Text

Diablo Canyon, Units 1 and 2 - License Amendment Request 15-03 Application of Alternative Source Term - Summary of Dose Analyses and Results. Part 2 of 8
	ML15176A528
Person / Time
Site:	Diablo Canyon
Issue date:	05/15/2015
From:	; Pacific Gas & Electric Co
To:	; Office of Nuclear Reactor Regulation
References
	DCL-15-069
	Download: ML15176A528 (280)
	v • d • e

{{#Wiki_filter:Enclosure Attachment 4 PG&E Letter DCL-15-069 ATTACHMENT 4 Diablo Canyon Power Plant Technical Assessment Prepared by Stone and Webster, Inc. (A CB&I Company)Implementation of Alternative Source Terms Summary of Dose Analyses and Results TECHNICAL REPORT CB&I Stone and Webster, Inc.IMPLMENTATION OF ALTERNATIVE SOURCE TERMS Summary of Dose Analyses and Results DIABLO CANYON POWER PLANT Prepared for: Pacific Gas & Electric Company May 15, 2015 QA CATEGORY I Safety Related Diablo Canyon Power Plant.Implementation ofAlternative Source Terms TABLE OF CONTENTS Section No. / Title Page

1.0 INTRODUCTION

................................................................................................................................ 3 2.0 REGULATORY APPROACH ........................................................................................................ 6 2.1 Proposed Changes to Current Licensing Basis .................................................................... 7 2.2 Planned Design Modifications ............................................................................................. 11 2.3 Planned Procedural Updates ............................................................................................. 12 2.4 Dose Acceptance Criteria .................................................................................................... 14 3.0 COM PUTER CODES ....................................................................................................................... 15 4.0 RADIATION SOURCE TERMS ................................................................................................. 20 4.1 Core Activity Inventory ........................................................................................................ 20 4.2 Coolant Activity Inventory .................................................................................................... 21 4.3 Gap Fractions for Non-LOCA Events ................................................................................. 23 5.0 ACCIDENT ATMOSPHERIC DISPERSION FACTORS (XIQ) ................................................... 30 5.1 Exclusion Area Boundary and Low Population Zone Atmospheric Dispersion Factors .......... 30 5.2 On-Site Atmospheric Dispersion Factors .......................................................................... 31 6.0 DOSE CALCULATION M ETHODOLOGY ................................................................................. 47 7.0 RADIOLOGICAL CONSEQUENCES USING AST ................................................................... 49 7.1 Control Room Design / Operation / Transport Model ........................................................ 50 7.2 Loss of Coolant Accident (LOCA) ..................................................................................... 52 7.3 Fuel Handling Accident (FHA) ............................................................................................. 81 7.4 Locked Rotor Accident (LRA) ............................................................................................. 85 7.5 Control Rod Ejection Accident (CREA) .............................................................................. 87 7.6 Main Steam Line Break (MSLB) .......................................................................................... 91 7.7 Steam Generator Tube Rupture (SGTR) ............................................................................ 94 7.8 Loss of Load Event (LOL) .................................................................................................... 98 8.0

SUMMARY

OF RESULTS: CONTROL ROOM / SITE BOUNDARY DOSES .............................. 136

9.0 CONCLUSION

S ............................................................................................................................. 138

10.0 REFERENCES

............................................................................................................................... 139 APPENDIX A: DCPP ARCON96 Atmospheric Dispersion Factor Inputs ................................... 144 APPENDIX B: Changes to Key Design Input Values (by accident): CLB vs AST ..................... 179 2 of 205 Diablo Canyon Power Plant Implementation ofAlternative Source Terms

1.0 INTRODUCTION

As holder of operating licenses issued prior to January 10, 1997, and in accordance with 10CFR50.67 (Reference

1) and Standard Review Plan 15.0.1 (Reference 2), Pacific Gas and Energy (PG&E) proposes to revise the accident source term used in the Diablo Canyon Power Plant (DCPP) Units I and 2 design basis site boundary and control room dose analyses, with the full implementation of Alternative Source Terms (AST) as defined in Regulatory Guide (RG)1.183, Section 1.2.1 (Reference 3).The first use of the AST for DCPP was a selective application per RG 1.183, Section 1.2.2, to revise the Fuel Handling Accident (FHA) in the Fuel Handling Building (FHB) in order to implement changes to the PG&E Design Class I ventilation filter testing program for the Control Room Ventilation System (CRVS), Auxiliary Building Ventilation System (ABVS) and the Fuel Handling Building Ventilation System (FHBVS). The FHA analysis demonstrated that acceptable doses would occur at both offsite locations and in the control room without taking credit for filtration by ventilation systems or control room isolation.

The application was reviewed and approved by the Nuclear Regulatory Commission (NRC) in its Safety Evaluation Report (SER) for License Amendment Nos. 163 and 165 to DCPP Facility Operating License Nos. DPR-80 and DPR-82, respectively. (Reference 4)With this application, and in the interest of evaluating DCPP design against a more realistic accident sequence, as well as in gaining dose analysis margin, the methodology / scenarios used in the following design basis accident (DBA) analyses discussed in the DCPP UFSAR (some of which utilize pre-NUREG-0800 assumptions), are being updated to reflect the AST guidance provided in RG 1.183.1. Loss of Coolant Accident (LOCA)2. FHA in the Containment

3. Locked Rotor Accident (LRA)4. Control Rod Ejection Accident (CREA)5. Main Steam Line Break (MSLB)6. Steam Generator Tube Rupture (SGTR)7. Loss-of Load (LOL) Event The Loss of Load Event is not addressed in Regulatory Guide 1.183 but has been selected by PG&E for re-evaluation to provide a bounding analysis for Condition II events that have no fuel damage but may have radioactivity in the environmental releases that occur as a result of reactor trip. The control room ventilation system is expected to remain in normal operation mode for the duration of these events.In addition, the dose consequence analysis associated with the FHA in the FHB has been revised to address updated design input as discussed below.The dose consequence analyses addressed in this application resolve the findings of a Licensing Basis Verification Project (LBVP) which was initiated by the licensee and has resulted 3 of 205 Diablo Canyon Power Plant Implementation of Alternative Source Terms in a total upgrade of the listed radiological post-accident dose consequence analyses.

The effect of the findings on current plant design, have been addressed in Prompt Operability Assessments which resulted in the implementation of several temporary compensatory measures.In addition, the listed dose consequence analyses have been revised to address: " Updated control room ventilation system parameters resulting from the installation of new back-draft dampers in the control room emergency ventilation filter recirculation lines. These dampers were installed to prevent reverse unfiltered flow into the control room." Updated control room unfiltered inleakage (including back-draft damper leakage)As part of an effort to improve the completeness and quality of design documentation and reflect current NRC guidance, PG&E has also elected to update the offsite atmospheric dispersion factors (X/Q) using recent meteorological data and RG 1.145, Revision 1 (Reference 5)methodology. The X/Q values applicable to on-site locations such as the Control Room and Technical Support Center, have been calculated using the "Atmospheric Relative CONcentrations in Building Wakes" (ARCON96) methodology (Ramsdell, 1997, Reference 6).The updated off-site and on-site atmospheric factors are utilized in the dose consequence analyses for the accident listed above.Also, and as part of the effort to clarify'and streamline the DCPP licensing basis related to post-accident dose consequences documented in DCPP UFSAR Section 15.5, PG&E is requesting NRC approval of the following proposed changes in licensing basis: 1. Removal of all the "expected" accident dose consequence assessments that were included in DCPP UFSAR Section 15.5.1 as part of the original license application. The original DCPP licensing basis included two evaluations, or cases, for each accident.The first case, called the expected case, used values for each factor involved in the accident, which were intended to be estimates of the actual values expected to occur if the accident took place. The resulting doses were close to the doses expected from an accident of this type. The second case, the DBA, used the customary conservative assumptions. The calculated doses for'the DBA, while not a realistic estimate of expected doses, provided the basis for determination of the design adequacy of the plant safety systems.Current NRC guidance related to expectations for a Safety Analysis Report for a nuclear power plant (e.g., NUREG-0800), does not require the inclusion of dose consequences from "expected" accident scenarios. Since these "expected accident scenario" evaluations are not relevant for determination of the design adequacy of the plant safety systems, PG&E is proposing to remove this information from its licensing basis.2. Elimination of the dose contribution of a containment purge via the Containment Hydrogen Purge System (CHPS) following a LOCA for purposes of hydrogen control.4 of 205 cDiablo Canyon Power Plant Implementation ofAlternative Source Terms The NRC revised 10CFR50.44 (Reference

9) to acknowledge that the amount of combustible gas generated for the design basis LOCA was not a risk significant threat to containment integrity.

