCREACUS emergency mode of operation, the SONGS 2 and 3 CRE has 67 standard cubic feet per minute (scfm) of unfiltered inleakage for train A and 65 scfm for train B. Regulatory Position C.1.4 of RG 1.197 provides guidance indicating that it is optional to include the uncertainty for facilities that demonstrate a CRE inleakage rate less than 100 scfm. Since the tracer gas tested unfiltered inleakage rate value is less than 100 scfm, the licensee complies with the guidance of RG 1.197 when using 67 scfm in their calculations for one train operation.

The licensee also determined that for the two-train operation of CREACUS, the unfiltered inleakage would be no more than the sum of the two single train measurements or 132 cfm, with an uncertainty of +/- 127 cfm. Consistent with RG 1.197, consideration of uncertainty indicates a maximum dual train unfiltered inleakage rate of 259 cfm. To add additional conservatism, the licensee's dose calculations assumed a CRE unfiltered inleakage rate of 990 cfm (which is greater than the tracer gas tested value) to evaluate the accident conditions, and added an additional 10 cfm for ingress and egress. Based on these conservatisms, the NRC staff finds that the assumption of total unfiltered air inleakage of 1000 cfm in the licensee's dose analysis is acceptable. To model the CREACUS mode of operation, the licensee conservatively assumed a greateroutside air intake rate than the nominal rate and a smaller CR recirculation rate than nominal. Filtration of particulates and iodine is credited for the emergency air conditioner (EAC) filters, but not for the emergency ventilation supply filters. The EAC filters are assumed to have 95 percent efficiency for elemental and organic iodine and 99 percent efficiency for particulateiodine and aerosols, consistent with the SONGS 2 and 3 current licensing basis and guidance in RG 1.183 and RG 1.52, as verified by SONGS 2 and 3 TS on filter testing. Major parameters and assumptions used by the licensee and found acceptable by the NRC staff for modeling the CR for DBA dose analyses are listed in Table 7. The licensee also considered the dose in the CR due to gamma shine from the externalradioactive plume (i.e., environmental cloud shine), radioactive material in the CREACUS filters in the CRE (i.e., CR filter shine), radioactive material in the containment (i.e., direct containment shine), and radioactive material in ESF recirculation loop piping outside the CRE (i.e., pipingshine). These shine doses are added to the CR dose for the LOCA. The licensee conservatively modeled the source for each shine dose based on the LOCA modeling discussed above for immersion and inhalation. The activity in the CR filters is maximized by assuming that the filters are 100 percent efficient at removing iodine and particulates from the air. The licensee modeled the shielding between the CR dose receptors and the source, including the concrete structures and air space within the walls, floor, and ceilings that are between the source and receptor. The shine dose in the CR from each source is the maximum dose calculated at one of several dose receptors within the CR board area. The NRC staff reviewed the licensee's description of the shine dose calculations and compared them to the analyses currently in the SONGS 2 and 3 licensing and design basis. Based on its evaluation, the NRC staff finds the licensee's modeling assumptions to be reasonable, consistent with the shielding dose calculations previously performed by the licensee and, therefore, acceptable.

The total shine dose results are a fraction of the doses due to immersion and inhalation, as listed in Table 4.5-11 of the licensee's letter dated March 10, 2006. 3.2.1.4 LOCA Radiological Consequences and Conclusion The licensee re-evaluated the radiological consequences resulting from the postulated LOCAusing the AST and concluded that the radiological consequences at the EAB, LPZ, and CR are within the dose criteria specified in 10 CFR 50.67. The results of the licensee's radiological consequence calculation are provided in Table 1 and the major parameters and assumptions used by the licensee and found acceptable by the NRC staff are listed in Tables 4 and 4a. The NRC staff performed an independent calculation of the dose consequences of LOCA usingthe licensee's assumptions and confirmed the licensee's dose results. The NRC staff found that the licensee used analysis assumptions and inputs consistent with applicable regulatory guidance identified in Section 2.0 of this SE and with those assumptions stated in the SONGS 2 and 3 UFSAR as design bases. Based on this, the NRC staff concluded that the LOCA radiological consequences are acceptable.3.2.2Main Steamline BreakThe MSLB accident considered is the complete severance of the largest main steamline outsidecontainment, downstream of an MSIV. Appendix E of RG 1.183 identifies acceptable radiological analysis assumptions for the MSLB. The radiological consequences of a MSLBoutside containment will bound the consequences of a break inside containment. Thus, only the MSLB outside of containment is considered with regard to radiological consequences.

