ML043650403

Proposed Change Number (PCN) 555 Alternative Source Term

ML043650403
Person / Time
Site:	San Onofre
Issue date:	12/27/2004
From:	Nunn D Southern California Edison Co
To:	Document Control Desk, Office of Nuclear Reactor Regulation
References
PCN 555, RG-1.183
Download: ML043650403 (121)

Text

VA SOUTHERN CALIFORNIA L# EDISON Dwigbt E. Nunn Vice President An EDISON INTERNATIONALC Company December 27, 2004 U.S. Nuclear Regulatory Commission Attn: Document Control Desk Washington, DC 20555-0001

SUBJECT:

San Onofre Nuclear Generating Station, Units 2 and 3 Docket Nos. 50-361 and 50-362 Proposed Change Number (PCN) 555 Alternative Source Term

Reference:

Letter from A. E. Scherer (SCE) to Document Control Desk (NRC),

dated September 17, 2004,

Subject:

Docket Nos. 50-361 and 50-362, Response to Generic Letter 2003-01, "Control Room Habitability," Tracer Gas Test Results, San Onofre Nuclear Generating Station, Units 2 and 3

Dear Sir or Madam:

Pursuant to 10 CFR 50.90, Southern California Edison (SCE) hereby requests approval of Amendment Application Numbers 231 and 215 for San Onofre Units 2 and 3, respectively, which consist of Proposed Change Number (PCN) 555: Revise the San Onofre Nuclear Generating Station (SONGS) Units 2 and 3 accident source term used in the design basis radiological consequences analyses. This license amendment is requested in accordance with the requirements of 10 CFR 50.67, which addresses the use of an Alternative Source Term (AST) at operating reactors, and relevant guidance of Regulatory Guide 1.183. This license amendment request represents full-scope implementation of the AST described in Regulatory Guide 1.183. SCE has evaluated this request under the standards set forth in 10 CFR 50.92(c) and determined that a finding of "no significant hazards consideration" is justified.

This license amendment request is being submitted to fulfill the commitment described in SCE's letter dated September 17, 2004 (Referenced). Approval of this license amendment request will establish use of an AST methodology that will document the acceptability of an assumed increase in SONGS Units 2 and 3 control room envelope (CRE) unfiltered inleakage rate to a value of 1,000 cfm (including ingress and egress related inleakage). As described in the referenced letter, this is necessary to restore the Control Room Envelope to full qualification.

P.O. Box 128 San Clemente. CA 92674-0128 4 \10O 949-368-1480 Fax 949-368-1490

Document Control Desk December 27, 2004 This license amendment request is also being submitted to expand the allowed use of fuel failure estimates by Departure from Nucleate Boiling (DNB) statistical convolution methodology from only the reactor coolant pump sheared shaft event to the UFSAR Chapter 15 non-LOCA events that assume a loss of flow (i.e., a loss of AC power) and that fail fuel.

Implementation of this amendment does not require any revision to the current Technical Specifications. Implementation of this amendment will necessitate revision to various sections of the SONGS Units 2 and 3 UFSAR. SCE requests that this amendment be issued effective as of the date of issuance, to be implemented within 180 days from the date of issuance.

If you have any questions or require additional information, please contact Mr. Jack Rainsberry at (949) 368-7420.

Sincerely,

Enclosures:

1. Notarized Affidavits

2. Licensee's Evaluation Attachments A. Acronyms B. List of Regulatory Commitments cc: B. S. Mallett, Regional Administrator, NRC Region IV C. C. Osterholtz, NRC Senior Resident Inspector, San Onofre Units 2 &3 B. M. Pham, NRC Project Manager, San Onofre Units 2 and 3 S. Y. Hsu, Department of Health Services, Radiologic Health Branch

ENCLOSURE I NOTARIZED AFFIDAVITS

UNITED STATES OF AMERICA NUCLEAR REGULATORY COMMISSION Application of SOUTHERN CALIFORNIA )

EDISON COMPANY, ET AL. for a Class 103 ) Docket No. 50-361 License to Acquire, Possess, and Use )

a Utilization Facility as Part of ) Amendment Application Unit No. 2 of the San Onofre Nuclear ) No. 231 Generating Station )

SOUTHERN CALIFORNIA EDISON COMPANY, et a[. pursuant to 10CFR50.90, hereby submit Amendment Application No. 231. This amendment application consists of Proposed Change No. 555 to Facility Operating License No. NPF-10. Proposed Change No. 555 is a request to revise the accident source term used in the design basis radiological consequences analyses to an Alternative Source Term in accordance with the requirements of 10 CFR 50.67.

State of Califomia County of San Diego ubscribed and sworn to (or affirmed) b fore me this ;k+h7 day of

( Lws ,'~

~20Q4 By:

ight E. Nunk Vice President I N

MAMME SANCHEZ

- Notay Pubic - Cdl omJo Notary Public e)

UNITED STATES OF AMERICA NUCLEAR REGULATORY COMMISSION Application of SOUTHERN CALIFORNIA )

EDISON COMPANY, ET AL. for a Class 103 ) Docket No. 50-362 License to Acquire, Possess, and Use )

a Utilization Facility as Part of ) Amendment Application Unit No. 3 of the San Onofre Nuclear ) No. 215 Generating Station SOUTHERN CALIFORNIA EDISON COMPANY, et al. pursuant to I OCFR50.90, hereby submit Amendment Application No. 215. This amendment application consists of Proposed Change No. 555 to Facility Operating License No. NPF-15. Proposed Change No. 555 is a request to revise the accident source term used in the design basis radiological consequences analyses to an Alternative Source Term in accordance with the requirements of 10 CFR 50.67.

State of California County of San Diego Subscribed and sworn to (or affirmed) before me this )211f*'- day of

.Satc /y. 20 t B3y:K E.id Nunn Vice President N

Notary Public

/htt Z~A,

ENCLOSURE 2 LICENSEE'S EVALUATION PCN 555 Alternative Source Term

4.0 TECHNICAL ANALYSIS

5.0 REGULATORY SAFETY ANALYSIS 5.1 No Significant Hazards Consideration 5.2 Applicable Regulatory Requirements/Criteria

7.0 REFERENCES

ATTACHMENTS A. Acronyms B. List of Regulatory Commitments

1.0 INTRODUCTION

This letter is a request to amend Operating Licenses NPF-10 and NPF-15 for San Onofre Nuclear Generating Station (SONGS) Units 2 and 3, respectively.

This license amendment request will revise the accident source term used in the design basis radiological consequences analyses to an Alternative Source Term (AST) in accordance with the requirements of 10 CFR 50.67.

This license amendment request will also expand the allowed use of fuel failure estimates by Departure from Nucleate Boiling (DNB) statistical convolution methodology from only the reactor coolant pump sheared shaft event to the UFSAR Chapter 15 non-LOCA events that assume a loss of flow (i.e., a loss of AC power) and that fail fuel.

2.0 PROPOSED CHANGE

This license amendment is requested 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 1.183 (Reference 1). This license amendment request represents full-scope implementation of the AST described in Regulatory Guide 1.183.

Full-scope implementation of an AST requires re-analysis of Updated Final Safety Analysis Report (UFSAR) Chapter 15 accident analyses, including the Loss-of-Coolant Accident (LOCA) and Fuel Handling Accident (FHA) at a minimum. Southern California Edison (SCE) has re-analyzed the LOCA, FHA inside containment (FHA-IC), and FHA in the Fuel Handling Building (FHA-FHB).

In addition, to ensure that the most limiting accident in terms of dose consequences has been included, SCE has re-analyzed the pre-trip Steam Line Break Outside Containment (SLB-OC) accident as well.

This license amendment request will also expand the allowed use of fuel failure estimates by DNB statistical convolution methodology from only the reactor coolant pump sheared shaft event to the UFSAR Chapter 15 non-LOCA events that assume a loss of flow (i.e., a loss of AC power) and that fail fuel.

Implementation of this license amendment will require changes to the SONGS UFSAR Chapter 15 control room and offsite radiological consequence analyses for these four Design Basis Accidents (DBAs). Following approval of this license amendment request, SCE will provide the revised UFSAR sections to the NRC as part of its normal UFSAR update required by 10 CFR 50.71 (e).

In summary, the requested license amendment will revise the accident source term used in the design basis radiological consequences analyses to an Page 1 of 110

Alternative Source Term in accordance with the requirements of 10 CFR 50.67, and allow fuel failure estimates for UFSAR Chapter 15 non-LOCA events that assume a loss of flow (i.e., a loss of AC power) and that fail fuel, to be based on the DNB statistical convolution methodology.

3.0 BACKGROUND

Generic Letter 2003-01 On June 12, 2003, the NRC issued Generic Letter (GL) 2003-01, uControl Room Habitability" (Reference 2) GL 2003-01 discussed the results of Control Room Envelope (CRE) inleakage testing at several plants. These test results indicate that current testing methods (positive pressure testing) do not give a good indication of CRE inleakage. GL 2003-01 requested that licensees confirm that the most limiting unfiltered inleakage into the CRE is no more than the value assumed in the design basis radiological analyses. SCE performed CRE inleakage testing in May, 2004 and transmitted the results of this testing to the NRC by letter dated September 17, 2004 (Reference 3).

As stated in SCE's August 5, 2003 letter (Reference 4) responding to GL 2003-01, the current value of assumed unfiltered inleakage into the CRE in the design basis radiological analyses is 0 cfm, plus an additional assumed 10 cfm for ingress and egress. As described in SCE's letter dated September 17, 2004, testing has shown that actual CRE inleakage exceeds that assumed in the current design basis radiological analyses. Since testing was completed, SONGS Units 2 and 3 have continued to operate based on operability assessments that rely on AST methodology. Approval of this proposed change will make the AST the SONGS Units 2 and 3 licensing and design basis and will restore the SONGS Units 2 and 3 CRE to full qualification.

As described above, this license amendment request represents full-scope implementation of the AST as described in Regulatory Guide 1.183. The LOCA, FHA-IC, FHA-FHB, and pre-trip SLB-OC analyses have been updated and show results within the acceptance criteria defined in 10CFR50.67 with an assumed CRE unfiltered boundary inleakage of 990 cfm plus 10 cfm assumed unfiltered inleakage due to CRE ingress and egress. Following approval of this license amendment request, future revisions to UFSAR Chapter 15 design basis accident control room and offsite radiological consequence analyses will be performed using AST methodology. In addition, following approval of this license amendment request, fuel failure estimates for UFSAR Chapter 15 non-LOCA events that assume a loss of flow (i.e., a loss of AC power) and that fail fuel may be based on the DNB statistical convolution methodology.

There are no physical changes to plant equipment or operation of the plant as a result of this proposed change. There are no changes to the Technical Specifications (TSs) as a result of this proposed change.

Page 2 of 110

Impact to the Site The following clarifications are provided to address source term implementation considerations of RG 1.183 that are not explicitly stated elsewhere in this proposed change:

1. Impact on Equipment Environmental Qualification Regulatory Guide 1.183 Regulatory Position 1.3.5 states that equipment Environmental Qualification (EQ) analyses that have assumptions or inputs affected by a proposed plant modification associated with the AST implementation should be updated to address these impacts. Regulatory Position 1.3.5 of RG 1.183 also states that the NRC staff is assessing the effect of increased cesium releases on EQ doses to determine whether licensee action is warranted, and that until such time as this generic issue (GSI-1 87) is resolved, licensees may use either the AST or the Technical Information Document (TID) 14844 assumptions for performing the required EQ analyses. The GSI has since been closed with the determination that a generic action was not warranted. Consistent with the RG 1.183 guidance and the GSI closure determination, since no plant modifications are required to address AST implementation, the existing equipment qualification analyses, which are based upon the TID-14844 source term, are considered acceptable. Future EQ analyses may use either the AST or TID-1 4844 source term.

2. Control Room Habitability NUREG-0737 (Reference 5), Task III.D.3.4, "Control-Room Habitability Requirements," requires that the control room (CR) operators be adequately protected against the effects of accidental release of radioactive gases, and that the nuclear power plant can be safely operated or shut down under design basis accident conditions (as required by 10 CFR 50, Appendix A, General Design Criterion 19). With approval of the Alternative Source Term methodology, the dose acceptance criterion for CR operators will become the 5 rem Total Effective Dose Equivalent (TEDE) dose criterion of 10 CFR 50.67.

As documented in Section 4, the CR operator post-accident dose for each of the evaluated events is less than 5 rem TEDE. Compliance with this dose acceptance criterion ensures that the CR operators are adequately protected against the effects of accidental release of radioactive gases, and that the nuclear power plant can be safely operated or shut down under design basis accident conditions.

Page 3 of 110

3. Emergency Response Facility Habitabilitv The Emergency Response Facilities (ERFs) consist of the Technical Support Center (TSC), the Operations Support Center (OSC), and the Emergency Operations Facility (EOF).

NUREG-0737, Task II.B.2, "Design Review of Plant Shielding and Environmental Qualification of Equipment for Spaces/Systems Which May Be Used in Postaccident Operations", provides dose criteria for CR and TSC occupants. This document states that the dose to individuals in the CR or TSC should not be in excess of 5 rem whole body, or its equivalent to any part of the body for the duration of the accident. The document also states the CR and TSC dose criteria should be based on the control room occupancy factors contained in Standard Review Plan 6.4.

The TSC is located within the control room envelope, in rooms that overlook the Units 2 and 3 control room operating areas. TSC occupants receive the same inhalation and immersion doses calculated for CR occupants. SCE has determined that although a given gamma shine dose to a TSC occupant may be higher or lower than the dose to a CR occupant, the net effect from all post-accident gamma shine sources is that the TSC shine dose is no more than the CR shine dose. Since the CR dose criterion of 5 rem TEDE is met using an AST methodology, the TSC dose criterion of 5 rem would also be met.

The OSC is located in the Auxiliary Building. This facility does not have isolation or filtration capabilities. Consistent with current emergency planning requirements, post-accident radiation dose rate surveys of the OSC would be performed by Health Physics personnel, and protective actions would be taken if necessary.

The EOF is located at the SONGS Mesa Facility across Interstate 5 from San Onofre Units 2 and 3. The EOF is protected by charcoal and HEPA filtration systems. Because of the distance from the various possible release points and the available filtration systems, any doses seen in the EOF would be bounded by those seen in the control room or TSC.

4. Impact on Emergency Planning Radioloqical Assessment Methodology Implementation of an AST will impact several aspects of Emergency Planning Radiological Dose Assessment Methodology. The behavior of radioactive iodine released under post-accident conditions, which is defined by the accident source term, is an input to emergency planning dose assessment calculations. Following approval of this license amendment request, the manual dose calculation methodology as described in Emergency Planning Implementation Procedures (EPIPs)

Page 4 of 110

and other Emergency Planning guidance documents will be revised to reflect AST methodology.

Raddose V dose assessment software will be evaluated by June 30, 2005, to determine what specific changes may be warranted in order to maintain consistency with the manual dose assessment calculation methodology.

5. Post Accident Sampling Capability Post Accident Sampling System (PASS) licensing requirements were deleted from the SONGS Units 2 & 3 Technical Specifications per Unit 2 License Amendment 178 and Unit 3 License Amendment 169 (Reference 6). Currently, the diluted depressurized grab sample portion of the PASS is retained for severe accident management only. Application of AST methodology has determined that, if required for severe accident management, an individual can draw a PASS sample approximately 17 hours1.967593e-4 days <br />0.00472 hours <br />2.810847e-5 weeks <br />6.4685e-6 months <br /> following the onset of a loss of coolant accident without exceeding the NUREG-0737 dose criteria for whole body and extremity exposures.

Application of AST methodology has also determined that an individual performing reactor coolant sample collection and analysis at the Normal Sampling Station (NSS) for boron, hydrogen, gas activity, and liquid activity at 3 hours3.472222e-5 days <br />8.333333e-4 hours <br />4.960317e-6 weeks <br />1.1415e-6 months <br /> following a non-LOCA event, and with up to 5% fuel clad failures, will not exceed the 5 rem whole body dose limit. This determination is consistent with the current licensing basis source term dose evaluation for this same exposure mechanism.

As described in the Safety Evaluation Report for License Amendments 178 and 169 for San Onofre Units 2 and 3, SCE is committed to maintain the capability for classifying fuel damage events at the Alert level threshold of 300 pCi/gram dose equivalent iodine-131. The value of 300 pCi/gram DE 1-131 is unaffected by use of an AST. Therefore, the capability for classifying fuel damage events at the Alert level threshold is unchanged.

6. Accident Monitoring Instrumentation A review of Accident Monitoring setpoint calculations was performed. This review determined that no setpoint changes will be required to implement the AST. Some setpoint calculations were unaffected. The remaining calculations were determined to be conservative relative to calculations that would be based on an AST, because the mix of isotopes predicted by the AST calculations is bounded by the mix of isotopes expected under the current licensing basis. Following approval of this license amendment request, future revisions to Accident Monitoring setpoint calculations will reflect the AST source term.

Page 5 of 110

7. Other Design Bases Not Affected This proposed change has been determined to have no affect on post-accident access, environmental reports, facility siting, or leakage control.

4.0 TECHNICAL ANALYSIS

To address the issue of measured CRE inleakage rate exceeding the currently assumed CRE inleakage rate, a series of new radiological dose analyses have been originated using the AST methodology of Regulatory Guide (RG) 1.183 to document the acceptability of an assumed increase in SONGS Units 2 and 3 CRE unfiltered inleakage rate to a value of 1,000 cfm (including ingress and egress related inleakage). This Section summarizes the analyses supporting the SONGS Units 2 and 3 license amendment request.

As recommended by RG 1.183, a complete LOCA dose analysis has been performed. Additionally, dose analyses have been performed to assess the radiological consequences of FHAs in both the containment and fuel handling buildings, and the radiological consequences of a pre-trip SLB-OC. The FHAs and SLB-OC have been re-analyzed since the current licensing basis analyses for these events challenge the offsite dose acceptance criteria. In addition, the pre-trip SLB-OC had not been previously evaluated for control room dose consequences.

With the exception of the Increased Main Steam Flow with single failure (IMSF-SF) event, all other design basis accidents that are currently evaluated in the SONGS Units 2 and 3 UFSAR Chapter 15 have control room and offsite dose consequences that are less severe than those of the LOCA, FHA, and pre-trip SLB-OC accidents. The IMSF-SF Exclusion Area Boundary (EAB) whole body gamma dose is slightly greater than for the pre-trip SLB-OC. The IMSF-SF EAB thyroid inhalation dose is significantly less than for the pre-trip SLB-OC. For this reason, the IMSF-SF has not been re-evaluated as part of this license amendment request.

Following approval of this license amendment request, future revisions to UFSAR Chapter 15 design basis accident control room and offsite radiological consequence analyses will be evaluated using AST methodology whenever a need arises for them to be updated.

Section 4.1 summarizes the core and fuel rod fission product inventories that were recalculated using the guidance in AST RG 1.183 as clarified in RG 1.195 (Reference 7). Section 4.1 also presents the recalculated activity profiles associated with operation at the primary and secondary activity concentration limits, with and without iodine spiking, as specified in Technical Specification Limiting Conditions For Operation (LCOs) 3.4.16 and 3.7.19.

Page 6 of 110

Section 4.2 summarizes the model used in evaluating offsite dose consequences at the EAB and at the outer boundary of the low population zone (LPZ). This model is generic to the dose analyses evaluating offsite dose consequences.

Section 4.3 summarizes the model used in evaluating control room dose consequences. This model is generic to the dose analyses evaluating control room dose consequences.

Adoption of the AST methodology guidance has imposed the need to recalculate atmospheric dispersion between various post-accident release points and the control room outside air ventilation intakes. The atmospheric dispersion analysis uses the ARCON96 computer code and guidance provided in RG 1.194 (Reference 8). Section 4.4 summarizes the ARCON96 analyses.

Section 4.5 summarizes the model used in evaluating the radiological consequences of a loss of coolant accident.

Section 4.6 summarizes the model used in evaluating the radiological consequences of a fuel handling accident inside the containment building.

Section 4.7 summarizes the model used in evaluating the radiological consequences of a fuel handling accident inside the fuel handling building.

Section 4.8 summarizes the model used in evaluating the radiological consequences of a pre-trip steam line break outside containment.

Page 7 of 110

Section 4.1 ACTIVITY INVENTORIES AND TECHNICAL SPECIFICATION ACTIVITY PROFILES The SONGS Units 2 and 3 core and fuel rod fission product inventories have been recalculated using the guidance in AST RG 1.183 as clarified in RG 1.195.

Activity profiles have also been recalculated for operation at the primary and secondary activity concentration limits, with and without iodine spiking.

Section 4.1.1 Core and Average Fuel Rod Activity Inventories Table 4.1-1 summarizes the parameters modeled in the evaluation of the reactor core activity inventory. The core inventory of fission products is based on the maximum full-power operation of the core with, as a minimum, currently licensed values for fuel enrichment, fuel burnup, and an assumed core power equal to the current licensed rated thermal power times the emergency core cooling system (ECCS) evaluation uncertainty. These parameters were examined parametrically to maximize the fission product inventory. The period of irradiation is of sufficient duration to allow the activity of dose significant radionuclides to reach equilibrium or to reach maximum values. The core inventory was developed using the SAS2H and ORIGEN-S modules of the NRC-sponsored SCALE code package, which is an appropriate isotope generation and depletion computer code.

TABLE 4.1-1: ACTIVITY INVENTORY MODEL PARAMETERS PARAMETER MODELED VALUE Maximum Core Average Burnup 40.0 GWD/T Maximum Core Average Enrichment 4.8 w/o U-235 Maximum Core Uranium Loading 95.5 MTU Core Rated Thermal Power 3,438 MW-t Core Thermal Power Uncertainty 0.58% actual, 2.0% modeled Analyzed Core Thermal Power 3,507 MW-t Minimum Number of Fuel Rods per Core 51,132 rods/core The ORIGEN-S code was executed for the various combinations of core average burnups (0, 10, 20, 30 and 40 GWD/T) and enrichments (3.8 and 4.8 w/o U-235).

Each ORIGEN-S code run evaluated the activity inventory in a single fuel assembly. In any code run, the maximum curie value of an isotope represents the sum of the ORIGEN-S code output identified as "Light Elements", "Fission Products", and "Actinides". For each isotope, the maximum curie value from the ORIGEN-S code runs was chosen to represent the inventory of that isotope in the composite fuel assembly. Activity inventories were originated for 540 isotopes of the elements listed in RG 1.183 Table 5. The maximum full core accident source term was determined by multiplying the composite maximum fuel assembly activity inventory by 217 fuel assemblies per core.

Page 8 of 110

Table 4.1-2 summarizes the full core accident source term. The original 540 isotopes were reduced to the Table 4.1-2 listing of 166 isotopes that are included in the Bechtel LocaDose code isotope library. Per the guidance of RG 1.183 Regulatory Position 4.1.1, the isotope libraries contain all radionuclides, including progeny from the decay of parent radionuclides, that are significant with regard to dose consequences and the released radioactivity. The 166 isotopes include all but one of the isotopes listed in the RADTRAD code isotope library as identified in NUREG/CR-6604 (Reference 9) Table 1.4.3.3-2. The missing isotope is Niobium-97m, which is a short-lived daughter of Zirconium-97, and which does not have a dose conversion factor in Federal Guidance Report 11 (Reference 10). Niobium-97m decays to Niobium-97. The Bechtel LocaDose code isotope library conservatively assumes that Zirconium-97 decays directly to Niobium-97.

Consistent with the guidance of RG 1.183 Regulatory Position 3.1, for events that do not involve the entire core, Table 4.1-3 summarizes the average fission product inventory of each damaged fuel rod as determined by dividing the Table 4.1-2 total core inventory by the minimum number of fuel rods in the core.

Per RG 1.183 Regulatory Positions 3.2 and 3.4, the only elements to be considered in design basis analyses for non-LOCA events, including fuel handling accidents, are xenon, krypton, iodine, bromine, cesium, and rubidium.

Page 9 of 110

TABLE 4.1-2: REACTOR CORE ISOTOPE INVENTORY AT SHUTDOWN ISOTOPE CORE OTOPE CORE ISOTOPE CORE ISOTOPE CORE INVENTORY INVENTORY INVENTORY INVENTORY

[curies] [curies] [curies] [curies]

XE-1 31 M 1.22E+06 TE-127M 1.44E+06 CO-58 2.21 E+05 CM-243 2.26E+03 XE-1 33M 6.05E+06 TE-127 8.48E+06 CO-60 4.60E+05 CM-244 3.91 E+05 XE-1 33 1.93E+08 TE-1 29M 5.95E+O6 LA-1 40 1.90E+08 CM-245 3.06E+O1 XE-1 35M 4.06E+07 TE-129 2.91 E+07 LA-141 1.67E+08 CM-246 8.05E+00 XE-135 7.05E+07 TE-131M 1.92E+07 LA-142 1.66E+08 CM-247 4.34E-05 XE-1 37 1.80E+08 TE-131 7.90E+07 LA-143 1.63E+08 CM-248 1.97E-04 XE-1 38 1.79E+08 TE-1 32 1.34E+08 ZR-93 1.56E+02 AM-241 1.57E+04 KR-83M 1.43E+07 TE-1 33M 9.20E+07 ZR-95 1.78E+08 AM-242M 1.06E+03 KR-85M 3.12E+07 TE-1 33 1.11 E+08 ZR-97 1.69E+08 AM-242 9.20E+06 KR-85 1.09E+06 TE-134 1.92E+08 ND-144 O.OOE+00 AM-243 2.93E+03 KR-87 6.38E+07 SB-124 8.79E+04 ND-147 6.55E+07 CE-141 1.67E+08 KR-88 8.98E+07 SB-125 1.03E+06 EU-152 9.37E+02 CE-142 3.19E-03 KR-89 1.15E+08 SB-126M 5.13E+04 EU-154 7.68E+05 CE-143 1.64E+08 1-129 3.62E+0O SB-126 4.37E+04 EU-155 3.12E+05 CE-144 1.29E+08 1-130 2.50E+06 SB-127 8.57E+06 EU-156 2.65E+07 PU-236 5.34E+01 1-131 9.37E+07 SB-1 29 3.06E+07 NB-93M 2.16E+02 PU-237 7.05E+02 1-132 1.36E+08 SE-79 7.86E+0O NB-95M 2.05E+06 PU-238 3.56E+05 1-133 1.98E+08 BA-136M 6.36E+05 NB-95 1.79E+08 PU-239 3.60E+04 1-134 2.26E+08 BA-1 37M 1.1 9E+07 NB-97 1.70E+08 PU-240 5.16E+04 1-135 1.87E+08 BA-139 1.82E+08 PM-147 1.87E+07 PU-241 1.53E+07 1-136 9.20E+07 BA-1 40 1.81 E+08 PM-148M 3.30E+06 PU-242 2.50E+02 1-137 9.46E+07 BA-141 1.66E+08 PM-148 1.84E+07 PU-243 4.95E+07 1-138 4.73E+07 SR-89 1.24E+08 PM-149 5.97E+07 PU-244 O.OOE+00 BR-82 3.43E+05 SR-90 9.48E+06 PM-1 51 1.95E+07 NP-236 1.24E-03 BR-83 1.42E+07 SR-91 1.53E+08 PR-143 1.58E+08 NP-237 4.04E+O1 BR-84 2.73E+07 SR-92 1.55E+08 PR-144M 1.81 E+06 NP-238 4.67E+07 BR-85 3.12E+07 SR-93 1.68E+08 PR-144 1.30E+08 NP-239 2.03E+09 BR-87 5.10E+07 SR-94 1.63E+08 SM-147 1.97E-04 GD-152 O.OOE+00 BR-88 5.1OE+07 SR-95 1.46E+08 SM-148 O.OOE+00 U-232 O.OOE+00 CS-134M 4.56E+06 RU-103 1.55E+08 SM-149 O.OOE+00 U-234 0.00E+00 CS-134 1.87E+07 RU-105 1.12E+08 SM-151 5.14E+04 U-236 O.OOE+00 CS-135 5.97E+O1 RU-106 6.08E+07 SM-153 5.01 E+07 U-237 O.OOE+00 CS-1 36 5.58E+06 RH-103M 1.55E+08 Y-89M 1.33E+05 U-238 O.OOE+00 CS-1 37 1.25E+07 RH-1 05 1.02E+08 Y-90M 5.82E+02 PA-233 O.OOE+00 CS-138 1.90E+08 RH-106 6.73E+07 Y--90 9.94E+06 TH-228 O.OOE+00 CS-1 39 1.79E+08 PD- 07 1.32E+O1 Y-91 M 8.85E+07 TH-230 O.OOE+00 RB-86 1.90E+05 PD- 09 4.06E+07 Y--91 1.51 E+08 TH-232 O.OOE+OO RB-87 2.54E-03 MO-99 1.80E+08 Y--92 1.56E+08 TH-234 O.OOE+O0 RB-88 9.20E+07 TC-99M 1.59E+08 Y--93 1.13E+08 U-233 O.OOE+00 RB-89 1.22E+08 TC-99 1.55E+03 Y--94 1.75E+08 TH-229 O.OOE+00 RB-90 1.14E+08 TC-1 01 1.59E+O8 Y--95 1.77E+08 I TE-125M 2.23E+05 -CO57O.OOE+OO CM-242 5.08E+06 -

Page l0of 110

TABLE 4.1-3: AVERAGE FUEL ROD ISOTOPE INVENTORY AT SHUTDOWN ISOTOPE AVG. ROD ISOTOPE AVG. ROD ISOTOPE AVG. ROD ISOTOPE AVG. ROD INVENTORY INVENTORY IjNVENTORY INVENTORY

[curies] _ [curies] j [curies] [curies]

XE-1 31 M 2.38E+01 TE-127M 2.81 E+01 CO-58 4.33E+OO CM-243 4.41 E-02 XE-1 33M 1.18E+02 TE-1 27 1.66E+02 CO-60 9.OOE+0O CM-244 7.64E+OO XE-133 3.78E+03 TE-129M 1.16E+02 LA-140 3.72E+03 CM-245 5.98E-04 XE-135M 7.94E+02 TE-129 5.69E+02 LA-141 3.26E+03 CM-246 1.57E-04 XE-135 1.38E+03 TE-131M 3.76E+02 LA-142 3.24E+03 CM-247 8.49E-10 XE-137 3.52E+03 TE-131 1.54E+03 LA-143 3.19E+03 CM-248 3.85E-09 XE-138 3.50E+03 TE-132 2.62E+03 ZR-93 3.04E-03 AM-241 3.06E-01 KR-83M 2.79E+02 TE-133M 1.80E+03 ZR-95 3.48E+03 AM-242M 2.07E-02 KR-85M 6.11 E+02 TE-1 33 2.17E+03 ZR-97 3.31 E+03 AM-242 1.80E+02 KR-85 2.13E+01 TE-134 3.76E+03 ND-144 O.OOE+0O AM-243 5.73E-02 KR-87 1.25E+03 SB-124 1.72E+O0 ND-147 1.28E+03 CE-141 3.26E+03 KR-88 1.76E+03 SB-1 25 2.01 E+01 EU-1 52 1.83E-02 CE-1 42 6.24E-08 KR-89 2.25E+03 SB-126M 1.OOE+0O EU-154 1.50E+01 CE-143 3.21E+03 I-129 7.09E-05 SB-126 8.55E-01 EU-155 6.11E+0O CE-144 2.52E+03 1-130 4.88E+01 SB-127 1.68E+02 EU-156 5.18E+02 PU-236 1.04E-03 1-131 1.83E+03 SB-129 5.98E+02 NB-93M 4.23E-03 PU-237 1.38E-02 1-132 2.67E+03 SE-79 1.54E-04 NB-95M 4.02E+01 PU-238 6.96E+O0 1-133 3.87E+03 BA-1 36M 1.24E+01 NB-95 3.50E+03 PU-239 7.04E-01 1-134 4.41 E+03 BA-1 37M 2.32E+02 NB-97 3.32E+03 PU-240 1.01 E+OO 1-135 3.65E+03 BA-1 39 3.56E+03 PM-1 47 3.65E+02 PU-241 2.98E+02 1-136 1.80E+03 BA-140 3.54E+03 PM-148M 6.45E+01 PU-242 4.88E-03 1-137 1.85E+03 BA-141 3.24E+03 PM-148 3.60E+02 PU-243 9.68E+02 1-138 9.25E+02 SR-89 2.42E+03 PM-149 1.17E+03 PU-244 O.OOE+O0 BR-82 6.71 E+00 SR-90 1.85E+02 PM-1 51 3.81 E+02 NP-236 2.42E-08 BR-83 2.78E+02 SR-91 2.98E+03 PR-1 43 3.1 OE+03 NP-237 7.89E-04 BR-84 5.35E+02 SR-92 3.03E+03 PR-144M 3.54E+01 NP-238 9.12E+02 BR-85 6.11 E+02 SR-93 3.28E+03 PR-1 44 2.53E+03 NP-239 3.97E+04 BR-87 9.97E+02 SR-94 3.19E+03 SM-147 3.85E-09 GD-152 O.OOE+00 BR-88 9.97E+02 SR-95 2.86E+03 SM-1 48 O.OOE+00 U-232 O.OOE+O0 CS-134M 8.91E+01 RU-103 3.04E+03 SM-149 0.OOE+00 U-234 O.OOE+00 CS-134 3.67E+02 RU-105 2.19E+03 SM-151 1.01E+OO U-236 O.OOE+00 CS-135 1.17E-03 RU-106 1.19E+03 SM-153 9.80E+02 U-237 O.OOE+00 CS-1 36 1.09E+02 RH-103M 3.03E+03 Y-89M 2.61 E+00 U-238 O.OOE+OO CS-1 37 2.44E+02 RH-1 05 2.OOE+03 Y-90M 1.14E-02 PA-233 O.OOE+0O CS-138 3.71 E+03 RH-106 1.32E+03 Y--90 1.94E+02 TH-228 O.OOE+OO CS-1 39 3.50E+03 PD-1 07 2.58E-04 Y-91 M 1.73E+03 TH-230 O.OOE+OO RB-86 3.72E+00 PD-1 09 7.94E+02 Y--91 2.95E+03 TH-232 O.OOE+00 RB-87 4.97E-08 MO-99 3.52E+03 Y--92 3.06E+03 TH-234 O.OOE+00 RB-88 1.80E+03 TC-99M 3.1 OE+03 Y--93 2.22E+03 U-233 O.OOE+00 RB-89 2.38E+03 TC-99 3.03E-02 Y--94 3.42E+03 TH-229 O.OOE+00 RB-90 2.24E+03 TC-101 3.11E+03 Y--95 3.47E+03 TE-125M 4.37E+00 CO-57 O.OOE+00 CM-242 9.93E+O1 Page 11 of 110

Consistent with RG 1.183 Appendix B Section 3.1, the fuel handling accident dose analyses model 72 hours8.333333e-4 days <br />0.02 hours <br />1.190476e-4 weeks <br />2.7396e-5 months <br /> of radioactive decay prior to an event, which is the minimum decay time required by SONGS Units 2 and 3 Licensee Controlled Specification (LCS) 3.9.101 prior to movement of irradiated fuel in the reactor vessel (Note: some licensees refer to the LCS as their Technical Requirements Manual). Table 4.1-4 summarizes the average fuel rod isotope inventory for use in the fuel handling accident AST dose analyses. Table 4.1-4 determines the fission product inventory of an average fuel rod by decaying the Table 4.1-3 average rod inventory for 72 hours8.333333e-4 days <br />0.02 hours <br />1.190476e-4 weeks <br />2.7396e-5 months <br />.