Thus, with the exception of demonstrating the capability of ensuring a mixed atmosphere within containment, the requirements for hydrogen control pertaining to the design basis LOCAs was eliminated. In the SER for License Amendment Nos. 168 and 169 to DCPP Facility Operating License Nos. DPR-80 and DPR-82, respectively (Reference 10), the NRC confirmed the elimination of hydrogen release concerns associated with a design-basis LOCA, and the associated requirements that necessitated the need for the hydrogen recombiners and backup hydrogen vent and purge systems.To ensure consistency with the current licensing basis, PG&E is proposing to eliminate the dose contribution due to the containment purge pathway currently included in the LOCA dose consequence analysis in support of hydrogen control.5 of 205 Diablo Canyon Power Plant Implementation ofAlternative Source Terms 2.0 REGULATORY APPROACH In 1962, the U.S. Atomic Energy Commission (AEC) published Technical Information Document (TID)-14844, "Calculation of Distance Factors for Power and Test Reactors" (Reference 7), which specified release fractions of the different categories of fission products from the core to the reactor containment, or "source term", to be considered in the event of a postulated design basis accident (DBA) involving a "substantial meltdown of the core" in a Light Water Reactor.These fractions were based on small-scale experiments performed in the late 1950s in which irradiated U02 pellets were heated to accident temperatures. The TID-14844 accident source term formed the basis for NRC Regulatory Guide 1.4 (Reference 8), which was used to determine compliance with the NRC reactor siting criteria, 10CFR100.11 (Reference 11). It was also used to evaluate other plant performance requirements including control room habitability requirements given in 10CFR50 Appendix A, GDC-19 (Reference

12) and expanded in NUREG-0800, Standard Review Plan 6.4 (Reference 13), to 5 rem whole body dose, 30 rem beta dose and 30 rem thyroid dose. The TID-14844 source term was also used for evaluation of equipment qualification (EQ) of class IE electrical components in accordance with 10CFR50.49 (Reference 14), and evaluation of vital area accessibility post-LOCA in accordance with NUREG-0737 (Reference 15).Since the publication of TID-14844, significant advances have been made in understanding the timing, magnitude, and chemical form of fission product releases from severe nuclear power plant accidents.

In 1995, the NRC published NUREG-1465, "Accident Source Terms for Light-Water Nuclear Power Plants" (Reference 16), which used this research to provide estimates of an accident source term that was more physically based and that could be applied to Light Water Reactors. This effort crystallized into regulatory acceptance for an alternative source term and the issuance of 10CFR50.67 and RG 1.183, which formalized the applicability of AST for operating reactors.As noted in Section 1ll.2.a, Standard Review Plan 15.0.1, July 2000, (Reference 2), a full implementation of AST addresses a) all the characteristics of AST (i.e., the radionuclide composition and magnitude, chemical and physical form of the radionuclides, and the timing of the release of these nuclides), b) replaces the previous accident source term used in all design basis radiological analyses, and c) incorporates the Total Effective Dose Equivalent (TEDE)criteria of 10CFR50.67, and Section II of Standard Review Plan 15.0.1.With this application, PG&E proposes a "full" implementation of AST as defined in RG 1.183, Section 1.2.1 for DCCP Units 1 and 2. To that end, the dose consequences of the eight accidents discussed in Section 1.0 have been re-analyzed using the guidance provided in RG 1.183. The dose consequences for the other events that have an accident source term, and are a part of the current DCPP licensing basis, are addressed by qualitative comparisons to the accidents that have been analyzed. (Refer to Section 2.1, Item 22).However, for the reasons outlined below, the post-LOCA integrated doses utilized for radiological environmental qualification of PG&E Design Class I equipment, and the estimated operator mission doses while performing vital functions post-LOCA, will continue to be based on TID-14844 assumptions and will remain unchanged. 6 of 205 Diablo Canyon Power Plant Implementation ofAlternative Source Terms Equipment Qualification (EQ): This above approach is acceptable based on Section 1.3.5 of RG 1.183 which indicates that though EQ analyses impacted by plant modifications should be updated to address the impact of the modification, no plant modification is required to address the impact of the difference in source term characteristics (i.e. AST vs TID-14844) on EQ doses until the generic issue associated with a potential increase in cesium releases is resolved. NUREG-0933, "Resolution of Generic Safety Issues", Section 3.0, Item 187 resolved the generic issue related to the effect of increased cesium releases on EQ doses. The NRC staff concluded that there is no clear basis for a requirement to modify the design basis for EQ to adopt AST since there would be no discernible risk reduction associated with adopting AST for EQ.(Reference 55)Post-LOCA Vital Area Mission Doses: This above approach is acceptable based on the AST benchmarking study reported in SECY-98-154 (Reference

17) which concluded that results of analyses based on TID-14844 would be more limiting earlier on in the event, after which time the AST results would be more limiting.

The NRC SER for Fort Calhoun Station's implementation of AST (Reference

18) referenced the SECY-98-154 study as the source for the conclusion that results of analyses based on TID-14844 would be more limiting for periods up to one to four months after which time the AST results would be more limiting.

Post-LOCA access to vital areas usually occur within the first one or two weeks when the original TID-14844 source term is more limiting.2.1 Proposed Changes to Current Licensing Basis PG&E proposes to revise the DCPP licensing basis to implement the AST described in RG 1.183 through a) reanalysis of the radiological consequences of the UFSAR Chapter 15.5 accidents listed in Chapter 1.0 in accordance with the guidance provided in RG 1.183, and b)implementation of the following changes in plant operations / licensing basis.1. The TEDE acceptance criterion of 10CFR50.67(b)(2) will replace the previous whole body and thyroid dose acceptance criteria of 1 OCFR1 00.11.2. Use of new offsite atmospheric dispersion factors (X/Q) based on recent 5-year meteorological data (2007 to 2011) and RG 1.145 methodology.

3. Use of new on-site atmospheric dispersion factors (X/Q) for locations such as the Control Room and the Technical Support Center based on recent 5-year meteorological data (2007 to 2011) and ARCON96 methodology.