Activity is introduced into the nuclear steam supply system (secondary side) through steam generator tube leakage, also called primary-to-secondary leakage. The licensee's dose analysis assumes a primary-to-secondary leakage rate into any single steam generator of 0.5 gpm, which is the maximum leak rate allowed by TS 3.4.13, and is consistent with RG 1.183 guidance. Activity is released to the outside environment through steaming and release through the steamline break, the MSSVs, the ADVs and the AFW turbine exhaust. The licensee assumed that the MSLB accident is terminated when shutdown cooling is initiated at13,659 seconds, and all steam releases from both steam generators cease. The SONGS 2 and 3 current licensing basis evaluates pre-trip and post-trip return to power forthe MSLB. The pre-trip return to power may result in no more than 7 percent of the core experiencing fuel failure by damage to the cladding, and is the case evaluated for the DBA dose analysis. The post-trip break does not result in fuel failure. The licensee's analysis conservatively assumes 10 percent fuel failure to bound the fuel failure by the design-basis deterministic departure from nucleate boiling ratio (DNBR) prediction of 7 percent or the fuel failure determined by the proposed DNB statistical convolution methodology. The proposed methodology is addressed in Section 3.4 of this SE. The licensee applied a radial peaking factor of 1.75 to account for differences in power level across the core. Consistent with RG 1.183, Appendix E, because more than minimal fuel failure is postulated, the licensee's dose analysis does not include primary coolant iodine spiking. The initial primary coolant activity concentration is assumed to be at the maximum TS 3.4.16 limit of 1.0 Ci/gm dose equivalentI-131. The licensee assumes that the secondary coolant activity concentration prior to the accident is at the maximum TS 3.7.19 limit of 0.10 Ci/gm dose equivalent I-131. Leakage from the RCS to the steam generators is assumed to be the maximum value permittedby TSs. Primary-to-secondary leakage is apportioned between faulted and intact steam generators in such a manner that the radiological consequence is maximized. The maximum TS limit for primary-to-secondary leakage to any one steam generator is 0.5 gpm. In accordance with RG 1.183 guidance, the licensee assumed that during periods when the steam generator tubes are covered, the primary-to-secondary leakage is mixed with the secondary water without flashing. The licensee used the CENTS computer code to analyze the secondary side response, including release masses and steam generator tube uncovery. The use of the CENTS code has been previously found acceptable by the NRC staff. The licensee's analysis determined that the tubes in one steam generator are uncovered from 17.3 seconds to 6,620 seconds after the break. The tubes in the other steam generator are uncovered from 17.2 seconds to 6,621 seconds after the break. The licensee conservatively assumed that the tubes in both steam generators are uncovered from 0 seconds to 6,621 seconds. Consistent with RG 1.183, during periods when the steam generator tubes are uncovered, aportion of the primary-to-secondary leakage flashes to vapor based on the thermodynamic conditions in the reactor coolant and secondary coolant. The licensee conservatively assumed a flashing fraction of 20 during uncovery, which bounds the calculated flashing for the event.

The flashed portion of primary-to-secondary leakage is available for release to the outside environment without further reduction for iodine scrubbing. The unflashed portion is assumed to mix with the bulk water. Consistent with RG 1.183, an iodine partition coefficient of 100 is applied to determine the amount of airborne iodine available for release to the environment. All noble gases released from the primary coolant are released to the environment without reduction. The licensee's dose analysis modeled the SONGS 2 and 3 steam generators moisturecarryover of 0.20 percent by assuming a particulate isotope partition coefficient of 500 to determine airborne particulate isotopes available for release to the environment. The release through the break begins at time zero and is terminated at 16.3 seconds when the MSIVs are fully closed. The dose analysis assumes a total mass release through the break of 115,103 pounds mass (lbm), which is 10 percent higher than the calculated mass release whichconsists of inventory loss from both steam generators and main feedwater flow for 3.83 seconds. The MSSV and ADV mass releases assumed in the licensee's dose analysis are also increasedby 10 percent over the calculated mass releases for the accident to provide analysis margin.

The MSSV release is assumed to begin when the MSSVs open at 1,200 seconds and terminate when the MSSVs close at 1,822 seconds. The ADV release is assumed to begin when the ADVs are opened by operator action at 30 minutes and continue for the duration of the accident. The licensee's dose analysis assumed that the total mass release occurs through the MSSVs from 1,200 seconds until 1,800 seconds, and through the ADVs from 1,800 seconds until the end of the accident. The NRC staff finds this assumption to be conservative because the ADV dispersion factors are higher than the MSSV factors, which results in a higher dose. For the time intervals during which the steam turbine AFW pump is operating, a release ofactivity occurs through that pathway. The licensee modeled two periods of AFW operation.

The first is from 89 seconds to 748 seconds, following the postulated accident. The second is from 1,921 seconds to the end of the accident. The AFW steam turbine mass releases assumed in the licensee's dose analysis are also increased by 10 percent over the calculated mass releases for the accident to provide additional margin in the analysis results. The licensee assumed that the CR is isolated at 180 seconds based on a high-radiation CRisolation signal. The high-radiation CR isolation signal is to be generated and the normalventilation system outside air dampers would be closed within 120 seconds per the licensee-controlled Specification 3.3.100 in the Technical Requirements Manual of the UFSAR..