Per RG 1.183 Regulatory Positions 3.2 and 3.4, the only elements considered in design basis analyses for fuel handling accidents are xenon, krypton, iodine, bromine, cesium, and rubidium. The limited number of elements listed in Table 4.1-4 is consistent with RG 1.183 Appendix B Section 3, which indicates that particulate radionuclides are retained by the water in the fuel storage pool or refueling water.

TABLE 4.1-4: AVERAGE FUEL ROD ISOTOPE INVENTORY AT 72 HOURS POST-SHUTDOWN AVERAGE FUEL ROD INVENTORY 72 HOURS AFTER SHUTDOWN ISOTOPE [curies]

BR-82 1.64E+00 BR-83 2.39E-07 1-129 7.09E-05 1-130 8.62E-01 1-131 1.41 E+03 1-132 8.76E-07 1-133 3.56E+02 1-134 7.57E-22 1-135 1.92E+00 KR-83m 4.01 E-10 KR-85m 8.89E-03 KR-85 2.13E+01 KR-87 1.13E-14 KR-88 4.66E-05 XE-131 m 2.OOE+01 XE-1 33m 4.57E+01 XE-1 33 2.54E+03 XE-1 35 5.74E+00 Page 12 of 110

Section 4.1.2 Primary and Secondary Coolant Activity Profiles Several of the AST dose analyses model primary and secondary activity profiles, with and without iodine spiking, associated with operation at the concentration limits specified in Technical Specification Limiting Conditions For Operation (LCOs) 3.4.16 and 3.7.19. These activity profiles have been recalculated for use in AST dose analyses to address changes in the maximum core activity profile specified in Section 4.1.1.

Table 4.1-5 summarizes the primary side equilibrium (no iodine spike) activity concentration profile determined for the conditions of 1.0 pCVgram lodine-131 dose equivalency and 100/E pCi/gram average activity concentration for other non-iodine isotopes, including tritium. These activity limits are consistent with LCO 3.4.16, "Reactor Coolant System Specific Activity". The primary side iodine activity concentration profile is based on the Technical Specification Section 1.1 definition for DOSE EQUIVALENT 1-131 (DE 1-131), using ICRP-30 thyroid inhalation dose conversion factors. The primary side non-iodine activity concentration profile is based on the Technical Specification Section 1.1 definition for E - AVERAGE DISINTEGRATION ENERGY, using total gamma and average beta disintegration energies provided in NUREG/CR-1413 (Reference 11).

Page 13 of 110

TABLE 4.1-5: ACTIVITY CONCENTRATIONS AT TECH. SPEC. LIMITS Isotope Primary Side Primary Side Secondary Side Secondary Side Equilibrium Pre-Accident Water Steam (No Spiking) Iodine Spike Concentration Concentration Concentration Concentration l [microClgm] [microCiLgm] [microCilgm] [microCigm]

1-131 8.24E-01 4.95E+01 8.33E-02 8.33E-04 1-132 2.28E-01 1.37E+01 1.55E-02 1.55E-04 1-133 9.54E-01 5.72E+01 9.19E-02 9.19E-04 1-134 9.15E-02 5.49E+00 4.04E-03 4.04E-05 1-135 4.31 E-01 2.59E+01 3.73E-02 3.73E-04 H-3 1 .80E+00 - 3.62E-02 3.62E-02 Br-84 3.81 E-02 2.85E-05 2.85E-07 Kr-85m 1.84E+00 O.OOE+00 6.52E-05 Kr-85 6.41 E+00 - O.OOE+00 2.27E-04 Kr-87 1.08E+00 - O.OOE+00 3.82E-05 Kr-88 3.36E+00 - O.OOE+O0 1.19E-04 Rb-88 3.40E+00 - 1.81 E-03 3.63E-06 Sr-89 9.16E-03 - 3.38E-05 6.76E-08 Sr-90 5.93E-04 - 2.19E-06 4.38E-09 Y-90 1.53E-03 - 5.51 E-06 1.1 OE-08 Sr-91 5.29E-03 1.65E-05 3.30E-08 Y-91 m 3.25E-03 - 3.85E-06 7.69E-09 Y-91 4.09E-02 1.51 E-04 3.02E-07 Zr-95 1.1OE-02 4.05E-05 8.09E-08 Mo-99 2.13E+00 7.69E-03 1.54E-05 Ru-1 03 1.45E-02 5.39E-05 1.08E-07 Ru-1 06 3.79E-03 1.41 E-05 2.82E-08 Te-129 4.41 E-02 6.48E-05 1.30E-07 Xe-131m 4.74E+00 O.OOE+00 1.67E-04 Te-132 6.53E-01 2.37E-03 4.73E-06 Xe-1 33 3.27E+02 O.OOE+00 1.16E-02 Cs-1 34 2.78E+00 - 1.04E-02 2.07E-05 Xe-135m 8. 1E-01 O.OOE+00 2.87E-05 Xe-1 35 1.41 E+01 O.OOE+00 4.99E-04 Cs-1 36 8.28E-01 3.07E-03 6.14E-06 Cs-137 1.87E+00 6.96E-03 1.39E-05 Xe-1 38 5.58E-01 O.OOE+00 1.97E-05 Ba-140 1.19E-02 4.38E-05 8.76E-08 La-1 40 1.20E-02 . 4.25E-05 8.50E-08 Pr-1 43 9.87E-03 . 3.63E-05 7.26E-08 Ce-144 8.01 E-03 2.96E-05 5.92E-08 Cr-51 3.36E-03 1.24E-05 2.48E-08 Mn-54 1.60E-03 5.92E-06 l 1.18E-08 Co-60 3.54E-03 1.31 E-05 2.62E-08 Fe-59 1.76E-03 6.50E-06 1.30E-08 Co-58 2.82E-02 -O1 .04E-04 2.08E-07 Page 14 of 110

Table 4.1-5 summarizes the primary side iodine activity concentration profile determined for the conditions of 60 pCi/gram DE 1-131 at full power operations (i.e., a pre-accident iodine spike). This activity limit is permitted by LCO 3.4.16.

The primary side pre-accident iodine spike activity concentration profile is a factor of 60 greater than the Table 4.1-5 profile for the normal operation conditions of 1.0 pCi/gram DE 1-131.

Table 4.1-5 summarizes the secondary side water iodine activity concentration profile determined for the condition of 0.1 pCi/gram lodine-131 dose equivalency.

This activity limit is consistent with LCO 3.7.19, "Secondary Specific Activity".

The secondary side iodine activity concentration profile is based on the Technical Specification Section 1.1 definition for DOSE EQUIVALENT 1-131, using ICRP-30 thyroid inhalation dose conversion factors. Table 4.1-5 also summarizes the secondary side water non-iodine activity concentration profile.

No Technical Specification limit exists for the secondary side water non-iodine activity concentration profile. The secondary side water non-iodine activity concentration profile was determined using a steady-state activity balance.

Primary side activity at LCO 3.4.16 concentrations was introduced into the steam generator liquid at the Technical Specification LCO 3.4.13, uReactor Coolant System Operational Leakage", total maximum primary-to-secondary leakage rate of 1.0 gallon/minute. Secondary side water activity was removed via partitioning into the secondary steam, demineralization by the full-flow condensate polisher demineralizer and the blowdown demineralizer, secondary side leakage, and radioactive decay. The concentration of noble gases in the secondary side water is negligible since all noble gas activity is assumed to be released to the steam generator gas space. This is modeled as a steam generator liquid to steam noble gas partition coefficient (i.e., liquid concentration divided by gas concentration) of 0.0.

Table 4.1-5 summarizes the secondary side steam activity concentration profile.

No Technical Specification limit exists for the secondary side steam activity concentration profile. The secondary side steam iodine and particulate activity concentrations were determined by considering partitioning and moisture carryover from the secondary side water activity concentration profile. The secondary side steam noble gas activity concentrations were determined using a steady-state activity balance and the assumption of a steam generator liquid to steam noble gas partition coefficient of 0.0. Primary side activity at LCO 3.4.16 concentrations was introduced into the steam generator water at the LCO 3.4.13 total maximum primary-to-secondary leakage rate of 1.0 gallon/minute.

Secondary side steam noble gas activity was removed at the total main steam flow rate.

In addition to the condition of a pre-accident iodine spike, AST dose analyses may model an accident induced (i.e., coincident or concurrent) iodine spike. Per RG 1.183 Appendices E and F, the concurrent iodine spike assumes that the iodine release rate from the fuel rods to the primary coolant increases to a value Page 15 of 110

of 335 or 500 times greater than the release rate corresponding to the iodine concentration at the equilibrium value of 1.0 pCi/gram DE 1-131 specified in the Technical Specifications. The calculation of the concurrent iodine spike release rate conservatively assumed maximum letdown flow, maximum allowable identified and unidentified primary coolant leak rates, maximum allowable primary-to-secondary leak rate, maximum reactor coolant pump seal controlled bleed-off flow rate, 100 percent removal of all iodine from the letdown stream by the purification ion exchanger, and minimum reactor coolant system mass.

Table 4.1-6 summarizes the concurrent iodine spike release rate in terms of escape rate coefficients that are to be modeled with the AST reactor core iodine inventory and an assumed 0.62 percent failed fuel. As an example, when the iodine spike release rate for the equilibrium case of 1.3E-08 sec 1 is modeled with the AST reactor core iodine inventory and 0.62 percent fuel failure, the resultant equilibrium primary coolant iodine activity concentration is 1.0 pCi/gram DE 1-131.

TABLE 4.1-6: CONCURRENT IODINE SPIKE ESCAPE RATE COEFFICIENTS Condition Iodine Iodine Escape Rate Escape Rate Coefficient Coefficient

[1/second] [1_hour]

Equilibrium (no spike) 1.3E-08 4.7E-05 Spiking Factor of 335 6.5E-06 2.4E-02 Spiking Factor of 500 4.4E-06 1.6E-02 Section 4.1.3 Radial Peaking Factor Consistent with the guidance of RG 1.183 Regulatory Position 3.1, to account for differences in power level across the core, Radial Peaking Factors (RPFs) are applied to the Section 4.1.1 Tables 4.1-3 and 4.1-4 average fuel rod isotope inventory in determining the activity inventory of the damaged fuel rods when only a portion of the core is damaged.

Per RG 1.183 Regulatory Position 3.1, the RPFs should be values from the facility's Core Operating Limits Report (COLR) or Technical Specifications.

SONGS Units 2 and 3 do not report RPFs in the facility's COLR or in the SONGS Technical Specifications. SONGS Units 2 and 3 calculate RPFs in unit and cycle specific reload physics analyses.

A review of the recent SONGS Units 2 and 3 Cycle 11 and 12 reload physics analyses identified RPFs with values no greater than 1.67 at 100 percent power.

For conservatism the non-LOCA AST dose calculations addressed in this AST license amendment request model an RPF of 1.75 for all damaged fuel rods. For the DBA LOCA, all fuel assemblies are damaged and the core average inventory (without peaking factor) is used.

Page 16of 110

Section 4.1.4 Fuel Damage in Non-LOCA Design Basis Accidents Per RG 1.183 Regulatory Position 3.6, the amount of fuel damage caused by non-LOCA design basis events should be analyzed to determine 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.

Consistent with the NRC approved SONGS Units 2 & 3 reload analysis methodology documented in Section 3.4.2.1 of SCE-9801 -P-A, "Reload Analysis Methodology for the San Onofre Nuclear Generating Station Units 2 and 3," fuel failure for the control element assembly (i.e., rod) ejection event is currently based on enthalpy deposition methodology. Per SCE-9801-P-A, fuel failure for the reactor coolant pump sheared shaft event is currently based on the Departure from Nucleate Boiling (DNB) statistical convolution methodology, and fuel failure for the remaining non-LOCA events that fail fuel are currently based on the DNB deterministic methodology.

Following approval of this license amendment request, in addition to the reactor coolant pump sheared shaft event, fuel failure estimates for UFSAR Chapter 15 non-LOCA events that assume a loss of flow (i.e., a loss of AC power) and that fail fuel (currently steam system piping failures and increased main steam flow with single failure) may be based on the DNB statistical convolution methodology.

The DNB statistical convolution technique is described in NRC approved Combustion Engineering document CENPD-1 83-A "C-E Methods for Loss of Flow Analysis" (Reference 18). The DNB statistical convolution technique estimates the amount of fuel failure by the probability density function with the DNB distribution. The DNB deterministic technique ignores the DNB distribution and uses a single value, the Departure from Nucleate Boiling Ratio (DNBR)

Specified Acceptable Fuel Design Limit (SAFDL), as the fuel failure criterion.

The DNB statistical convolution technique is widely used to determine fuel failure for events at Combustion Engineering designed reactors. The SONGS Units 2 &

3 current licensing basis uses the DNB statistical convolution technique for predicting fuel failure for the reactor coolant pump sheared shaft evaluation. In addition, the Palo Verde Nuclear Generating Station uses the DNB statistical convolution technique for the calculation of fuel failures for transients that result in fuel failure.

Page 17of 110

Section 4.2 OFFSITE DOSE MODEL Regulatory Guide 1.183 Regulatory Position 4.1 provides guidance to be used in determining the TEDE for persons located at or beyond the boundary of the exclusion area, including the outer boundary of the LPZ. This Section addresses the applicability of this guidance to the SONGS Units 2 and 3 AST dose analyses as it relates to the offsite dose exposure parameters.

The characteristics of the offsite dose exposure parameters as modeled in the AST dose analyses are summarized in Tables 4.2-1 and 4.2-2 for the EAB and LPZ dose receptors, respectively.

TABLE 4.2-1: EAB DOSE EXPOSURE PARAMETERS

[ EXCLUSION AREA BOUNDARY PARAMETER ll MODELED VALUE EAB dose acceptance criterion, Rem TEDE Varies by event EAB dose exposure duration, hours 2-hour window Per Federal Guidance Committed effective dose equivalent (CEDE) dose conversion factors Report (FGR)-l 1 Effective dose equivalent (EDE) dose conversion factors Per FGR-12 EAB breathing rate, event duration, m3 /second 3.5E-04 EAB atmospheric dispersion factor, event duration, seconds/M 3 2.72E-04 TABLE 4.2-2: LPZ DOSE EXPOSURE PARAMETERS l LOW POPULATION ZONE PARAMETER MODELED VALUE LPZ dose acceptance criterion, Rem TEDE Varies by event LPZ dose exposure duration, hours Event Duration Committed effective dose equivalent (CEDE) dose conversion factors Per FGR-1 1 Effective dose equivalent (EDE) dose conversion factors Per FGR-12 LPZ breathing rates, m3 /second 0 to 8 hours9.259259e-5 days <br />0.00222 hours <br />1.322751e-5 weeks <br />3.044e-6 months <br /> 3.5E-04 8 to 24 hours2.777778e-4 days <br />0.00667 hours <br />3.968254e-5 weeks <br />9.132e-6 months <br /> 1.8E-04 1 day to end of event 2.3E-04 LPZ atmospheric dispersion factors, seconds/M 3 0 to 8 hours9.259259e-5 days <br />0.00222 hours <br />1.322751e-5 weeks <br />3.044e-6 months <br /> 7.72E-06 8 to 24 hours2.777778e-4 days <br />0.00667 hours <br />3.968254e-5 weeks <br />9.132e-6 months <br /> 4.74E-06 1 to 4 days 3.67E-06 4 days to end of event 2.67E-06 Consistent with RG 1.183 Regulatory Position 4.1.1, the offsite dose calculations determine TEDE, which is the sum of the committed effective dose equivalent (CEDE) from inhalation and the deep dose equivalent (DDE) from external exposure. Consistent with RG 1.183 Regulatory Position 4.1.4, the EDE from external exposure is used in lieu of DDE in determining the contribution of external dose to the TEDE. The calculation of the CEDE and EDE components of the TEDE consider all radionuclides identified in Section 4.1 of this license Page 18of 110

amendment request, including progeny from the decay of parent radionuclides, that are significant with regard to dose consequences and the released radioactivity.

Consistent with RG 1.183 Regulatory Position 4.1.2, the AST analyses model CEDE dose conversion factors taken from the column headed Ueffective" in Table 2.1 of Federal Guidance Report (FGR) 11, "Limiting Values of Radionuclide Intake and Air Concentration and Dose Conversion Factors for Inhalation, Submersion, and Ingestion".

Consistent with RG 1.183 Regulatory Position 4.1.3, for the first 8 hours9.259259e-5 days <br />0.00222 hours <br />1.322751e-5 weeks <br />3.044e-6 months <br />, the breathing rate of persons at the outer boundary of the LPZ is assumed to be 3.5E-4 cubic meters per second. From 8 to 24 hours2.777778e-4 days <br />0.00667 hours <br />3.968254e-5 weeks <br />9.132e-6 months <br /> following the accident, the breathing rate is assumed to be 1.8E-4 cubic meters per second. After that and until the end of the accident, the rate is assumed to be 2.3E-4 cubic meters per second. The breathing rate for persons at the EAB is assumed to be 3.5E-4 cubic meters per second for the event duration.

Consistent with RG 1.183 Regulatory Position 4.1.4, the AST analyses model EDE dose conversion factors taken from the column headed "effective" in Table 111.1 of Federal Guidance Report (FGR) 12, "External Exposure to Radionuclides in Air, Water, and Soil" (Reference 12).

Consistent with RG 1.183 Regulatory Positions 4.1.5, 4.1.6 and 4.4, the radiological criteria for the EAB and for the outer boundary of the LPZ are in 10 CFR 50.67. These criteria are stated for evaluating reactor accidents of exceedingly low probability of occurrence and low risk of public exposure to radiation, e.g., a large-break LOCA. For events with a higher probability of occurrence, postulated EAB and LPZ doses should not exceed the criteria tabulated in RG 1.183 Table 6.

Consistent with RG 1.183 Regulatory Position 4.1.5, the maximum EAB TEDE for any two-hour period following the start of the radioactivity release is determined and used in determining compliance with the dose criteria. The Bechtel LocaDose code used in the AST dose analyses determines the maximum two-hour TEDE at the EAB by calculating the postulated dose for a series of small time increments and performing a 'sliding" sum over the increments for successive two-hour periods. The time increments appropriately reflect the progression of the accident to capture the peak dose interval between the start of the event and the end of radioactivity release.

The AST dose analyses for exposure to individuals at the EAB and LPZ consider immersion of the individual in the radioactive plume released from the facility.

Consistent with RG 1.183 Regulatory Position 5.3, the atmospheric dispersion values for the EAB and the LPZ that were approved by the NRC staff during initial facility licensing are used in performing the AST radiological analyses.

These atmospheric dispersion factors for the EAB and LPZ are the five percentile Page 19 of 110

values listed in SONGS Units 2 and 3 UFSAR Appendix 15B Table 15B-4.

Consistent with RG 1.183 Regulatory Position 4.1.7, no correction is made for depletion of the effluent plume by deposition on the ground.

Radioactive material contained in a plant structure is assumed to be a negligible radiation shine source to the offsite dose receptors relative to the dose associated with immersion in the radioactive plume (i.e., environmental cloud) released from the facility. To evaluate the conservatism present in this modeling when using an alternative source term, the post-LOCA reactor containment building shine doses at the EAB and at the outer boundary of the LPZ were compared to the post-LOCA offsite immersion/inhalation doses. As shown in Table 4.2-3, the EAB and LPZ doses due to containment shine are at least three orders of magnitude (a factor of 1,000) smaller than the EAB and LPZ doses due to immersion in the radioactive plume released from the containment.

TABLE 4.2-3: SIGNIFICANCE OF PLANT STRUCTURE SHINE DOSE AST POST-LOCA CONTAINMENT LEAKAGE RADIATION SOURCE TEDE DOSE (REM)

Maximum 2-hour EAB dose due to immersion and inhalation 3.547E+00 Maximum 2-hour EAB dose due to Containment Building shine 1.204E-03 Event duration LPZ dose due to immersion and inhalation 2.309E-01 Event duration LPZ dose due to Containment Building shine 1.810E-09 Page 20 of 110

Section 4.3 CONTROL ROOM DOSE MODEL SONGS Units 2 and 3 share a combined control room. RG 1.183 Regulatory Position 4.2 provides guidance to be used in determining the TEDE for persons located in the control room. Section 4.3.1 addresses the applicability of this guidance to the SONGS Units 2 and 3 AST dose analyses as it relates to the control room dose exposure parameters. Section 4.3.2 addresses the applicability of the RG 1.183 guidance as it relates to the control room response to radiation sources that may cause exposure to control room personnel.

Section 4.3.1 Control Room Dose Exposure Parameters The characteristics of the control room dose exposure parameters as modeled in the AST dose analyses are summarized in Table 4.3-1.

TABLE 4.3-1: CONTROL ROOM DOSE EXPOSURE PARAMETERS CONTROL ROOM PARAMETER l MODELED VALUE CR dose acceptance criterion, Rem TEDE 5 Committed effective dose equivalent (CEDE) dose conversion factors Per FGR-1 1 Effective dose equivalent (EDE) dose conversion factors Per FGR-12 CR occupancy factors, percent of time present in CR 0 to 1 day 100 1 to 4 days 60 4 to 30 days 40 CR breathing rate, event duration, m3 /second 3.5E-04 Consistent with RG 1.183 Regulatory Position 4.2.2, the radioactive material releases and radiation levels modeled in the control room dose analyses are determined using the same source term, transport, and release assumptions used for determining the EAB and LPZ TEDE values. These parameters are detailed in the later sections of this license amendment request that describe the various accident scenarios. These parameters do not result in non-conservative results for the control room.

Consistent with RG 1.183 Regulatory Position 4.2.7, the control room doses are calculated using the dose conversion factors identified in RG 1.183 Regulatory Position 4.1 for use in offsite dose analyses. The control room dose calculations determine the TEDE, which is the sum of the CEDE from inhalation and the DDE from external exposure. Consistent with RG 1.183 Regulatory Position 4.1.4, the EDE from external exposure is used in lieu of DDE in determining the contribution of external dose to the TEDE. The calculation of the CEDE and EDE components of the TEDE consider all radionuclides identified in Section 4.1 of this license amendment request, including progeny from the decay of parent radionuclides, that are significant with regard to dose consequences and the released radioactivity.

Page 21 of 110

Consistent with RG 1.183 Regulatory Position 4.1.2, the AST analyses model CEDE dose conversion factors taken from the column headed "effective" in Table 2.1 of Federal Guidance Report 11, "Limiting Values of Radionuclide Intake and Air Concentration and Dose Conversion Factors for Inhalation, Submersion, and Ingestion".

Consistent with RG 1.183 Regulatory Position 4.1.4, the AST analyses model EDE dose conversion factors taken from the column headed "effective" in Table 111.1 of Federal Guidance Report 12, "External Exposure to Radionuclides in Air, Water, and Soil."

Consistent with RG 1.183 Regulatory Position 4.2.7, the DDE from photons (i.e.,

the EDE) is corrected for the difference between finite cloud geometry in the control room and the semi-infinite cloud assumption used in calculating the dose conversion factors. The Bechtel LocaDose code used in this analysis employs the following RG 1.183 Equation 1 to correct the semi-infinite cloud dose, DDE.,

to a finite cloud dose, DDEfinite, where the control room is modeled as a hemisphere that has a volume, V, in cubic feet, equivalent to that of the control room:

DDEfinite = (DDE.0 x V 338 ) / 1173 Consistent with RG 1.183 Regulatory Position 4.4, as an AST dose analysis acceptance criterion the postulated control room dose is evaluated to ensure that that it does not exceed the 5 Rem TEDE criterion established in 10 CFR 50.67.

Consistent with RG 1.183 Regulatory Position 4.2.6, the control room dose receptor for the AST analyses is the hypothetical maximum exposed individual who is present in the control room for 100% of the time during the first 24 hours2.777778e-4 days <br />0.00667 hours <br />3.968254e-5 weeks <br />9.132e-6 months <br /> after an event, 60% of the time between 1 and 4 days, and 40% of the time from 4 days to 30 days. These occupancy factors are not modeled in the ARCON96 atmospheric dispersion factors discussed in Section 4.4 of this license amendment request.

Consistent with RG 1.183 Regulatory Position 4.2.6, for the duration of any event, the breathing rate of the hypothetical maximum exposed individual who is present in the control room is assumed to be 3.5E-04 cubic meters per second.

Consistent with RG 1.183 Regulatory Position 4.2.5, credit is not taken for the control room personnel use of personal protective equipment (e.g., protective beta radiation resistant clothing, eye protection, or self-contained breathing apparatus [SCBA]) or prophylactic drugs (i.e., potassium iodide [KI] pills).

Page 22 of 110

Section 4.3.2 Control Room Response to Radiation Sources Consistent with RG 1.183 Regulatory Position 4.2.3, the models used to transport radioactive material into and through the control room, and the shielding models used to determine radiation dose rates from external sources, are structured to provide suitably conservative estimates of the exposure to control room personnel. The control room response to these radiation sources is discussed in this section.

Consistent with RG 1.183 Regulatory Position 4.2.1, the AST dose analyses consider the following sources of radiation that may cause exposure to control room personnel:

• Contamination of the control room atmosphere by the intake or infiltration of radioactive material contained in the radioactive plume released from the facility,

• Contamination of the control room atmosphere by the intake or infiltration of airborne radioactive material from areas and structures adjacent to the control room envelope,

• Radiation shine from the external radioactive plume released from the facility,

• Radiation shine from radioactive material in the reactor containment building,

• Radiation shine from radioactive material in systems and components inside or external to the control room envelope.

The characteristics of the control room as modeled in the AST dose analyses are summarized in Table 4.3-2.

Page 23 of 110

TABLE 4.3-2: CONTROL ROOM MODEL PARAMETERS CONTROL ROOM PARAMETER l NOMINAL MODELED CR net free volume, cubic feet 266,920 266,920 CR unfiltered outside air inleakage, event duration ingress and egress, cfm not applicable 10 Boundary and system inleakage, cfm 132 990 total unfiltered inleakage, cfm not applicable 1,000 CR Normal mode of operation unfiltered outside air makeup, cfm 5820 6402 filtered outside air makeup, cfm 0 0 filtered CR air recirculation, cfm 0 0 Control Room Emergency Air Cleanup System (CREACUS) initiation time Safety Injection Actuation Signal (SIAS) induced <10 0 Control Room Isolation Signal (CRIS), seconds High CR Heating, Ventilation, and Air-Conditioning 120.0 180 (HVAC) intake radiation induced CRIS, seconds CREACUS Emergency mode of operation, one train operation filtered outside air makeup, cfm 2,050 2,200 filtered CR air recirculation, cfm 33,505 29,934 CREACUS Emergency mode of operation, two train operation filtered outside air makeup, cfm 4,100 4,400 filtered CR air recirculation, cfm 67,010 59,869 CREACUS Emergency mode of operation, Emergency Ventilation Supply (EVS) filter efficiencies Elemental iodine, percent removal > 90 0 organic iodide, percent removal >90 0 Particulate iodine and aerosols, percent removal > 99.95 0 CREACUS Emergency mode of operation, Emergency Air Conditioner (EAC) filter efficiencies elemental iodine, percent removal > 99 95 organic iodide, percent removal > 99 95 Particulate iodine and aerosols, percent removal > 99.95 99 Section 4.3.2.1 Control Room Intake and Infiltration of Contaminated Air Only the west side of the control room envelope is exposed to the radioactive plumes released from the facility. The 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 only become contaminated with air introduced via intake or infiltration of radioactive material contained in the radioactive plumes released from the facility. Consequently, the resultant activity concentrations in the adjacent areas and structures will be less contaminated than any radioactive plume. For this reason, the AST dose analyses conservatively assume that all Page 24 of 110

intake and infiltration (i.e., inleakage) into the CRE is from the radioactive plumes released from the facility as they pass west of the control room envelope.

The control room Normal Mode Heating, Ventilation, and Air Conditioning (HVAC) outside air intake is located near the northwest corner of the control room envelope, and the control room emergency air cleanup system (CREACUS) Emergency Mode HVAC outside air intakes are located near the northwest and southwest corners of the control room envelope. Per the ARCON96 atmospheric dispersion analysis detailed in Section 4.4, the maximum atmospheric dispersion factor for any activity release location to any of these three outside air intakes is modeled in the evaluation of contaminated air intake and infiltration (i.e., inleakage).