4. Use of inhalation dose conversion factors from Federal Guidance Reports (FGR) No.11 (Reference 19)5. Proposed changes to the DCPP Plant Technical Specifications

/ Technical Specifications Bases include TS 1.1 -Dose Equivalent 1-131 concentrations will be developed based on the committed thyroid dose equivalent conversion factors provided in EPA FGR No. 11 7 of 205 Diablo Canyon Power Plant Implementation ofAlternative Source Terms TS 3.4.16 -The Noble gas activity shall be limited to 5 270 pci/gm DE Xe-133.The current limit of 600 pCi/gm DE Xe-1 33 corresponds to -1% fuel defects which is the DCPP design basis value for system and shielding design. The limit is being reduced to control the noble gas activity in the coolant to levels below the design basis values.' TS 3.6.3 -The 48 inch containment purge valves will meet NUREG 0737, Item II.E.4.2, Position 6 (Reference 15), by being sealed closed as defined by SRP 6.2.4, item 11.6.f (Reference

22) during MODE's 1 through 4 (currently allows the containment purge system to be operable for less than 200 hours per year during MODE's 1 through 4).* TS 5.5.11 -The allowable methyl iodine penetration for the Auxiliary Building Ventilation charcoal filter will be changed to 5% (currently the allowable methyl iodine penetration for the Auxiliary Building Ventilation charcoal filter is 15%)" TS 5.5.9 -Primary to secondary accident induced steam generator tube leakage will be limited to a maximum of 0.75 gpm at standard temperature and pressure TS 5.5.19 -The dose acceptance criterion to demonstrate compliance with the control room habitability program will be changed from 5 rem whole body or its equivalent to any part of the body for the duration of the accident to 5 rem TEDE.* TS Bases 3.4.17 -Primary to Secondary leakage (total for all 4 Steam Generators (SGs)) will be limited to a maximum of 0.75 gpm at STP which includes accident induced leakage (note that TS 3.4.13d limits the maximum allowable operational leakage to 150 gpd, per SG)* TS Bases 3.6.6 -will be updated to require initiation of containment spray in -the recirculation mode within 12 minutes of termination of injection spray inclusive of a description of the updated valve alignments.

The related TS Bases, including those that reference 10CFR100 or define recently irradiated fuel, will also be affected.6. Use of an increased value for control room (CR) unfiltered air inleakage.

7. Credit for the dual ventilation intake design of the CR pressurization air intakes. Based on the availability of redundant PG&E Design Class I radiation monitors at each pressurization intake location, the DCPP design has the capability of initial selection of the cleaner intake, but does not have the capability of automatic selection of the clean intake throughout the event. Based on the CRVS pressurization intake design, and the expectation that the operator will manually make the proper intake selection throughout the event, and per RG 1.194, June 2003, Regulatory Position C.3.3.2.3, (Reference 21), when the CRVS is in Mode 4, the X/Q values for the more favorable CR intake is reduced by a factor of 4 and utilized to estimate the dose consequences.

See Section 5.2 for further detail.8. To support flexibility in future DCPP fuel management schemes with respect to the potential of having fuel rods that exceed the RG 1.183 linear heat generation rate criteria, the gap fractions used for the DCPP Non-LOCA events (with the exception of 8 of 205 Diablo Canyon Power Plant Implementation of Alternative Source Terms the CREA) are based on the bounding (higher) values per isotope / isotope class provided in the following documents: o Safety Guide (SG) 25, March 1972 (Reference 23)o NUREG/CR 5009 (Reference 24), and o RG1.183Rev0.

9. MSLB / SGTR / LRA I CREA / LOL: Credit for a reduction factor of 5 applied to the calculated ARCON96 7IQ values for energetic releases from the DCPP Main Steam Safety Valves (MSSVs) and the 10% Atmospheric Dump Valves (ADVs). This approach has been deemed acceptable per Regulatory Position C.6 of RG 1.194 for uncapped and vertically-oriented relief valves. See Section 5.2 for further detail.10. MSLB / SGTR / LRA / CREA / LOL: Credit for the fact that as a result of the close proximity of the MSSVs/10%

ADVs and the normal operation CR intake of the affected unit, and due to the high vertical velocity of the steam discharge from the MSSVs/10%ADVs, the resultant post-accident plume from the MSSVs/10% ADVs will not contaminate the normal operation CR intake of the affected unit. See Section 5.2 for further detail.11. FHA: Credit for the PG&E Design Class I radiation monitors located at the control room (CR) normal intakes to switch the CRVS from Mode 1 (normal operation) to Mode 4 (pressurized filtered mode) following a FHA in the Fuel Handling Building or Containment. See Section 7.3 for further detail.12. FHA: Credit for the following administrative controls reflected in plant procedures that ensure the FHB is maintained at a negative pressure relative to atmosphere during movement of irradiated fuel thus ensuring that the environmental releases occur via the Unit vent: See Section 7.3 for further detail.* The movable wall is in place and secured* No exit door from the FHB is propped open* At least one FHBVS exhaust fan is operating. The supply fan flow has been confirmed by design to have less flow than the exhaust fan; thus environmental releases are via the Unit Vent 13. FHA: A minimum decay time of 72 hrs prior to fuel movement. See Section 7.3 for further detail.14. LOCA: Removal of all the "expected" accident dose consequence assessments that were included in DCPP UFSAR Section 15.5.1 as part of the original license application.

15. LOCA: Credit for containment spray in the recirculation mode -To address the delayed core damage sequence of a post-LOCA AST scenario and support fission product removal from the containment atmosphere, credit is taken for manual initiation of containment spray in the recirculation mode within 12 minutes of termination of containment spray in the injection mode. Containment spray in the recirculation mode is credited until 6.25 hours after accident initiation.