After 180 seconds, the CREACUS emergency mode of operation provides filtered pressurization and recirculation in the CRE as discussed above for the LOCA. The licensee re-evaluated the radiological consequences resulting from the postulated MSLBaccident using the AST and concluded that the radiological consequences at the EAB, LPZ, and CR are within the dose criteria specified in 10 CFR 50.67 and the accident specific dose acceptance criteria specified in SRP Section 15.0.1 and RG 1.183. These accident-specific dose acceptance criteria for the MSLB with fuel failure are a TEDE of 25 rem at the EAB for any two hours, 25 rem at the outer boundary of the LPZ and 5 rem in the CR for the duration of the accident. The NRC staff found that the licensee used analysis assumptions and inputs consistent with applicable regulatory guidance identified in Section 2.0 of this SE and with those stated in the SONGS 2 and 3 UFSAR as design bases. The NRC staff also performed an independent calculation of the dose consequences of theMSLB using the licensee's assumptions for input to the RADTRAD computer code. The NRC staff's calculation confirmed the licensee's dose results. The major parameters and assumptions used by the licensee and found acceptable to the NRC staff are presented in Table 5. The results of the licensee's radiological consequence calculation are provided inTable 1. The EAB, LPZ, and CR doses estimated for the MSLB meet the accident dose criteria in 10 CFR 50.67 and are, therefore, acceptable. 3.2.3Fuel Handling AccidentsThe licensee analyzed the radiological consequences of an FHA in two different locations;inside the containment, and in the FHB. RG 1.183, Appendix B, identifies acceptable radiological analysis assumptions for the FHA. The FHA inside containment involves the inadvertent dropping of a fuel assembly during fuelhanding operations inside the reactor vessel and the consequent rupture of fuel rods in both the dropped and impacted fuel assemblies. The number of fuel rods damaged during the accident is based on a conservative analysis that considers the most limiting case, as is discussed in the SONGS 2 and 3 UFSAR. Per the UFSAR, a maximum of 226 fuel rods will fail, which includes 16 rods in the dropped assembly and 210 rods in the impacted assemblies. This fuel failure assumption was previously found acceptable, and nothing in the current license amendment request affects this assumption. Consistent with guidance in RG 1.183, the licensee applied a radial peaking factor of 1.75 tothe average fuel rod isotope inventory to determine the activity inventory in each of damaged fuel rods. The dose analysis models 72 hours8.333333e-4 days <br />0.02 hours <br />1.190476e-4 weeks <br />2.7396e-5 months <br /> of radioactive decay, consistent with the minimum decay time required by SONGS 2 and 3 licensee controlled specification 3.9.101 prior to movement of irradiated fuel in the reactor vessel. The licensee's analysis used the RG 1.183, Table 3, fission product gap fraction values. The licensee stated that the fuel at SONGS 2 and 3 meets the burnup and maximum linear heat generation rate limitations for use of RG 1.183, Table 3, as given in the table's footnote. Consistent with RG 1.183, the licensee assumed that the gap activity is instantaneouslyreleased into the refueling water. Because the depth of water above the damaged fuel is greater than 23 feet, the licensee applied an iodine effective decontamination factor of 200 per the RG 1.183 guidance. Retention of noble gases in the refueling water was assumed to be negligible, and particulate radionuclides are assumed to be retained by the refueling water. The licensee assumed that the iodine released from the water is composed of 57 percent elemental and 43 percent organic species, consistent with RG 1.183. SONGS 2 and 3 TS 3.9.3 allows the containment personnel airlock to be open under certainconditions during core alterations and movement of irradiated fuel in containment. In addition, the licensee has submitted a license amendment request to allow the containment equipment hatch to be open during core alterations and movement of irradiation fuel in containment. Since the containment may be open during fuel handling operations, the activity that escapes the refueling water is assumed to be released to the environment over a 2-hour period without credit for containment closure. No credit is taken for activity dilution within the air of the containment dome space. Activity may be released to the environment through the containment purge system or by leakage through open containment penetrations, including the personnel airlock or equipment hatch. The CR atmospheric dispersion factors are different for each release point, and are discussed above in Section 3.1.2 of this SE. Since one set of atmospheric dispersion factors for a given source-receptor pair does not consistently give less dispersion than the others over time, the licensee assumed a composite set of limiting atmospheric dispersion factors to give the maximum dose result in the CR. The FHA in the FHB involves the inadvertent dropping of a fuel assembly during fuel handingoperations in the spent fuel pool and the consequent rupture of fuel rods in both the dropped fuel assembly. The number of fuel rods damaged during the accident is based on a conservative analysis that considers the most limiting case, as is discussed in SONGS 2 and 3 UFSAR, Section 15.7.3.4.2.2. Per the UFSAR, a maximum of 60 fuel rods will fail. This fuel failure assumption was previously found acceptable, and nothing in the current LAR affects this assumption. Consistent with guidance in RG 1.183, the licensee applied a radial peaking factor of 1.75 tothe average fuel rod isotope inventory to determine the activity inventory in each of damaged fuel rods. The dose analysis models 72 hours8.333333e-4 days <br />0.02 hours <br />1.190476e-4 weeks <br />2.7396e-5 months <br /> of radioactive decay, consistent with the minimum decay time required by SONGS 2 and 3 licensee controlled specification 3.9.101 prior to movement of irradiated fuel in the reactor vessel. The licensee's analysis used the RG 1.183, Table 3, fission product gap fraction values. The licensee stated that the fuel at SONGS 2 and 3 meets the burnup and maximum linear heat generation rate limitations for use of RG 1.183, Table 3, as given in the table's footnote.Consistent with RG 1.183, the licensee assumed that the gap activity is instantaneouslyreleased into the spent fuel pool water. Because the depth of water above the damaged fuel is greater than 23 feet as assured by TS 3.7.16, the licensee applied an iodine effective decontamination factor of 200 per the RG 1.183 guidance. Retention of noble gases in the spent fuel pool water was assumed to be negligible, and particulate radionuclides are assumed to be retained by the spent fuel pool water. The licensee assumed that the iodine released from the water is composed of 57 percent elemental and 43 percent organic species, consistent with RG 1.183. The licensee does not take credit for FHB closure for the FHA in the FHB or for operation of theFHB post-accident cleanup unit filter system. Consistent with RG 1.183, the licensee's dose analysis assumes that the activity that escapes the spent fuel pool is released to the environment over a 2-hour period. Activity may be released to the environment by the fuel handling normal ventilation exhaust system through the main plant vent or as leakage through the FHB penetrations such as doors. The CR atmospheric dispersion factors are different for each release point, and are discussed above in Section 3.1.2 of this SE. Since one set of atmospheric dispersion factors for a given source-receptor pair does not consistently give less dispersion than the others over time, the licensee assumed a composite set of limiting atmospheric dispersion factors to give the maximum dose result in the CR. For both cases, the licensee assumed that the CR is isolated at 180 seconds based on ahigh-radiation CR isolation signal. The high-radiation CR isolation signal is to be generated andthe normal ventilation system outside air dampers would be closed within 120 seconds per licensee-controlled Specification 3.3.100. After 180 seconds, the CREACUS emergency mode of operation provides filtered pressurization and recirculation in the CRE as discussed above for the LOCA. The licensee re-evaluated the radiological consequences resulting from the postulated FHAinside the containment and the postulated FHA in the FHB using the AST and concluded that the radiological consequences at the EAB, LPZ, and CR are within the dose criteria specified in 10 CFR 50.67 and the accident-specific dose acceptance criteria given in SRP 15.0.1 and RG 1.183. These accident-specific dose acceptance criteria for the FHA are a TEDE of6.3 rem at the EAB for any 2 hours2.314815e-5 days <br />5.555556e-4 hours <br />3.306878e-6 weeks <br />7.61e-7 months <br /> and 6.3 rem at the outer boundary of the LPZ and 5 rem in the CR for the duration of the accident. The results of the licensee's radiological consequence calculation are provided in Table 1 and the major parameters and assumptions used by the licensee and found acceptable by the NRC staff are listed in Table 6. The NRC staff found that the licensee used analysis assumptions and inputs consistent withapplicable regulatory guidance identified in Section 2.0 of this SE and with those stated in the SONGS 2 and 3 UFSAR as design bases. The NRC staff also performed an independent calculation of the dose consequences of the FHA using the licensee's assumptions for input to the RADTRAD computer code. The NRC staff's calculation confirmed the licensee's dose