Section 4.3.2.1.1 Control Room Isolation Signal Consistent with RG 1.183 Regulatory Position 4.2.4, the AST analyses credit engineered safety features (ESF) that mitigate airborne radioactive material within the control room, such as control room isolation actuated by ESF signals and radiation monitors.

The control room Normal Mode HVAC systems can be shifted to CREACUS Emergency Mode, which is an operational mode in which the control room is isolated and pressurized to protect operational personnel from radiation exposure. The CREACUS Emergency mode of operation can be actuated either automatically following a Control Room Isolation Signal (CRIS) or manually. The CRIS may be generated automatically by a Safety Injection Actuation Signal (SIAS) or by the detection of high radioactivity concentrations in the control room outside air inflow.

A SIAS-induced CRIS is credited in the evaluation of the LOCA. A SIAS-induced CRIS is capable of initiating CREACUS Emergency mode of operation within 10 seconds. The SIAS is generated in response to high containment pressure within seconds of the onset of the LOCA event. Since the gap release activity is not released into the containment until 30 seconds after the onset of the LOCA event, and since a SIAS induced CRIS is capable of initiating CREACUS Emergency mode of operation in less than 30 seconds, the AST LOCA model credits CREACUS Emergency mode of operation initiation at time zero (i.e., prior to the arrival of any contaminated air reaching the control room outside air intakes) due to a SlAS-induced CRIS.

Per LCS 3.3.100, Table 3.3.100-2, a high radiation induced CRIS is to be generated and the normal HVAC outside air dampers are to be closed, within 120.0 seconds. The non-LOCA and FHA dose analyses conservatively assume that a high-radiation-induced CRIS initiates the CREACUS Emergency mode of operation at 180 seconds.

Page 25 of 110

Section 4.3.2.1.2 Control Room Unfiltered Inleakage The AST dose analyses model the introduction of an assumed 1,000 cfm of unfiltered outside air into the CRE beginning at time zero and continuing for the event duration. This inleakage rate includes 10 cfm as a reasonable estimate for ingress and egress, and an assumed 990 cfm for inleakage via other paths. The 10 cfm estimate for ingress and egress inleakage is consistent with guidance provided in RG 1.197 (Reference 13) Regulatory Position 2.5.

The CRE inleakage testing to verify actual inleakage was conducted from May 18, 2004 to May 25, 2004. As described in SCE's letter to the NRC dated September 17, 2004, CRE inleakage testing has shown the actual inleakage via other paths, including uncertainty, is less than 990 cfm.

Section 4.3.2.1.3 Control Room HVAC Flow Rates and Filtration During the control room Normal Mode of HVAC operation, there is no filtered outside air makeup flow nor is there filtered control room air recirculation flow.

During the control room Normal Mode of HVAC operation, the AST dose analyses model an outside air makeup flow rate that is conservatively greater than the nominal outside air makeup flow rate. The outside air introduced into the control room during the normal mode of operation is unfiltered. The total unfiltered inleakage rate of 1,000 cfm is added to this Normal Mode of operation unfiltered outside air makeup flow rate.

Consistent with RG 1.183 Regulatory Position 4.2.4, the AST dose analyses credit ESFs that mitigate airborne radioactive material within the control room.

Such features include control room pressurization, and intake and recirculation filtration.

The CREACUS Emergency mode of operation is facilitated by two 100%

redundant subsystems. As shown in Figure 4.3-1, for each CREACUS Emergency mode of operation flow path, the control room outside makeup air passes through intake filters of an emergency ventilation supply (EVS) unit and then through recirculation filters of an emergency air conditioner (EAC) unit prior to being discharged into the control room envelope.

FIGURE 4.3-1: CREACUS EMERGENCY MODE OF OPERATION FLOW PATH {SINGLE TRAIN)

OUTSIDE OUTID l EVS l dFILTERS

• EAC FILTERS CONTROL ROOM Page 26 of 110

During the CREACUS Emergency mode of operation, the AST dose analyses model an outside air makeup flow rate that is conservatively greater than the nominal outside air makeup flow rate. The outside air introduced into the control room is filtered. Consistent with the current SONGS Units 2 and 3 licensing basis, filtration credit is not taken for outside air iodine and particulate removal by the EVS unit filters. Filtration credit is only taken for outside air makeup iodine and particulate removal by the EAC unit filters. The total unfiltered inleakage rate of 1,000 cfm is added to this CREACUS Emergency mode of operation flow rate.

During the CREACUS Emergency mode of operation, the AST dose analyses model a control room recirculation flow rate that is conservatively smaller than the nominal recirculation flow rate. This flow rate is calculated by subtracting the maximum outside air inflow rate entering the EAC unit from the minimum total flow rate (i.e., outside air inflow rate plus control room recirculation air flow rate) through the EAC unit.

Consistent with the current SONGS Units 2 and 3 licensing basis, filtration credit is taken for iodine and particulate removal by the EAC filters. The EAC charcoal filters are credited with the removal of 95 percent of the elemental iodine and organic iodide in the HVAC air flow. The EAC HEPA filters are credited with the removal of 99 percent of the particulate iodine and other aerosols in the HVAC air flow.

In the AST dose analyses the potential exists for one or two trains of HVAC to function during the CREACUS Emergency mode of operation. The SONGS Units 2 and 3 current licensing basis models Operator action at eight hours to secure one of the two trains of CREACUS that are modeled as being in operation at the onset of an event. To evaluate the conservatism present in this modeling when using an AST source term, the post-LOCA containment leakage path was evaluated for three scenarios: (1) one CREACUS train operating throughout the event, (2) two CREACUS trains operating throughout the event, and (3)two CREACUS trains operating for the first eight hours, and a single CREACUS train operating for the remainder of the event. As shown in Table 4.3-3, the first scenario with its operation of a single CREACUS train throughout the event results in the largest control room dose. For this reason, the AST dose analyses conservatively assume the failure of one CREACUS train and model single CREACUS train operation throughout an event.

TABLE 4.3-3: CONTROL ROOM CREACUS MODEL OPTIONS

[ LOCA CONTROL ROOM CREACUS MODEL l CONTAINMENT LEAKAGE l TEDEDOSE (REM)

Single CREACUS train operation throughout the event 9.112E-01 Two CREACUS train operation throughout the event 7.303E-01 Two CREACUS train operation for 8 hours9.259259e-5 days <br />0.00222 hours <br />1.322751e-5 weeks <br />3.044e-6 months <br />, 7.650E-01 one CREACUS train operation thereafter ____ ____

Page 27 of 110

Section 4.3.2.2 Environmental Cloud Gamma Radiation Shine Model Consistent with RG 1.183 Regulatory Position 4.2.1, the AST dose analyses consider exposure to control room personnel due to radiation shine from the external radioactive plume (i.e., environmental or outside cloud shine). The following discussion elaborates on the modeling described in SONGS Units 2 and 3 UFSAR Appendix 15.106.

Activity releases to the environment from sources such as post-LOCA containment building leakage will result in the formation of a radioactive cloud.

Radioactivity concentrations in the radioactive cloud surrounding the control room are the product of the building leak rate and the control room atmospheric dispersion factor.

For conservatism it is assumed that this cloud surrounds the control room, entering adjacent areas that are not part of the control room envelope. Gamma radiation from this cloud can penetrate the control room ceiling and walls resulting in a whole body gamma dose to control room personnel. The cloud is modeled as a cylinder with a 4000 foot radius and a 4000 foot height. The radius and height values ensure that dose contributions from the outer portions of the cloud are considered. The radioactivity present in the outside cloud is assumed to be uniformly distributed in the cylindrical source.

The environmental cloud radiation shine dose is the maximum dose calculated at one of several dose receptors modeled within the control room board area.

Shielding modeled between the control room dose receptors and the outside cloud include the concrete containment structures that lie within the 4000 foot radius cloud, the concrete safety equipment building wall adjacent to the control building, the control room concrete walls, floor and ceiling, the auxiliary/radwaste building outer concrete walls, floors and roof, several of the internal control room fire partition walls, and the air spaces between these walls, floor and ceilings.

Section 4.3.2.3 Control Room Filter Gamma Radiation Shine Model Consistent with RG 1.183 Regulatory Position 4.2.1, the AST dose analyses consider exposure to control room personnel due to radiation shine from radioactive material in the CREACUS filters inside the CRE (i.e., control room filter shine). The following discussion elaborates on the modeling described in SONGS Units 2 and 3 UFSAR Appendix 15B, Section 15B.5.

The activity released to the environment during an event and dispersed to the CREACUS outside air intake is assumed to accumulate onto the CREACUS filters for the duration of an activity release. For those events in which the release terminates prior to 30 days (e.g., fuel handling accident), the activity Page 28 of 110

accumulated on the charcoal filter is allowed to decay for the remainder of the 30-day event duration to facilitate determination of a 30-day control room dose.

As previously shown in Figure 4.3-1, for each CREACUS Emergency mode of operation flow path, the control room outside makeup air passes through intake filters of an EVS intake unit and then through recirculation filters of an EAC unit prior to being discharged into the CRE. Per Section 4.3.2.1, the AST dose analyses conservatively assume the failure of one CREACUS train and model single CREACUS train operation throughout an event.

In determining the filter shine dose, the charcoal and HEPA filters of the EVS intake units are assumed to be 100 percent efficient at removing iodine and particulates from the incoming air, thereby maximizing the shine dose from the EVS intake unit filters. In reality, iodine and particulates that are not trapped on the intake filters (such as the activity present in the 1,000 cfm of unfiltered air inleakage) will eventually be trapped on the charcoal and HEPA filters of the downstream EAC recirculation unit. The EAC recirculation units are located in the vicinity of the EVS intake units, with an EAC recirculation unit filter direct shine to control room pathway geometry that is similar to that of the EVS intake unit filters. Consequently, the results of the EVS intake unit filter shine dose calculation address the EAC recirculation unit filter shine.

To address the potential shine from unfiltered inleakage that is eventually trapped on the EAC recirculation unit filters, the filter shine model includes an additional 1,000 cfm of contaminated outside air inflow to the CREACUS air intake flow rate. The CREACUS filter shine dose is the maximum dose calculated at one of several dose receptors modeled within the control room board area.

Shielding modeled between the control room dose receptors and the CREACUS filter units include the control room fire partition walls and the air spaces between these walls.

Section 4.3.2.4 Containment Building Gamma Radiation Shine Model Consistent with RG 1.183 Regulatory Position 4.2.1, the AST dose analyses consider exposure to control room personnel due to radiation shine from radioactive material in the reactor containment building (i.e., direct containment shine). The following discussion elaborates on the modeling described in SONGS Units 2 and 3 UFSAR Appendix 15.10B.

The containment is modeled by an equivalent volume cylindrical source having a diameter of 150 feet and a height of 129.25 feet. The radioactivity in the containment is modeled as being uniformly distributed in the cylindrical source.

Page 29 of 110

The containment radiation shine dose is the maximum dose calculated at one of several dose receptors modeled within the control room board area.

Shielding modeled between the control room dose receptors and the containment air includes the 1/4-inch steel containment liner, the 4-foot concrete containment wall and the 3-foot 9-inch concrete containment dome. No credit is taken for shielding afforded by the internal containment concrete structure. The penetration building lies between the containment and the control building.

Modeled shielding includes the 2-foot penetration building concrete wall and the adjacent 2-foot 6-inch control building concrete wall, a 2-inch fire partition wall that separates the control room from the cable riser gallery adjacent to the penetration building, and the air spaces between these walls.

Section 4.3.2.5 Post-LOCA Piping Gamma Radiation Shine Model Consistent with RG 1.183 Regulatory Position 4.2.1, the AST dose analyses consider exposure to control room personnel due to radiation shine from radioactive material in recirculation loop piping outside the CRE (i.e., piping shine). The following discussion elaborates on the modeling described in SONGS Units 2 and 3 UFSAR Appendix 15B, Section 15B.5.

This piping is modeled as a series of finite length shielded cylinders filled with post-LOCA containment sump recirculation liquid radiation source. The piping shine model considers those pipes in the Auxiliary Building Penetration Area that are outside the containment penetration area shield walls at plant elevation 30-foot (i.e., at the same plant elevation as the control room). The dose contributions from other pipes that are either behind the shield walls or below the 30-foot concrete floor are much less than the dose contributions from the modeled pipes.

The piping shine dose is the maximum dose calculated at one of several dose receptors modeled within the control room board area.

Shielding modeled between the control room dose receptors and the piping includes Auxiliary Building walls and floor slabs, the concrete wall separating the control room from the cable riser gallery, the steel door in the concrete wall separating the cable riser gallery from the Penetration Area, and the air spaces between these walls, floor and ceilings.

Page 30 of 110

Section 4.4 ARCON96 ATMOSPHERIC DISPERSION ANALYSIS UFSAR Section 2.3.4.2 and UFSAR Appendix 15B.5 discuss the Current Licensing Basis (CLB) methodology used in evaluating atmospheric dispersion between the post-accident containment building release point and the control room outside air ventilation intakes. The CLB applies the atmospheric dispersion factors for the release from the containment to the control room HVAC intakes for all potential release points. The CLB methodology for evaluating this atmospheric dispersion utilizes the Murphy-Campe diffuse source point receptor model.

Per RG 1.183 Regulatory Position 5.3, atmospheric dispersion values for the control room that were approved by the staff during initial facility licensing or in subsequent licensing proceedings (i.e., the CLB) may be used in performing the AST radiological analyses.

The limiting condition for Control Room Habitability (CRH) is the event configuration that results in the maximum consequences to the control room operators. Per RG 1.196 (Reference 14) Regulatory Position 2.3.2, determining the limiting condition for CRH requires consideration of the location of the activity release points for the various accidents relative to the control room intakes.

Although RG 1.183 allows continued use of the CLB atmospheric dispersion values, to comply with the guidance of RG 1.196 it is necessary to calculate atmospheric dispersion values between the various post-accident release points and the control room outside air ventilation intakes. The new atmospheric dispersion analysis uses the ARCON96 computer program and guidance provided in RG 1.194. This section summarizes the ARCON96 analysis.

Section 4.4.1 ARCON96 Background Information The ARCON96 computer program was developed for the U.S. Nuclear Regulatory Commission Office of Nuclear Reactor Regulation by Pacific Northwest National Laboratory (PNNL) for potential use in control room habitability assessments. This code is documented in NUREG/CR-6331 Revision 1 (Reference 15), which includes a users guide, a programmers guide, and a description of the technical basis for the code. The ARCON96 code uses hourly meteorological data and recently developed methods for estimating dispersion in the vicinity of buildings to calculate relative concentrations at control room intakes that would be exceeded no more than 5 percent of the time. RG 1.194 provides guidance on the use of the ARCON96 computer program for determining atmospheric relative concentrations to be used in design basis evaluations of control room radiological habitability.

Bechtel Power Corporation (BPC) originated the ARCON96 calculations under contract to Southern California Edison. The ARCON96 code was obtained by Page 31 of 110

BPC, and is maintained as Bechtel Standard Computer Program (SCP) number EV138. Bechtel SCPs are those computer programs that have been developed, validated, documented and controlled in accordance with Bechtel Engineering Department procedures so that they may be used without detailed description and validation in a calculation package.

Section 4.4.2 Summary of Evaluated Release Point to Intake Combinations The ARCON96 computer program with the guidance of RG 1.194 has been used to determine the atmospheric dispersion factors for a combination of nine activity release point locations and three control room ventilation outside air intake locations. The release point locations are:

• Main plant vent

• Containment shell diffusion

• Containment equipment hatch

• Main Steam Safety Valves (MSSV)

• Atmospheric Dump Valves (ADV)

• Steam Line Break Outside Containment (SLB-OC)

• Auxiliary Feedwater (AFW) Turbine steam discharge

• Refueling Water Storage Tank (RWST) Vent

• Fuel Handling Building The three control room ventilation outside air intake locations are:

• Control room normal air intake

• Control room Unit 2 emergency air intake

• Control room Unit 3 emergency air intake Each of the 27 release-to-intake combinations has been investigated. Release locations from both of SONGS Units 2 and 3 are considered.

Section 4.4.3 Meteorological Data Input The ARCON96 atmospheric dispersion analysis uses actual site hourly meteorological data spanning ten full years from 1993 through 2002. Full year meteorology is used to eliminate bias due to seasonal fluctuations. RG 1.194 Regulatory Position 3.1 states that 5 years of hourly observations are considered to be representative of long-term trends at most sites. The use of ten years of meteorological data satisfies this recommendation, while enhancing the statistical Page 32 of 110

basis for the calculated control room atmospheric dispersion factors due to the expanded meteorological data set.

The input meteorological data identify invalid data by coding such data as either "999" or "9999". In each year, more than 99 percent of the lower level wind speed data are valid. Overall, about 99.8 percent of the lower level wind speed data are valid. Except for year 1994, more than 95 percent of each years upper level wind speed data are valid. Overall, about 96.5 percent of the upper level wind speed data are valid. Therefore, the meteorological input is representative.

The meteorological towers lower wind instrument is at elevation 10 meters. The meteorological towers upper wind instrument is at elevation 40 meters.

The meteorological data was converted to the ARCON96 format presented in NUREG/CR-6331 Section 4.4.2 and RG 1.194 Appendix A, Table A-1.

Consistent with RG 1.194 Regulatory Position 3.1, wind direction is expressed as the direction from which the wind is blowing (i.e., the upwind direction from the center of the site) referenced from true north. A north wind (wind from the north) is entered as 360 degrees, and a south wind is entered as 180 degrees.

Consistent with RG 1.194 Regulatory Position 3.1, atmospheric stability is entered as a number from 1 through 7. A stability class of 1 represents extremely unstable conditions, and a stability class of 7 represents extremely stable conditions. Atmospheric stability classes are determined from the AT given in the meteorological data.

Section 4.4.4 Non-Meteorological Data Input RG 1.194 Appendix A Table A-2 discusses input parameters for ARCON96. Per Table 4.4-1, the ARCON96 analysis complies with the regulatory guidance presented in Table A-2.

The following subsections summarize the ARCON96 non-meteorological data input for each of the release point and receptor location combinations.

Table 4.4-1: ARCON96 Input Parameters for Design Basis Assessments Parameter Acceptable Input Comments Lower Use the actual instrumentation height Used actual measurement height, which Measurement when known. Otherwise, assume 10 is 10 meters above bluff grade. The bluff Height, meters meters. grade is above the plant grade.

Upper Use the actual instrumentation height Used actual measurement height of 40 Measurement hen known. Otherwise, use the height meters above the bluff grade.

Height, meters of the containment or the stack height, as appropriate. Ifwind speed measurements are available at more than two elevations, the instrumentation at the height closest to the release height l_ should be used.

Page 33 of 110

Parameter I Acceptable Input Comments Wind Speed Units Use the wind speed units that correspond The raw meteorological data expresses to the units of the wind speeds in the wind speeds in miles per hour. However, meteorological data file. these data are pre-processed to convert the wind speeds to meters per second in the resulting meteorological input files.

The ARCON96 input files (*.RSF) are set for wind speeds in units of meters per second. Thus, the units used for wind speeds in the analysis are applied consistently.

Release Height, Use the actual release heights whenever As clarified below, actual release heights meters available. Plume rise from buoyancy and above plant grade are used.

mechanical jet effects may be considered in establishing the release height if the For the containment diffuse area source, analyst can demonstrate with reasonable the release is assumed to be from the assurance that the vertical velocity of the containment mid-height of 80.5 feet release will be maintained during the above grade (i.e., Elevation 110.5'). This course of the accident. elevation allows for unimpeded flow above the Auxiliary Building roof.

If actual release height is not available, Because the control room intakes are on set release height equal to intake height. the Auxiliary Building wall opposite the Unit 2 and Unit 3 containments, the release cloud can only flow to the intakes by first passing over the Auxiliary Building roof.

This assumption is consistent with the NRC recommendation to set the release height for a diffuse area source at the vertical center of the projected plane of the above-grade cross-sectional area perpendicular to the line of sight from the building center to the control room intake (Regulatory Guide 1.194, section 3.2.4.5).

(There is also a pathway between the containment and the intakes via grated openings into the Turbine Building and then into the corrugated metal-sided passageway west of the Auxiliary Building. This pathway is longer and more tortuous than the pathway over the Auxiliary Building roof; therefore, it is not used.)

Page 34 of 110

Parameter I Acceptable Input l Comments Steam from a steam line break outside containment (SLB-OC) is assumed to be released via the blowout panels mounted on the roof of the respective Main Steam Isolation Valve (MSIV)/Main Feedwater Isolation Valve (MFIV) enclosure directly above the main steam line.

The x/Qs for an MSSV release credit plume rise due to jet effects in

+

accordance with section 6.0 of RG 1.194.

Buildin9 Area, Use the actual building vertical cross- The cross-sectional area of the meters sectional area perpendicular to the wind containment is used for all release points direction. Use default of 2000 m2 if the except for the Fuel Handling Building area is not readily available. Do not enter (FHB). For the FHB, the FHB east zero. Use 0.01 m2 if a zero entry is cross-section area and one half of the desired. containment cross-section area is used.

Only one half of the containment is Note: This building area is for the conservatively considered since it is building(s) that has the largest impact on partially offset from the release to intake the building wake within the wind direction axis. All other intervening buildings, such window. This is usually, but need not as the auxiliary building, are always be, the reactor containment. With conservatively ignored.

regard to the diffuse area source option, the building area entered here may be different from that used to establish the diffuse source.

4-Vertical Velocity, Note: the vent release model should not The vent release model is not used for meters/seconds be used for DBA accident calculations. DBA accident calculations.

For stack release calculations only, use For all vent stack releases, the vertical the actual vertical velocity if the licensee velocity is set to zero.

can demonstrate with reasonable assurance that the value will be maintained during the course of the accident (e.g., addressed by technical specifications), otherwise, enter zero. If the vertical velocity is set to zero, ARCON96 will reduce the stack height by 6 times the stack radius for all wind speeds. If this reduction is not desired, the stack radius should also be set to zero.

Page 35 of 110

Parameter Acceptable Input I Comments Stack Flow, Use actual flow if it can be demonstrated Stack flow is set to zero in all cases.

meters3 /s with reasonable assurance that the value will be maintained during the course of the accident (e.g., addressed by technical specifications). Otherwise, enter zero.

The flow is used in both elevated and ground-level release modes to establish a maximum X1Q value. This value is significant only if the flow is large and the distance from the release point to the receptor is small.

Stack Radius, Use the actual stack internal radius when Stack radius is set to zero in all cases.

meters both the stack radius and vertical velocity are available. If the stack flow is zero, the radius should be set to zero.

Distance to Use the actual straight line horizontal The actual straight line horizontal Receptor, meters distance between the release point and distance between the release point and he control room intake. the control room intake is used in all cases other than the Containment For ground-level releases, it may be Equipment Hatch release.

appropriate to consider flow around an intervening building if the building is Except for the Containment Equipment sufficiently tall that it is unrealistic to Hatch release, flow around intervening expect flow from the release point to go buildings is not considered. The over the building. Equipment Hatches are on the opposite side of the Containment structure from Note: If the distance to receptor is less the Control Room Intakes and are located than about 10 meters, ARCON96 should at ground level. The top of the not be used to assess relative Containment is 161 feet above plant concentrations. grade and the top of equipment hatch is 17.5 feet above plant grade; therefore, it is unrealistic to expect flow from the Equipment Hatch to go over the Containment building. The Equipment Hatch to receptor distances are measured as the shortest path around the Containment ("taut string length"), as allowed by section 3.4 of RG 1.194.

No source-receptor distance is less than 10 meters.

Intake Height, Use the actual intake height. If the intake The actual heights at the centerline of the meters height is not available for ground level control room intakes are used.

releases, assume the intake height is equal to the release height. For elevated releases, assume the height of the tallest site building.

Elevation Use zero unless it is known that the The release and receptor heights are Difference, release heights are reported relative to reported with respect to the same grade meters different grades or reference datum. datum.

Page 36 of 110

Parameter l Acceptable Input I Comments Direction to Use the direction FROM the intake back SONGS' met. data is given relative to true Source, degrees TO the release point. (Wind directions north. SONGS' site arrangement are reported as the direction from which drawings do have a "Plant North' the wind is blowing. Thus, if the direction designation that is 57 degrees west of from the intake to the release point is "true north;" consequently, directions north, a north wind will carry the plume entered into the ARCON96 code are from the release point to the intake.) corrected to model true north as the point of reference.

Note: some facilities have a "plant north" shown on site arrangement drawings that For the scenario of an equipment hatch is different from "true north." The direction release, the x1Q is calculated assuming entered must have the same point of flow both around and over (through) the reference as the wind directions reported containment building, and the higher of in the meteorological data. the X/Q values is used.

For ground level releases, if the plume is assumed to flow around a building rather than over it, the direction may need to be modified to account for the redirected flow. In this case, the x/Q should be calculated assuming flow around and flow over (through) the building and the higher of the two X/O s should be used.

Surface Use a value of 0.2 in lieu of the default Used value of 0.2. SONGS is a seaside Roughness value of 0.1 for most sites. (Reasonable site with low surface vegetation.

Length, meters values range from 0.1 for sites with low surface vegetation to 0.5 for forest covered sites.)

Wind Direction Use the default window of 90 degrees (45 Used 90 degrees.

Window, degrees degrees on either side of line of sight from the source to the receptor).

Code Default Minimum Wind Use the default wind speed of 0.5 m/s Used the default wind speed of 0.5 m/s.

Speed, (regardless of the wind speed units The minimum SONGS site meteorological meters/second entered earlier), unless there is some tower wind speed reported is 0.3 mph, or indication that the anemometer threshold 0.13 m/s. Thus, the anemometer Code Default is greater than 0.6 m/s. threshold is less than 0.6 m/s.

Averaging Sector Although the default value is 4, a value of Used 4.3.

Width Constant 4.3 is preferred. (A future revision to ARCON96 will change the default to 4.3)

Code Default Initial Diffusion These values will normally be set to zero. For containment releases, a diffuse Coefficients, If the diffuse source option is being used, source is modeled in accordance with meters see Regulatory Position 2.2.4. Regulatory Guide 1.194, section 3.2.4.4.

For the steam line break outside containment, the releases from the MSIV/MFIV enclosure are modeled as a area source in accordance with Regulatory Guide 1.194, section 3.2.4.7.

Page 37 of 110

I Parameter Acceptable Input I Comments Hours in Use the default values. Used the default values.

Averages Code Default Minimum Number Use the default values. Used the default values.

of Hours Code Default Section 4.4.4.1 Control Room HVAC Intakes Three control room HVAC intake locations are modeled in the ARCON96 analysis:

• Control room normal air intake

• Control room Unit 2 emergency air intake

• Control room Unit 3 emergency air intake The center of the control room normal air intake is at plant elevation 35.50 feet (10.82 meters). The center of each control room emergency air intake is at plant elevation 43.00 feet (13.11 meters).

Section 4.4.4.2 Main Plant Vent Release Atmospheric dispersion between the main plant vent and the control room HVAC intakes is modeled as a point source using the ARCON96 ground level release option.

The plant vent release height is at plant elevation 206 feet (62.79 meters).

For the plant vent release, only the cross-sectional area of the containment is used to determine the building wake area. All other intervening buildings, such as the auxiliary building, are conservatively ignored. The building wake area of 2123.33 square meters is the projected area of the containment cylindrical lower portion and the containment upper dome portion.

Table 4.4-2 presents the separation distances and wind directions that characterize the releases from the two plant vent activity release point locations to the three control room HVAC intake locations.

Page 38 of 110

TABLE 4.4-2: PLANT VENT TO CONTROL ROOM MODELING Release Point Control Room Receptor Separation Distance Wind Direction I _I_(meters) (degrees, North = 0)

U2 Plant Vent Normal Air Intake 62.83 348 U2 Plant Vent U2 emergency air intake 60.18 351 U2 Plant Vent U3 emergency air intake 101.6 329 U3 Plant Vent Normal Air Intake 98.15 96 U3 Plant Vent U2 emergency air intake 101.6 97 U3 Plant Vent U3 emergency air intake 60.18 75 The results of the ARCON96 analysis show that the Unit 2 plant vent to Unit 2 emergency air intake release path has the more conservative atmospheric dispersion (i.e., the maximum atmospheric dispersion factor). The resultant 95th percentile control room atmospheric dispersion factors for this release path are presented in Section 4.4.5.

Section 4.4.4.3 Containment Shell Release Atmospheric dispersion between the containment shell (surface) and the control room HVAC intakes is modeled as an area (diffuse) source using the ARCON96 ground level release option.

Consistent with RG 1.194 Sections 3.2.4.4 and 3.2.4.5, the height and width of the area source (i.e., the containment shell surface) are taken as the maximum vertical and horizontal dimensions of the above-grade building cross-sectional area perpendicular to the line of sight from the building center to the control room intake. The initial horizontal diffusion coefficient (ayo) is determined to be 8.06 meters, based on the 158.66 foot containment diameter. The initial vertical diffusion coefficient (yz,o) is determined to be 8.18 meters, based on the 161.00 foot containment above-grade height.

The containment shell diffuse release is assumed to be from its mid-height of 80.5 feet (24.54 meters) above grade. This elevation allows for unimpeded flow above the Auxiliary Building roof. Because the control room intakes are on the Auxiliary Building wall opposite the U2 and U3 containments, the release cloud will flow to the intakes by first passing over the Auxiliary Building roof. This assumption is consistent with the RG 1.194 Section 3.2.4.5 recommendation to set the release height for a diffuse area source at the vertical center of the projected plane of the above-grade cross-sectional area perpendicular to the line of sight from the building center to the control room intake.

For the containment shell release, only the cross-sectional area of the containment is used to determine the building wake area. All other intervening buildings, such as the auxiliary building, are conservatively ignored. The building wake area of 2123.33 square meters is the projected area of the containment cylindrical lower portion and the containment upper dome portion.

Page 39 of 110

Table 4.4-3 presents the separation distances and wind directions that characterize the releases from the two containment shell release point locations to the three control room HVAC intake locations:

TABLE 4.4-3: CONTAINMENT TO CONTROL ROOM MODELING II Release Point 1Control Room Receptor l Separation Distance (meters)

Wind Direction (degrees, North = 0)

U2 Containment Normal Air Intake 38.6 348 U2 Containment U2 emergency air intake 36 351 U2 Containment U3 emergency air intake 77.4 329 U3 Containment Normal Air Intake 74 96 U3 Containment U2 emergency air intake 77.4 97 U3 Containment U3 emergency air intake 36 75 The results of the ARCON96 analysis show that the Unit 2 containment shell to Unit 2 emergency air intake release path has the more conservative atmospheric dispersion (i.e., the maximum atmospheric dispersion factor). The resultant 95th percentile control room atmospheric dispersion factors for this release path are presented in Section 4.4.5.

Section 4.4.4.4 Containment Equipment Hatch Release Atmospheric dispersion between the containment equipment hatch and the control room HVAC intakes is modeled as an area (diffuse) source using the ARCON96 ground level release option.

The Containment Equipment Hatch is a large circular opening through the containment wall. The equipment hatch meets the conditions for a diffuse source as set forth in RG 1.194 Section 3.2.4.8: (1) the release from the hatch will be essentially equally dispersed over the entire opening, and (2) assumptions of mixing, dilution and transport within Containment necessary to meet condition 1 are supported by the interior containment arrangement. Consistent with RG 1.194 Section 3.2.4.4, the initial horizontal and vertical diffusion coefficients (Gy,0 and oz,o) are each determined to be 0.97 meters, based on the clear 19-foot diameter of the hatch opening.