See Section 7.2.1 for more detail.9 of 205 Diablo Canyon Power Plant Implementation ofAlternative Source Terms 16. LOCA: Updated allowable ESF System leakage values and associated release points.See Section 7.2.3.3 for more detail.17. LOCA: Updated allowable ESF System leakage into the Miscellaneous Equipment Drain Tank (MEDT) and associated releases / release points. See Section 7.2.3.6 for more detail.18. LOCA: Inclusion of environmental releases from the Refueling Water Storage Tank (RWST) vent due to sump water back-leakage. See Section 7.2.3.5 for more detail.19. LOCA: Elimination of the dose contribution due to the containment purge pathway currently included in the dose consequence analysis in support of hydrogen control.20. LOCA: Inclusion of environmental releases via the 12 inch Containment Vacuum I pressure relief pathway prior to containment isolation. See Section 7.2.3.1 for more detail.21. LOCA: Defining the portion of Room 506 of the Control Room which serves as a control room foyer adjacent to the Shift Supervisor's office, as a low occupancy area, and conservatively assigning the referenced area with an occupancy factor of less than 5%of the total time spent daily in the control room. See Section 7.2.5.2 for more detail.22. LOCA/ FHA / MSLB / SGTR / LRA / CREA / LOL: Update of the list of computer codes that support the dose consequence analyses.23. .Dose Consequences Associated with Other Chapter 15 Accidents: The DCPP licensing basis includes dose assessments at offsite locations for several Condition Ill and Condition IV events which are not addressed in Regulatory Guide 1.183 and have therefore not been re-analyzed with this application. The demonstration of compliance with the accident-specific regulatory limits at the EAB and LPZ for the referenced accidents will be addressed using engineering judgment and by comparison to the accident sequence, predicted fuel damage (if applicable), and resultant dose consequences of the DBAs discussed in Section 1.0 and re-analyzed in support of this application.(i) Small Break LOCA (SBLOCA) inside Containment (Condition III event): The possible radiological consequence of a SBLOCA inside containment (defined in UFSAR Chapter 15.3.1 as a break that is large enough to actuate the emergency core cooling system), is expected to be bounded by the "containment release" scenario of the control rod ejection accident (CREA) since the CREA is postulated to result in 10% fuel damage, whereas the SBLOCA has no fuel damage. Refer to Section 7.5 for the CREA.(ii) Minor Secondary System Pipe Break (Condition III event): The possible radiological consequence of a minor secondary system line break (defined n UFSAR Section 15.3 as 6 inch diameter or smaller) is expected to be significantly less than a MSLB since in both cases there is no fuel damage and the steam 10 of 205 Diablo Canyon Power Plant Implementation ofAlternative Source Terms releases following a minor secondary line break is expected to be significantly less than that associated with a MSLB. Refer to Section 7.6 for the MSLB.(iii) Complete Loss of Forced Reactor Coolant Flow (Condition III event): The possible radiological consequence of a complete loss of forced reactor coolant flow (postulated to result from a simultaneous loss of electrical supplies to all reactor coolant pumps resulting in increased coolant.temperature; no predicted fuel damage; dose consequences resulting from the atmospheric steam dumping required for plant cooldown), is expected to be bounded by the conservative Loss-of-Load scenario with a coincident Loss of Offsite power analyzed herein.Refer to Section 7.8 for the LOL event.(iv) Underfrequency Event (Condition III event): The possible radiological consequence of an underfrequency event (no predicted fuel damage; dose consequences resulting from the atmospheric steam dumping required for plant cooldown), is expected to be bounded by the conservative Loss-of-Load scenario evaluated for DCPP which includes a coincident Loss of Offsite Power. Refer to Section 7.8 for the LOL event.(v) A Single Rod Cluster Control Assembly Withdrawal (Condition II event): The possible radiological consequence of a single rod cluster control assembly withdrawal will be less than a CREA since the CREA is postulated to result in 10% fuel damage, whereas the condition of one rod cluster control assembly fully withdrawn with the rest of the bank fully inserted, at full power has only 5% fuel damage. Refer to Section 7.5 for the CREA.(vi) Maior Rupture of a Main Feedwater Pipe (Condition IV event): The. possible radiological consequence of a main feedwater line break (no predicted fuel damage; dose consequences resulting from airborne activity released to the environment from the break location and via the MSSVs / 10% ADVs) is expected to be bounded by the MSLB since the airborne environmental release via the break point is expected to be less than the MSLB. Per Standard Review Plan 15.2.8, Section III, Item 6 (Reference 32), the evaluation of the radiological consequences of a design basis feedwater line break may be based on a qualitative comparison to the results of the design basis MSLB. Refer to Section 7.6 for the MSLB.Tank Rupture Events: The tank rupture events represent the accidental release of the radioactivity accumulated in tanks resulting from normal plant operations and are not affected by the accident source term associated with AST. Therefore, the tank rupture events are not reanalyzed in support of this Licensing Application Request.2.2 Planned Design Modifications The following design modifications will be implemented prior to implementation of AST: 1. The external west wall of the Control Room has a 6'-8.5" x 10'-1" area where the concrete was replaced by other material as a result of a design change. The affected 11 of 205 Diablo Canyon Power Plant Implementation ofAlternative Source Terms area is the CR briefing room. PG&E will install shielding material at this location equivalent to that provided by the CR outer walls.2. The setpoints for the redundant PG&E Design Class I gamma sensitive area radiation monitors l-RE-25/26, 2-RE-25/26 will be updated. These monitors are designed to automatically isolate the control room normal air intakes and shift to CRVS Mode 4 (filtered I pressurized emergency ventilation) on detection of high radiation at the location of the CR normal intakes. The FHA analyses developed in support of this License Application Request utilize an "analytical limit" of 1 mR/hr for the gamma radiation environment at the CR normal operation air intakes prior to taking credit for CRVS Mode 4 actuation. The actual monitor trip setpoint will include instrument loop uncertainty; i.e., Trip setpoint (or High Alarm Setpoint (HASP) = Analytical limit / (1 +uncertainty %)).3. The 40 inch GE/GW Containment Penetration Area Ventilation line and the 2 inch gaseous radwaste system line which connect to the PG&E Design Class I Plant Vent are currently classified as PG&E Design Class II. In support of this application:

a. The portion of the 2 inch gaseous radwaste system line that connects to the Plant Vent will be re-classified as PG&E Design Class I.b. The portion of the GE/GW 40 inch Containment Penetration Area Ventilation line that connects to the Plant Vent up to and including the isolating damper solenoid valves, the associated damper actuators and the pressure switches will be re-classified as PG&E Design Class I (Refer to Section 5.2 for detail).4. Install a high efficiency particulate (HEPA) filter in the TSC normal ventilation system intake.2.3 Planned Procedural Updates Provided below are the key plant operating procedures that will be updated prior to implementation of AST. The full set of impacted procedures will be addressed in the AST Engineering Change Package (ECP).1. Update of Equipment Control Guideline ECG 42.1 to lower the restriction on fuel movement from 100 hrs to 72 hrs post-shutdown.

2. Review / update (as deemed necessary) of the EOPs and operator training procedures to ensure that the requisite steps to select the least contaminated outside air intake, and the provisions for monitoring to ensure the least contaminated intake is in use throughout the event, are provided.3. Update of Surveillance Test Procedure STP M-57, Control Room Ventilation System Tracer Gas Test Procedure, to include the new CR inleakage test acceptance criteria and the range of CRVS ventilation flows deemed acceptable by the AST dose consequence analyses.12 of 205 Diablo Canyon Power Plant Implementation of Alternative Source Terms 4. Update of the EOPs to include valve alignment information to manually initiate containment spray in the recirculation mode within 12 minutes of termination of injection spray, and to ensure, that at a minimum, spray operation continues until 6.25 hours after accident initiation.

An associated Time Critical Operator Action (TCOA) will be implemented upon receipt of NRC approval of this application.

5. Update of the ESF system leak testing procedures that are part of the Boundary Leakage Program to establish administrative acceptance criteria to ensure that: a) The total as-tested leakage from ESF systems that recirculate sump fluid outside containment is < 126 cc/min, and with the following breakdown:

o In areas covered by the Auxiliary Building ventilation the as-tested leakage is< 120 cc/min o In the containment penetration area the as-tested leakage is < 6 cc/min b) The total as-tested back leakage into the RWST from the containment recirculation sump is < 1 gpm.c) The total as-tested flow hard piped to the MEDT is 5 the following values: o Leakage from systems carrying non-radioactive fluids < 484 cc/min o Leakage from ESF systems that recirculate sump fluids < 950 cc/min 6. Update the TSC administrative procedures to ensure that o The nominal normal operation TSC ventilation air intake flowrate is 500 cfm o Following a LOCA, the TSC will be manually placed in Mode 4 operation such that filtered pressurization and recirculation can be credited within 2 hours of accident initiation. o The nominal post-LOCA TSC ventilation filtered pressurization and recirculation flowrates are 500 cfm, respectively.

7. Review and update, as necessary, administrative procedure OP B-8H (which is a compilation of all the precautions and prerequisites involved with moving irradiated fuel, fuel components and other items over / within the spent fuel pool) to include the requirements and / or restrictions imposed by the FHA dose consequence analyses on the FHB door closure status, the operation of the FHB ventilation system and the ability of the CRVS to actuate Mode 4 operation.