results. 3.2.4 SummaryAs described above, the NRC staff reviewed the assumptions, inputs, and methods used by thelicensee to assess the radiological consequences of the proposed full implementation of an AST and TS changes requested. The NRC staff finds that the licensee used analysis methods and assumptions consistent with the conservative regulatory requirements and guidance identified in Section 2.0 above. The NRC staff finds reasonable assurance that SONGS 2 and 3, as modified by this license amendment, will continue to provide sufficient safety margins with adequate defense-in-depth to address unanticipated events and to compensate for uncertainties in accident progression and analysis assumptions, and parameters. The NRC staff also finds, with reasonable assurance, that the licensee's estimates of the EAB, LPZ, and CR doses will comply with the dose criteria in 10 CFR 50.67. Therefore, the NRC staff concludes that the proposed license amendment is acceptable.This licensing action is considered a full implementation of the AST. With this approval, theprevious accident source term in the SONGS 2 and 3 design basis is superseded by the AST proposed by the licensee. The previous offsite and CR accident dose criteria expressed in terms of whole body, thyroid, and skin doses are superseded by the TEDE criteria of 10 CFR 50.67 or fractions thereof, as defined in SRP 15.0.1. All future radiological accident analyses performed to show compliance with regulatory requirements shall address all characteristics of the AST and the TEDE criteria as defined the SONGS 2 and 3 design basis, and modified by the present amendment.3.3pH Control in the Containment Sump 3.3.1BackgroundIn its request for the approval of the license amendment for the application of the AST methodology for SONGS 2 and 3, the licensee included an analysis of its ability to maintain the pH of the sump water at or above 7 for a period of 30 days following a LOCA in order to minimize the amount of radioactive iodine released to the environment. The control of pH in the sump water is to retain radioiodine in the sump water, which effects the radioactivity removal inside containment, is addressed in Section 3.2.1.1.1 of this SE. This subsection reviews the methodology and supporting calculations provided by the licensee. 3.3.2EvaluationAccording to NUREG-1465, "Accident Source Terms for Light-Water Nuclear Power Plants,"the iodine entering the containment from the damaged core during an accident contains at least 95 percent cesium iodide (CsI). Upon its dissolution in sump water the iodine will be,therefore, predominantly in an iodide form (I- ). However, in the radiation field existing in thecontainment, some of this iodine will be converted into the molecular form (I 2 ). This conversionis strongly dependent on pH and will increase with the decreasing value of pH. Since molecular iodine is scarcely soluble in water, some of it will be released to the containment atmosphere and leak to the outside, contributing to the radiation doses. However, if the pH of sump water is maintained at or above 7, formation of the molecular iodine will be impeded and its release to the outside considerably reduced. At SONGS 2 and 3, this is achieved by adding to the sump water, TSP (trisodium phosphate - dodecahydrate, Na 3 PO 4 12 H 2 O). The TSP is a bufferingagent and if a sufficient amount added, its buffering action will prevent pH from dropping below 7 with addition of strong acids. The two strong acids of concern that are generated in the containment and released to the containment sump water are hydrochloric acid, which is generated by radiolytic decomposition of Hypalon cable insulation and nitric acid, which is produced by irradiation of the water and air in the containment. Since the concentration of these acids is zero at the beginning of a LOCA and is building up with time, immediately after a LOCA, the TSP has to neutralize only boric acid which accumulates in the containment sump from the RCS, accumulators, and RWST. However, as the concentration of strong acids in the sump is increasing, more TSP is need to maintain pH basic. Therefore, the amount of TSP stored in the containment should be determined by the amount of total acids (boric and strong acids) present in the containment at 30 days after a LOCA. The licensee calculated this amount using the Polestar STARpH code. The code is based onwork performed at Oak Ridge National Laboratory and reported in NUREG/CR-5732, "Iodine Chemical Forms in LWR [Light-Water Reactor] Severe Accidents," April 1992, and NUREG/CR-5950, "Iodine Evolution and pH Control," December 1992. The code has been used for safety-related analyses of numerous operating plants under accident conditions.