The Unit 2 and 3 Containment Equipment Hatches are on the opposite side of their respective Containment structures from the control room air intakes. The containment equipment hatch diffuse release is assumed to be from its mid-height at plant elevation 38.00 feet (11.58 meters). The top of the Containment is 161 feet above grade; therefore, it is unrealistic to expect flow from the Equipment Hatch to go over the Containment building. The Equipment Hatches to receptor distances are measured as the shortest path around the Containment ("taut string length"), as allowed by RG 1.194 Section 3.4. To determine the taut string length, a tangent is drawn from each intake to the side Page 40 of 110

of the containment closest to the equipment hatch. That distance is added to the length of the arc around the containment from the tangent line intersection to the centerline of the hatch.

As requested by RG 1.194 Appendix A Table A-2, since the plume is assumed to flow around the containment building rather than over it, the atmospheric dispersion value is calculated assuming flow both around and over (through) the building, and the higher of the atmospheric dispersion values is used.

For the containment equipment hatch release, only the cross-sectional area of the containment is used to determine the building wake area. All other intervening buildings, such as the auxiliary building, are conservatively ignored.

The building wake area of 2123.33 square meters is the projected area of the containment cylindrical lower portion and the containment upper dome portion.

Table 4.4-4 presents the separation distances and wind directions that characterize the releases from the two containment equipment hatch release point locations to the three control room HVAC intake locations:

TABLE 4.4-4: CONTAINMENT EQUIPMENT HATCH TO CONTROL ROOM MODELING Wind Direction Release Point Control Room Receptor Separation Distance Over I Around (meters) Containment (degrees, North = 0)

U2 Ctmt Equip. Hatch Normal Air Intake 98.1 353/11 U2 Ctmt Equip. Hatch U2 emergency air intake 96.8 355 / 15 U2 Ctmt Equip. Hatch U3 emergency air intake 126.9 336 / 343 U3 Ctmt Equip. Hatch Normal Air Intake 124 89 / 82 U3 Ctmt Equip. Hatch U2 emergency air intake 126.9 90/ 83 U3 Ctmt Equip. Hatch U3 emergency air intake 96.8 71 / 51 The results of the ARCON96 analysis show that the Unit 2 equipment hatch to Unit 2 emergency air intake release path modeling flow over (through) the containment building has the more conservative atmospheric dispersion (i.e., the maximum atmospheric dispersion factor) during the 8 to 24 hour2.777778e-4 days <br />0.00667 hours <br />3.968254e-5 weeks <br />9.132e-6 months <br /> time period. For all other time periods, the results of the ARCON96 analysis show that the Unit 2 equipment hatch to Unit 2 emergency air intake release path modeling flow around the containment building has the more conservative atmospheric dispersion. The atmospheric dispersion factors for this release are a conservative composite of these two flow paths. The resultant 95th percentile control room atmospheric dispersion factors are presented in Section 4.4.5.

Section 4.4.4.5 Main Steam Safety Valve (MSSV) Stack Release Atmospheric dispersion between the MSSV stack and the control room HVAC intakes is modeled as a point source using the ARCON96 ground level release Page41 of 110

option. Consistent with RG 1.194 Section 6.0 (and as justified in the following text), a reduction factor of 5 is applied to the ARCON96 results to allow credit for buoyant plume rise in determining the control room atmospheric dispersion factors associated with the energetic release from MSSVs.

Each reactor has two sets of nine MSSV stacks arrayed around a Main Steam Isolation Valve (MSIV). As an average location, the center of the MSIV (X and Y dimensions only) is modeled as the MSSV release location. MSIV 8205 is located north of the Unit 2 containment centerline (south for Unit 3). MSIV 8204 is located south of the Unit 2 containment centerline (north for Unit 3).

The MSSV release height is at plant elevation 73.42 feet (22.38 meters).

For the MSSV release, only the cross-sectional area of the containment is used to determine the building wake area. All other intervening buildings, such as the auxiliary building, are conservatively ignored. The building wake area of 2123.33 square meters is the projected area of the containment cylindrical lower portion and the containment upper dome portion.

Table 4.4-5 presents the separation distances and wind directions that characterize the releases from the MSSV release point locations (i.e., Unit 2 MSSVs centered at MSIVs 8204 and 8205, and Unit 3 MSSVs centered at MSIVs 8204 and 8205) to the three control room HVAC intake locations:

TABLE 4.4-5: MSSV TO CONTROL ROOM MODELING Release Point Control Room Receptor Separation Distance Wind Direction (meters) (degrees, North = )

U2 MSSV 8204 Normal Air Intake 35.71 339 U2 MSSV 8204 U2 emergency air intake 32.65 343 U2 MSSV 8204 U3 emergency air intake 78.96 318 U2 MSSV 8205 Normal Air Intake 60.17 322 U2 MSSV 8205 U2 emergency air intake 56.54 323 U2 MSSV 8205 U3 emergency air intake 105.87 314 U3 MSSV 8204 Normal Air Intake 75.25 107 U3 MSSV 8204 U2 emergency air intake 78.96 108 U3 MSSV 8204 U3 emergency air intake 32.65 83 U3 MSSV 8205 Normal Air Intake 102.08 112 U3 MSSV 8205 U2 emergency air intake 105.87 112 U3 MSSV 8205 U3 emergency air intake 56.54 103 The results of the ARCON96 analysis show that the Unit 2 MSSVs centered at MSIV 8204 to Unit 2 emergency air intake release path has the more conservative atmospheric dispersion (i.e., the maximum atmospheric dispersion factor).

RG 1.194 allows credit for buoyant plume rise in determining the Control Room atmospheric dispersion factors associated with an energetic release from main Page 42 of 110

steam safety valves. RG 1.194 Section 6.0 states that in lieu of mechanistically addressing the amount of buoyant plume rise:

"...the ground level X/O value calculated with ARCON96 (on the basis of the physical height of the release point) may be reduced by a factor of 5.

This reduction may be taken only if (1) the release point is uncapped and vertically oriented and (2)the time-dependent vertical velocity exceeds the 95th-percentile wind speed (at the release point height) by a factor of 5."

The MSSVs are uncapped and vertically oriented, thereby satisfying the first criterion required for plume rise credit per RG 1.194, Section 6.0.

Since the MSSV stack exit is at plant elevation 73.4 ft and grade is at plant elevation 30 ft, the height of the stack exit above grade is 43.4 ft, or 13.2 meters.

This is reasonably close to the height of the lower meteorological tower wind measurement instrumentation at 10 meters; therefore, the MSSV stack exit velocity is compared with the 95th percentile 10-m wind speed of 5.8 mrs.

For purposes of calculating the minimum flow velocity at the exit of the MSSV stack, the following conservative assumptions are made:

1. All MSSVs lift at the lowest set pressure of all the valves.

2. The pressure in the MSSV stack is equal to the maximum backpressure.

3. No credit is taken for head loss due to elevation changes or pipe friction.

4. No credit is taken for expansion of the steam through the stack.

The calculated minimum MSSV stack exit velocity is 72 meters/second. This exit velocity exceeds five times the 95th percentile wind speed (i.e., exceeds 5 x 5.8 m/s = 29 m/s); thereby satisfying the second criterion required for plume rise credit per RG 1.194 Section 6.0.

Since both criteria are satisfied, the ground level atmospheric dispersion factors calculated with ARCON96 (on the basis of the physical height of the release point) for MSSV releases are reduced by a factor of 5. The resultant 95th percentile control room atmospheric dispersion factors for the MSSV release path with credit for plume rise are presented in Section 4.4.5.

Section 4.4.4.6 Atmospheric Dump Valve (ADV) Stack Release Atmospheric dispersion between the ADV and the control room HVAC intakes is modeled as a point source using the ARCON96 ground level release option. Per RG 1.194 Section 6.0 (and as justified in the following text), a reduction factor of 5 may be applied to the ARCON96 results to allow credit for buoyant plume rise in determining the Control Room atmospheric dispersion factors associated with the energetic release from atmospheric dump valves.

Page 43 of 110

Each reactor has two ADV stacks. ADV 606 is located north of the Unit 2 containment centerline (south for Unit 3). ADV 607 is located south of the Unit 2 containment centerline (north for Unit 3).

The ADV release height is at plant elevation 113.92 feet (34.72 meters).

For the ADV release, only the cross-sectional area of the containment is used to determine the building wake area. All other intervening buildings, such as the auxiliary building, are conservatively ignored. The building wake area of 2123.33 square meters is the projected area of the containment cylindrical lower portion and the containment upper dome portion.

Table 4.4-6 presents the separation distances and wind directions that characterize the releases from the ADV release point locations (i.e., Unit 2 ADVs 606 and 607, and Unit 3 ADVs 606 and 607) to the three control room HVAC intake locations:

TABLE 4.4-6: ADV TO CONTROL ROOM MODELING Release Point Control Room Receptor Release PoinICotrolRoo _Recpto Separation Distance (meters) J Wind Direction (degrees, North =0)

U2 ADV 606 Normal Air Intake 57.45 325 U2 ADV 606 U2 emergency air intake 53.88 326 U2 ADV 606 U3 emergency air intake 102.76 315 U2 ADV 607 Normal Air Intake 37.69 343 U2 ADV 607 U2 emergency air intake 34.84 348 U2 ADV 607 U3 emergency air intake 79.66 321 U3 ADV 606 Normal Air Intake 98.98 111 U3 ADV 606 U2 emergency air intake 102.76 111 U3 ADV 606 U3 emergency air intake 53.88 100 U3 ADV 607 Normal Air Intake 75.99 104 U3 ADV 607 U2 emergency air intake 79.66 105 U3 ADV 607 U3 emergency air intake 34.84 78 The results of the ARCON96 analysis show that the Unit 2 ADV 607 to Unit 2 emergency air intake release path has the more conservative atmospheric dispersion (i.e., the maximum atmospheric dispersion factor).

RG 1.194 allows credit for buoyant plume rise in determining the Control Room atmospheric dispersion factors associated with an energetic release from atmospheric dump valves. RG 1.194 Section 6.0 states that in lieu of mechanistically addressing the amount of buoyant plume rise:

"...the ground level X/Q value calculated with ARCON96 (on the basis of the physical height of the release point) may be reduced by a factor of 5.

This reduction may be taken only if (1)the release point is uncapped and vertically oriented and (2) the time-dependent vertical velocity exceeds the 95th-percentile wind speed (at the release point height) by a factor of 5."

Page 44 of 110

The ADVs are uncapped and vertically oriented, thereby satisfying the first criterion required for plume rise credit per RG 1.194 Section 6.0.

Since the ADV stack exit is at plant elevation 113.92 ft and grade is at plant elevation 30 ft, the height of the stack exit above grade is 83.92 ft, or 25.6 meters. Since the ADV stack exit is closer in height to the upper meteorological tower wind measurement instrumentation at 40 meters than to the lower meteorological tower wind measurement instrumentation at 10 meters, the ADV stack exit velocity is compared with the 95th percentile 40-m wind speed of 6.8 m/s.

The accident analyses assume that an ADV is operated manually; therefore, the flow velocity at the ADV stack exit will decrease over time, as the steam generator blows down. Thus, in order to credit plume rise in an ADV release dose analysis, the period for which the ADV stack exit vertical flow velocity exceeds five times the 95th percentile upper level wind speed of 6.8 m/s (i.e.,

exceeds 5 x 6.8 m/s = 34 m/s) would need to be determined.

Since the second criterion is not necessarily satisfied for the duration of a dose analysis, the ground level atmospheric dispersion factors calculated with ARCON96 (on the basis of the physical height of the release point) for ADV releases may or may not be reduced by a factor of 5 for the duration of a dose analysis. The resultant 95th percentile control room atmospheric dispersion factors for the ADV release path with and without credit for plume rise are presented in Section 4.4.5. The use of the lower values crediting plume rise will be evaluated on an event-specific basis.

Section 4.4.4.7 Steam Line Break Outside Containment (SLB-OC)

Release Atmospheric dispersion between the steam line break outside containment and the control room HVAC intakes is modeled as an area (diffuse) source using the ARCON96 ground level release option.

The SLB-OC is postulated to occur outboard of the main steam line restraint/anchor downstream of the main steam isolation valve. Thus, the location of the postulated break is in the walkway between the east wall of the Turbine Building and the Main Steam Isolation Valve/Main Feedwater Isolation Valve (MSIVIMFIV) enclosure structures. The enclosure structures are open to the walkway, which is then open to the atmosphere above. Several blowout panels are present on the roofs of the enclosure structures. There are also blowout panels on the walls of the MSIV/MFIV enclosure. The blowout panels open during a large SLB-OC to protect the enclosure structures from overpressurizing. Thus, depending on the size of the steam line break, there are multiple pathways for steam blowdown.

Page 45 of 110

Steam from an SLB-OC is assumed to be released via the blowout panels mounted on the roof of the respective MSIV/MFIV enclosure directly above the main steam line. For a small SLB-OC, the pressure may not exceed the relief setting of the blowout panels; in such an instance, steam will escape via the enclosure opening to the walkway between the enclosure and the Turbine Building. For a large SLB-OC, steam may also escape via the roof blowout panel above the main feedwater line and via the enclosure wall-mounted blowout panels. Nevertheless, the assumption that the release is solely via the roof blowout panels is conservative, because it results in a smaller flow area, which means a smaller initial horizontal diffusion coefficient, than if all the potential vent paths were considered.

The SLB-OC release via the roof blowout panels meets the conditions for a diffuse source in RG 1.194 Section 3.2.4.7, which states that the application of the diffuse area source model to determine atmospheric dispersion factors for multiple (i.e., 3 or more) roof vents is:

"...appropriate for configurations in which (1)the vents are in close arrangement, (2) no individual vent is significantly closer to the control room intake than the center of the area source, (3) the release rate from each vent is approximately the same, and (4) no credit is taken for plume rise."

Condition 1 is satisfied since the 3 roof-mounted blowout panels directly above each main steam line are in close proximity with each other (spaced from 2.0 to 3.5 feet apart). Condition 2 is satisfied since no individual blowout panel is significantly closer to the control room intake than the center of the area source.

Although there are other vent paths (i.e., through the enclosure opening adjacent to the walkway, the roof-mounted blowout panels over the main feedwater lines, and the wall-mounted blowout panels), it is conservative for purposes of determining control room atmospheric dispersion factors to minimize the initial dispersion area by accounting only for the area source presented by the blowout panels above the main steam lines. Condition 3 is satisfied since the steam is assumed to rise evenly through the three adjacent blowout panels. Condition 4 is satisfied since no credit is taken for plume rise. Therefore, the SLB-OC release meets the conditions for a diffuse source per Regulatory Guide 1.1 94, Section 3.2.4.7.

The area width is measured across the area formed by the three blowout panels mounted on the roof of the MSIV enclosure perpendicular to the line of sight from the MSIVs to the respective control room intake. Table 4.4-7 presents the initial horizontal diffusion coefficients (ayo) that characterize the releases from the SLB-OC release point locations. Consistent with RG 1.194 Section 3.2.4.7, because the blowout panel openings are in a horizontal configuration on the MSIV enclosure roof, there is no initial vertical dispersion coefficient (i.e., azO is zero).

Page 46 of 110

The distance between the SLB-OC release point and the control room intakes is measured from the closest point on the perimeter of the roof blowout panels above the MSIVs, as allowed by Section 3.2.4.7 of RG 1.194.

The SLB-OC release height via the MSIV/MFIV enclosure roof blowout panels is at plant elevation 63.50 feet (19.35 meters).

For the SLB-OC release, only the cross-sectional area of the containment is used to determine the building wake area. All other intervening buildings, such as the auxiliary building, are conservatively ignored. The building wake area of 2123.33 square meters is the projected area of the containment cylindrical lower portion and the containment upper dome portion.

Table 4.4-7 presents the separation distances, wind directions and initial horizontal diffusion coefficients (ayo) that characterize the releases from the SLB-OC release point locations (i.e., Unit 2 North and South MSIV/MFIV enclosure roof blowout panels, and Unit 3 North and South MSIV/MFIV enclosure roof blowout panels) to the three control room HVAC intake locations:

TABLE 4.4-7: SLB-OC TO CONTROL ROOM MODELING Separation Wind Direction aypo Release Point Control Room Receptor Distance (degrees, [Cro = 0]

(meters) North = 0) (meters)

U2 N Panels Normal Air Intake 58.5 322 0.96 U2 N Panels U2 emergency air intake 54.9 323 0.96 U2 N Panels U3 emergency air intake 104.2 314 0.99 U2 S Panels Normal Air Intake 32.3 339 0.56 U2 S Panels U2 emergency air intake 29.3 343 0.54 U2 S Panels U3 emergency air intake 76.9 318 0.7 U3 N Panels Normal Air Intake 100.4 112 0.99 U3 N Panels U2 emergency air intake 104.2 112 0.99 U3 N Panels U3 emergency air intake 54.9 103 0.96 U3S Panels Normal Air Intake 73.2 107 1.42 U3 S Panels U2 emergency air intake 76.9 108 1.44 U3 S Panels U3 emergency air intake 29.3 83 1.12 The results of the ARCON96 analysis show that the Unit 2 South MSIV/MFIV enclosure structure (housing Unit 2 MSIV 8204) roof blowout panels to Unit 2 emergency air intake release path have the more conservative atmospheric dispersion (i.e., the maximum atmospheric dispersion factor). The resultant 95th percentile control room atmospheric dispersion factors for this release path are presented in Section 4.4.5.

Page 47 of 110

Section 4.4.4.8 Auxiliary Feedwater (AFW) Turbine Stack Release Atmospheric dispersion between the AFW turbine stack and the control room HVAC intakes is modeled as a point source using the ARCON96 vent release option.

The AFW turbine stack is located near the south wall of the Unit 2 refueling water storage and condensate storage tank farm (north wall for Unit 3).

The AFW turbine stack release height is at plant elevation 59 feet (17.98 meters).

For the AFW turbine stack release, only the cross-sectional area of the containment is used to determine the building wake area. All other intervening buildings, such as the auxiliary building, are conservatively ignored. The building wake area of 2123.33 square meters is the projected area of the containment cylindrical lower portion and the containment upper dome portion.

Table 4.4-8 presents the separation distances and wind directions that characterize the releases from the two AFW turbine stack activity release point locations to the three control room HVAC intake locations:

TABLE 4.4-8: AFW TURBINE STACK TO CONTROL ROOM MODELING Release Point Control Room Receptor Separation Distance (Wind Direction Cotolom eepo (meters) ~(degrees, North =0 U2 AFW Turbine Stack Normal Air Intake 86.89 332 U2 AFW Turbine Stack U2 emergency air intake 83.53 333 U2 AFW Turbine Stack U3 emergency air intake 130.22 322 U3 AFW Turbine Stack Normal Air Intake 126.57 104 U3 AFW Turbine Stack U2 emergency air intake 130.22 104 U3 AFW Turbine Stack U3 emergency air intake 83.53 93 The results of the ARCON96 analysis show that the Unit 2 AFW turbine stack to Unit 2 emergency air intake release path has the more conservative atmospheric dispersion (i.e., the maximum atmospheric dispersion factor). The resultant 95th percentile control room atmospheric dispersion factors for this release path are presented in Section 4.4.5.

Section 4.4.4.9 Refueling Water Storage Tank (RWST) Vent Release Atmospheric dispersion between the RWST vent and the control room HVAC intakes is modeled as a point source using the ARCON96 vent release option.

There are two RWST tanks for each unit (T005 and T006). The two tanks are located in the Units 2 and 3 refueling water storage and condensate storage tank farms, which are mirror images of each other. Unit 2 RWST T005 is located northeast of the Unit 2 containment centerline (southeast for Unit 3). Unit 2 Page 48 of 110

RWST T006 is located northwest of the containment centerline (southwest for Unit 3).

The RWST release location is assumed to be the center of the roof vent on each RWST. These vents are offset from the center of the tank roofs. The Unit 2 RWST T005 and T006 vent release height is at plant elevation 71.57 feet (21.81 meters). The Unit 3 RWST T005 and T006 vent release height is at plant elevation 71.73 feet (21.86 meters).

For the RWST vent release, only the cross-sectional area of the containment is used to determine the building wake area. All other intervening buildings, such as the auxiliary building, are conservatively ignored. The building wake area of 2123.33 square meters is the projected area of the containment cylindrical lower portion and the containment upper dome portion.

Table 4.4-9 presents the separation distances and wind directions that characterize the releases from the four RWST vent activity release point locations (i.e., Unit 2 RWSTs T005 and T006, and Unit 3 RWSTs T005 and T006) to the three control room HVAC intake locations:

TABLE 4.4-9: RWST TO CONTROL ROOM MODELING Release Point Control Room Receptor Separation Distance l Wind Direction I Co (meters) J(degrees, North =0)

U2 RWST T005 Normal Air Intake 106.66 330 U2 RWST T005 U2 emergency air intake 103.23 331 U2 RWST T005 U3 emergency air intake 150.24 322 U2 RWST T006 Normal Air Intake 103.79 327 U2 RWST T006 U2 emergency air intake 100.26 327 U2 RWST T006 U3 emergency air intake 148.21 319 U3 RWST T005 Normal Air Intake 146.59 104 U3 RWST T005 U2 emergency air intake 150.24 104 U3 RWST T005 U3 emergency air intake 103.23 95 U3 RWST T006 Normal Air Intake 144.51 106 U3 RWSTT006 U2 emergency air intake 148.21 107 U3 RWST T006 U3 emergency air intake 100.26 99 The results of the ARCON96 analysis show that the Unit 2 RWST T006 vent to Unit 2 emergency air intake release path has the more conservative atmospheric dispersion (i.e., the maximum atmospheric dispersion factor) for the 0 to 2 hour2.314815e-5 days <br />5.555556e-4 hours <br />3.306878e-6 weeks <br />7.61e-7 months <br /> time period. For time periods after 2 hours2.314815e-5 days <br />5.555556e-4 hours <br />3.306878e-6 weeks <br />7.61e-7 months <br />, the results of the ARCON96 analysis show that the Unit 2 RWST T005 vent to Unit 2 emergency air intake release path has the more conservative atmospheric dispersion. The atmospheric dispersion factors for this release are a conservative composite of these two flow paths. The resultant 95th percentile control room atmospheric dispersion factors for these release paths are presented in Section 4.4.5.

Page 49 of 110

Section 4.4.4.10 Fuel Handling Building (FHB) Release Atmospheric dispersion between the fuel handling building and the control room HVAC intakes is modeled as a point source using the ARCON96 vent release option.

The Units 2 and 3 FHBs are each located to the east of the Units 2 and 3 containment buildings. The FHB release location is assumed to be the spent fuel cask hatch over the south end of the Unit 2 railroad access bay (north end of Unit 3 bay). This hatch is larger and closer to the control room air intake locations than the smaller cask hatch over the opposite ends of the respective railroad bays. The centerline of the spent fuel cask is south and east of the Unit 2 containment centerline (north and east of the Unit 3 containment centerline).

The FHB spent fuel cask hatch release height is at plant elevation 63.50 feet (19.35 meters).

For the FHB release, the fuel handling building east cross-section area and one half of the containment cross-section area is used. Only one half of the containment is conservatively considered since it is partially offset from the release to intake axis. All other intervening buildings, such as the auxiliary building, are conservatively ignored. The building wake area for the FHB release is 2076.78 square meters.

Table 4.4-10 presents the separation distances and wind directions that characterize the releases from the two FHB activity release point locations to the three control room HVAC intake locations:

TABLE 4.4-10: FHB TO CONTROL ROOM MODELING Release Point ControlRoom Rece Separation Distance Wind Direction I(meters) ptor (degrees, North = 0)

U2 FH3 Normal Air Intake 91.82 20 U2 FHB U2 emergency air intake 91.06 23 U2 FHB U3 emergency air intake 112 356 U3 FHB Normal Air Intake 109.73 68 U3 FHB U2 emergency air intake 112 70 U3 FHB U3 emergency air intake 91.06 43 The results of the ARCON96 analysis show that the Unit 2 FHB to Unit 2 emergency air intake release path has the more conservative atmospheric dispersion (i.e., the maximum atmospheric dispersion factor). The resultant 95th percentile control room atmospheric dispersion factors for this release path are presented in Section 4.4.5.

Page 50 of 110

Section 4.4.5 ARCON96 Results - 95th Percentile Control Room Atmospheric Dispersion Factors For each of the release locations, the maximum atmospheric dispersion factor from either unit to any of the three control room air intakes is determined. The resultant 95th percentile control room atmospheric dispersion factors are presented in Table 4.4-11.

As discussed in Section 4.4.4.5, the control room atmospheric dispersion factor results for the MSSV stack release include credit for plume rise. MSSV plume rise credit may be modeled for any accident with an MSSV release.

The control room atmospheric dispersion factor results for the ADV stack release are presented with and without credit for plume rise. The determination of whether ADV plume rise credit can be taken must be determined on an accident-specific basis.

Control room occupancy factors are not included in the 95th percentile atmospheric dispersion factors reported in Table 4.4-11. Control room occupancy factors are modeled as separate input parameters to the dose analyses.

TABLE 4.4-11: 9 5 th Percentile Control Room X/Qs (sec/m 3 )

[without CR Occupancy Factors]

Main Containment Equipment ADV ADV Time Interval Plant Vent Shell Hatch (no plume (with plume rise credit) rise credit) 0 to 2 hrs 1.14E-03 9.94E-04 7.99E-04 3.70E-03 7.40E-04 2 to 8 hrs 6.11 E-04 6.32E-04 6.30E-04 1.97E-03 3.94E-04 8 to 24 hrs 2.1 OE-04 1.77E-04 1.77E-04 6.86E-04 1.37E-04 1 to 4 days 2.20E-04 2.34E-04 2.23E-04 6.97E-04 1.39E-04 4 to 30 days 1.98E-04 2.1 8E-04 2.03E-04 6.33E-04 1.27E-04 MSSV AFW Fuel Time Interval (with plume SLB-OC Turbine RWST Handling rise credit) Exhaust Building 0 to 2 hrs 1.21 E-03 7.74E-03 8.55E-04 5.65E-04 9.45E-04 2 to 8 hrs 7.48E-04 4.79E-03 3.60E-04 2.17E-04 7.48E-04 8 to 24 hrs 2.50E-04 1 .62E-03 1.56E-04 8.67E-05 1 .93E-04 1 to 4 days 2.86E-04 1.83E-03 1.60E-04 8.88E-05 2.64E-04 4 to 30 days 2.60E-04 1 .68E-03 1.30E-04 7.33E-05 2.43E-04 Page 51 of 110

Section 4.5 LOSS OF COOLANT ACCIDENT ANALYSIS Regulatory Guide 1.183 Appendix A provides assumptions for use in evaluating the radiological consequences of a LOCA. These assumptions supplement the guidance provided in the main body of RG 1.183.

The SONGS Units 2 and 3 LOCA is characterized by the following activity release paths:

• Containment leakage

• Engineered Safety Feature (ESF) recirculation loop leakage

• Refueling Water Storage Tank (RWST) release

• Post Accident Sampling System (PASS) leakage The SONGS Units 2 and 3 LOCA is also characterized by the following gamma radiation shine dose contributors which are discussed in Section 4.3:

• Environmental cloud gamma radiation shine

• Control room filter gamma radiation shine

• Containment building gamma radiation shine

• Post-LOCA piping gamma radiation shine The characteristics of the LOCA model are summarized in Table 4.5-1.

This Section presents the assumptions, design input, methodology employed in evaluating, and the radiological consequences of each of the SONGS Units 2 and 3 LOCA dose contributors.

TABLE 4.5-1: LOCA DOSE ANALYSIS PARAMETERS LOCA PARAMETER l MODELED VALUE Dose acceptance criteria, Rem TEDE Control Room 5 EAB 25 LPZ 25 Containment leakage parameters per Table 4.5-2 ESF system leakage parameters per Table 4.5-8 RWST release parameters per Table 4.5-9 PASS leakage parameters per Table 4.5-10 Offsite dose evaluation model per Section 4.2 Control Room dose evaluation model l per Section 4.3 Section 4.5.1 Containment Leakage Regulatory Guide 1.183 Appendix A provides assumptions for use in evaluating the radiological consequences of the containment leakage pathway for a LOCA.

Page 52 of 110

These assumptions supplement the guidance provided in the main body of RG 1.183.

This Section presents the assumptions, design input, and methodology employed in evaluating the radiological consequences of the SONGS Units 2 and 3 LOCA containment leakage path. The characteristics of the LOCA containment leakage model are summarized in Table 4.5-2.

The control room and offsite doses associated with containment leakage are summarized in Section 4.5.7.

TABLE 4.5-2: LOCA CONTAINMENT LEAKAGE ANALYSIS PARAMETERS LOCA CONTAINMENT LEAKAGE PARAMETER l MODELED VALUE LOCA source term l Reactor core isotope inventory at shutdown, curies per Section 4.1 Timing of core activity release into containment Gap release phase 0.5 to 30 minutes Early in-vessel phase 0.5 to 1.8 hours9.259259e-5 days <br />0.00222 hours <br />1.322751e-5 weeks <br />3.044e-6 months <br /> Core inventory fraction released into containment, gap release phase Noble gases (Xe, Kr) 0.05 Halogens (I, Br) 0.05 Alkali Metals (Cs, Rb) 0.05 Other Elements 0.00 Core inventory fraction released into containment, early in-vessel phase Noble gases (Xe, Kr) 0.95 Halogens (I, Br) 0.35 Alkali metals (Cs, Rb) 0.25 Tellurium metals (Te, Sb, Se) 0.05 Barium, Strontium (Ba, Sr) 0.02 Noble metals (Ru, Rh, Pd, Mo, Tc, Co) 0.0025 Cerium group (Ce, Pu, Np) 0.0005 Lanthanides (La, Zr, Nd, Eu, Nb, Pm, Pr, Sm, Y, Cm, Am) 0.0002 Chemical form of Iodine released into containment, percent of iodine Cesium iodide (Csl) (particulate iodine) 95.00 Elemental iodine 4.85 Organic iodide 0.15 Containment net free volume, cubic feet Total 2.284E+06 Sprayed volume 1.907E+06 Unsprayed volume 3.770E+05 Containment mechanical air mixing Number of operating emergency cooling unit (ECU) trains 1 Number of operating emergency cooling units 2 ECUs per train Emergency cooling unit flow rate, ft3/minute per ECU 31,000 Number of operating dome air circulator unit (DACU) trains 1 Number of operating dome air circulator units 2 DACUs per train Dome air circulator unit flow rate, ft3/minute per DACU 37,000 Page 53 of 110

LOCA CONTAINMENT LEAKAGE PARAMETER MODELED VALUE Natural deposition (plateout) removal of airborne radionuclides Per Section 4.5.1.3 Containment spray removal of airborne radionuclides Per Section 4.5.1.3 Containment leakage rate, ft3 /minute 0 to 1 day, Total leakage 1.6 0 to 1 day, Sprayed volume leakage 1.34 0 to 1 day, Unsprayed volume leakage 0.26 1 to 30 days, Total leakage 0.8 1 to 30 days, Sprayed volume leakage 0.67 1 to 30 days, Unsprayed volume leakage 0.13 Atmospheric Dispersion Factors for the Containment Leakage release via per Section 4.4 the Containment Shell to Control Room, seconds/M 3 Section 4.5.1.1 Containment Leakage Source Term Regulatory Guide 1.183 Appendix A Section 1 states that acceptable assumptions regarding core inventory and the release of radionuclides from the fuel are provided in RG 1.183 Regulatory Position 3.

Consistent with RG 1.183 Regulatory Position 3.4 and its Table 5, the core isotopes released into the containment are grouped into chemically similar groups. The elements of each group are as listed in Table 4.5-2 Consistent with RG 1.183 Regulatory Position 3.2, the core inventory release fractions, by radionuclide groups, for the gap release and early in-vessel phases for DBA LOCAs are as listed in Table 4.5-2.

Consistent with RG 1.183 Regulatory Position 3.2, the core inventory release fractions are applied to the equilibrium reactor core isotope inventory at shutdown as described in Section 4.1.