13 of 205 Diablo Canyon Power Plant Implementation ofAlternative Source Terms 2.4 Dose Acceptance Criteria PG&E has utilized the following acceptance criteria for the AST DCPP site boundary and control room dose analyses supporting this application: The acceptance criteria for the Exclusion Area Boundary (EAB) and Low Population Zone (LPZ)Dose are based on 10CFR50.67, and Section 4.4 Table 6 of Regulatory Guide 1.183: (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, shall not receive a radiation dose in excess of the accident-specific TEDE value noted in Reference 3, Table 6.(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), shall not receive a radiation dose in excess of the accident-specific TEDE value noted in Reference 3, Table 6.EAB and LPZ Dose Acceptance Criteria -Condition 11 and Condition Ill events: RG 1.183 (regulatory guidance for accident analyses using AST), does not specifically address Condition II and Condition Ill scenarios. However, per RG 1.183, Section 1.2.1, a full implementation of AST allows a licensee to utilize the dose acceptance criteria of 10CFR50.67 in all dose consequence analyses. In addition, Section 4.4 of RG 1.183 indicates that for events with a higher probability of occurrence than those listed in Table 6 of RG 1.183, the postulated EAB and LPZ doses should not exceed the criteria tabulated in Table 6. Thus, the dose consequences at the EAB and LPZ for Condition II and Condition III events will be limited to the lowest value reported in Table 6, i.e., a small fraction (10%)of the limit imposed by 10CFR50.67. The acceptance criterion for the Control Room Dose is based on 10 CFR 50.67: 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) TEDE for the duration of the accident.This criteria ensures that the dose criteria of GDC 19, 1999 and NUREG-0737, November 1980, Item I11.D.3.4 is met.The acceptance criterion for the Technical Support Center Dose is based on Section 8.2.1, Item f of NUREG-0737, Supplement 1 (Reference 27), as amended by RG 1.183, Section 1.2.1, which states that a full implementation of AST allows a licensee to utilize the dose acceptance criteria of 10CFR50.67 in all dose consequence analyses: The dose to an operator in the TSC should not exceed 5 rem TEDE for the duration of the accident.14 of 205 A6 Diablo Canyon Power Plant Implementation ofAlternative Source Terms 3.0 COMPUTER CODES The computer codes programs have been Assurance Program, discussed below.utilized in support of this application are listed below. These computer verified and validated under the CB&I S&W Inc. NRC approved Quality and have been shown to be accurate and acceptable for the use Code Name Description of Code Use SCALE 4.3 / SCALE 4.3, "Modular Code System for Performing Standardized Computer SAS2/ ORIGEN-S Analyses for Licensing Evaluation for Workstations And Personal Computers," Control Module SAS2 -CB&I S&W Inc. QA Category I Computer Code NU-230, V04, L03. (Reference 60)ORIGEN-S is part of the SCALE 4.3 suite of codes which was developed by Oak Ridge National Laboratory (ORNL) for the NRC to perform standardized computer analyses for licensing evaluations. SAS2 is a control module that provides a sequence to calculate the nuclide inventory in a fuel assembly by calling various neutron cross section treatment modules and the exponential matrix point-depletion module ORIGEN-S. SAS2/ORIGEN-S calculates the time-dependent neutron flux and the buildup of fissile trans-uranium nuclides. It properly accounts for all major nuclear interactions including fission, activation, and various neutron absorption reactions. It can calculate accurately the neutron-activated products, the actinides and the fission products in a reactor core.SAS2/ORIGEN-S is used to develop the DCPP equilibrium core activity inventory and the decayed fuel inventories after shutdown (for the FHA).SAS2/ORIGEN-S has been used in prior AST licensing applications (e.g., Beaver Valley Power Station [ML032530204], Fort Calhoun Power Station[ML013030027]) to develop the core inventory, and its results accepted by the NRC.ACTIVITY2 ACTIVITY2, "Fission Products in a Nuclear Reactor" -CB&I S&W Inc.Proprietary QA Category I Computer Code NU-014, V01, L03. (Reference 61)ACTIVITY2 calculates the concentration of fission products in the fuel, coolant, waste gas decay tanks, ion exchangers, miscellaneous tanks, and release lines to the atmosphere for a pressurized water reactor system. The program uses a library of properties of more than 100 significant fission products and may be modified to include as many as 200 nuclides. The program output presents the activity and energy spectrum at the selected part of the system for any specified operating time.ACTIVITY2 is used to develop the DCPP reactor coolant activity inventory (design and as limited by the plant Technical Specifications) ACTIVITY2 has been used in prior AST licensing applications (e.g., Beaver Valley Power Station[ML032530204], Fort Calhoun Power Station [ML013030027]) to develop the primary coolant inventory, and its results accepted by the NRC.15 of 205 A6 Diablo Canyon Power Plant Implementation of Alternative Source Terms Code Name Description of Code Use IDNEXCHANGER IONEXCHANGER, -CB&I S&W Inc. Proprietary QA Category I Computer Code NU-009, Ver. 01, Lev. 03. (Reference 62)IONEXCHANGER calculates the activity of nuclides in an ion exchanger or tank of a nuclear reactor plant by solving the appropriate growth-decay-purification equations. Based on a known feed rate of primary coolant or other fluid with known radionuclide activities, it calculates the activity of each nuclide and its products in the ion exchanger or tank at some later time. The program also calculates the specific gamma activity for each of the seven fixed energy groups.IONEXCHANGER is used to develop the DCPP secondary coolant activity inventory (design and as limited by the plant Technical Specifications) IONEXCHANGER has been used in prior AST licensing applications (e.g., Beaver. Valley Power Station [ML032530204], Fort Calhoun Power Station[ML013030027]) to develop the secondary coolant inventory, and its results accepted by the NRC.Atmospheric EN-113, "Atmospheric Dispersion Factors" -CB&I S&W Inc. Proprietary QA Dispersion Factors Category I Computer Code EN-1 13, V06, L08. (Reference 63)EN-1 13 calculates %/Q values at the Exclusion Area Boundary (EAB) and Low Population Zone (LPZ) following the methodology and logic outlined in NRC Regulatory Guide 1.145. The program can handle single or multiple release points for a specified time period and set of site-specific and plant-specific parameters. A release point can be identified as either of two types of release (i.e., ground or elevated), time periods for which sliding averages are calculated (i.e., 1 to 624 hours and/or annual average), applicable short-term building wake effect, meandering plume, long-term building height wake effect, and a wind speed value to be assigned to calm conditions. Downwind distances can be assigned for each of the sixteen 22.5-degree sectors for two irregular boundaries and for ten additional concentric boundaries used only in the annual average calculation. EN-1 13 performs the same calculations as the NRC PAVAN code except that EN-113 calculates X/Q values for the various averaging periods directly using hourly meteorological data whereas PAVAN uses a joint frequency distribution of wind speed, wind direction, and stability class.EN-1 13 is used to develop the DCPP site boundary atmospheric dispersion factors. EN-113 has been used in prior AST licensing applications (e.g., Fort Calhoun Power Station [ML0130300271), and its results accepted by the NRC.ARCON96 ARCON96, "Atmospheric Relative Concentrations in Building Wakes" -CB&I S&W Inc. QA Category I Computer Code EN-292, VO0, LOD. (Reference 64)ARCON96 was developed by Pacific Northwest National Laboratory (PNNL) for the Nuclear Regulatory Commission (NRC) to calculate relative concentrations in plumes from nuclear power