Using this code, the licensee determined that 2.0x10 4 pounds of TSP are required formaintaining basic or neutral pH throughout the 30-day post-LOCA period. In the plant, this TSP is stored in metal baskets located in the sump. After a LOCA, the water collected in the sump dissolves the TSP and forms a solution. Because of the buffering action of the TSP, the sump pH remains at an approximate value of 7, regardless of the concentrations of strong acids present. The NRC staff reviewed the licensee's analysis and confirmed that the amount of TSP calculated by the licensee will maintain a neutral or basic environment in the sump during the 30-day post-LOCA period.3.3.3SummaryThe NRC staff reviewed the licensee's assumptions and calculations for controlling sump pH inorder to maintain a neutral or basic environment (pH 7) in the containment sump. Thiscondition is necessary to keeping dissolved iodine in the containment water for 30 days after a LOCA. The NRC staff's review indicates that the assumptions and methodologies used by the licensee in its analysis are valid and consistent with the accepted methods, and adequately assure proper control of the sump water pH. 3.4Departure from Nucleate Boiling (DNB) Statistical Convolution Methodology forEstimating Fuel Failure for non-LOCA Events This methodology for estimating fuel failures is used in the calculation of the doseconsequences for the MSLB in Section 3.2.2 of this SE.3.4.1Introduction and BackgroundIn addition to its request to apply an AST methodology for SONGS 2 and 3, the licensee's LARrequested expanded use of fuel failure estimates by DNB statistical convolution methodology to all UFSAR Chapter 15 non-LOCA events that assume a loss of flow (i.e., a loss of alternating current (AC) power) and that fail fuel. The DNB statistical convolution methodology for estimating fuel failures in use at SONGS 2 and 3, is contained in CENPD-183-A, "C-E Methods for Loss of Flow Analysis." Currently, the licensee is only licensed to use the DNB statistical convolution methodology for the reactor coolant pump (RCP) sheared shaft event analysis. 3.4.2 Regulatory EvaluationPart 50 of 10 CFR, Appendix A, GDC 17, "Electric power systems," prohibits anticipatedoperational occurrences (AOOs) initiated by or resulting in a loss of flow from exceeding any specified acceptable fuel design limit (SAFDL). Therefore, AOOs initiated by or resulting in a loss of flow are not permitted to result in fuel failures.According to RG 1.183, Secti on 3.6, "[t]he amount of fuel damage caused by non-LOCA designbasis events should be analyzed to determine, for the case resulting in the highest radioactivity release, the fraction of the fuel that reaches or exceeds the initiation temperature of fuel melt and the fraction of fuel elements for which the fuel clad is breached." The DNB statistical convolution methodology is an NRC-approved method for estimating the amount of fuel failures for non-LOCA design-basis events.As previously stated, the licensee is currently only licensed to use the DNB statisticalconvolution methodology for its RCP sheared shaft event analysis. The proposed change would extend the use of the methodology to all non-LOCA events that assume a loss of flow (i.e., a loss of AC power) and that fail fuel.3.4.3 Evaluation DNB is the point at which the heat transfer from a fuel rod rapidly decreases due to theinsulating effect of a steam blanket that forms on the rod surface when the temperature continues to increase. Fuel failure is conservatively assumed to occur whenever a fuel rod experiences DNB, even for a short time. To determine when a fuel rod would experience DNB, a DNBR SAFDL is established. DNBR is the ratio of the heat flux to cause DNB to the actual local heat flux of a fuel rod. The DNBR SAFDL is defined such that there is a 95-percent probability with a 95-percent confidence level that the fuel rod will not experience DNB whenever its DNBR is above the DNBR SAFDL.Historically, a fuel rod was considered to experience DNB if its DNBR fell below the DNBRSAFDL. The number of failed fuel rods was the number with a DNBR below the DNBR SAFDL. All fuel rods with a DNBR at or above the DNBR SAFDL were considered to not experienceDNB and therefore did not fail.The CENPD-183-A DNB statistical convolution methodology for estimating fuel failures makesuse of the fact that the DNBR SAFDL is defined such that there is a 95-percent probability with a 95-percent confidence level that the fuel rod will not experience DNB whenever its DNBR is above the DNBR SAFDL. In other words, there is a 5-percent probability, with a 95-percent confidence level, that DNB will occur at the DNBR SAFDL. Extrapolating further, as DNBR increases above the DNBR SAFDL the probability decreases that DNB will occur, and as DNBR decreases below the DNBR SAFDL, the probability increases that DNB will occur.The CENPD-183-A procedure for a DNB statistical convolution is to group fuel rods with respectto radial peaking factors; calculate the minimum DNBR in each radial peaking group; and determine the probability of experiencing DNB corresponding to a DNBR value. The number of fuel rods, within a radial peaking group, that are predicted to experience DNB and fail, is the product of the number of fuel rods in the radial peaking group and the probability of experiencing DNB associated with the corresponding minimum DNBR. The total number of fuel failures is the summation of each group's failed fuel. The key to the use of the DNB statistical convolution methodology is developing the probabilitydistribution of exceeding DNB with respect to DNBR. In the NRC SE report (SER) approving CENPD-183-A, the NRC staff stated: Since experimental evidence (Ref. 11) indicates that fuel cladding failure is notnecessarily coincident with a short duration of DNB, we conclude that the statistical convolution technique is conservative and acceptable provided that the probability distribution for DNB is acceptable. CENPD-183-A contained probability distributions for the Combustion Engineering (CE) 14x14and 16x16 rod matrixes. Those distributions were established using the TORC computer code and the CE-1 heat flux correlation. The NRC SER for CENPD-183-A establishes a clear link between the computer code, critical heat flux correlation, and the probability distributions and requires NRC staff approval for any combination other than that specifically approved in the SER. The NRC SER states: The staff has reviewed the Loss-of-Flow [LOF] topical report CENPD-183. Thecomputer codes used for the LOF analysis are acceptable for their assigned purposes.