Consistent with RG 1.183 Regulatory Position 3.3 and its Table 4, the onset and duration of the sequential gap release and early in-vessel release phases for the LOCA are as specified in Table 4.5-2. The gap release phase begins at 30 seconds and the early in-vessel release phase begins at 30 minutes.

Although the gap release duration is to be 30 minutes, it is conservatively assumed to end at 30 minutes so that there is no overlap with the start of the early in-vessel release phase (i.e., the gap release phase is modeled with a duration of 29 minutes and 30 seconds). The activity released from the core during each release phase is modeled as increasing in a linear fashion over the duration of the phase.

Consistent with RG 1.183 Appendix A Section 2, an evaluation of post-LOCA containment sump pH has considered the effect of acids and bases created during the LOCA event (e.g., radiolysis products). As discussed in Section 4.5.1.3.4, the containment sump pH is at a value of 7 or greater at the start of the LOCA when iodine evolution from the containment sump is a concern, and for Page 54 of 110

worst case conditions the containment sump pH is not lower than approximately 6.9 at the end of 30 days. Consistent with RG 1.183 Appendix A Section 2, since the containment sump pH is controlled at a value of 7 or greater during the release of radioiodine to the containment, the chemical form of radioiodine released to the containment is assumed to be 95 percent cesium iodide (Csl),

4.85 percent elemental iodine, and 0.15 percent organic iodide.

Consistent with Regulatory Guide 1.183 Appendix A Section 2, with the exception of elemental and organic iodine and noble gases, fission products are assumed to be in particulate form (i.e., subject to particulate spray and deposition removal as well as by HEPA filtration).

Section 4.5.1.2 Containment Leakage Activity Release Model Consistent with RG 1.183 Appendix A Section 3.1, the radioactivity released from the fuel is assumed to mix instantaneously and homogeneously throughout the free air volume of the primary containment. The activity release is terminated at the end of the early in-vessel phase.

Consistent with RG 1.183 Appendix A Section 3.2, reduction in containment airborne radioactivity by natural deposition within the containment is credited. As discussed in Section 4.5.1.3, the removal of iodines and aerosols by natural processes is calculated using the models presented in Standard Review Plan (SRP) Section 6.5.2 and NUREG/CR-6189, "A Simplified Model of Aerosol Removal by Natural Processes in Reactor Containments" (Reference 16).

Consistent with RG 1.183 Appendix A Section 3.3, reduction in containment airborne radioactivity by the containment spray system is credited. As discussed in Section 4.5.1.3, removal of iodines and aerosols by containment sprays are calculated using the models presented in SRP Section 6.5.2 and NUREG/CR-5966 "A Simplified Model of Aerosol Removal by Containment Sprays" (Reference 17).

Consistent with RG 1.183 Appendix A Section 3.3, the containment building atmosphere is not considered a single, well-mixed volume because the containment sprays cover less than 90% of the containment net free air volume.

The total primary containment net free air volume of 2,284,000 cubic feet consists of a sprayed volume of 1,907,000 cubic feet (83.5 percent) and an unsprayed volume of 377,000 cubic feet (16.5 percent).

Consistent with RG 1.183 Appendix A Section 3.7, the containment is assumed to leak at the peak pressure technical specification leak rate of 0.1 percent of the containment air weight per day for the first 24 hours2.777778e-4 days <br />0.00667 hours <br />3.968254e-5 weeks <br />9.132e-6 months <br /> of the LOCA event.

Consistent with RG 1.183 Appendix A Section 3.7, after the first 24 hours2.777778e-4 days <br />0.00667 hours <br />3.968254e-5 weeks <br />9.132e-6 months <br />, the containment leak rate is halved to 0.05 percent of the containment air weight per day.

Page 55 of 110

The containment leakage is assumed to originate from the containment sprayed and unsprayed regions in flow rates that are proportional to the total volume of each region. A well-mixed containment is necessary to justify this modeling. The Containment Emergency Cooling Units (ECUs or air coolers) and the Containment Dome Air Circulator Units (DACUs) provide the necessary air mixing action. Due to an assumed failure, only one of the two Containment ECU trains and one of the two Containment DACU trains are modeled as being operational. The containment ECUs and DACUs are assumed to start operation one minute after the start of the LOCA. Assuming that the air mixing removal of containment unsprayed region activity can be approximated by an exponential relationship, 99 percent of the contaminated air in the containment unsprayed region will be replaced with air from the sprayed region within 13 minutes, ensuring well-mixing of the containment air with more than 8 change-outs of the containment unsprayed region activity prior to the cessation of activity releases at the conclusion of the early in-vessel phase.

The containment mini-purge represents a potential containment airborne activity release path to the environment. Per TS LCO 3.6.3 Surveillance Requirement 3.6.3.2, the 8-inch mini-purge valves are closed except when the valves are open for pressure control, As-Low-As-Reasonably Achievable (ALARA), or air quality considerations for personnel entry, or for surveillances that require the valves to be open. Consistent with the guidance of RG 1.183 Appendix A Section 3.8, because the containment is not routinely purged with the mini-purge system, activity releases through the containment 8-inch mini-purge valves are not analyzed.

The containment purging system represents another potential containment airborne activity release path to the environment. Per TS LCO 3.6.3 Surveillance Requirement 3.6.3.1, the 42-inch purge valves are sealed closed. Consistent with the guidance of RG 1.183 Appendix A Section 7, since the installed containment purging system is not credited in any design basis analysis, activity releases through the containment 42-inch purge valves are not evaluated.

The activity released from the containment is transported by atmospheric dispersion to the control room HVAC intake and to the offsite EAB and LPZ dose receptors. Section 4.4 and its Table 4.4-11 present the San Onofre site-specific 95th percentile meteorology atmospheric dispersion factors used for the containment shell to control room release pathway. No credit is taken for radioactive decay of the isotopes during this atmospheric dispersion transit to the control room or offsite dose locations. Consistent with RG 1.183 Regulatory Positions 4.1.7 and 4.2.2, no correction is made for depletion of the effluent plume by deposition on the ground.

Page 56 of 110

Section 4.5.1.3 Containment Aerosol and Elemental Iodine Removal Consistent with RG 1.183 Appendix A Sections 3.2 and 3.3, reductions in containment airborne radioactivity by natural deposition within the containment and by the containment spray system are credited. This section discusses the modeling employed in determining the aerosol and iodine natural deposition and spray removal rates in the containment sprayed and unsprayed regions.

Section 4.5.1.3.1 addresses the natural deposition of aerosols, including particulate iodine (i.e., cesium iodide). Section 4.5.1.3.2 addresses the natural deposition of elemental iodine. Natural deposition is not considered for the removal of organic iodide and noble gases.

Section 4.5.1.3.3 addresses the spray removal of aerosols, including particulate iodine. Section 4.5.1.3.4 addresses the spray removal of elemental iodine.

Spray removal is not considered for the removal of organic iodide and noble gases.

Section 4.5.1.3.5 presents the combined time-dependent natural deposition plus spray removal rates in the containment sprayed and unsprayed regions for elemental iodine, halogens and other particulates.

Section 4.5.1.3.1 Containment Natural Deposition of Aerosols Consistent with RG 1.183 Appendix A Section 3.2, reduction in containment airborne aerosol radioactivity by natural deposition is credited in the SONGS Units 2 and 3 AST dose analyses. The SONGS natural deposition analysis uses a simplified natural deposition model for aerosols developed by Powers et.al. in NUREG/CR-6189 for the different types of reactors and various operating power levels using a Monte Carlo uncertainty analysis. The results of these analyses for PWR design basis accidents are summarized in correlations provided in NUREG/CR-6189 Table 36. Per RG 1.183 Appendix A Section 3.2, the NUREG/CR-6189 model is incorporated into the RADTRAD analysis code of NUREG/CR-6604.

The aerosol natural deposition removal rates for the gap release and the early in-vessel release phases for the different time periods are conservatively calculated using the Powers model with 10th percentile natural deposition correlations (i.e., minimum deposition) and the analysis models the core rated thermal power of 3438 MWt specified in Section 4.1.1. These deposition rates are weighted by each chemical group release rate reflecting the core release fractions and release phase durations specified in Table 4.5-2, as required by NUREG/CR-6604 Section 2.2.2.1.2. The resultant aerosol natural deposition rates for halogens, alkali metals, and other particulates are shown in Table 4.5-3.

Page 57 of 110

TABLE 4.5-3: AEROSOL NATURAL DEPOSITION REMOVAL RATES HALOGENS ALKALI METALS OTHER PARTICULATES TIME PERIOD 10th PERCENTILE 10th PERCENTILE 10th PERCENTILE P

(hour) NATURAL DEPOSITION NATURAL DEPOSITION NATURAL DEPOSITION (hours) REMOVAL RATE REMOVAL RATE REMOVAL RATE (1/hour) (1/hour) (1/hour) 0 to 0.5 2.94E-02 2.94E-02 N/A 0.5 to 1.8 3.95E-02 4.17E-02 3.12E-02 1.8 to 3.8 8.93E-02 8.93E-02 8.93E-02 3.8 to 13.8 1.16E-01 1.16E-01 1.16E-01 13.8 to 22.2 8.60E-02 8.60E-02 8.60E-02 22.2 to 720 0.OOE+00 0.OOE+00 0.OOE+00 Section 4.5.1.3.2 Containment Natural Deposition of Elemental Iodine Consistent with RG 1.183 Appendix A Section 3.2, reduction in containment airborne elemental iodine radioactivity by natural deposition is credited in the SONGS Units 2 and 3 AST dose analyses. The SONGS natural deposition analysis uses a model provided in NUREG-0800 SRP Section 6.5.2.

Removal of elemental iodine by wall deposition is estimated using the methodology of SRP 6.5.2 Section 111.4.c.(1). Input to this methodology includes a mass transfer coefficient, the wetted surface area inside containment, and the containment building net free volume.

The mass transfer coefficient is modeled with a value of 0.137 cm/sec, consistent with the value of 4.9 m/hr recommended by SRP 6.5.2 Section 111.4.c.(1).

As discussed in NUREG/CR-0009 Section 5.1.2, the natural deposition model for elemental iodine assumes that the bulk gas in the containment atmosphere is well-mixed by natural convection, by steam flows, and by spray operations.

Therefore, all surfaces within the containment are available for elemental iodine aerosol deposition. The wetted surface area modeled for natural deposition is the surface area of 601,519 square feet that is used for the passive heat sinks in the containment pressure/temperature response analyses for LOCA and Main Steam Line Break (MSLB).

The total primary containment net free air volume of 2,284,000 cubic feet is the sum of the containment sprayed and unsprayed volumes.

Based on these parameters, the resultant elemental iodine natural deposition rate is 4.26 inverse hours.

Page 58 of 110

Section 4.5.1.3.3 Containment Spray Removal of Aerosols Consistent with RG 1.183 Appendix A Section 3.3, reduction in containment airborne aerosol radioactivity by the Containment Spray System (CSS) is credited in the SONGS Units 2 and 3 AST dose analyses. The SONGS spray removal analysis uses a simplified spray removal model for aerosols that was developed by Powers et.al. in NUREG/CR-5966 using a Monte Carlo uncertainty analysis. Per RG 1.183 Appendix A Section 3.3, the NUREG/CR-5966 model is incorporated into the RADTRAD analysis code of NUREG/CR-6604.

The aerosol spray removal model input parameters are the CSS spray water flux, the fall height of the spray droplets, and the aerosol mass fraction, which is defined as the aerosol mass in the atmosphere at a given time divided by the total aerosol mass released into the compartment atmosphere until this time.

The CSS has two phases of operation, an injection phase and a recirculation phase. During the injection phase the CSS draws water from the refueling water storage tank until this tank source is exhausted. Following the injection phase, the CSS enters the recirculation phase where water is drawn from the containment sump and recirculated through the CSS. The minimum flow rate per spray header is 1,606 gpm during the injection phase and 1,991 gpm during the recirculation phase. The SONGS spray removal analysis conservatively models the lower injection phase flow rate, rounded down to 1,600 gpm per spray header, throughout the CSS operation. The lower flow rate and resultant spray flux minimizes the activity removal by the sprays and maximizes the airborne radionuclide concentrations. For further conservatism, the SONGS spray removal analysis assumes that only one of the two CSS headers is in operation.

The Powers model is valid for spray water fluxes ranging from 0.001 to 0.25 cm3 H20/cm2-s. The total spray flux for the SONGS Units 2 and 3 spray system has been determined to fall within this applicability range. The spray flux is 0.00615 cm3 H20/ cm2-sbased on a minimum spray system flow rate for one spray header of 1,600 gpm, and the circular floor coverage area for the containment inner diameter of 150 feet.

The Powers model is valid for fall heights ranging from 500 to 5,000 cm. The range of fall heights for the SONGS Units 2 and 3 spray system has been determined to fall within this applicability range. The SONGS Units 2 and 3 containment spray system is designed with multiple spray rings at varying heights within the containment. The fall height of the spray droplets from the various rings to the operating floor ranges from a minimum of 2,433 cm to a maximum of 3,611 cm.

Per NUREG/CR-5966 page 99, the value of the aerosol removal coefficient at an aerosol mass fraction of 0.9 is indicative of the initial rate of decontamination when aerosol is first exposed to the action of a spray.

Page 59 of 110

Unique aerosol spray removal rates were calculated for each spray ring in each of the two CSS spray headers. The calculations address the fact that each spray ring is located at a different height with its own unique spray flux due to different coverage areas and number of spray nozzles per spray ring. The aerosol spray removal rates were conservatively calculated for an aerosol mass fraction of 0.9 using the Powers model with 10th percentile spray removal correlations (i.e.,

minimum deposition). The aerosol spray removal rates for each ring in a spray header were then summed together, and the lowest header value of 4.73 inverse hours for an aerosol mass fraction of 0.9 was determined.

The aerosol spray removal rate will vary from the value of 4.73 inverse hours as a function of the aerosol mass fraction. The Powers model was originally developed for a puff release of aerosol into a system. In those cases where there is a continuous release, such as the AST LOCA with its gap and early in-vessel releases, the injected aerosols will continually renew the aerosol size distribution. For this reason, the aerosol mass fraction has been modeled with a constant value of one until the release stops at 1.8 hours9.259259e-5 days <br />0.00222 hours <br />1.322751e-5 weeks <br />3.044e-6 months <br />. The SONGS spray removal analysis applies the Powers model equations to determine the elapsed time since the end of the in-vessel phase to reach a given aerosol mass fraction, and the aerosol spray removal rate corresponding to that aerosol mass fraction.

The Bechtel LocaDose code is used in the AST dose analyses. The LocaDose code treats the aerosol spray removal rates as constants for any given time period. Therefore, to model the time dependency, average aerosol spray removal rates have been calculated for specified LocaDose code time periods.

The resultant average aerosol spray removal rates are shown in Table 4.5-4.

TABLE 4.5-4: AEROSOL SPRAY REMOVAL RATES AEROSOL TIME PERIOD 10th PERCENTILE (hours) SPRAY REMOVAL RATE (1/hour)

Oto 1.8 5.15 1.8 to 2 3.79 2 to 3.8 1.32 3.8 to 4 0.79 4 to 8 0.62 8 to 13.8 0.52 13.8 to 24 0.50 24 to 48 0.50 48 to 96 0.50 96 to 720 0.50 Page 60 of 110

Section 4.5.1.3.4 Containment Spray Removal of Elemental Iodine Consistent with RG 1.183 Appendix A Section 3.3, reduction in containment airborne elemental iodine radioactivity by the CSS is credited in the SONGS Units 2 and 3 AST dose analyses. The SONGS spray removal analysis uses a model provided in NUREG-0800 SRP Section 6.5.2, "Containment Spray as a Fission Product Cleanup System" that is based on NUREG/CR-0009, "Technological Bases for Models of Spray Washout of Airborne Contaminants in Containment Vessels".

Elemental iodine removal rates by sprays are calculated using the Bechtel REMOVE code, which incorporates the models of SRP 6.5.2 and NUREG/CR-0009. The removal rates are calculated as a function of time. Spray removal rates are calculated for each spray header and each ring individually, and then added together to determine an overall elemental iodine spray removal rate.

Input to the Bechtel REMOVE code includes the CSS characteristics as described in Section 4.5.1.3.3. As in the aerosol spray removal analysis, the SONGS elemental iodine spray removal analysis conservatively assumes that only one of the two containment spray system headers is in operation, and it is operating at 1,600 gpm throughout the CSS injection and recirculation phases.

A spectrum of drop sizes produced by the containment spray system nozzle is used in the REMOVE code model to determine a spray removal rate constant.

The SONGS containment spray system is designed with SPRACO 1713A nozzles. The drop size distribution for this nozzle is built into the Bechtel REMOVE code.

The elemental iodine spray removal rates are dependent on the temperature of the air-steam mixture in the containment. A parametric study was performed to determine the appropriate temperatures to model that will result in conservative removal rate estimates. The study temperatures ranged from the minimum containment air temperature of 89.9 degrees Fahrenheit (32.2 C)to the maximum containment air temperature of 267.4 degrees Fahrenheit (130.8 C),

as determined in the containment pressure/temperature response analysis for the LOCA. The elemental iodine removal rate was found to increase with temperature until about 100 C, and then decrease with further increasing temperature. Within the studied temperature range, the maximum removal rate is 5.314 inverse hours, and the minimum removal rate is 4.569 inverse hours, for a ratio of 1.2 between these extremes. Since the removal rates determined in the spray removal calculation are dependent on a post-LOCA environment that may be revised due to changes in the reactor system or steam generator characteristics, the resultant removal rates calculated with the Bechtel REMOVE code for modeling in the AST containment leakage dose analysis have been conservatively reduced by a factor of 1.2.

Page 61 of 110

The elemental iodine spray removal rates are calculated as a function of time for two phases, the CSS injection phase and the CSS recirculation phase. Input to the Bechtel REMOVE code includes the partition coefficient of iodine between the water and air.

The CSS injection phase begins with initiation of spray and lasts from a minimum of approximately 20 minutes (with two CSS trains in operation) to a maximum of approximately 40 minutes (with one CSS train in operation). During the injection phase, borated water from the RWST is used in the spray system. Consistent with NUREG/CR-0009 page 61, a partition coefficient (i.e., liquid concentration divided by gas concentration) of 200 for boric acid solutions is modeled.

The CSS recirculation phase begins at the end of the CSS injection phase.

During the recirculation phase the spray system draws water from the containment sump. The partition coefficients of iodine between the water and air during the recirculation phase are calculated with the Bechtel ICONC code, using the methodology developed by L.P. Parsly in ORNL-TM-2412 Part IV. Key input parameters for this methodology include the containment sump water temperature and pH, the volumes of the gaseous and liquid phases, the initial iodine inventory in the containment sump water, and the elemental iodine equilibrium constants K1 and K3.

The partition coefficient increases with increasing containment sump water temperature. The containment sump water temperature varies with time during the CSS recirculation phase. Therefore, for each time interval of interest, the minimum containment sump water temperature for that time interval has been conservatively modeled.

The partition coefficient increases with increasing containment sump pH.

Therefore, the minimum containment sump pH level of 7 has been conservatively modeled. SONGS Units 2 and 3 TS LCO 3.5.5 provides for periodic surveillances of the ECCS Trisodium Phosphate (TSP) Dodecahydrate to ensure compliance with the Standard Review Plan 6.5.2 requirement of a pH of at least seven by the onset of recirculation after a LOCA. The required amount of TSP has been calculated based upon the extreme cases of water volume and pH possible in the containment sump. Combustion Engineering Owners Group (CEOG) Task 1178 evaluated the effect of non-traditional acid formers on postaccident containment sump pH for CEOG plants, including SONGS. The evaluation showed that for worst-case conditions, the inclusion of additional acid formers resulted in a slightly lower containment sump pH (approximately 6.9) at the end of 30 days. The evaluation concluded that this pH will continue to ensure that radioiodines remain in the sump solution and are not re-evolved to the containment atmosphere.

The volume of the gaseous phase is the containment free air volume of 2.284E+06 ft3. The volume of the liquid phase is 46,647 ft3, representing the minimum containment emergency sump volume available at the start of the Page 62 of 110

post-LOCA CSS and Safety Injection System (SIS) recirculation mode of operation. The minimum sump volume is appropriate for this calculation since larger liquid volumes will retain more iodine thus increasing the partition coefficient.

The initial iodine inventory in the containment sump water is 2.03 moles, based on the elemental iodine available for release during the gap and the early in-vessel phases.

The elemental iodine equilibrium constants K1 and K3 are obtained from Tables 1 and 2 of ORNL-TM-2412 Part IV.

Unique elemental iodine spray removal rates were calculated for each spray ring in each of the two CSS spray headers. The calculations address the fact that each spray ring is located at a different height 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 then summed together, and the lowest header value at each time interval was determined.

The resultant elemental iodine spray removal rates and decontamination factors are presented in Table 4.5-5. The tabulated removal rates include the previously discussed reduction factor of 1.2.

Per Standard Review Plan 6.5.2 Section 111.4.c.(1), the elemental spray removal rate is limited to a maximum value of 20 inverse hours. The elemental iodine spray removal analysis predicts CSS recirculation phase elemental iodine removal rates in excess of 20 inverse hours during the first 4 hours4.62963e-5 days <br />0.00111 hours <br />6.613757e-6 weeks <br />1.522e-6 months <br /> of the LOCA.

Table 4.5-5 reflects this SRP 6.5.2 limitation.

TABLE 4.5-5: ELEMENTAL IODINE SPRAY REMOVAL RATES TIME PERIOD ELEMENTAL IODINE ELEMENTAL IODINE TIME SPRAY REMOVAL RATE DECON. FACTOR l1(hours) (1/hour) (unitless)

Prior to start of CSS injection phase 0.00 N/A Duration of 1.02 110 CSS injection phase Start of CSS recirculation 20.00 170 phase to 2 hours2.314815e-5 days <br />5.555556e-4 hours <br />3.306878e-6 weeks <br />7.61e-7 months <br /> 2 to 4 20.00 160 4 to 8 18.86 140 8 to 13.8 16.01 110 13.8 to 24 12.99 84 24 to 48 10.09 64 48 to 96 7.83 48 96 to 720 3.78 25 Page 63 of 110

Section 4.5.1.3.5 Combined Natural Deposition Plus Containment Spray Removal Rates The preceding sections present the natural deposition and containment spray removal rates. In this section the individual removal rates are combined into total containment removal rates for use in Bechtel LocaDose code dose assessment calculations. Individual removal rates as a function of time for four groups are determined: elemental iodines, particulate iodines including other halogens, alkali metals, and all other particulates.

The total removal rate is calculated by combining the individual removal rates from the following sources:

1. The aerosol natural deposition rates per Section 4.5.1.3.1

2. The elemental iodine natural deposition rate per Section 4.5.1.3.2

3. The aerosol spray removal rates per Section 4.5.1.3.3

4. The elemental iodine spray removal rates per Section 4.5.1.3.4 Removal rates are determined for the sprayed and the unsprayed containment regions. Inthe sprayed containment region the total removal rate consists of a combination of natural deposition and spray removal. In the unsprayed containment region the total removal rate consists only of the natural deposition rates. The resultant total containment removal rates are shown in Tables 4.5-6 and 4.5-7.

Aerosol natural deposition removal is assumed to begin coincident with the start of the gap release phase at 30 seconds.

CSS injection phase aerosol and elemental iodine removal are assumed to begin with the onset of full flow containment spray, which is conservatively modeled as beginning one minute after the start of the LOCA.

The CSS recirculation phase is entered when the ESF recirculation system begins operation. The ESF recirculation system circulates containment sump liquid via the High Pressure Safety Injection (HPSI), Low Pressure Safety Injection (LPSI), and CSS pumps following a Recirculation Actuation Signal (RAS). This analysis assumes only one train of CSS is in operation, and that the onset of CSS recirculation mode of operation is thereby delayed until 42 minutes after the start of the LOCA. This assumption is conservative since the CSS injection phase spray removal rates for elemental iodine are much lower than the removal rates during CSS recirculation phase. Therefore, a longer CSS injection phase results in slower removal of elemental iodines from the containment atmosphere.

The containment spray is assumed to operate for four hours after the onset of the LOCA. This ensures that a significant portion of the aerosol and elemental iodine Page 64 of 110

releases are removed from the containment atmosphere. Since there is no Decontamination Factor (DF) cut-off for spray removal of aerosols, limiting the spray operation to a maximum of four hours results in conservative estimates of the airborne activity in the containment. The elemental iodine DF cutoff values are reached prior to four hours.

After four hours, the sprayed region is modeled with the unsprayed region natural deposition removal rates.

TABLE 4.5-6: CONTAINMENT SPRAYED REGION REMOVAL RATES ELEMENTAL ELEMENTAL PARTICULATE ALKALI OTHER IODINE IODINE IODINE AND METALS PARTICULATES TIME PERIOD REMOVAL DECON HALOGENS REMOVAL REMOVAL RATE FACTOR REMOVAL RATE RATE (1/hour) (unitless) (llhour) (1/hour)(llhour) 0 to 0.5 minutes 0 N/A 0 0 0 0.5 to 1 minute 4.26 110 0.0294 0.0294 0 1 to 5 minutes 5.28 110 5.18 5.18 5.15 5 to 20 minutes 5.28 110 5.18 5.18 5.15 20 to 30 minutes 5.28 110 5.18 5.18 5.15 30 to 42 minutes 5.28 110 5.19 5.19 5.18 42 minutes to 1 hour1.157407e-5 days <br />2.777778e-4 hours <br />1.653439e-6 weeks <br />3.805e-7 months <br /> 24.26 170 5.19 5.19 5.18 1 to 1.8 hours9.259259e-5 days <br />0.00222 hours <br />1.322751e-5 weeks <br />3.044e-6 months <br /> l24.26 170 5.19 5.19 5.18 1.8 to 2 hours2.314815e-5 days <br />5.555556e-4 hours <br />3.306878e-6 weeks <br />7.61e-7 months <br /> 24.26 170 3.88 3.88 3.88 2 to 3.8 hours9.259259e-5 days <br />0.00222 hours <br />1.322751e-5 weeks <br />3.044e-6 months <br /> 24.26 160 1.41 1.41 1.41 3.8 to 4 hours4.62963e-5 days <br />0.00111 hours <br />6.613757e-6 weeks <br />1.522e-6 months <br /> 24.26 160 0.91 0.91 0.91 4 to 8 hours9.259259e-5 days <br />0.00222 hours <br />1.322751e-5 weeks <br />3.044e-6 months <br /> 4.26 140 0.116 0.116 0.116 8to 13.8 hours9.259259e-5 days <br />0.00222 hours <br />1.322751e-5 weeks <br />3.044e-6 months <br /> 4.26 110 0.116 0.116 0.116 13.8 to 22.2 hours2.314815e-5 days <br />5.555556e-4 hours <br />3.306878e-6 weeks <br />7.61e-7 months <br /> 4.26 84 0.086 0.086 0.086 22.2 to 24 hours2.777778e-4 days <br />0.00667 hours <br />3.968254e-5 weeks <br />9.132e-6 months <br /> 4.26 84 0 0 0 1 to 2 days 4.26 64 0 0 2 to 4 days 4.26 48 0 0 0 4to30days 4.26 25 0 0 0 Page 65 of 110

TABLE 4.5-7: CONTAINMENT UNSPRAYED REGION REMOVAL RATES ELEMENTAL ELEMENTAL PARTICULATE ALKALI OTHER IODINE IODINE IODINE AND METALS PARTICULATES TIME PERIOD REMOVAL DECON HALOGENS REMOVAL REMOVAL RATE FACTOR REMOVAL RATE RATE (1/hour) (unitless) (1/hour) (1/hour) 0 to 0.5 minutes 0 N/A 0 0 0 0.5 to 1 minute 4.26 110 0.0294 0.0294 0 1 to 5 minutes 4.26 110 0.0294 0.0294 0 5 to 20 minutes 4.26 110 0.0294 0.0294 0 20 to 30 minutes 4.26 110 0.0294 0.0294 0 30 to 42 minutes 4.26 110 0.0395 0.0417 0.0312 42 minutes to 1 hour1.157407e-5 days <br />2.777778e-4 hours <br />1.653439e-6 weeks <br />3.805e-7 months <br /> 4.26 170 0.0395 0.0417 0.0312 i to 1.8 hours9.259259e-5 days <br />0.00222 hours <br />1.322751e-5 weeks <br />3.044e-6 months <br /> 4.26 170 0.0395 0.0417 0.0312 1.8 to 2 hours2.314815e-5 days <br />5.555556e-4 hours <br />3.306878e-6 weeks <br />7.61e-7 months <br /> 4.26 170 0.0893 0.0893 0.0893 2 to 3.8 hours9.259259e-5 days <br />0.00222 hours <br />1.322751e-5 weeks <br />3.044e-6 months <br /> 4.26 160 0.0893 0.0893 0.0893 3.8 to 4 hours4.62963e-5 days <br />0.00111 hours <br />6.613757e-6 weeks <br />1.522e-6 months <br /> 4.26 160 0.116 0.116 0.116 4 to 8 hours9.259259e-5 days <br />0.00222 hours <br />1.322751e-5 weeks <br />3.044e-6 months <br /> 4.26 140 0.116 0.116 0.116 8 to 13.8 hours9.259259e-5 days <br />0.00222 hours <br />1.322751e-5 weeks <br />3.044e-6 months <br /> 4.26 110 0.116 0.116 0.116 13.8 to 22.2 hours2.314815e-5 days <br />5.555556e-4 hours <br />3.306878e-6 weeks <br />7.61e-7 months <br /> 4.26 84 0.086 0.086 0.086 22.2 to 24 hours2.777778e-4 days <br />0.00667 hours <br />3.968254e-5 weeks <br />9.132e-6 months <br /> 4.26 84 0 0 0 1 to 2 days 4.26 64 0 0 0 2 to 4 days 4.26 48 0 0 0 4 to 30 days 4.26 25 0 0 0 Section 4.5.2 ESF Recirculation Loop Leakage The ESF recirculation loop circulates containment sump water outside of the containment via the SIS and CSS pumps following receipt of a RAS. Per RG 1.183 Appendix A Section 5, ESF systems that recirculate 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.

Regulatory Guide 1.183 Appendix A provides assumptions for use in evaluating the radiological consequences of the ESF recirculation loop leakage pathway for a LOCA. These assumptions supplement the guidance provided in the main body of RG 1.183.

This Section presents the assumptions, design input, and methodology employed in evaluating the radiological consequences of SONGS Units 2 and 3 LOCA ESF recirculation loop leakage path. The characteristics of the LOCA ESF recirculation loop leakage model are summarized in Table 4.5-8.

Page 66 of 110

The control room and offsite doses associated with ESF leakage are summarized in Section 4.5.7.

TABLE 4.5-8: LOCA ESF RECIRCULATION LOOP LEAKAGE ANALYSIS PARAMETERS LOCA ESF RECIRCULATION LOOP LEAKAGE PARAMETER MODELED VALUE Timing of core activity release into containment sump Gap release phase time zero l Early in-vessel phase time zero Core inventory fraction released into containment sump Noble gases (Xe, Kr) 0 Halogens (I, Br) 0.40 Alkali metals (Cs, Rb) 0.30 Tellurium metals (Te, Sb, Se) 0.05 Barium, Strontium (Ba, Sr) 0.02 Noble metals (Ru, Rh, Pd, Mo, Tc, Co) 0.0025 Cerium group (Ce, Pu, Np) 0.0005 Lanthanides (La, Zr, Nd, Eu, Nb, Pm, Pr, Sm, Y, Cm, Am) 0.0002 ESF recirculation loop dilution volume, cubic feet 46,647 ESF recirculation loop leakage rate to the environment, ft3/minute 0 to 20 minutes 0 20 minutes to 30 days 7.00E-03 ESF recirculation loop leakage flashing fractions and partition coefficients Iodine isotopes flashing fraction, percent 10 Noble gases (Xe, Kr) partition coefficient 1.OE-06 Particulate isotopes partition coefficient 1E+06 Chemical form of Iodine released to environment, percent of iodine Cesium iodide (Csl) (particulate iodine) 0 Elemental iodine 97 Organic iodide 3 ESF Leakage release point (Main Plant Vent) to Control Room per Section 4.4 Atmospheric Dispersion Factors, seconds/m Section 4.5.2.1 ESF Leakage Source Term Regulatory Guide 1.183 Appendix A Section 1 states that acceptable assumptions regarding core inventory and the release of radionuclides from the fuel are provided in RG 1.183 Regulatory Position 3.