plants at control room air intakes in the vicinity of the release point. ARCON96 has the ability to evaluate ground-level, vent, and elevated stack releases; it implements a straight-line Gaussian dispersion model with dispersion coefficients that are modified to account for low wind meander and building wake effects. The methodology is also able to evaluate diffuse and area source releases using the virtual point source technique, wherein initial 16 of 205 Diablo Canyon Power Plant Implementation ofAlternative Source Terms Code Name Description of Code Use values of the dispersion coefficients are assigned based on the size of the diffuse or area source. Hourly, normalized concentrations (X/Q) are calculated from hourly meteorological data. The hourly values are averaged to form XIQs for periods ranging from 2 to 720 hours in duration. The calculated values for each period are used to form cumulative frequency distributions. ARCON96 is used to develop the DCPP on-site control room and technical support center atmospheric dispersion factors. ARCON96 has been used extensively by the nuclear power industry in prior AST licensing applications and its results accepted by the NRC.SWNAUA SWNAUA, "Aerosol Behavior in Condensing Atmosphere", CB&I S&W Inc.Proprietary QA Category I Computer Code NU-1 85, V02, LOD. (Reference 65)SWNAUA is a derivative of industry computer code NAUA/Mod 4 which was originally developed, in Germany and was based on experimental data.NAUA/Mod 4 addressed particulate aerosol transport and removal following a LOCA at an LWR. It. developed removal coefficients to address physical phenomena such as gravitational settling (also called gravitational sedimentation), diffusion, particle growth due to agglomeration, etc. using time-dependent airborne aerosol mass. NAUA4 (included in the NRC Source Term Code Package) was used by NRC during the initial evaluations of post-TMI data.S&W modified NAUA/Mod 4 to include spray removal and diffusiophoretic effects suitable for design basis accident analyses. A version of SWNAUA (SWNAUA-HYGRO) was proven to be the most reliable of more than a dozen international entries, in making predictions of aerosol removal for the LWR Aerosol Containment Experiments (LACE) series.SWNAUA is used to develop the time dependent post LOCA particulate aerosol removal coefficients in the sprayed and unsprayed regions of DCPP containment. SWNAUA has been used in prior AST applications (e.g., the design certification of CE System 80+, and for operating nuclear plants Beaver Valley Power Station [ML0325302041 and Fort Calhoun Station [ML013030027]) and its results accepted by the NRC.RADTRAD 3.03 RADTRAD 3.03 "A Simplified Model for RADionuclide Transport and Removal And Dose Estimates" -CB&I S&W Inc. QA Category I computer code No. NU-232, Version 3.03, Level (NA). (Reference 66)RADTRAD 3.03 is a NRC sponsored program, developed by Sandia National Labs (SNL). It can be used to calculate radiological doses to the public, plant operators and emergency personnel due to environmental releases that resulting from postulated design basis accidents at light water reactor (LWR) power plants The RADTRAD 3.03 (GUI Interface Mode) includes models for a variety of processes that can attenuate and/or transport radionuclides. It can model sprays and natural deposition that reduce the quantity of radionuclides suspended in the containment or other compartments. It can model the flow of radionuclides between compartments within a building, from buildings into the environment, and from the environment into a Control Room (CR)). These flows can be through filters, piping, or simply due to air leakage. RADTRAD 3.03 can also model radioactive decay and in-growth of daughters. Ultimately the program 17 of 205 "ýi Diablo Canyon Power Plant Implementation ofAlternative Source Terms Code Name Description of Code Use calculates the Thyroid and TEDE dose (rem) to the public located offsite and to onsite personnel located in the CR due to inhalation and submersion in airborne radioactivity based on user specified, fuel inventory, nuclear data, dispersion coefficients, and dose conversion factors.RADTRAD is used to develop the TEDE dose to the public located offsite and to onsite personnel located in the CR due to inhalation and submersion in airborne radioactivity following the DCPP accidents listed in Chapter 1.0. RADTRAD has been used extensively by the nuclear power industry in prior AST licensing applications and its results accepted by the NRC.PERC2 PERC2, "Passive Evolutionary Regulatory Consequence Code" -CB&I S&W Inc. Proprietary QA Category I Computer Code, NU-226, VOO, L02. (Reference 67)PERC2 is a multi-region activity transport and radiological dose consequence program. It includes the following major features: " Provision of time-dependent releases from the reactor coolant system to the containment atmosphere." Provision for airborne radionuclides for both TID and AST release assumptions, including daughter in growth.* Provision for calculating the CEDE to individual organs as well as EDE from inhalation, DDE and beta from submersion, and TEDE." Provisions for tracking time-dependent inventories of all radionuclides in all control regions of the plant model.* Provision for calculating instantaneous and integrated gamma radiation source strengths as well as activities for the inventoried radionuclides to permit direct assessment of the dose from contained / or external sources for equipment qualification, vital area access and CR/TSC and EAB direct shine dose estimates. PERC2 is used to calculate the accident energy release rates and integrated gamma energy releases versus time for the various DCPP post-LOCA external and contained radiation sources. This source term information is input into SW QADCGGP to develop the direct shine dose to the CR. PERC2 is also used to develop the decay heat in the RWST and MEDT and develop the TEDE dose to personnel located in the technical support center due to inhalation and submersion in airborne radioactivity following LOCA. PERC2 has been used in prior AST licensing applications (e.g., Beaver Valley Power Station[ML032530204], Fort Calhoun Power Station [ML013030027]), and its results accepted by NRC.SW-QADCGGP SW-QADCGGP, "A Combinatorial Geometry Version of QAD-5A" -CB&I S&W Inc. Proprietary QA Category I Computer Code, NU-222, VOO, L02. (Reference 68)SW-QADCGGP is a variant of the QAD point kernel shielding program originally written at the Los Alamos Scientific Laboratory by R. E. Malenfant. The QADCGGP version implements combinatorial geometry and the geometric progression build-up factor algorithm. The SW-QADCGGP implements a graphical indication of the status of the computation process.18 of 205 Diablo Canyon Power Plant Implementation ofAlternative Source Terms Code Name Description of Code Use SW-QADCGGP is used to develop the direct shine dose to the DCPP CR, TSC and EAB. SW-QADCGGP has been used in prior AST licensing applications (e.g., Beaver Valley Power Station [ML032530204], Fort Calhoun Power Station[ML013030027]), and its results accepted by NRC.GOTHIC GOTHIC, "Generation of Thermal-Hydraulic Information for Containments", CB&I S&W Inc. QA Category I computer code No. ME-376, Version 8.0, Lev (NA).