The COAST code can be used for transient system flow rate calculations, and the QUIX code can be used for axial power distributions and reactivity calculations. Either COSMO/W-3 or TORC/CE-1 can be used for steady state hot channel minimum DNBR calculations. The CESEC code is still under review and the comparison, to date, of its analysis with ANO-2 [Arkansas Nuclear One, Unit 2] startup test data indicates that CESEC is acceptable for LOF NSSS [Nuclear Steam Supply System] transient response analysis. The statistical convolution technique is acceptable for fuel rod failure calculations. Thefuel damage probability distributions for CE-1 correlation are listed in Tables 2 and 3, respectively for the standard 16x16 and 14x14 fuel assemblies. If COSMO/W-3 is used for DNBR calculations, the applicant is required to submit a fuel damage probabilitydistribution for staff's approval. In summary, the LOF analysis procedure using the static method of hot channel DNBRcalculation is acceptable.The LAR did not request the use of alternate computer codes or critical heat flux correlations,nor has the licensee provided sufficient information to warrant the approval to use alternate computer codes or critical heat flux correlations. A review of the licensee's UFSAR reveals that the CENPD-183-A listed computer codes are part of the licensee's current licensing basis and that TORC/CE-1 is being used for the DNB analysis for the single RCP sheared shaft transient.The SER approving CENPD-183-A did not limit it to the sheared shaft event analysis. CENPD-183-A was intended for use with loss of forced flow (LOF) transients at CE plants.

Further, the abstract for CENPD-183-A does not limit it to any particular LOF transient. In demonstrating the methodology, CENPD-183-A provided sample analysis for a four pump LOF coast down and a seized shaft LOF. Additionally, while the concept of a DNB statistical convolution methodology is approved in CENPD-183-A, the probability distributions provided therein are computer code, critical heat flux correlation, and CE fuel design 14x14 and 16x16 specific. Therefore, based on the review stated above, the NRC staff is approving the licensee's request to expand the use of of fuel failure estimates by DNB statistical convolution methodology to all USFAR Chapter 15 non-LOCA events that assume a loss of flow. However, the use of any combination of computer code, critical heat flux correlation, or fuel design, other than that explicitly approved by CENPD-183-A, will require submittal of revised probability distributions for NRC staff review and approval.4.0REGULATORY COMMITMENTSIn its letter dated December 27, 2004, the licensee has made the following regulatory commitments:1. Following approval of this license amendment request, future revisions toUFSAR Chapter 15 design basis accident control room and offsite radiological consequence analyses will be performed using AST methodology.2. Following approval of this license amendment request, the manual dosecalculation methodology as described in Emergency Planning Implementation Procedures (EPIPs) and other Emergency Planning guidance documents will be revised to reflect AST methodology.3. Raddose V dose assessment software will be evaluated by June 30, 2005, todetermine what specific changes may be warranted in order to maintain consistency with the manual dose assessment calculation methodology.The licensee, in an e-mail dated December 27, 2006 (to be added to ADAMS), stated that itcompleted an evaluation of Raddose V assessment software in April, 2006, to determine the changes necessary to implement the AST PCN. That evaluation identified 6 changes, listed below, needed prior to implementing the AST LAR. 1.Derive new SGTR Iodine removal factors for AST environment.2.Update Technical Team Notebook.