Consistent with RG 1.183 Regulatory Position 3.4, the core isotopes released into the containment are grouped into chemically similar groups in accordance with RG 1.183 Table 5. The elements of each group are as listed in Table 4.5-8.

Consistent with RG 1.183 Regulatory Position 3.2, the core inventory release Page 67 of 110

fractions are applied to the equilibrium reactor core isotope inventory at shutdown as described in Section 4.1.

Consistent with RG 1.183 Appendix A Section 5.1, all fission products, with the exception of noble gases, released from the fuel to the containment during the gap and early in-vessel activity release phases are assumed to instantaneously and homogeneously mix in the primary containment sump coincident with the start of the LOCA. The dose analysis conservatively assumes that all fission product activity is released to the containment sump prior to the start of ESF recirculation operation at 20 minutes, which is sooner than the conclusion of the in-vessel activity release phase at 1.8 hours9.259259e-5 days <br />0.00222 hours <br />1.322751e-5 weeks <br />3.044e-6 months <br />.

The AST ESF leakage dose analysis assumes that the ESF recirculation loop dilutes the core activity release into a volume of 348,946 gallons (46,647 cubic feet). This volume represents the minimum containment emergency sump volume available at the start of the post-LOCA SIS and CSS recirculation mode of operation. The modeling of the minimum sump volume conservatively maximizes the ESF recirculation loop activity concentration.

Consistent with Regulatory Guide 1.183 Appendix A Section 2, with the exception of elemental and organic iodine and noble gases, fission products are assumed to be in particulate form (i.e., subject to HEPA filtration).

Section 4.5.2.2 ESF Leakage Activity Release Model Consistent with RG 1.183 Appendix A Section 5.2, the ESF leakage is assumed to start at the earliest time the recirculation flow starts in the ESF recirculation system, and end at the latest time the releases from these systems are terminated. Per the containment pressure/temperature response analysis for the LOCA event, the CSS and SIS recirculation mode of operation begins as early as 20.2 minutes after the start of the LOCA for a two-train recirculation mode. This start time has been conservatively rounded down to 20 minutes after the start of the LOCA for use in the AST ESF leakage dose analysis. The ESF leakage is assumed to continue for the duration of the 30-day LOCA event.

The maximum expected leakage rates from all components in the recirculation systems are 1,770 cc/hr from the HPSI system, 3,000 cc/hr from the LPSI system, and 1,180 cc/hr from the CSS. Consistent with RG Guide 1.183 Appendix A, Section 5.2, the modeled ESF recirculation loop leakage rate of 11,900 cc/hr (7.OOE-03 cfm) represents two times the sum of the simultaneous maximum expected leakage from these systems.

The ESF leakage AST dose analysis assumes that 10 percent of the iodine in the ESF Recirculation Loop leakage flashes to vapor and is therefore capable of migrating to the outside environment. The ten percent flashing fraction is consistent with the guidance of RG 1.183 Appendix A Section 5.5, which states Page 68 of 110

that if the water temperature is less than 212 Fahrenheit, then 10 percent of the iodine in the leakage is assumed to become airborne unless a smaller amount is justified based on actual sump pH history and ventilation rates. Per the containment pressure/temperature response analysis for the LOCA event, the temperature of the containment sump liquid has been reduced to below 212 Fahrenheit when the CSS and SIS recirculation mode of operation begins at 20 minutes.

Consistent with RG 1.183 Appendix A, Section 5.6, the composition of iodine available for release to the environment is assumed to be 97% elemental iodine, 3% organic iodide, and 0%particulate iodine.

Consistent with RG 1.183 Appendix A Section 5.3, all particulate isotopes in the recirculating liquid are retained in the liquid phase. Radioactive decay of some particulate isotopes in the recirculating liquid yields noble gas isotopes. In the dose program model, a particulate partition coefficient (i.e., liquid concentration divided by gas concentration) of 1.OE+06 is used in order to retain particulates within the recirculating liquid where noble gas daughters are then formed by decay.

Per RG 1.183 Appendix A Section 5.3, with the exception of iodine, all radioactive materials in the recirculating liquid are retained in the liquid phase.

This guidance implies that a release of noble gases (krypton and/or xenon) via the ESF leakage need not be considered (most likely because noble gas isotopes present in the containment sump water would be released into the containment air prior to being recirculated). However, noble gas isotopes are formed by the decay of other isotopes that are present in the ESF recirculating liquid. Therefore, the ESF leakage AST dose analysis conservatively assumes that 100 percent of the noble gases in the ESF Recirculation Loop leakage will become airborne (i.e., a noble gas partition coefficient of 1.OE-06).

Once airborne, the noble gases can migrate to the outside environment.

The activity released from the ESF leakage into the Penetration Areas and Safety Equipment Building is exhausted to the environment via the main plant vent, and transported by atmospheric dispersion to the control room HVAC intake and to the offsite EAB and LPZ dose receptors. Section 4.4 and its Table 4.4-11 present the San Onofre site-specific 95th percentile meteorology atmospheric dispersion factors used for the main plant vent to control room release pathway.

No credit is taken for radioactive decay of the isotopes during this atmospheric dispersion transit to the control room or offsite dose locations. Consistent with RG 1.183 Regulatory Positions 4.1.7 and 4.2.2, no correction is made for depletion of the effluent plume by deposition on the ground.

Page 69 of 110

Section 4.5.3 RWST Release The ESF recirculation loop circulates containment sump water outside of the containment via the SIS and CSS pumps following receipt of a RAS. Per RG 1.183 Appendix A Section 5, ESF systems that recirculate sump water outside of the primary containment may leak through valves isolating interfacing systems, including the RWST.

Regulatory Guide 1.183 Appendix A provides assumptions for use in evaluating the radiological consequences of ESF leakage to the RWST and its subsequent release pathway to the environment. These assumptions supplement the guidance provided in the main body of RG 1.183.

This Section presents the assumptions, design input, and methodology employed in evaluating the radiological consequences of SONGS Units 2 and 3 LOCA RWST release path. The characteristics of the LOCA RWST release model are summarized in Table 4.5-9.

The control room and offsite doses associated with the RWST release are summarized in Section 4.5.7.

TABLE 4.5-9: LOCA RWST RELEASE ANALYSIS PARAMETERS LOCA RWST RELEASE PARAMETER MODELED VALUE iming of core activity release into containment sump Gap release phase Gap elese pasetime instantaneous zero at Early in-vessel phase time zero Core inventory fraction released into containment sump Noble gases (Xe, Kr) 0 Halogens (1,Br) 0.40 Alkali metals (Cs, Rb) 0.30 Tellurium metals (Te, Sb, Se) 0.05 Barium, Strontium (Ba, Sr) 0.02 Noble metals (Ru, Rh, Pd, Mo, Tc, Co) 0.0025 Cerium group (Ce, Pu, Np) 0.0005 Lanthanides (La, Zr, Nd, Eu, Nb, Pm, Pr, Sm, Y, Cm, Am) 0.0002 ESF recirculation loop dilution volume, cubic feet 46,647 RWST dilution volumes, cubic feet per tank RWST air region volume 35,880 RWST water region volume 7,345 Page 70 of 110

LOCA RWST RELEASE PARAMETER MODELED VALUE RWST inflow due to ESF pump mini-flow isolation valve leakage Mini-flow valve leakage flashing fraction, percent 10 Mini-flow valve leakage rates, cf m Total mini-flow leakage rate to the RWST 0 to 20 minutes 0 20 minutes to 30 days 0.4010 Mini-flow valve leakage rate to the RWST water region 0 to 20 minutes 0 20 minutes to 30 days 0.3609 Mini-flow valve leakage rate to the RWST air region 0 to 20 minutes 0 20 minutes to 30 days 0.0401 RWST water region inflow due to RWST discharge check valve leakage, cfm 0 to 1.08 hours9.259259e-5 days <br />0.00222 hours <br />1.322751e-5 weeks <br />3.044e-6 months <br /> 0 1.08 to 2 hours2.314815e-5 days <br />5.555556e-4 hours <br />3.306878e-6 weeks <br />7.61e-7 months <br /> 1.2859 2 to 8 hours9.259259e-5 days <br />0.00222 hours <br />1.322751e-5 weeks <br />3.044e-6 months <br /> 1.2778 8 to 24 hours2.777778e-4 days <br />0.00667 hours <br />3.968254e-5 weeks <br />9.132e-6 months <br /> 0.9622 24 to 96 hours0.00111 days <br />0.0267 hours <br />1.587302e-4 weeks <br />3.6528e-5 months <br /> 0.5103 96 to 119.72 hours8.333333e-4 days <br />0.02 hours <br />1.190476e-4 weeks <br />2.7396e-5 months <br /> 0.1078 119.72 to 720 hours0.00833 days <br />0.2 hours <br />0.00119 weeks <br />2.7396e-4 months <br /> 0 Partition coefficients between ESF recirculation loop and RWST air region Iodine 1.0 Particulates 1E+06 Mixing and partition coefficients between RWST air and water regions Mixing rate between RWST air and water regions, cfm 0.4010 Iodine partition coefficient 200 Noble gases (Xe, Kr) partition coefficient 1E-06 Particulates partition coefficient 1E+06 RWST air region release rate to the environment, cfm 0 to 20 minutes 0 20 minutes to 1.08 hours9.259259e-5 days <br />0.00222 hours <br />1.322751e-5 weeks <br />3.044e-6 months <br /> 0.4010 1.08 to 2 hours2.314815e-5 days <br />5.555556e-4 hours <br />3.306878e-6 weeks <br />7.61e-7 months <br /> 1.6869 2 to 8 hours9.259259e-5 days <br />0.00222 hours <br />1.322751e-5 weeks <br />3.044e-6 months <br /> 1.6788 8 to 24 hours2.777778e-4 days <br />0.00667 hours <br />3.968254e-5 weeks <br />9.132e-6 months <br /> 1.3632 24 to 96 hours0.00111 days <br />0.0267 hours <br />1.587302e-4 weeks <br />3.6528e-5 months <br /> 0.9113 96 to 119.72 hours8.333333e-4 days <br />0.02 hours <br />1.190476e-4 weeks <br />2.7396e-5 months <br /> 0.5088 119.72 to 720 hours0.00833 days <br />0.2 hours <br />0.00119 weeks <br />2.7396e-4 months <br /> 0.4010 Chemical form of Iodine released to environment, percent of iodine Cesium iodide (CsI) (particulate iodine) 0 Elemental iodine 97 Organic iodide 3 RWST release via RWST vent to Control Room Atmospheric Dispersion per Section 4.4 Factors, seconds/in 3 Page 71 of 110

Section 4.5.3.1 RWST Release Source Term The RWST release source term is the same as the ESF leakage source term described in Section 4.5.2.1.

ESF back-leakage into the RWST occurs after the RWST water level has drained down, and a RAS has been generated. In evaluating the post-LOCA RWST release AST dose consequences, the most conservative scenario for leakage out of the RWST is when the RWST air volume is minimized (and the water volume is maximized). Minimizing the air volume will minimize dilution of the activity entering the RWST air space prior to its release to the atmosphere.

The RWST level setpoint for a RAS is 18.5% + 3.8%. The minimum RWST air volume occurs when RAS is quickly initiated at the 22.3% upper tolerance of the RWST level setpoint. At this setpoint, the RWST minimum air volume is 35,880 cubic feet, and the RWST maximum water volume is 7,345 cubic feet.

Section 4.5.3.2 RWST Activity Release Model During the ESF recirculation operation, ESF leakage may enter the RWST via various pathways. The pathways are dependent on whether the LOCA occurs with or without an assumed Diesel Generator (DG) failure. Due to the time dependency associated with the pathway leakage rates, the Current Licensing Basis reports the maximum 0 to 2 hour2.314815e-5 days <br />5.555556e-4 hours <br />3.306878e-6 weeks <br />7.61e-7 months <br /> EAB dose for the scenario of the post-LOCA RWST release with an assumed DG failure, and the maximum event duration control room and LPZ doses for the scenario of the post-LOCA RWST release without an assumed DG failure. The AST RWST release dose analysis evaluated the two leakage scenarios and determined that the variable 2-hour window EAB dose and the event duration control room and LPZ doses are more severe for the scenario of the post-LOCA RWST release without an assumed DG failure rather than with an assumed DG failure. The characteristics of the RWST release without an assumed DG failure are detailed in this section.

Two ESF leakage pathways to the RWST characterize the scenario of a RWST release without an assumed DG failure. The first pathway is RWST air region inflow due to ESF pump minimum flow (mini-flow) isolation valve leakage. The second pathway is RWST water region inflow due to RWST discharge check valve leakage. ESF leakage to the RWST for potential release paths with three or more normally closed isolation valves in series is assumed to be negligible.

Section 4.5.3.2.1 RWST Inflow Due to ESF Pump Mini-Flow Isolation Valve Leakage The SIS and CSS pumps minimum flow return paths to the RWST are isolated following a RAS by two sets of 4-inch mini-flow isolation valves. Valves Page 72 of 110

1204-HV-9306 and 9307 are in series in one flow path, and valves 1204-HV-9347 and 9348 are in series in a second flow path. The maximum allowable leakage rate for each of the valves is 0.75 gpm. Consequently, for either path with its two valves in series, the maximum allowable path leakage rate is assumed to be 0.75 gpm. For the scenario of a LOCA without an assumed DG failure, the total leakage rate past the valves in the two parallel flow paths is 1.5 gpm. Consistent with RG 1.183 Appendix A, Section 5.2, the modeled ESF recirculation loop leakage rate of 3.0 gpm (0.4010 cfm) represents two times the sum of the simultaneous maximum allowable leakage from these systems. The 0.4010 cfm leakage past the mini-flow isolation valves is released into the RWST air space.

Consistent with RG 1.183 Appendix A Section 5.2, ESF leakage to the RWST is assumed to start at the earliest time the recirculation flow starts in the ESF recirculation system, and end at the latest time the releases from these systems are terminated. Per the containment pressure/temperature response analysis for the LOCA event, the CSS and SIS recirculation mode of operation begins as early as 20.2 minutes after the start of the LOCA for a two-train recirculation mode. This start time has been conservatively rounded down to 20 minutes after the start of the LOCA for use in the AST RWST release dose analysis. The ESF leakage and coincident RWST release to the environment are assumed to continue for the duration of the 30-day LOCA event.

The RWST release AST dose analysis assumes that 10 percent of the iodine in the ESF Recirculation Loop leakage into the RWST air space flashes to vapor.

The ten percent flashing fraction is consistent with the guidance of RG 1.183 Appendix A Section 5.5, which states that if the water temperature is less than 212 Fahrenheit, then 10 percent of the iodine in the leakage is assumed to become airborne unless a smaller amount is justified based on actual sump pH history and ventilation rates. Per the containment pressure/temperature response analysis for the LOCA event, the temperature of the containment sump liquid has been reduced to below 212 Fahrenheit when the CSS and SIS recirculation mode of operation begins at 20 minutes.

To address the 10 percent ESF leakage flashing, 10 percent of the ESF pump mini-flow leakage is modeled as entering the RWST air region, and 90 percent of the ESF pump mini-flow leakage is modeled as entering the RWST water region.

Section 4.5.3.2.2 RWST Inflow Due to RWST Discharge Check Valve Leakage Following a RAS, each of the two RWSTs is isolated from the ESF recirculation loop by an RWST isolation valve located near the RWST. In the event that the isolation valve fails to close following a RAS, an RWST discharge check valve (1204-MU-001 or 1204-MU-002) is the only barrier between the containment and the RWST.

Page 73 of 110

The maximum allowable leakage rate through both check valves is 5 gpm.

Consistent with RG 1.183 Appendix A, Section 5.2, the modeled ESF recirculation loop leakage rate of 10 gpm represents two times the maximum allowable leakage. The leakage past the check valves is released into the RWST water space.

Should the check valve leak, a certain amount of containment sump water will enter the line and start moving toward the RWST. The flow rate is simply the leak rate of the check valve, but the speed at which the hot water from the sump moves toward the RWST depends on the pressure differential between the post-LOCA containment and the RWST and on the temperature difference between the cold RWST water and the hot containment sump water.

A valve seat leak analysis has determined that with the RWST isolation valve open and the check valve leaking at 10 gpm, the sump water will reach the RWST isolation valve and enter the RWST in approximately 1.08 hours9.259259e-5 days <br />0.00222 hours <br />1.322751e-5 weeks <br />3.044e-6 months <br /> after the start of the LOCA. Neglecting the effect of thermal buoyancy (i.e., assuming instant cooling of the leaking sump water to the RWST temperature) would extend the travel time to approximately 8.7 hours8.101852e-5 days <br />0.00194 hours <br />1.157407e-5 weeks <br />2.6635e-6 months <br />.

As the LOCA event progresses, the containment pressure decreases, and the leakage rate past the check valve decreases. Table 4.5-9 summarizes the time-dependent RWST water region inflow due to RWST discharge check valve leakage. The RWST water region inflow due to RWST discharge check valve leakage is maximized by modeling the post-LOCA containment pressure and temperature history that yields the highest containment pressure. Eventually, the driving force for the check valve leakage ends.

Section 4.5.3.2.3 Mixing of RWST Water and Air Activity Inventories Because the iodine concentrations in the RWST air space and water space are not at their equilibrium values at the onset of ESF recirculation loop leakage, the mixing rate between the RWST water and air regions will impact the dose consequences.

At equilibrium conditions, the ratio of the iodine concentration in the RWST water space to the iodine concentration in the RWST air space is defined by an iodine partition coefficient (i.e., liquid concentration divided by gas concentration). Per NUREG/CR-0009, the partition coefficient is numerically determined by the physical sorption of 12 and by rapid ionization reactions that occur in solution. Per NUREG/CR-0009 page 61, for borated solutions (such as the RWST) a partition coefficient of 200 is achievable. Therefore, a partition coefficient of 200 can be used to establish the equilibrium between the iodine concentrations in the RWST air and water spaces.

Page 74 of 110

As RWST water and air region mixing occurs, the ratio of the iodine concentration in the RWST water space to the iodine concentration in the RWST air space tends to a partition coefficient of 200. Mixing is facilitated by the turbulence added to the RWST water by the RWST discharge check valve leakage into the RWST water space, by the turbulence added to the RWST water by the mini-flow valve leakage into the RWST air space that drops down into the RWST water, and by the thermal gradients associated with the RWST water warmed by the introduction of the hot ESF leakage fluid relative to colder RWST air. Per NUREG/CR-0009 page 64, very small thermal sources are sufficient to mix a large vessel.

The higher the mixing rate between the air space and water space, the more rapid the approach to equilibrium. Initially the RWST air activity concentration is greater than when at equilibrium conditions. Minimizing the mixing rate delays the transfer of iodine activity from the RWST air region to the RWST water region, thereby conservatively maximizing the RWST air activity available for leakage to the environment. Inthe RWST leakage AST analysis, the mixing rate between the RWST air and water spaces is conservatively modeled as the mini-flow valve leakage rate into the RWST. No credit is taken for RWST air and water space mixing associated with the RWST inflow due to RWST discharge check valve leakage. No credit is taken for mixing by thermal gradients.

Consistent with RG 1.183 Appendix A Section 5.3, all particulate isotopes in the recirculating liquid are retained in the liquid phase. Radioactive decay of some particulate isotopes in the RWST liquid space yields noble gas isotopes. A particulate partition coefficient (i.e., liquid concentration divided by gas concentration) of 1.OE+06 is used in order to retain particulates within the RWST liquid space where noble gas daughters are then formed by decay.

Per RG 1.183 Appendix A Section 5.3, with the exception of iodine, all radioactive materials in the recirculating liquid are retained in the liquid phase.

This guidance implies that a release of noble gases (krypton and/or xenon) via the RWST release need not be considered (most likely because noble gas isotopes present in the containment sump water would be released into the containment air prior to being recirculated). However, noble gas isotopes are formed by the decay of other isotopes that are present in the RWST liquid.

Therefore, the RWST release AST dose analysis conservatively assumes that 100 percent of the noble gases formed by the decay of the isotopes in the RWST liquid will become airborne (i.e., a noble gas partition coefficient of 1.OE-06).

Once airborne, the noble gases can migrate to the outside environment.

Section 4.5.3.2.4 RWST Releases to the Environment Consistent with RG 1.183 Appendix A, Section 5.6, the composition of iodine available for release to the environment from the RWST air space is assumed to be 97% elemental iodine, 3% organic iodide, and 0% particulate iodine.

Page 75 of 110

The activity released from the RWST is exhausted to the environment via the RWST vent, and transported by atmospheric dispersion to the control room HVAC intake and to the offsite EAB and LPZ dose receptors. Section 4.4 and its Table 4.4-11 present the San Onofre site-specific 95th percentile meteorology atmospheric dispersion factors used for the RWST vent to control room release pathway. No credit is taken for radioactive decay of the isotopes during this atmospheric dispersion transit to the control room or offsite dose locations.

Consistent with RG 1.183 Regulatory Positions 4.1.7 and 4.2.2, no correction is made for depletion of the effluent plume by deposition on the ground.

Section 4.5.4 Post-Accident Sampling System (PASS) Leakage PASS licensing requirements were deleted from the SONGS Units 2 and 3 Technical Specifications per Unit 2 License Amendment 178 and Unit 3 License Amendment 169. Currently, the PASS is maintained at SONGS Units 2 and 3 for severe accident management only. The PASS is capable of analyzing samples of containment air, containment sump water, and reactor coolant.

Until such time that the PASS is isolated from post-LOCA radiation sources, portions of the PASS that are outside of the containment present the potential for a release path due to leakage through valve packing glands, pump shaft seals, flanged connections, and other similar components. Should a design modification be implemented to isolate PASS and thereby eliminate the potential release paths, the dose contribution from PASS leakage will be removed from the LOCA dose analysis.

Although RG 1.183 Appendix A does not provide a specific requirement to evaluate PASS leakage, the SONGS Units 2 and 3 AST dose analysis has applied related RG 1.183 Appendix A ESF leakage guidance to an evaluation of the radiological consequences of the PASS leakage activity release path for a LOCA.

This Section presents the assumptions, design input, and methodology employed in evaluating the radiological consequences of SONGS Units 2 and 3 LOCA PASS leakage path. The characteristics of the LOCA PASS leakage model are summarized in Table 4.5-10.

The control room and offsite doses associated with PASS leakage are summarized in Section 4.5.7.

Page 76 of 110

TABLE 4.5-10: LOCA PASS LEAKAGE ANALYSIS PARAMETERS LOCA PASS LEAKAGE PARAMETER MODELED VALUE]

Timing of core activity release into reactor coolant system Gap release phase instantaneous at time zero Early in-vessel phase instantaneous at Core inventory fraction released into reactor coolant system Noble gases (Xe, Kr) 1.00 Halogens (I, Br) 0.40 Alkali metals (Cs, Rb) 0.30 Tellurium metals (Te, Sb, Se) 0.05 Barium, Strontium (Ba, Sr) 0.02 Noble metals (Ru, Rh, Pd, Mo, Tc, Co) 0.0025 Cerium group (Ce, Pu, Np) 0.0005 Lanthanides (La, Zr, Nd, Eu, Nb, Pm, Pr, Sm, Y, Cm, Am) 0.0002 Reactor coolant system dilution volume, cubic feet 10,179 PASS leakage rate to the environment, ft3/minute 0 to 30 minutes 0 30 minutes to 30 days 4.12E-04 PASS leakage partition coefficients and flashing fractions Iodine isotopes flashing fraction, percent 10 Noble gases (Xe, Kr) partition coefficient 1.OE-06 Particulate isotopes partition coefficient 1E+06 Chemical form of Iodine released to environment, percent of iodine Cesium iodide (Csl) (particulate iodine) 0 Elemental iodine 97 Organic iodide 3 PASS Leakage release point (Main Plant Vent) to Control Room Per Section 4.4 Atmospheric Dispersion Factors, seconds/m 3 PerISection_4.4 Section 4.5.4.1 PASS Leakage Source Term The PASS samples containment sump liquid, reactor coolant, and containment air. The PASS leakage AST dose analysis assumes that reactor coolant is the PASS fluid leaking during the LOCA. Reactor coolant has been modeled since the reactor coolant iodine activity concentration is greater than the containment sump liquid or containment air iodine activity concentration.

Regulatory Guide 1.183 Appendix A Section 1 states that acceptable assumptions regarding core inventory and the release of radionuclides from the fuel are provided in RG 1.183 Regulatory Position 3.

Consistent with RG 1.183 Regulatory Position 3.4, the core isotopes released into the reactor coolant are grouped into chemically similar groups in accordance with RG 1.183 Table 5. The elements of each group are as listed in Page 77 of 110

Table 4.5-10. Consistent with RG 1.183 Regulatory Position 3.1, the core inventory release fractions are applied to the equilibrium reactor core isotope inventory at shutdown as described in Section 4.1.

All fission products released from the fuel to the reactor coolant are assumed to instantaneously and homogeneously mix in the reactor coolant coincident with the start of the LOCA.

The AST PASS leakage dose analysis assumes that the reactor coolant dilutes the core activity release into a volume of 10,179 cubic feet. This volume has conservatively omitted the primary side volume that is present in an assumed 2,000 plugged U-tubes in each of the two original steam generators.

Consistent with Regulatory Guide 1.183 Appendix A Section 2, with the exception of elemental and organic iodine and noble gases, fission products are assumed to be in particulate form (i.e., subject to HEPA filtration).

Section 4.5.4.2 PASS Leakage Activity Release Model Consistent with RG 1.183 Appendix A Section 5.2, the PASS leakage is assumed to start at the earliest time that PASS sampling could start, and end at the latest time the releases from the PASS are terminated. Although the PASS is currently maintained for severe accident management only, this analysis assumes the PASS leakage begins 30 minutes after the start of the LOCA.

Chemistry Procedures address the steps associated with collecting a post-LOCA reactor coolant sample. One procedure requires the Nuclear Chemistry Technician to obtain permission from the affected unit Control Operator to operate the PASS. A second procedure addresses steps that must be taken prior to operation of the PASS during accident conditions. These steps include requesting that Health Physics survey the PASS laboratory, assembling the Chemistry sampling team, and implementing health physics requirements.

These actions should require more than the 30 minute delay assumed prior to PASS operation. The PASS leakage is assumed to continue for the duration of the 30-day LOCA event.

The maximum expected leakage rate from all components in the PASS is 350 cc/hr. Consistent with RG 1.183 Appendix A, Section 5.2, the modeled PASS leakage rate of 700 cc/hr (4.12E-04 cfm) represents two times the maximum expected leakage from the PASS.

The PASS leakage AST dose analysis assumes that 10 percent of the iodine in the PASS leakage flashes to vapor and is therefore capable of migrating to the outside environment. The ten percent flashing fraction is consistent with the guidance of RG 1.183 Appendix A Section 5.5, which states that ifthe water temperature is less than 212 Fahrenheit, then 10 percent of the iodine in the leakage is assumed to become airborne. Per the PASS Technical Manual, Page 78 of 110

Sample Vessel Heat Exchanger SA1 212ME752 is used to cool the reactor coolant sample flow from the maximum reactor coolant temperature to allow for low temperature (120 Fahrenheit) sample analysis. The bulk of the PASS reactor coolant leakage from PASS sample station fittings will be at a low temperature.

Consistent with RG 1.183 Appendix A, Section 5.6, the composition of iodine available for release to the environment is assumed to be 97% elemental iodine, 3% organic iodide, and 0% particulate iodine.

Consistent with RG 1.183 Appendix A Section 5.3, all particulate isotopes in the reactor coolant are retained in the liquid phase. Radioactive decay of some particulate isotopes in the reactor coolant yields noble gas isotopes. Inthe Bechtel LocaDose code model, a particulate partition coefficient (i.e., liquid concentration divided by gas concentration) of 1.OE+06 is used in order to retain particulates within the reactor coolant where noble gas daughters are then formed by decay.

The PASS leakage AST dose analysis conservatively assumes that 100 percent of the noble gases (krypton and/or xenon) either initially present or formed byithe decay of the isotopes in the PASS reactor coolant leakage will become airborne (i.e., a noble gas partition coefficient of 1.OE-06) and migrate to the outside environment.

The activity released from the PASS leakage into the Radwaste Building is exhausted to the environment via the main plant vent, and transported by atmospheric dispersion to the control room HVAC intake and to the offsite EAB and LPZ dose receptors. Section 4.4 and its Table 4.4-11 present the San Onofre site-specific 95th percentile meteorology atmospheric dispersion factors used for the main plant vent to control room release pathway. No credit is taken for radioactive decay of the isotopes during this atmospheric dispersion transit to the control room or offsite dose locations. Consistent with RG 1.183 Regulatory Positions 4.1.7 and 4.2.2, no correction is made for depletion of the effluent plume by deposition on the ground.

Section 4.5.5 LOCA EAB and LPZ Model Regulatory Guide 1.183 Regulatory Position 4.1 provides guidance to be used in determining TEDE for persons located at or beyond the boundary of the exclusion area, including the outer boundary of the low population zone.

Section 4.2 addresses the applicability of this guidance to the SONGS Units 2 and 3 AST LOCA dose analysis as it relates to the offsite dose exposure parameters.

As discussed in Section 4.2, the LOCA dose analysis considers the dose consequences of inhalation and immersion. Radioactive material contained in a Page 79 of 110

plant structure is assumed to be a negligible radiation shine source to the offsite dose receptors relative to the dose associated with immersion in the radioactive plume released from the facility.

Consistent with RG 1.183 Regulatory Positions 4.1.5, 4.1.6 and 4.4 and Table 6, the LOCA event radiological criterion for the EAB and for the outer boundary of the LPZ is 25 Rem TEDE.

Section 4.5.6 LOCA Control Room Model Regulatory Guide 1.183 Regulatory Position 4.2 provides guidance to be used in determining the total effective dose equivalent for persons located in the control room. Section 4.3 addresses the applicability of this guidance to the SONGS Units 2 and 3 AST LOCA dose analysis as it relates to the control room dose exposure parameters.

The CREACUS Emergency mode of operation can be actuated either manually or automatically following a CRIS. The CRIS may be generated automatically by a SIAS or by the detection of high radioactivity concentrations in the control room, outside air inflow. Per Section 4.3.2.1.1, the LOCA model credits CREACUS Emergency mode of operation initiation at time zero (i.e., prior to the arrival of any contaminated air reaching the control room outside air intakes) due to a SIAS-induced CRIS.

As discussed in Section 4.3, the LOCA dose analysis considers the dose consequences of inhalation, immersion, and radiation shine from the environmental (or outside) cloud, the control room emergency HVAC filters, the post-LOCA piping, and the containment building.

Consistent with RG 1.183 Regulatory Position 4.4, as an AST dose analysis acceptance criterion the postulated control room dose is evaluated to ensure that it does not exceed the 5 Rem TEDE criterion established in 10 CFR 50.67.

Section 4.5.7 LOCA Dose Consequences The resulting LOCA offsite and control room operator doses are listed in Tables 4.5-11 and 4.5-12. The analysis demonstrates that the LOCA event 25 Rem TEDE radiological criterion for the EAB and for the outer boundary of the LPZ is met. The analysis also demonstrates that the LOCA event 5 Rem TEDE radiological criterion for the control room is met.