(Reference 69)GOTHIC is developed and maintained by Numerical Applications Incorporated (NAI) and is an integrated, general purpose thermal-hydraulics software package for design, licensing, safety and operating analysis of nuclear power plant containments and other confinement buildings. GOTHIC solves the conservation equations for mass, momentum and energy for multicomponent, multi-phase flow in lumped parameter and/or multi-dimensional geometries. The phase balance equations are coupled by mechanistic models for interface mass, energy and momentum transfer that cover the entire flow regime from bubbly flow to film/drop flow, as well as single phase flows. The interface models allow for the possibility of thermal non equilibrium between phases and unequal phase velocities, including countercurrent flow. Other phenomena include models for commonly available safety equipment, heat transfer to structures, hydrogen burn and isotope transport. GOTHIC is used to estimate the containment and sump pressure and temperature response with recirculation spray, the temperature transient in the DCPP RWST / MEDT gas and liquid due to incoming sump water leakage /inflow / decay heat from the RWST / MEDT fission product inventory, and the volumetric release fraction transient from the RWST / MEDT gas space to the environment. GOTHIC has been used in prior AST licensing applications for this purpose (Beaver Valley Power Station [ML072780163]), and its results accepted by NRC.19 of 205 Diablo Canyon Power Plant Implementation ofAlternative Source Terms 4.0 RADIATION SOURCE TERMS 4.1 Core Activity Inventory The guidance provided in Section 3.1 of RG 1.183 indicates that the inventory of fission products in the reactor core available for release to the containment following an accident should reflect maximum full power operation of the core with the current licensed values for fuel enrichment, fuel burnup, and an assumed core power equal to the current licensed rated thermal power times the ECCS evaluation uncertainty in the 10CFR50 Appendix K analysis (typically 1.02).In accordance with the above guidance, the isotopic inventory of fission products in the DCPP reactor core developed in support of implementation of AST is conservatively based on plant operation at 105% of the current licensed rated thermal power of 3411 MWth, with current licensed values of fuel enrichment and fuel burnup.The DCPP equilibrium core inventory is calculated using computer code ORIGEN-S. The calculation is performed by utilizing Control Module SAS2 of the SCALE 4.3 computer code package. The SAS2 control module provides a sequence to calculate the nuclide inventory in a fuel assembly by calling various neutron cross section treatment modules and the exponential matrix point-depletion module ORIGEN-S. It calculates the time-dependent neutron flux and the buildup of fissile trans-uranium nuclides. It accounts for all major nuclear interactions including fission, activation, and various neutron absorption reactions with materials in the core. It calculates the neutron-activated products, the actinides, and the fission products in a reactor core.The DCPP reactor core consists of 193 fuel assemblies with various Uranium-235 enrichments. Per DCPP core-reload design documentation, the peak rod burnup limit at the end of cycle is not allowed to exceed 62,000 MWD/MTU. The current licensed maximum value for fuel enrichment is 5.0%. To account for variation of U-235 enrichment in fresh fuel, the radionuclide inventories were calculated for a 4.2% average enriched core (representing minimum enrichment at DCPP), and 5% average enriched core (representing maximum enrichment). The higher activity for each isotope from the above two enrichment cases is chosen to represent the inventory of that isotope in the equilibrium core.The equilibrium core at the end of a fuel cycle is assumed to consist of fuel assemblies with three different burnups, i.e., approximately 1/3 of the core is subjected to one fuel cycle, 1/3 of the core to two fuel cycles and 1/3 of the core to three fuel cycles. This approach has been demonstrated to develop an isotopic core inventory that is a reasonable and conservative approximation of a core inventory developed using DCPP specific fuel management history data.A 19 month fuel cycle length was utilized in the analysis. The 19-month average fuel cycle is an artifact of the current DCPP fuel management scheme which specifies 3 fuel cycles every 5 years and refueling outages in Spring or Fall. Minor variations in fuel irradiation time and duration of refueling outages will have a slight impact on the estimated inventory of long-lived isotopes in the core. However, these inventory changes will have an insignificant impact on the radiological consequences of postulated accidents. Regardless, a 4% margin has been 20 of 205 -Diablo Canyon Power Plant Implementation of Alternative Source Terms included in the final isotopic radioactive inventories in support of bounding analyses and to address minor changes in future fuel management schemes.In summary, the equilibrium isotopic core average inventory is based on* A power level of 3580 MWth inclusive of power uncertainty" A range of enrichment of 4.2 to 5.0 w % U-235. Use of a few assemblies with lower enrichment is a common industry practice when replacing assemblies previously irradiated but proven unsuitable for continued irradiation. As these assemblies are designed to replace higher enrichment assemblies with ones of similar reactivity for the remainder of the fuel cycle, their inventory is enveloped by the isotopic core average inventory developed to support the dose consequence analyses A maximum core average burnup of 50 GWD/MTU The core inventory developed by ORIGEN-S using the above methodology includes over 800 isotopes. The DCPP equilibrium core fission product inventory of dose significant isotopes relative to LWR accidents is presented in Table 4.1-1.4.2 Coolant Activity Inventory Desiqn Basis Primary and Secondary Coolant Activity Concentrations CB&I S&W Inc. computer code ACTIVITY2, is used to calculate the design basis primary coolant activity concentrations for both DCPP Units 1 and 2 based on the core inventory developed using ORIGEN-S and discussed in Section 4.1. The source terms for the primary coolant fission product activity include leakage from 1% fuel defects and the decay of parent and second parent isotopes. The depletion terms of the primary coolant fission product activity include radioactive decay, purification of the letdown flow and neutron absorption when the coolant passes the reactor core. The nuclear library includes 3 rd order decay chains of approximately 200 isotopes.CB&I S&W Inc. computer code IONEXCHANGER, is used to calculate the design basis halogen and remainder activity concentrations in the secondary side liquid. The source terms for the secondary side activity include the primary-to-secondary leakage in steam generators and the decay products of parent and second parent isotopes. The depletion terms of the secondary side liquid activity include radioactive decay, and purification due to the steam generator blowdown flow, and continuous condensate polishing. The design basis noble gas concentrations in the secondary steam are calculated by dividing the appearance rate (pCi/sec) by the steam flow rate (gm/sec). The noble gas appearance rate in the steam generator steam space includes the primary-to-secondary leak contribution and the noble gas generation due to decay of halogens in the SG liquid. The activity concentrations of the other isotopes in the steam are determined by the SG liquid concentrations and the partition coefficients recommended in NUREG-0017, Revision 1. (Reference 25)21 of 205 Diablo Canyon Power Plant Implementation ofAlternative Source Terms Technical Specification Primary and Secondary Coolant Activity Concentrations DCPP Technical Specifications Limiting Condition for Operation (LCO) 3.4.16 limits the specific activity for iodines and noble gases in the primary coolant to 1 pCi/gm Dose Equivalent (DE) I-131, and 600 pCi/gm DE Xe-133 (corresponds to -1% fuel defects), respectively. Technical Specifications LCO 3.7.18 limits the specific activity for iodines in the secondary liquid to 0.1 pCi/gm DE 1-131.In accordance with current licensing basis, the primary coolant technical specification activities for iodines are based on 1.0 pCi/gm DE 1-131. As discussed earlier, with this application, PG&E proposes to reduce the TS LCO limit for the noble gases to 270 pCi/gm DE Xe-133 (corresponds to -0.5 % fuel defects).The Technical Specification (TS) based primary coolant isotopic activity reflects the following: " Isotopic compositions based on the design basis primary coolant equilibrium concentrations at 1% fuel defects." Iodine concentrations based on the thyroid inhalation weighting factors* for 1-131, 1-132, 1-133, 1-134, and 1-135 obtained from Federal Guidance Report 11 (Reference 19).