3.Update 40.200 "don't use Source term capability in a SGTR."

4.Update 40.100 - "new Iodine removal factors."

5.Issue Emergency Planning Bulletin when AST implemented to appropriate Health PhysicsEmergency Response Organization members.6.Place Note on RadDose V computers stating "don't use Source term capability in a SGTR."The licensee sated that these changes will be made prior to implementation of the AST. 4. Following approval of this license amendment request, future revisions to AccidentMonitoring setpoint calculations will reflect the AST.5. Following approval of this license amendment request, SCE will provide the revisedUFSAR sections to the NRC as part of its normal UFSAR update required by 10 CFR 50.71 (e).The NRC staff finds that reasonable controls for the implementation and for subsequentevaluation of proposed changes pertaining to the regulatory commitments are best provided by the licensee's administrative processes, including its commitment management program. The regulatory commitments do not warrant the creation of regulatory requirements (items requiring prior NRC approval of subsequent changes).

5.0STATE CONSULTATION

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

6.0ENVIRONMENTAL CONSIDERATION

The amendments change the requirements with respect to installation or use of a facilitycomponent located within the restricted area as defined in 10 CFR Part 20. The NRC staff has determined that the amendments involve no significant increase in the amounts, and no significant change in the types, of any effluents 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 the amendments involve no significant hazards consideration and there has been no public comment on such finding (70 FR 5248, dated February 1, 2005). Accordingly, the amendments meet 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 the amendments.

7.0CONCLUSION

The Commission has concluded, based on the considerations discussed above, that: (1) thereis 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 amendments will not be inimical to the common defense and security or to the health and safety of the public.Principal Contributors:M. HartK. Parczewski B. Harvey K. WoodDate: December 29, 2006 Table 1Licensee Calculated Radiological ConsequencesTEDE (rem) (1)DBA EAB (2)LPZ (3)CRLOCA5.21.92.8Dose acceptance criteria(4)25255.0MSLB, 10% Fuel Failure4.10.12.2Dose acceptance criteria25255.0FHA, Inside Containment0.8<0.10.3Dose acceptance criteria6.36.35.0FHA, FHB0.2<0.1<0.1Dose acceptance criteria6.36.35.0(1) Total effective dose equivalent (2) Exclusion area boundary(3) Low-population zone, maximum 2-hour dose(4) From RG 1.183 and SRP Section 15.0.1 (a)For the FHA-IC event, the highest /Q value for each time interval among the main plant vent,containment shell, and equipment hatch release pathways is used.(b)For the FHA-FHB event, the highest /Q value for each time interval between the main plantvent and FHB release pathways is used.Table 2SONGS 2 and 3CR Atmospheric Dispersion Factors ReleasePathwayDBA/Q Value (sec/m 3)0-2 hrs2-8 hrs 8-24 hrs1-4 days4-30days Main Plant Vent!LOCA ESF RecircLoop Leakage