Page 80 of 110

TABLE 4.5-11: LOCA RELEASE PATH DOSE CONSEQUENCES CONTAINMENT ESF RWST PASS PIPING DOSE RECEPTOR LEAKAGE LEAKAGE RELEASE LEAKAGE SHINE DSREETRDOSE DOSE DOSE DOSE DOSE (REM TEDE) (REM TEDE) (REM TEDE) (REM TEDE) (REM TEDE)

Control Room (30-day accident duration)

Immersion and Inhalation 7.225E-01 4.697E-01 8.420E-01 1.461 E-01 Control Room Filter Shine 1.446E-01 7.55E-02 1.263E-01 2.02E-02 Environmental Cloud Shine 4.373E-02 3.44E-03 1.1 34E-03 3.31 E-03 Containment Building Shine 4.304E-04 Piping Shine 1.06E-01 TOTAL 9.112E-01 5.487E-01 9.694E-01 1.696E-01 1.06E-01 Exclusion Area Boundary 3.548E+00 3.398E-01 1.103E+00 1.370E-01 (Maximum 2-hour dose) (0.6 to 2.6 hrs) (0.4 to 2.4 hrs) (94 to 96 hrs) (0.5 to 2.5 hrs) _

(30-day accident duration) 2.309E-01 2.381 E-01 1.311 E+00 6.897E-02 l TABLE 4.5-12: LOCA DOSE CONSEQUENCES LOCA ACCEPTANCE 1 DOSE RECEPTOR DOSE CRITERION 1 (REM TEDE) (REM TEDE)_l Control Room (30-day accident duration) 2.7 l 5 EAB (Maximum 2-hour dose) 5.1 25 LPZ (30-day accident duration) 1.8 25 Page 81 of 110

Section 4.6 FUEL HANDLING ACCIDENT INSIDE CONTAINMENT (FHA-IC)

ANALYSIS Regulatory Guide 1.183 Appendix B provides assumptions for use in evaluating the radiological consequences of an FHA-IC. These assumptions supplement the guidance provided in the main body of RG 1.183.

This Section presents the assumptions, design input, methodology employed in evaluating, and the radiological consequences of, the SONGS Units 2 and 3 FHA-IC. The characteristics of the FHA-IC model are summarized in Table 4.6-1.

TABLE 4.6-1: FHA-IC ANALYSIS PARAMETERS FHA-IC PARAMETER MODELED VALUE Dose acceptance criteria, Rem TEDE Control Room 5 EAB 6.3 LPZ 6.3 FHA-IC source term Decay time after reactor shutdown, hours 72 Average fuel rod isotope inventory at 72 hours8.333333e-4 days <br />0.02 hours <br />1.190476e-4 weeks <br />2.7396e-5 months <br />, curies/rod per Section 4.1 Radial peaking factor applied to all failed fuel rods 1.75 Number of failed fuel rods - in dropped fuel bundle 16 Number of failed fuel rods - in impacted fuel bundles 210 Core fission product fractions in fuel rod gaps lodine-1 31 0.08 Krypton-85 0.10 Other noble gases (Krypton, Xenon) 0.05 Other Halogens (Iodine, Bromine) 0.05 Alkali Metals (Cesium, Rubidium) 0.12 Fraction of gap activity released to the refueling water 1.00 Minimum water depth above reactor vessel flange (and above the damaged fuel rods), feet 23 Refueling water decontamination factors lodines (effective DF) 200 Noble Gases 1 Particulates Infinite Iodine composition above the refueling water, percent of iodine Elemental iodine 57 Organic iodide 43 Page 82 of 110

FHA-IC PARAMETER MODELED VALUE Containment model Containment dome air circulators not modeled Engineered Safety Features Actuation System (ESFAS)

- containment purge isolation signal (CPIS) not modeled ESFAS - containment isolation actuation signal (CIAS) not modeled Containment personnel airlock closure not modeled Containment equipment hatch closure not modeled Activity release duration from containment, hours 2 Containment net free volume without dome, cubic feet 1.422E+06 Containment air exhaust flow rate, ft3 /minute 82,000 Offsite dose evaluation model per Section 4.2 Control Room dose evaluation model per Section 4.3 FHA-IC Release Points to Control Room Atmospheric Dispersion Factors, seconds/i 3 per Section 4.4 Section 4.6.1 FHA-IC Source Term The fuel handling accident inside containment involves the inadvertent dropping of a fuel assembly during fuel handling operations inside the reactor vessel, and the consequent rupture of fuel pins in both the dropped and impacted fuel assemblies. Consistent with RG 1.183 Appendix B Section 1.1, the number of fuel rods damaged during the accident is based on a conservative analysis that considers the most limiting case. UFSAR Section 15.10.7.3.9 details the structural evaluation of dropped fuel assembly damage. Per the UFSAR, a maximum of 226 fuel rods will fail as a result of a vertical drop of a fuel assembly on to the fuel bundles remaining in the partially loaded core. The 226 failed fuel rods represent 16 failed fuel rods in the dropped fuel assembly and 210 failed fuel rods in the impacted fuel assemblies.

Table 4.1-3 presents the fission product inventory of an average fuel rod in the reactor core. Consistent with the guidance of RG 1.183 Regulatory Position 3.1, to account for differences in power level across the core, a radial peaking factor of 1.75 is applied to the Table 4.1-3 average fuel rod isotope inventory to determine the activity inventory in each of the 226 failed fuel rods as described in Section 4.1.3. Consistent with RG 1.183 Regulatory Position 3.1, the FHA-IC 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 prior to the event, which is consistent with the minimum decay time required by SONGS Units 2 and 3 LCS 3.9.101 prior to movement of irradiated fuel in the reactor vessel.

Consistent with RG 1.183 Appendix B Section 1.2, the fission product release from the breached fuel is based on RG 1.183 Regulatory Position 3.2.

Consistent with RG 1.183 Footnote 11, the release fractions are acceptable for use since the fuel has a peak burnup of less than 62,000 MWD/MTU, and a Page 83 of 110

maximum linear heat generation rate that does not exceed 6.3 kw/ft peak rod average power for burnups exceeding 54 GWD/MTU.

All gap activity in the damaged rods is instantaneously released into the refueling water. Radionuclides that are considered are xenons, kryptons, iodines, bromines, cesiums, and rubidiums. Cesium and rubidium are particulates that are retained in the refueling pool water; therefore, these radionuclides do not contribute to the FHA doses.

Consistent with RG 1.183 Appendix B Section 1.3, the chemical form of radioiodine released from the fuel to the refueling water is assumed to be 95 percent cesium iodide (CsI), 4.85 percent elemental iodine, and 0.15 percent organic iodide. The Csl released from the fuel is assumed to completely dissociate in the refueling water and instantaneously re-evolve as elemental iodine. Consequently, the chemical form of radioiodine in the refueling water, prior to application of a decontamination factor, is 99.85 percent elemental iodine and 0.15 percent organic iodide.

Per Units 2 and 3 Technical Specification LCO 3.9.6, during movement of irradiated fuel assemblies within containment, the refueling water level above the top of the reactor vessel flange shall be greater than or equal to 23 feet. Since the damaged fuel assemblies would be lower than the reactor vessel flange, the water depth above the damaged fuel would be greater than 23 feet. Consistent with RG 1.183 Appendix B Section 2, the 23 foot water depth requirement allows for an overall effective decontamination factor of 200 (i.e., 99.5% of the total iodine released from the damaged rods is retained by the water). The difference in decontamination factors for elemental (99.85%) and organic iodine (0.15%)

species results in the iodine above the water being composed of 57% elemental and 43% organic species.

Consistent with RG 1.183 Appendix B Section 3, the retention of noble gases in the refueling water is negligible (i.e., decontamination factor of 1). Particulate radionuclides are assumed to be retained by the refueling water (i.e., infinite decontamination factor).

Section 4.6.2 FHA-IC Activity Release Model Per SONGS Units 2 and 3 TS LCO 3.9.3, the containment personnel airlock may, under certain conditions be open during core alterations and movement of irradiated fuel in containment. In addition, SONGS Units 2 and 3 have submitted license amendment request PCN-534 that will allow the containment equipment hatch to be open during Mode 6 core alterations and movement of irradiated fuel in containment. Consistent with RG 1.183 Appendix B Section 5.3, since the containment may be open during fuel handling operations, the radioactive material that escapes from the refueling water to the containment is assumed to Page 84 of 110

be released to the environment over a 2-hour time period (i.e., containment closure is not modeled during the FHA-IC event).

Consistent with the 2-hour release model requirement, the FHA-IC AST dose analysis does not model the generation of an Engineered Safety Features Actuation System (ESFAS) containment purge isolation signal (CPIS) or containment isolation actuation signal (CIAS). The containment purge is assumed to remain operational throughout the FHA-IC event. The containment personnel airlock and the containment equipment hatch are assumed to remain open throughout the FHA-IC event.

The containment air volume dilutes the gaseous activity released from the damaged fuel rods. During Mode 6 refueling operations there is no SONGS Units 2 and 3 TS requirement for the containment dome air circulators or containment cooling train fans to be operable. Therefore, no credit is taken for activity dilution within the air of the containment dome space.

The FHA-IC AST dose analysis does not model a reduction in the amount of radioactive material available for release from the containment by any containment engineered safety feature. In addition, airborne activity removal by containment purge filters is not credited.

The release of activity to the environment within the required 2-hour time period is established by specifying a containment air exhaust flow rate that ensures that at least 99.9 percent of the airborne activity will be released to the environment.

Activity released during the FHA-IC event is transported by atmospheric dispersion to the control room HVAC intake and to the offsite EAB and LPZ dose receptors. Activity may be released to the environment via the containment purge system or as leakage through containment penetrations, including the containment personnel airlock or the containment equipment hatch. Leakage from the containment personnel airlock would be exhausted via the main plant vent. Table 4.6-2 presents the San Onofre site-specific 95th percentile meteorology atmospheric dispersion factors for these release pathways as discussed in Section 4.4. Since one set of atmospheric dispersion factors does not consistently yield less dispersion than the others over time, a composite maximum of the three release points is utilized for assessing control room dose consequences. No credit is taken for radioactive decay of the isotopes during atmospheric dispersion transit to the control room or offsite dose locations.

Consistent with RG 1.183 Regulatory Positions 4.1.7 and 4.2.2, no correction is made for depletion of the effluent plume by deposition on the ground.

Page 85 of 110

TABLE 4.6-2: FHA-IC CR ATMOSPHERIC DISPERSION FACTORS FHA-IC to CR 95th Percentile Atmospheric Dispersion Factors (seconds/m 3 )

Containment Equipment Plant Vent Modeled Time Interval Shell Hatch Stack Vale Release Point Release Point Release Point Value 0 to 2 hours2.314815e-5 days <br />5.555556e-4 hours <br />3.306878e-6 weeks <br />7.61e-7 months <br /> 9.94E-04 7.99E-04 1.14E-03 1.14E-03 2 to 8 hours9.259259e-5 days <br />0.00222 hours <br />1.322751e-5 weeks <br />3.044e-6 months <br /> 6.32E-04 6.30E-04 6.11 E-04 6.32E-04 8 to 24 hours2.777778e-4 days <br />0.00667 hours <br />3.968254e-5 weeks <br />9.132e-6 months <br /> 1.77E-04 1.77E-04 2.10E-04 2.10E-04 1 to 4 days 2.34E-04 2.23E-04 2.20E-04 2.34E-04 4 to 30 days 2.18E-04 2.03E-04 1.98E-04 2.18E-04 Section 4.6.3 FHA-IC EAB and LPZ Model Regulatory Guide 1.183 Regulatory Position 4.1 provides guidance to be used in determining the TEDE for persons located at or beyond the boundary of the exclusion area, including the outer boundary of the low population zone.

Section 4.2 of this license amendment request addresses the applicability of this guidance to the SONGS Units 2 and 3 AST FHA-IC dose analysis as it relates to.

the offsite dose exposure parameters.

As discussed in Section 4.2, the FHA-IC dose analysis considers the dose consequences of inhalation and immersion. Radioactive material in the containment is assumed to be a negligible radiation shine source to the offsite dose receptors relative to the dose associated with immersion in the radioactive plume released from the facility.

Consistent with RG 1.183 Regulatory Positions 4.1.5, 4.1.6 and 4.4 and Table 6, the FHA-IC event radiological criterion for the EAB and for the outer boundary of the LPZ is 6.3 Rem TEDE.

Section 4.6.4 FHA-IC Control Room Model RG 1.183 Regulatory Position 4.2 provides guidance to be used in determining the total effective dose equivalent for persons located in the control room.

Section 4.3 addresses the applicability of this guidance to the SONGS Units 2 and 3 AST FHA-IC dose analysis as it relates to the control room dose exposure parameters.

The CREACUS Emergency mode of operation can be actuated either automatically following a CRIS or manually. The CRIS may be generated automatically by a SIAS or by the detection of high radioactivity concentrations in the control room outside air inflow. Per Section 4.3.2.1.1, the FHA-IC model credits CREACUS Emergency Mode of operation initiation 3 minutes following Page 86 of 110

the start of the event, due to detection of high radioactivity concentrations in the control room outside air inflow.

As discussed in Section 4.3, the FHA-IC dose analysis considers the dose consequences of inhalation, immersion, and radiation shine from the environmental (or outside) cloud, the control room emergency HVAC filters, and the containment building.

Consistent with RG 1.183 Regulatory Position 4.4, as an AST dose analysis acceptance criterion the postulated control room dose is evaluated to ensure that that it does not exceed the 5 Rem TEDE criterion established in 10 CFR 50.67.

Section 4.6.5 FHA-IC Dose Consequences The resulting FHA-IC offsite and control room operator doses are listed in Table 4.6-3. The analysis demonstrates that the FHA-IC event 6.3 Rem TEDE radiological criterion for the EAB and for the outer boundary of the LPZ is met.

The analysis also demonstrates that the FHA-IC event 5 Rem TEDE radiological criterion for the control room is met.

TABLE 4.6-3: FHA-IC DOSE CONSEQUENCES DOSE RECEPTOR J1 FHA-IC DOSE (REM TEDE)

ACCEPTANCE CRITERION (REM TEDE)

Control Room (30-day accident duration) 0.3 5 EAB (Maximum 2-hour dose -- 0.0 to 2.0 hours0 days <br />0 hours <br />0 weeks <br />0 months <br />) l 0.8 6.3 LPZ (30-day accident duration) < 0.1 6.3 Page 87 of 110

Section 4.7 FUEL HANDLING ACCIDENT INSIDE FUEL HANDLING BUILDING (FHA-FHB) ANALYSIS Regulatory Guide 1.183 Appendix B provides assumptions for use in evaluating the radiological consequences of an FHA-FHB. These assumptions supplement the guidance provided in the main body of RG 1.183.

This Section presents the assumptions, design input, methodology employed in evaluating, and the radiological consequences of, the SONGS Units 2 and 3 FHA-FHB. The characteristics of the FHA-FHB model are summarized in Table 4.7-1.

TABLE 4.7-1: FHA-FHB ANALYSIS PARAMETERS FHA-FHB PARAMETER llMODELED VALUE Dose acceptance criteria, Rem TEDE Control Room 5 EAB 6.3 LPZ 6.3 FHA-FHB source term Decay time after reactor shutdown, hours 72 Average fuel rod isotope inventory at 72 hours8.333333e-4 days <br />0.02 hours <br />1.190476e-4 weeks <br />2.7396e-5 months <br />, curies/rod per Section 4.1 Radial peaking factor applied to all failed fuel rods 1.75 Number of failed fuel rods 60 Core fission product fractions in fuel rod gaps Iodine-131 0.08 Krypton-85 0.10 Other noble gases (Krypton, Xenon) 0.05 Other Halogens (Iodine, Bromine) 0.05 Alkali Metals (Cesium, Rubidium) 0.12 Fraction of gap activity released to the fuel storage pool 1.00 Minimum water depth above damaged fuel rods, feet 23 Fuel storage pool decontamination factors lodines (effective DF) 200 Noble Gases 1 Particulates infinite Iodine composition above the fuel storage pool, percent of iodine Elemental iodine 57 Organic iodide 43 Fuel Handling Building model ESFAS - Fuel Handling [building] Isolation Signal (FHIS) not modeled Post-Accident Cleanup Units (PACUs) not modeled Activity release duration from FHB, hours 2 FHB net free volume, cubic feet 365,305 FHB air exhaust flow rate, ft3/minute 22,000 Offsite dose evaluation model per Section 4.2 Control Room dose evaluation model per Section 4.3 Page 88 of 110

FHA-FHB PARAMETER I MODELED VALUE FHA-FHB Release Points to Control Room Atmospheric Dispersion Factors, seconds/m 3 per Section 4.4l Section 4.7.1 FHA-FHB Source Term The FHA-FHB involves the inadvertent dropping of a fuel assembly during fuel handling operations, and the consequent rupture of fuel pins in the dropped assembly. Consistent with RG 1.183 Appendix B Section 1.1, the number of fuel rods damaged during the accident is based on a conservative analysis that considers the most limiting case. UFSAR Section 15.7.3.4.2.2 details the structural evaluation of dropped fuel assembly damage. Per the UFSAR, a maximum of 60 fuel rods will fail as a result of a fuel assembly drop in the spent fuel pool.

Table 4.1-3 presents the fission product inventory of an average fuel rod in the reactor core. Consistent with the guidance of RG 1.183 Regulatory Position 3.1, to account for differences in power level across the core, a radial peaking factor of 1.75 is applied to the Table 4.1-3 average fuel rod isotope inventory to determine the activity inventory in each of the 60 failed fuel rods as described in Section 4.1.3. Consistent with RG 1.183 Regulatory Position 3.1, the FHA-FHB 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 prior to the event, which is also consistent with the minimum decay time required by SONGS Units 2 and 3 LCS 3.9.101 prior to movement of irradiated fuel in the reactor vessel.

Consistent with RG 1.183 Appendix B Section 1.2, the fission product release from the breached fuel is based on RG 1.183 Regulatory Position 3.2.

Consistent with RG 1.183 Footnote 11, the release fractions are acceptable for use since the fuel has a peak burnup of less than 62,000 MWD/MTU, and a maximum linear heat generation rate that does not exceed 6.3 kw/ft peak rod average power for burnups exceeding 54 GWD/MTU.

All gap activity in the damaged rods is instantaneously released into the fuel storage pool. Radionuclides that are considered are xenons, kryptons, iodines, bromines, cesiums, and rubidiums. Cesium and rubidium are particulates that are retained in the spent fuel pool water; therefore, these radionuclides do not contribute to the FHA doses.

Consistent with RG 1.183 Appendix B Section 1.3, the chemical form of radioiodine released from the fuel to the fuel storage pool is assumed to be 95 percent cesium iodide (Csl), 4.85 percent elemental iodine, and 0.15 percent organic iodide. The Csl released from the fuel is assumed to completely dissociate in the fuel storage pool water and instantaneously re-evolve as elemental iodine. Consequently, the chemical form of radioiodine in the fuel storage pool, prior to application of a decontamination factor, is 99.85 percent elemental iodine and 0.15 percent organic iodide.

Page 89 of 110

Per Units 2 and 3 TS LCO 3.7.16, during movement of irradiated fuel assemblies in the fuel storage pool, the fuel storage pool water level shall be at least 23 feet over the top of the irradiated fuel assemblies seated in the storage racks. As noted in the LCO 3.7.16 Bases, there would be less than 23 feet of water above the top of a dropped single bundle laying horizontally on top of the spent fuel racks. However, as also noted in the LCO 3.7.16 Bases, when the potential of a dropped fuel assembly exists (which is when fuel is being moved) a water level is maintained that would ensure that there would be greater than 23 feet above the fuel assembly laying on top of the racks. This increased water level is required by Units 2 and 3 TS LCO 3.9.6 when the fuel storage pool is connected to the refueling cavity and by station procedures whenever fuel is being moved.

Consistent with RG 1.183 Appendix B Section 2, the 23 foot water depth requirement allows for elemental and organic iodine decontamination factors of 500 and 1, respectively, giving an overall effective decontamination factor of 200 (i.e., 99.5% of the total iodine released from the damaged rods is retained by the water). This difference in decontamination factors for elemental (99.85%) and organic iodine (0.15%) species results in the iodine above the water being composed of 57% elemental and 43% organic species.

Consistent with RG 1.183 Appendix B Section 3, the retention of noble gases in the water in the fuel storage pool is negligible (i.e., decontamination factor of 1).

Particulate radionuclides are assumed to be retained by the water in the fuel storage pool (i.e., infinite decontamination factor).

Section 4.7.2 FHA-FHB Activity Release Model Consistent with RG 1.183 Appendix B Section 4.1, the radioactive material that escapes from the fuel storage pool to the FHB is released to the environment over a 2-hour time period (i.e., FHB closure is not modeled during the FHA-FHB event).

Consistent with the 2-hour release model requirement, the FHA-FHB AST dose analysis does not model the generation of an ESFAS fuel handling [building]

isolation signal (FHIS). The FHB normal ventilation exhaust is assumed to remain in operation throughout the FHA-FHB event.

The FHB air volume dilutes the gaseous activity released from the damaged fuel rods.

The FHA-FHB AST dose analysis does not model a reduction in the amount of radioactive material available for release from the FHB by the fuel handling building Post-Accident Cleanup Unit (PACU) filter system. The FHB PACU system consists of two independent, redundant trains that each consists of Page 90Of 110

charcoal and HEPA filters for the removal of airborne gaseous and particulate activity following an FHA.

The release of activity to the environment within the required 2-hour time period is established by specifying a FHB air exhaust flow rate that ensures that at least 99.9 percent of the gaseous activity will be released to the environment.

Activity released during the FHA-FHB event is transported by atmospheric dispersion to the control room HVAC intake and to the offsite EAB and LPZ dose receptors. Activity may be released to the environment via the FHB normal ventilation exhaust system through the main plant vent, or as leakage through FHB penetrations (e.g., doors). Table 4.7-2 presents the San Onofre site-specific 95th percentile meteorology atmospheric dispersion factors for these release pathways as discussed in Section 4.4. Since one set of atmospheric dispersion factors does not consistently yield less dispersion than the others over time, a composite maximum of the two release points is utilized for assessing control room dose consequences. No credit is taken for radioactive decay of the isotopes during atmospheric dispersion transit to the control room or offsite dose locations. Consistent with RG 1.183 Regulatory Positions 4.1.7 and 4.2.2, no.

correction is made for depletion of the effluent plume by deposition on the ground.

TABLE 4.7-2: FHA-FHB CR ATMOSPHERIC DISPERSION FACTORS FHA-FHB to CR 95th Percentile Atmospheric Dispersion Factors (seconds/mr3) 1 Tim Tme Inr ntervaRelease FHBPoint Main PlantPoint Release Vent ll Modeled Value o to 2 hours2.314815e-5 days <br />5.555556e-4 hours <br />3.306878e-6 weeks <br />7.61e-7 months <br /> 9.45E-04 1.1 4E-03 1.14E-03 2 to 8 hours9.259259e-5 days <br />0.00222 hours <br />1.322751e-5 weeks <br />3.044e-6 months <br /> 7.48E-04 6.11 E-04 7.48E-04 8 to 24 hours2.777778e-4 days <br />0.00667 hours <br />3.968254e-5 weeks <br />9.132e-6 months <br /> 1.93E-04 2.1 OE-04 2.1 OE-04 1 to 4 days 2.64E-04 2.20E-04 2.64E-04 4 to 30 days 2.43E-04 1.98E-04 2.43E-04 Section 4.7.3 FHA-FHB EAB and LPZ Model Regulatory Guide 1.183 Regulatory Position 4.1 provides guidance to be used in determining the TEDE for persons located at or beyond the boundary of the exclusion area, including the outer boundary of the low population zone.

Section 4.2 addresses the applicability of this guidance to the SONGS Units 2 and 3 AST FHA-FHB dose analysis as it relates to the offsite dose exposure parameters.

As discussed in Section 4.2, the FHA-FHB dose analysis considers the dose consequences of inhalation and immersion. Radioactive material in the FHB is assumed to be a negligible radiation shine source to the offsite dose receptors Page 91 of 110

relative to the dose associated with immersion in the radioactive plume released from the facility.

Consistent with RG 1.183 Regulatory Positions 4.1.5, 4.1.6 and 4.4 and Table 6, the FHA-FHB event radiological criterion for the EAB and for the outer boundary of the LPZ is 6.3 Rem TEDE.

Section 4.7.4 FHA-FHB Control Room Model RG 1.183 Regulatory Position 4.2 provides guidance to be used in determining the TEDE for persons located in the control room. Section 4.3 addresses the applicability of this guidance to the SONGS Units 2 and 3 AST FHA-FHB dose analysis as it relates to the control room dose exposure parameters.

The CREACUS Emergency mode of operation can be actuated either automatically following a CRIS or manually. The CRIS may be generated automatically by a SIAS or by the detection of high radioactivity concentrations in the control room outside air inflow. Per Section 4.3.2.1.1, the FHA-FHB model credits CREACUS Emergency mode of operation initiation 3 minutes following the start of the event, due to detection of high radioactivity concentrations in the control room outside air inflow.

As discussed in Section 4.3, the FHA-FHB dose analysis considers the dose consequences of inhalation, immersion, and radiation shine from the environmental (or outside) cloud, and the control room emergency HVAC filters.

Radiation shine from contaminated air in the FHB is considered negligible due to the presence of numerous intervening concrete walls and the geometric attenuation due to the distance between the FHB and the control room.

Consistent with RG 1.183 Regulatory Position 4.4, as an AST dose analysis acceptance criterion the postulated control room dose is evaluated to ensure that that it does not exceed the 5 Rem TEDE criterion established in 10 CFR 50.67.

Section 4.7.5 FHA-FHB Dose Consequences The resulting FHA-FHB offsite and control room operator doses are listed in Table 4.7-3. The analysis demonstrates that the FHA-FHB event 6.3 Rem TEDE radiological criterion for the EAB and for the outer boundary of the LPZ is met.

The analysis also demonstrates that the FHA-FHB event 5 Rem TEDE radiological criterion for the control room is met.

Page 92 of 110

TABLE 4.7-3: FHA-FHB DOSE CONSEQUENCES FHA-FHB ACCEPTANCE DOSE RECEPTOR DOSE CRITERION (REM TEDE) (REM TEDE)

Control Room (30-day accident duration) < 0.1 5 EAB (Maximum 2-hour dose -- 0.0 to 2.0 hours0 days <br />0 hours <br />0 weeks <br />0 months <br />) 0.2 6.3 LPZ (30-day accident duration) < 0.1 6.3 Page 93 of 110

Section 4.8 MAIN STEAM LINE BREAK (MSLB) ANALYSIS RG 1.183 Appendix E provides assumptions for use in evaluating the radiological consequences of a Pressurized Water Reactor (PWR) MSLB. These assumptions supplement the guidance provided in the main body of RG 1.183.

A MSLB may occur either inside or outside containment. A steam line break inside containment will release contaminated steam via the break location to the containment air space, where it will be diluted within the containment net free air volume and then slowly leaked to the outside environment at the design basis containment leakage rate. A more severe scenario is that of a steam line break outside containment (SLB-OC) that will release contaminated steam via the break location directly to the environment. This section evaluates a SLB-OC event consistent with the guidance in RG 1.183 Appendix E.

The SONGS Units 2 and 3 CLB evaluates pre-trip and post-trip return-to-power steam line break events. The pre-trip SLB event may result in fuel failure (i.e.,

clad damage). The post-trip SLB event does not result in fuel failure. This section specifically evaluates a pre-trip SLB-OC event.

The transient response of the pre-trip SLB-OC event is analyzed using the CENTS computer code for the NSSS response, including mass releases and steam generator tube uncovery, and the CETOP computer code for the DNBR response.

This Section presents the assumptions, design input, methodology employed in evaluating, and the radiological consequences of, the SONGS Units 2 and 3 pre-trip SLB-OC. The characteristics of the pre-trip SLB-OC model are summarized in Table 4.8-1.

TABLE 4.8-1: PRE-TRIP SLB-OC ANALYSIS PARAMETERS

[ PRE-TRIP SLB-OC PARAMETER MODELED VALUE]

Dose acceptance criteria, Rem TEDE Control Room 5 EAB 25 LPZ 25 SLB-OC source term Core isotope inventory at reactor shutdown, curies per Section 4.1 Failed Fuel (clad damage), percent of core 10 Radial peaking factor 1.75 Core fission product fractions in fuel rod gaps Iodine-131 0.08 Krypton-85 0.10 Other noble gases (Krypton, Xenon) 0.05 Other Halogens (Iodine, Bromine) 0.05 Alkali Metals (Cesium, Rubidium) 0.12 Page 94 of 110

PRE-TRIP SLB-OC PARAMETER l MODELED VALUE Initial Primary Coolant Activity Profile per Section 4.1 Initial Secondary Coolant Activity Profile per Section 4.1 Dilution Volumes and Masses Reactor Coolant dilution volume, cubic feet 10,179 Reactor Coolant dilution mass, grams 2.015E+08 Secondary dilution water mass, Ibm 1.59E+05 Primary-to-Secondary leakage rate, gpm per Steam Generator (SG) 0.5 SG Water to Steam flashing fractions and partition coefficients SG tube uncovery period, seconds Oto 6,621 Iodine flashing factor during SG tube uncovery, percent 20 Iodine partition coefficient 100 Noble gases (Xe, Kr) partition coefficient 1E-06 Particulate isotopes partition coefficient 500 Steam Line Break Mass Release, Ibm 0 to 16.3 seconds 115,103 16.3 seconds to shutdown cooling at 13,659 seconds 0 Main Steam Safety Valve (MSSV) Mass Release, Ibm 0 to 30 minutes 47,553 30 minutes to 2 hours2.314815e-5 days <br />5.555556e-4 hours <br />3.306878e-6 weeks <br />7.61e-7 months <br /> 555.5 2 hours2.314815e-5 days <br />5.555556e-4 hours <br />3.306878e-6 weeks <br />7.61e-7 months <br /> to shutdown cooling at 13,659 seconds 0 Atmospheric Dump Valve (ADV) Mass Release, Ibm o to 30 minutes 0 30 minutes to 2 hours2.314815e-5 days <br />5.555556e-4 hours <br />3.306878e-6 weeks <br />7.61e-7 months <br /> 374,719 2 hours2.314815e-5 days <br />5.555556e-4 hours <br />3.306878e-6 weeks <br />7.61e-7 months <br /> to shutdown cooling at 13,659 seconds 356,610 Auxiliary Feedwater (AFW) Steam Turbine Mass Release, Ibm 0 to 30 minutes 8,078 30 minutes to 2 hours2.314815e-5 days <br />5.555556e-4 hours <br />3.306878e-6 weeks <br />7.61e-7 months <br /> 64,522 2 hours2.314815e-5 days <br />5.555556e-4 hours <br />3.306878e-6 weeks <br />7.61e-7 months <br /> to shutdown cooling at 13,659 seconds 78,944 Iodine composition released to the environment, percent of iodine Elemental iodine 97 Organic iodide 3 Offsite dose evaluation model per Section 4.2 Control Room dose evaluation model per Section 4.3 SLB-OC Release Points to Control Room Atmospheric Dispersion Factors, seconds/iM3 per Section 4.4 Section 4.8.1 Pre-Trip SLB-OC Source Term The pre-trip SLB-OC transient analysis is characterized by fuel failure (i.e., clad damage). Using the current licensing basis deterministic DNBR fuel failure prediction methodology, the radiological consequences for the pre-trip SLB-OC event have been characterized by no more than 7 percent fuel failure with a Core Operating Limits Supervisory System (COLSS) Required Overpower Margin (ROPM) of 123 percent power. Application of the DNB statistical convolution methodology described in Section 4.1.4 will result in a gain in COLSS DNBR Page 95 of 110

plant operating margin of 6% (to 117% ROPM) for an equivalent amount of fuel failure as described in Table 4.8-2. The methodology used to offset predicted fuel failure with COLSS required DNBR plant operating margin is per the standard NRC approved Westinghouse methodology for Combustion Engineering digital protection plants as discussed in Section 5.5.4 of the NRC-approved SONGS Reload Analysis Methodology Topical Report SCE-9801 -P-A.

The pre-trip SLB-OC AST dose analysis conservatively assumes 10 percent fuel failure to bound future operating cycle fuel failure predictions.

Table 4.8-2 Typical SONGS Pre-Trip SLB-OC Fuel Failures COLSS Required Overpower Margin (% Power) 117 119 121 123 Deterministic Fuel Failure (%) 21.7 l 16.0 l 11.3 6.4 Statistical Convolution Fuel Failure (%) 6.9 4.5 2.8 1.6 Consistent with RG 1.183 Appendix E Section 2, because more than minimal fuel damage is postulated, the pre-trip SLB-OC AST activity release model does not address primary coolant iodine spiking.

The 10 percent fuel failure estimate is applied to the reactor core fission product inventory presented in Table 4.1-3. Consistent with the guidance of RG 1.183 Regulatory Position 3.1, to account for differences in power level across the core, a radial peaking factor of 1.75 is also applied as described in Section 4.1.3.

Consistent with RG 1.183 Appendix E Section 1, the fission product release from the breached fuel is based on RG 1.183 Regulatory Position 3.2. Consistent with RG 1.183 Footnote 11,the release fractions are acceptable for use since the fuel has a peak burnup of less than 62,000 MWD/MTU, and a maximum linear heat generation rate that does not exceed 6.3 kw/ft peak rod average power for burnups exceeding 54 GWD/MTU.

Consistent with RG 1.183 Appendix E Section 3, the activity released from the fuel is instantaneously and homogeneously released into the reactor coolant system. Radionuclides that are considered are xenons, kryptons, iodines, bromines, cesiums, and rubidiums.

The initial reactor coolant concentration prior to the introduction of the fission product release from the breached fuel is assumed to be at the maximum TS LCO 3.4.16 limiting condition as specified in Section 4.1.

The AST pre-trip SLB-OC dose analysis assumes that the reactor coolant dilutes the core activity release into a volume of 10,179 cubic feet. This volume has conservatively omitted the primary side volume that is present in an assumed 2,000 plugged U-tubes in each of the two steam generators.

Page 96 of 110

Consistent with RG 1.183 Appendix E Section 4, the chemical form of radioiodine released from the steam generators to the environment is 97 percent elemental and 3 percent organic.

Section 4.8.2 Pre-Trip SLB-OC Activity Release Model Activity is introduced into the secondary side via steam generator tube leakage.

Consistent with RG 1.183 Appendix E Section 5.1, the pre-trip SLB-OC AST dose analysis models a primary-to-secondary leak rate into any single steam generator of 0.5 gallon/minute consistent with the maximum leak rate allowed by TS LCO 3.4.13.

The initial secondary side activity concentration prior to the introduction of the primary-to-secondary leakage is assumed to be at the maximum TS LCO 3.7.19 limit of 0.10 pCigm dose equivalent lodine-131.

Consistent with RG 1.183 Appendix E Sections 5.5.1 and 5.6, the primary-to-secondary leakage is assumed to mix with the secondary water without flashing during periods of total tube submergence. 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 pre-trip SLB-OC AST dose analysis conservatively assumes that the tubes in both steam generators are uncovered from 0 seconds to 6,621 seconds.

Consistent with RG 1.183 Appendix E Sections 5.5.1 and 5.6, during periods where the tubes are uncovered, a portion of the primary-to-secondary leakage flashes to vapor based on the thermodynamic conditions in the reactor coolant and the secondary coolant. The maximum flashing fraction is 14.41%, which occurs at the start of the event. Conservatively, the pre-trip SLB-OC AST dose analysis models a bounding flashing fraction of 20% during periods of steam generator tube uncovery.

The portion of primary-to-secondary leakage that flashes to steam enters the steam generator steam space, with no credit taken for iodine scrubbing.

Consistent with RG 1.183 Appendix E Section 5.5.3, the unflashed portion of primary-to-secondary leakage mixes with the bulk water. Consistent with RG 1.183 Appendix E Section 5.5.4, an iodine partition coefficient (i.e., liquid concentration divided by gas concentration) of 100 is modeled when evaluating the vaporization of the secondary side water (steam generator liquid). Consistent with RG 1.183 Appendix E Section 5.4, all noble gases released from the primary coolant are released to the environment without reduction or mitigation.

The SONGS Units 2 and 3 steam generators have a maximum full-power steam generator moisture carryover (steam quality) of 0.20 percent. The pre-trip Page 97 of 110

SLB-OC AST dose analyses address this carryover by modeling a particulate isotope partition coefficient of 500 when evaluating the vaporization of the secondary side water.

Activity is released to the environment via the steam line break location, the MSSVs, the ADVs, and the AFW turbine exhaust.

Consistent with the guidance in Branch Technical Position (BTP) MEB 3-1 Section B.1.b, the SLB-OC is modeled downstream of a MSIV. A break upstream of a MSIV is not postulated since the SONGS Units 2 and 3 design complies with the BTP MEB 3-1 (ASME Section III and design stress and fatigue limit requirements) for crediting break exclusion zones.

The release through the break begins at time zero and is terminated at 16.3 seconds, the time when MSIVs are fully closed. The total mass release through the break is 104,639 Ibm, consisting of inventory loss from both steam generators, and main feedwater flow for duration of 3.83 seconds. The pre-trip SLB AST dose analysis increased the break mass release predicted by the pre-trip SLB mass release analysis by 10 percent to 115,103 Ibm to provide margin for any potential increased mass release that may be determined in future cycle-specific transient analysis.

The MSSV and ADV mass releases are as shown in Table 4.8-1. The pre-trip SLB AST dose analysis increased the MSSV and ADV mass release predicted by the pre-trip SLB mass release analysis by 10 percent to provide margin for any potential increased mass release that may be determined in future cycle-specific transient analysis. The MSSV mass release begins when the MSSVs open at 1,200 seconds and terminates when the MSSVs close at 1,822 seconds. The ADV mass release begins when the ADVs are opened (by operator action) at 30 minutes, and stay open for the duration of the event. The pre-trip SLB-OC AST dose analysis models the mass releases as being from the MSSV from 1,200 seconds to 1,800 seconds and from the ADVs from 1,800 seconds until the end of the event. This is conservative since, as shown in Section 4.4, the ADV atmospheric dispersion factors are greater than the MSSV atmospheric dispersion factors, thus resulting in higher doses.

The time intervals during which the steam turbine AFW pump is operating, and the mass released during those intervals, are as shown in Table 4.8-1. Two periods of AFW operation are modeled. The first is from 89 seconds to 748 seconds. The second is from 1,921 seconds to the end of the event. The pre-trip SLB AST dose analysis increased the AFW steam turbine mass release predicted by the pre-trip SLB mass release analysis by 10 percent to provide margin for any potential increased mass release that may be determined in future cycle-specific transient analysis.

Page 98 of 110

The pre-trip SLB-OC event is terminated when shutdown cooling is initiated at 13,659 seconds. After this time all steam releases from both steam generators cease.

Activity released during the pre-trip SLB-OC event is transported by atmospheric dispersion to the control room HVAC intake and to the offsite EAB and LPZ dose receptors. San Onofre site-specific 95th percentile meteorology atmospheric dispersion factors for the pre-trip SLB-OC release pathways are discussed in Sections 4.2 and 4.4 for the offsite and control room dose receptors, respectively.

No credit is taken for plume rise dispersion associated with the ADV release pathway. No credit is taken for radioactive decay of the isotopes during atmospheric dispersion transit to the control room or offsite dose locations.

Consistent with RG 1.183 Regulatory Positions 4.1.7 and 4.2.2, no correction is made for depletion of the effluent plume by deposition on the ground.

Section 4.8.3 Pre-Trip SLB-OC EAB and LPZ Model RG 1.183 Regulatory Position 4.1 provides guidance to be used in determining the TEDE for persons located at or beyond the boundary of the exclusion area, including the outer boundary of the low population zone. Section 4.2 of this license amendment request addresses the applicability of this guidance to the SONGS Units 2 and 3 AST pre-trip SLB-OC dose analysis as it relates to the offsite dose exposure parameters.

As discussed in Section 4.2, the pre-trip SLB-OC dose analysis considers the dose consequences of inhalation and immersion.

Consistent with RG 1.183 Regulatory Positions 4.1.5, 4.1.6 and 4.4 and Table 6, the SLB event radiological criterion for the EAB and for the outer boundary of the LPZ is 25 Rem TEDE for an event scenario with fuel damage.

Section 4.8.4 Pre-Trip SLB-OC Control Room Model RG 1.183 Regulatory Position 4.2 provides guidance to be used in determining the TEDE for persons located in the control room. Section 4.3 of this license amendment request addresses the applicability of this guidance to the SONGS Units 2 and 3 AST pre-trip SLB-OC dose analysis as it relates to the control room dose exposure parameters.

The CREACUS Emergency mode of operation can be actuated either automatically following a CRIS or manually. The CRIS may be generated automatically by a SIAS or by the detection of high radioactivity concentrations in the control room outside air inflow. Per Section 4.3.2, the pre-trip SLB-OC model credits CREACUS Emergency mode of operation initiation 3 minutes following Page 99 of 110

the start of the event, due to detection of high radioactivity concentrations in the control room outside air inflow.

As discussed in Section 4.3, the pre-trip SLB-OC dose analysis considers the dose consequences of inhalation, immersion, and radiation shine from the environmental (or outside) cloud, and the control room emergency HVAC filters.

Consistent with RG 1.183 Regulatory Position 4.4, as an AST dose analysis acceptance criterion the postulated control room dose is evaluated to ensure that that it does not exceed the 5 Rem TEDE criterion established in 10 CFR 50.67.

Section 4.8.5 Pre-Trip SLB-OC Dose Consequences The resulting pre-trip SLB-OC offsite and control room operator doses are listed in Table 4.8-3. The analysis demonstrates that the SLB event 25 Rem TEDE radiological criterion for the EAB and for the outer boundary of the LPZ is met.

The analysis also demonstrates that the SLB event 5 Rem TEDE radiological criterion for the control room is met.

TABLE 4.8-3: PRE-TRIP SLB-OC DOSE CONSEQUENCES PRE-TRIP ACCEPTANCE DOSE RECEPTOR SLB-OC CRITERION DOSE (E EE (REM TEDE) (REM TEDE)

Control Room (30-day accident duration) 2.1 5 EAB (Maximum 2-hour dose -- 0.0 to 2.0 hours0 days <br />0 hours <br />0 weeks <br />0 months <br />) 4.1 25 LPZ (30-day accident duration) l 0.1 25 Page 100 of 110

5.0 REGULATORY SAFETY ANALYSIS 5.1 No Significant Hazards Consideration Southern California Edison (SCE) has evaluated whether or not a significant hazards consideration is involved with the proposed amendments by focusing on the three standards set forth in 10CFR50.92, uIssuance of Amendment," as discussed below:

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

Response: No.

The proposed changes to the Facility Operating Licenses for San Onofre Units 2 and 3 credit an Alternative Source Term (AST) for the design basis radiological site boundary and control room dose analyses. This change represents full scope implementation of the AST as described in Regulatory Guide 1.183. The proposed changes to the Facility Operating Licenses also expand the allowed use of fuel failure estimates by Departure from Nucleate Boiling (DNB) statistical convolution methodology from only the reactor coolant pump sheared shaft event to the Updated Final Safety Analysis Report (UFSAR) Chapter 15 non-Loss-of-Coolant-Accident (LOCA) events that assume a loss of flow (i.e., a loss of AC power) and that fail fuel. The proposed changes reflect the parameters used in the radiological consequences calculations for the LOCA, Fuel Handling Accident inside containment (FHA-IC), Fuel Handling Accident in the Fuel Handling Building (FHA-FHB) and pre-trip Steam Line Break Outside Containment (SLB-OC).

The purpose of this proposed change is to change the design requirements for the Control Room Envelope (CRE). This proposed change will allow an increase in the assumed amount of unfiltered air inleakage through the CRE. Currently, design basis radiological consequence analyses assume CRE inleakage of 0 cfm, plus an assumed 10 cubic feet per minute (cfm) inleakage due to ingress and egress into the Control Room. Analyses to support this change demonstrate acceptable post-accident dose consequences in the Control Room assuming 990 cfm of CRE inleakage (plus 10 cfm due to ingress and egress for a total of 1000 cfm).

This proposed change does not affect the precursors for accidents or transients analyzed in Chapter 15 of the San Onofre Units 2 and 3 UFSAR. Therefore, there is no increase in the probability of accidents previously evaluated. The probability remains the same because the accident analyses performed involve no change to a system, component Page 101 of 110

or structure that affects initiating events for any UFSAR Chapter 15 accident evaluated.

A re-analysis of the UFSAR Chapter 15 LOCA, SLB-OC, FHA-IC, and FHA-FHB events was conducted with respect to radiological consequences. This re-analysis was performed in accordance with AST methodology provided in Regulatory Guide (RG) 1.183 and with ARCON96 atmospheric dispersion methodology provided in RG 1.194.

The reanalysis consequences were expressed in terms of Total Effective Dose Equivalent (TEDE) dose.

Implementation of the AST methodology, as described in 10CFR50.67, specifies control room, exclusion area boundary (EAB), and low population zone (LPZ) dose acceptance criteria in terms of TEDE dose.

The dose acceptance criteria for specific events are specified in RG 1.183.

The revised analyses for all evaluated events meet the applicable RG 1.183 TEDE dose acceptance criteria for AST implementation.

The previous dose calculations analyzed the dose consequences to thyroid and whole body as a result of postulated design basis events. The previous control room dose calculations were shown to be within the regulatory limits of 10CFR50 Appendix A General Design Criterion 19 with respect to thyroid, beta-skin and whole body dose. The previous LOCA and SLB offsite dose calculations were shown to be within the regulatory limits of 10CFR1 00.1 1 with respect to thyroid and whole body dose. The previous FHA-IC and FHA-FHB offsite dose calculations were shown to be well within (i.e., less than 25 percent of) the regulatory limits of 10CFR100.1 1 with respect to thyroid and whole body dose.

RG 1.183 Footnote 7 provides a means to compare the thyroid and whole body dose results of the previous calculations with the TEDE results of the AST calculations. This methodology requires multiplying the previous thyroid dose by 0.03 and adding the product to the previous whole body dose. The resultant "effective" TEDE is then compared to the AST TEDE result.. This comparison is presented in Table 5-1.

The Table 5-1 comparison shows a decrease in dose consequences when evaluated using AST methodology for all but the LOCA offsite dose receptors. The LOCA EAB dose using AST methodology has increased due to the requirement to calculate the maximum 2-hour window EAB dose versus the previous requirement to calculate the 0 to 2 hour2.314815e-5 days <br />5.555556e-4 hours <br />3.306878e-6 weeks <br />7.61e-7 months <br /> window EAB dose. The LOCA LPZ dose using AST methodology has increased primarily due to changes in the AST Refueling Water Storage Tank (RWST) iodine transport model. Although the LOCA EAB and LPZ doses using AST methodology have increased, they remain significantly below the 25 Rem TEDE offsite dose acceptance criterion.

Page 102 of 110

Table 5 Comr arison of Previous and AST Doses Event - Dose Receptor "Effective" TEDE of AST TEDE Previous Dose (Rem)

Analyses (Rem)

FHA-IC Control Room 1.0 2.7 E-01 EAB 2.0 8.0 E-01 LPZ 5.6 E-02 2.3 E-02 FHA-FHB Control Room 3.7 E-01 7.3 E-02 EAB 6.6 E-01 2.1 E-01 LPZ 1.9 E-02 6.1 E-03 LOCA Control Room 4.5 2.7 EAB 3.7 5.1 LPZ 1.2 1.8 SLB-OC Control Room Not evaluated 2.1 EAB 8.0 4.1 LPZ Not evaluated 0.1 The proposed changes do not increase the probability of an accident previously evaluated. The proposed changes result in dose consequences that, if compared to previous ones, are in most cases decreased and in other cases only slightly increased (using guidance in footnote 7 of RG 1.183). However, the dose consequences of the revised analyses are below the AST regulatory acceptance criteria.

Therefore, the propsosed change does not involve a significant increase in the probability or consequences of any accident previously evaluated.

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

Response: No.

The implementation of this proposed change does not create the possibility of an accident of a different type than was previously evaluated in the UFSAR. The proposed change credits the AST for the design basis radiological site boundary and control room dose analyses and expands the allowed use of fuel failure estimates by DNB statistical convolution methodology from only the reactor coolant pump sheared shaft event to the UFSAR Chapter 15 non-LOCA events that assume a loss of flow (i.e.,

a loss of AC power) and that fail fuel.

Page 103 of 110

The changes proposed do not change how Design Basis Accident (DBA) events were postulated nor do the changes themselves initiate a new kind of accident with a unique set of conditions. The changes proposed are based on a re-analysis of offsite and control room doses for four design basis accidents. The revised analyses are consistent with the regulatory guidance established in RG 1.183. The revised analyses utilize the most current understanding of source term timing and chemical forms. Through this re-analysis, no new accident initiator or failure mode was identified.

Therefore, this proposed change does not create the possibility of a new or different kind of accident from any accident previously evaluated.

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

Response: No.

The implementation of this proposed amendment does not reduce the margin of safety. The alternative source term radiological dose consequence analyses utilize the regulatory acceptance criteria of 10 CFR 50 Appendix A General Design Criterion (GDC) 19 and 10 CFR 50.67, as specified in RG 1.183. These acceptance criteria have been developed for the purpose of use in design basis accident analyses such that meeting these limits demonstrates adequate protection of public health and safety. An acceptable margin of safety is inherent in these licensing limits. The radiological analyses results remain within these regulatory acceptance criteria.

Therefore, there is no significant reduction in the margin of safety as a result of the proposed amendment.

Based on the above, SCE concludes that the proposed amendments present no significant hazards consideration under the standards set forth in 10CFR50.92(c), and, accordingly, a finding of "no significant hazards consideration" is justified.

5.2 Applicable Regulatory Requirements/Criteria GDC 19 Control Room - A control room shall be provided from which actions can be taken to operate the nuclear power unit safely under normal conditions and to maintain it in a safe condition under accident conditions, including loss-of-coolant accidents. Adequate radiation protection shall be provided to permit access and occupancy of the control room under accident conditions without personnel receiving radiation exposures in excess of 5 rem whole body, or its equivalent to any part of the body, for the duration Page 104 of 110

of the accident. Equipment at appropriate locations outside the control room shall be provided (1)with a design capability for prompt hot shutdown of the reactor, including necessary instrumentation and controls to maintain the unit in a safe condition during hot shutdown, and (2)with a potential capability for subsequent cold shutdown of the reactor through use of suitable procedures. Applicants for and holders of construction permits and operating licenses under this part who apply on or after January 10, 1997, applicants for design certifications under part 52 of this chapter who apply on or after January 10, 1997, applicants for and holders of combined licenses under part 52 of this chapter who do not reference a standard design certification, or holders of operating licenses using an alternative source term under §50.67 shall meet the requirements of this criterion, except that with regard to control room access and occupancy, adequate radiation protection shall be provided to ensure that radiation exposures shall not exceed 0.05 Sv (5 rem) total effective dose equivalent (TEDE) as defined in §50.2 for the duration of the accident.

10CFR50, Appendix A, General Design Criterion 19 requires that the control room be designed such that the radiological dose to the operators following a design basis accident be less than 5 rem whole body, or its equivalent to any part of the body.

GDC-19 is the current licensing basis for the San Onofre Units 2 and 3 control room. Radiological consequences of design basis accidents are currently shown to be less than the criterion of 5 rem whole body, or its equivalent to any part of the body. Following approval of this license amendment request, the provisions of GDC-19 will continue to apply to San Onofre Units 2 and 3 except that with regard to control room access and occupancy, adequate radiation protection shall be provided to ensure that radiation exposures shall not exceed 0.05 Sv (5 rem) total effective dose equivalent (TEDE) as defined in §50.2 for the duration of the accident.

10CFR100.11 (a)

...(1) An exclusion area of such size that an individual located at any point on its boundary for two hours immediately following onset of the postulated fission product release would not receive a total radiation dose to the whole body in excess of 25 rem or a total radiation dose in excess of 300 rem to the thyroid from iodine exposure.

(2)A low population zone of such size that an individual located at any point on its outer boundary who is exposed to the radioactive cloud resulting from the postulated fission product release (during the entire period of its passage) would not receive a total radiation dose to the whole body in excess of 25 rem or a total radiation dose in excess of 300 rem to the thyroid from iodine exposure....

Page 105 of 110

Paragraphs (a)(1) and (a)(2) of 10CFR100.1 1 describe the current accident analysis dose acceptance criteria for the exclusion area boundary and the low population zone for San Onofre Units 2 and 3. Following approval of this license amendment request, the dose acceptance criteria for the exclusion area boundary and low population zone will be the 25 rem TEDE criteria specified by 10CFR50.67.

10CFR50.67 (a)Applicability. The requirements of this section apply to all holders of operating licenses issued prior to January 10, 1997, and holders of renewed licenses under part 54 of this chapter whose initial operating license was issued prior to January 10, 1997, who seek to revise the current accident source term used in their design basis radiological analyses.

(b) Requirements. (1)A licensee who seeks to revise its current accident source term in design basis radiological consequence analyses shall apply for a license amendment under §50.90. The application shall contain an evaluation of the consequences of applicable design basis accidents previously analyzed in the safety analysis report.

(2) The NRC may issue the amendment only if the applicant's analysis demonstrates with reasonable assurance that:

(i) An individual located at any point on the boundary of the exclusion area for any 2-hour period following the onset of the postulated fission product release, would not receive a radiation dose in excess of 0.25 Sv (25 rem) total effective dose equivalent (TEDE).

(ii) An individual located at any point on the outer boundary of the low population zone, who is exposed to the radioactive cloud resulting from the postulated fission product release (during the entire period of its passage), would not receive a radiation dose in excess of 0.25 Sv (25 rem) total effective dose equivalent (TEDE).

(iii) Adequate radiation protection is provided to permit access to and occupancy of the control room under accident conditions without personnel receiving radiation exposures in excess of 0.05 Sv (5 rem) total effective dose equivalent (TEDE) for the duration of the accident.

Regulatory Guide 1.183, "Alternative Radiological Source Terms for Evaluating Design Basis Accidents at Nuclear Power Reactors," provides guidance to licensees of operating power reactors on acceptable applications of alternative source terms. Regulatory Guide 1.194, "Atmospheric Relative Concentrations for Control Room Radiological Habitability Assessments at Nuclear Power Plants,"

describes methods acceptable to the NRC staff for determining atmospheric relative concentration (X/Q) values that will be used in control room radiological habitability assessments performed in support of applications for license amendment requests. The LOCA, FHA-IC, FHA-FHB, and SLB-OC were re-analyzed consistent with the guidance of RGs 1.183 and 1.194.

Page 106 of 110

Using the methods described in RG 1.183 and 1.194, the results of the new analyses for the LOCA, FHA-IC, FHA-FHB, and SLB-OC meet the criteria of 10 CFR 50.67 as shown in Table 5-2. These results demonstrate that the 10CFR50.67 dose acceptance criteria for exclusion area boundary, low population zone, and control room are met for these four events. In addition, the analysis results described in Section 4 above also show that the exclusion area boundary and low population zone dose acceptance criteria from Regulatory Guide 1.183, Table 6 are met.

Table 5 Comparison of AST Doses with AST Dose Criteria Event - Dose Receptor AST TEDE AST TEDE Dose (Rem) Acceptance Criteria

_ __ _ _ _ _ _ _ __ _ __ __ _ _ _ _ _ _ _ _ _ _ _ _(R em)

FHA-IC Control Room 0.3 5 EAB 0.8 6.3 LPZ < 0.1 6.3 FHA-FHB Control Room < 0.1 5 EAB 0.2 6.3 LPZ < 0.1 6.3 LOCA Control Room 2.7 5 EAB 5.1 25 LPZ 1.8 25 SLB-OC Control Room 2.1 5 EAB 4.1 25 LPZ 0.1 25 In conclusion, based on the considerations discussed above, (1)there is reasonable assurance that the health and safety of the public will not be endangered by operation in the proposed manner, (2)such activities will be conducted in compliance with the Commission's regulations, and (3) the issuance of the amendment will not be inimical to the common defense and security or to the health and safety of the public.

6.0 ENVIRONMENTAL CONSIDERATION

A review has determined that the proposed amendment would change a requirement with respect to installation or use of a facility component located within the restricted area, as defined in 10CFR20, or would change an inspection or surveillance requirement. However, the proposed amendment does not involve (i) a significant hazards consideration, (ii) a significant change in the types or significant increase in the amounts of any effluent that may be released offsite, or (iii) a significant increase in individual or cumulative occupational Page 107 of 110

radiation exposure. Accordingly, the proposed amendment meets the eligibility criterion for categorical exclusion set forth in 10CFR51.22(c)(9). Therefore, pursuant to 10CFR51 .22(b), no environmental impact statement or environmental assessment need be prepared in connection with the proposed amendment.

Page 108 of 110

7.0 REFERENCES

1. Regulatory Guide 1.183, "Alternative Radiological Source Terms for Evaluating Design Basis Accidents at Nuclear Power Reactors" July 2000

2. Generic Letter 2003-01, "Control Room Habitability," dated June 12, 2003

3. Letter from A. E. Scherer (SCE) to Document Control Desk (NRC), dated September 17, 2004,

Subject:

Response to Generic Letter 2003-01, "Control Room Habitability" Tracer Gas Test Results, San Onofre Nuclear Generating Station Units 2 and 3

4. Letter from A. E. Scherer (SCE) to Document Control Desk (NRC), dated August 5, 2003,

Subject:

Response to Generic Letter 2003-01, "Control Room Habitability," San Onofre Nuclear Generating Station Units 2 and 3

5. NUREG 0737, "Post-TMI Requirements"

6. Letter from L. Raghavan (NRC) to Harold B. Ray (SCE), dated March 26, 2001,

Subject:

"San Onofre Nuclear Generating Station, Units 2 and 3, Issuance of Amendments on Post-Accident Sampling Program"

7. Regulatory Guide 1.195, "Methods and Assumptions for Evaluating Radiological Consequences of Design Basis Accidents at Light-Water Nuclear Power Reactor," May 2003

8. Regulatory Guide 1.194, "Atmospheric Relative Concentrations for Control Room Radiological Habitability Assessments at Nuclear Power Plants,"

June 2003

9. NUREG/CR-6604, USNRC, April 1998. S.L. Humphreys et al.,

"RADTRAD: A Simplified Model for Radionuclide Transport and Removal and Dose Estimation" 10.K.F. Eckerman et al., 'Limiting Values of Radionuclide Intake and Air Concentration and Dose Conversion Factors for Inhalation, Submersion, and Ingestion," Federal Guidance Report 11, EPA-520/1-88-020, Environmental Protection Agency, 1988 11.NUREG/CR-1413, D. C. Kocher, May 1980, "A Radionuclide Decay Data Base-index and Summary Table" 12.K.F. Eckerman and J.C. Ryman, "External Exposure to Radionuclides in Air, Water, and Soil," Federal Guidance Report 12, EPA-402-R-93-081, Environmental Protection Agency, 1993 Page 109 of 110

13. Regulatory Guide 1.197, "Demonstrating Control Room Envelope Integrity at Nuclear Power Reactors," May 2003

14. Regulatory Guide 1.196, "Control Room Habitability at Light-Water Nuclear Power Reactors," May 2003

15. NUREG/CR-6331, Revision 1, USNRC, May 1997, J. V. Ramsdell, Jr.,

and C.A. Simonen, "Atmospheric Relative Concentrations in Building Wakes"

16. NUREG/CR-6189, "A Simplified Model of Aerosol Removal by Natural Processes in Reactor Containments"

17. NUREG/CR-5966 UA Simplified Model of Aerosol Removal by Containment Sprays"

18. CENPD-1 83-A, "C-E Methods for Loss of Flow Analysis", June 1984 (PROPRIETARY)

Page 110 of 110

ATTACHMENT A ACRONYMS

ACRONYMS Acronym Meaning ADV Atmospheric Dump Valve AFW Auxiliary Feedwater AST Alternative Source Term BPC Bechtel Power Corporation CEDE Committed Effective Dose Equivalent CEOG Combustion Engineering Owners Group CIAS Containment Isolation Actuation Signal CLB Current Licensing Basis COLR Core Operating Limits Report COLSS Core Operating Limits Supervisory System CPIS Containment Purge Isolation Signal CR Control Room CRE Control Room Envelope CREACUS Control Room Emergency Air Cleanup System CRH Control Room Habitability CRIS Control Room Isolation Signal CSS Containment Spray System DACU Dome Air Circulator Unit DBA Design Basis Accident DDE Deep Dose Equivalent DE 1-131 Dose Equivalent Iodine-131 DF Decontamination Factor DG Diesel Generator DNB Departure from Nucleate Boiling DNBR Departure from Nucleate Boiling Ratio E_ Average Disintegration Energy EAB Exclusion Area Boundary EAC Emergency Air Conditioner ECCS Emergency Core Cooling System ECU Emergency Cooling Unit EDE Effective Dose Equivalent EPIP Emergency Planning Implementing Procedure EQ Environmental Qualification ESF Engineered Safety Features ESFAS Engineered Safety Features Actuation System EVS Emergency Ventilation Supply FGR Federal Guidance Report FHA Fuel Handling Accident FHA-FHB Fuel Handling Accident in the Fuel Handling Building FHA-IC Fuel Handling Accident - Inside Containment Page 1 of 2

FHB Fuel Handling Building FHIS Fuel Handling Isolation Signal GL Generic Letter HPSI High Pressure Safety Injection HVAC Heating, Ventilation, and Air-Conditioning IMSF-SF Increased Main Steam Flow with Single Failure LCO Limiting Condition for Operation LCS Licensee Controlled Specification LOCA Loss of Coolant Accident LPSI Low Pressure Safety Injection LPZ Low Population Zone MFIV Main Feedwater Isolation Valve MSIV Main Steam Isolation Valve MSLB Main Steam Line Break MSSV Main Steam Safety Valve PACU Post-Accident Cleanup Unit PASS Post-Accident Sampling System PNNL Pacific Northwest National Laboratory PWR Pressurized Water Reactor RAS Recirculation Actuation Signal RG Regulatory Guide ROPM Required Overpower Margin RPF Radial Peaking Factor RWST Refueling Water Storage Tank SAFDL Specified Acceptable Fuel Design Limit SCE Southern California Edison SCP Standard Computer Program SIAS Safety Injection Actuation Signal SIS Safety Injection System SLB Steam Line Break SLB-OC Steam Line Break - Outside Containment SONGS San Onofre Nuclear Generating Station SRP Standard Review Plan TEDE Total Effective Dose Equivalent TID Technical Information Document TS Technical Specification TSP Tri-Sodium Phosphate UFSAR Updated Final Safety Analysis Report Page 2 of 2

ATTACHMENT B LIST OF REGULATORY COMMITMENTS

LIST OF REGULATORY COMMITMENTS

1. Following approval of this license amendment request, future revisions to UFSAR 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 dose calculation 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, to determine what specific changes may be warranted in order to maintain consistency with the manual dose assessment calculation methodology.

4. Following approval of this license amendment request, future revisions to Accident Monitoring setpoint calculations will reflect the AST source term.

5. Following approval of this license amendment request, SCE will provide the revised UFSAR sections to the NRC as part of its normal UFSAR update required by 10 CFR 50.71 (e).