!LOCA PASSLeakage!FHA-IC (a)!FHA-FHB (b)1.15x10-3 6.23x10-4 2.14x10-4 2.22x10-4 2.02x10-4Containment Shell!LOCAContainment Leakage!FHA-IC (a)1.01x10-3 6.41x10-4 1.77x10-4 2.36x10-4 2.20x10-4EquipmentHatch!FHA-IC (a)8.01x10-4 6.35x10-4 1.78x10-4 2.23x10-4 2.03x10-4 MSSV!MSLB (1200 to1822 sec)1.22x10-3 7.52x10-4 2.48x10-4 2.86x10-4 2.60x10-4 ADV!MSLB (1800 sec toend of event) 3.70x10-3 1.99x10-3 6.95x10-4 7.04x10-4 6.34x10-4SLB-OC!MSLB (0 to 16.3sec)7.78x10-3 4.81x10-3 1.62x10-3 1.83x10-3 1.68x10-3AFW TurbineExhaust!MSLB (89 to 748sec)!MSLB (1921 sec toend of event) 8.60x10-4 3.70x10-4 1.56x10-4 1.61x10-4 1.30x10-4RWST!LOCA RWSTRelease 5.67x10-4 2.25x10-4 8.84x10-5 8.97x10-5 7.37x10-5 FHB!FHA-FHB (b)9.48x10-4 7.61x10-4 1.92x10-4 2.65x10-4 2.43x10-4 Table 3SONGS 2 and 3EAB and LPZ Atmospheric Dispersion FactorsReceptorTime Interval/Q Value (sec/m 3)EAB0-2 hrs2.72x10-4 LPZ 0-8 hrs7.72x10-6 8-24 hrs4.74x10-61-4 days3.67x10-64-30 days2.67x10-6 Table 4Parameters and Assumptions Used inRadiological Consequence Calculations forLOCAParameter ValueReactor power, MWt3,507Containment volume, ft 3Total2,284,000Sprayed area1,907,000 Unsprayed area377,000Containment leak rates, % per day0 to 24 hour0.1 24 to 720 hour0.00833 days <br />0.2 hours <br />0.00119 weeks <br />2.7396e-4 months <br />s0.05Containment mechanical mixing rate, cfmSprayed to unsprayed31,000 Unsprayed to sprayed31,000Containment iodine and aerosol removalVariable SprayTable 4aAerosol natural depositionPowers 10 th percentileElemental iodine deposition, per hr4.26ESF recirculation volume, ft 346,647ESF leak rates, cfm0 to 20 minutes0 20 minutes to 30 days0.007RWST volume, ft 3Air space35,880Liquid space7,345RWST flow rates, cfmMini-flow into RWST, total after 20 min0.4010 Discharge check valve into RWST0 - 1.08 hr0 1.08 - 2 hr1.2859 2 - 8 hr1.2778 8 - 24 hr0.9622 24 - 96 hr0.5103 96 - 119.72 hr0.1078 119.73 - 720 hr0 Table 4 (cont.)Parameters and Assumptions Used inRadiological Consequence Calculations forLOCARWST flow rates, cfm (cont.)From RWST water to RWST air space0.4010 From RWST air space to environment0 - 20 min0 20 min - 1.08 hr0.4010 1.08 - 2 hr1.6869 2 - 8 hr1.6788 8 - 24 hr1.3632 24 - 96 hr0.9113 96 - 119.72 hr0.5088 119.72 - 720 hr0.4010RCS volume, ft 310,179PASS leak rates, cfm0 - 30 min0 30 min - 30 days4.12E-04Iodine flashing factors for leakage, %ESF recirculation 10 RWST mini-flow leakage to air space10 PASS10RWST iodine partition coefficient200 Atmospheric dispersion factorsTables 2 and 3 Table 4aElemental Iodine and Aerosol Spray Removal RatesElemental Iodine Spray Removal RatesTime Period (hr)Elemental Iodine SprayRemoval Rate (1/hour)Prior to injection phase0During injection phase1.02Start of recirculation to 2 hr20.002 to 420.004 to 818.868 to 13.816.0113.8 to 2412.9924 to 4810.09 48 to 967.8396 to 7203.78Aerosol Spray Removal RatesTime Period (hr)Aerosol Spray RemovalRate (1/hour)0 to 1.85.151.8 to 23.792 to 3.81.32 3.8 to 40.794 to 80.628 to 13.80.5213.8 to 7200.50 Table 5Parameters and Assumptions Used in Radiological Consequence Calculations forMSLB Accident Parameter ValueInitial RCS activity, Ci/gm DEI-1311.0Secondary coolant activity, Ci/gm DEI-1310.1Failed fuel, percent of core10Radial peaking factor1.75Primary-to-secondary leakage per SG, gpm0.5RCS volume, ft 310,179RCS liquid mass, gm2.015E+08 Secondary coolant mass, lbm1.59E+05Shutdown cooling initiation time, sec13,659 Steamline break mass release, lbm0 to 16.3 sec115,103 16.3 to 13,659 sec0Steam release from MSSV, lbm0 to 30 min47,553 30 min to 2 hr555.5 2 hr to 13,659 sec0Steam release from ADV, lbm0 to 30 min8,078 30 min to 2 hr64,522 2 hr to 13,659 sec78,944Steam release from AFW steam turbine, lbm0 to 30 min47,553 30 min to 2 hr555.5 2 hr to 13,659 sec0SG flashing factors and partition coefficientsSG tube uncovery period, sec0 to 6,621 Iodine flashing factor during uncovery, %20 Iodine partition coefficient100 Particulate partition coefficient500Atmospheric dispersion factorsTables 2 and 3 Table 6Parameters and Assumptions Used inRadiological Consequence Calculations forFHAsParameter ValueNumber of failed fuel rodsFHA inside containment226 FHA in FHB60Fraction of Core Inventory in Gap Kr-850.10 I-1310.08 Alkali metals0.12 Other noble gases / iodines0.05Decay time after reactor shutdown, hr72Radial peaking factor1.75 Minimum water depth above damaged fuel, ftReactor vessel23 Spent fuel pool23Pool effective iodine decontamination factor 200 Iodine species above water, % of iodineElemental iodine57 Organic iodine43Release duration, hr2 FHA in containmentContainment closureNot credited Release pointBounding for containment openings FHA in FHBFHB closureNot credited Release pointBounding for FHBAtmospheric dispersion factorsTables 2 and 3 Table 7Parameters and Assumptions Used inRadiological Consequence Calculations forCR HabitabilityParameter ValueCR net free volume, ft 326,920CR total unfiltered inleakage, cfm1,000(Includes 10 cfm ingress/egress)CR ventilation normal modeUnfiltered outside air intake, cfm6402CREACUS emergency mode initiationLOCA (SIAS-induced), sec0 MSLB, FHA (high-radiation), sec180CREACUS emergency mode, one trainFiltered outside air intake, cfm2,200 Filtered recirculation, cfm29,934CREACUS emergency mode, two trainsFiltered outside air intake, cfm4,400 Filtered recirculation, cfm59,869CREACUS emergency ventilation supply filtersnot credited CREACUS emergency air conditioner filter efficiencyElemental iodine, %95 Organic iodine, % 95 Particulate iodine and aerosols, %99