* Noble gas concentrations based on the submersion weighting factors for Xe-133, Xe-133m, Xe-135m, Xe-135, Xe-138, Kr-85m, Kr-87 and Kr-88 obtained from Federal Guidance Report 12 (Reference 20)The isotopic iodine concentrations in the primary and secondary coolant allowable by the Plant Technical Specifications are based on committed thyroid dose equivalent conversion factors from Table 2.1 of Federal Guidance 11. This selection is made in recognition of the fact that available physical data with respect to radiation damage resulting from inhalation of radioactive iodine is associated with a specific organ, i.e., the thyroid.It is noted that use of the committed effective dose equivalent conversion factors from Table 2.1 of Federal Guidance 11 would predict slightly lower Technical Specification primary and secondary coolant iodine concentrations (by -2%), which, when used in the accident analyses to estimate the releases, would result in slightly lower dose consequences (-2%).It is acknowledged that defining the dose equivalent 1-131 based on the committed effective dose has the advantage of being consistent with the post-accident dose consequences since with the implementation of AST, the dose is estimated in terms of TEDE. However, the approach used in defining DE 1-131 based on the thyroid dose conversion factors is more conservative, and believed to be more appropriate since the thyroid dose is a more precisely defined physical quantity for the radio-toxicity of iodines.In accordance with the requirements of Item 10 of RIS 2006-04 (Reference 70), the TS definition for Dose Equivalent 1-131 will be updated to reflect the use of the committed thyroid dose equivalent conversion factors.The Technical Specification 1 pCi/gm DE 1-131 concentrations per nuclide in the primary coolant are calculated with the following equation: 22 of 205 Diablo Canyon Power Plant Implementation of Alternative Source Terms C (i) x CTo, DEJ 1 3 , (i) (ki/gn= {F(ix C(i)}Where: F(i) = DCF(i) / DCF 1-131 DCF(i)= FGR-1 1 Table 2-1 Thyroid Dose Conversion Factor per Nuclide (Rem/Cl)C(i) = design basis primary coolant equilibrium iodine concentration per nuclide (pCi/gm)CTot = primary coolant total (DE 1-131) technical specification iodine concentration (pCi/gm).The CTtot for the pre-accident iodine spike is 60 pCi/gm (transient TS limit for full power operation), or 60 times the primary coolant total iodine technical specification concentration. The accident initiated iodine spike activities are based on an accident dependent multiplier times the equilibrium iodine appearance rate. The equilibrium appearance rates are conservatively calculated based on the technical specification reactor coolant activities, along with the maximum design letdown rate, maximum technical specification based allowed primary coolant leakage, and an assumed ion-exchanger iodine removal efficiency of 100%.The TS secondary liquid iodine concentration is determined using methodology similar to that described above for the primary coolant where CTtot is 0.1 pCi/gm DE 1-131, and C(i) is the design basis secondary coolant equilibrium concentrations per nuclide.The TS noble gas concentrations for the primary coolant are based on 270 pci/gm DE Xe-133.The DE Xe-133 for noble gases is calculated as follows: DE X 1 3 3 = _{F(i) x C(i)}Where: F(i) = DCF(i) / DCF Xe-1 33 DCF(i) = EPA FGR-12 (1993), Table I11.1, Dose Coefficient per Nuclide [(rem-m 3)/(Ci-sec)] C(i) = design basis primary coolant equilibrium noble gas concentration per nuclide (pCi/gm)The noble gas and halogen primary and secondary coolant Technical Specification Activity Concentrations for DCPP Units 1 and 2 are presented in Table 4.2-1. The pre-accident iodine spike concentrations and the equilibrium iodine appearance rates (utilized to develop accident initiated iodine spike values), are presented in Table 4.2-2.4.3 Gap Fractions for Non-LOCA Events RG 1.183, Rev 0, Table 3 provides the gap fractions for Non-LOCA events (with the exception of the CREA) for AST applications. The referenced gap fractions are contingent upon meeting Note 11 of RG 1.183. Note 11 indicates that the release fractions listed in Table 3 are"acceptable for use with currently approved LWR fuel with a peak burnup of 62,000 MWD/MTU provided that the maximum linear heat generation rate does not exceed 6.3 kw/ft peak rod average power for burnups exceeding 54 GWD/MTU." As documented in NRC communications with other licensees, (Millstone, ML041320350), the burnup criterion associated with the maximum allowable linear heat generation rate is applicable to the peak rod average burnup in 23 of 205 Diablo Canyon Power Plant Implementation ofAlternative Source Terms any assembly and is not limited to assemblies with an average burnup that exceeds 54 GWD/MTU.Based on NRC acceptance of licensing applications from Millstone Unit 2 (SER: ML042360671) and Indian Point Unit 3 (SER: ML030760135) for fuel rods that exceed the RG 1.183 linear heat generation rate criteria, and to support flexibility in future DCPP fuel management schemes, the gap fractions used for the DCPP Non-LOCA events (with the exception of the CREA) are based on the bounding (higher) values per isotope / isotope class provided in the following documents: 0 S S Safety Guide (SG) 25, March 1972 (Reference 23)NUREG/CR 5009 (Reference 24), and RG 1.183 Rev 0.SG25 was traditionally the regulatory guidance used for gap fractions at LWRs for Non-LOCA events with the exception of the CREA and is the current DCPP licensing basis. NUREG/CR 5009 Section 3.2.2 addresses the impact of extended burn fuel on the gap fractions. RG 1.183 DCPP Nuclide Group Gap Fraction for Gap Fraction for Non-LOCA events Non-LOCA events Gap Fraction Source Document 1-131 0.08 0.12 NUREG/CR 5009 Kr-85 0.10 0.30 Safety Guide 25 Other Noble Gases 0.05 0.10 Safety Guide 25 Other Halogens 0.05 0.10 Safety Guide 25 Alkali Metals 0.12 0.17 NUREG/CR 5009 In accordance with RG 1.183, the gap fraction associated with the CREA is as follows: Noble Gases: Halogens: 10%10%Refer to the following Tables for the isotopic activity in the gap for Non-LOCA events Tables 4.3-1 -LRA and CREA Table 7.3-2 -FHA 24 of 205 Diablo Canyon Power Plant Implementation ofAlternative Source Terms Table 4.1-1 DCPP Equilibrium Core Inventory (Power Level: 3580 MWth)Dose Significant Isotopes including the Parent, Grandparent and 2nd Parent Isotopes ACTIVITY ACTIVITY ACTIVITY ISOTOPE* (CURIES) ISOTOPE* (CURIES) ISOTOPE* (CURIES)AG-110 2.67E+07 IN-125 8.46E+05 SB-125 9.63E+05 AG-110M 6.92E+05 IN-127 1.86E+06 SB-127 9.14E+06 AG-111 7.09E+06 IN-129 3.55E+06 SB-129 3.25E+07 AG-111M 7.09E+06 IN-131 1.09E+06 SB-130 1.08E+07 AG-112 3.16E+06 IN-132 2.85E+05 SB-130M 4.38E+07 AG-115 6.21E+05 KR-83M 1.14E+07 SB-131 7.67E+07 AG-115M 2.60E+05 KR-85 1.11 E+06 SB-132 4.70E+07 AM-239 4.90E-01 KR-85M 2.33E+07 SB-132M 4.37E+07 AM-241 1.32E+04 KR-87 4.65E+07 SB- 33 6.32E+07 AM-242 9.40E+06 KR-88 6.43E+07 SB-134 1.14E+07 AM-242M 8.54E+02 KR-89 7.94E+07 SB-135 5.46E+06 AM-243 5.28E+03 KR-90 8.48E+07 SB-136 8.63E+05 AM-244 3.79E+07 KR-91 5.83E+07 SE-83 5.38E+06 AM-245 1.12E-03 KR-92 3.12E+07 SE-83M 5.65E+06 AS-76 3.05E+03 KR-93 1.07E+07 SE-84 2.04E+07 AS-83 7.02E+06 KR-94 5.00E+06 SE-85 9.54E+06 BA-137M 1.30E+07 LA-1 40 1.85E+08 SE-87 1.32E+07 BA-139 1.76E+08 LA-141 1.61E+08 SE-88 7.15E+06 BA-1 40 1.78E+08 LA-1 42 1.57E+08 SE-89 2.49E+06 BA-141 1.59E+08 LA-143 1.48E+08 SM-153 6.04E+07 BA-142 1.51 E+08 LA- 144 1.31 E+08 SM-155 4.30E+06 BA-143 1.29E+08 MO-99 1.84E+08 SM-156 2.66E+06 BA-144 9.93E+07 MO-101 1.69E+08 SM-157 1.70E+06 BR-82 4.44E+05 MO-103 1.62E+08 SN-121 8.43E+05 BR-82M 3.88E+05 MO-104 1.33E+08 SN-123 6.43E+04 BR-83 1.13E+07 MO-105 9.97E+07 SN-125 5.25E+05 BR-84 2.1OE+07 MO-106 5.86E+07 SN-125M 1.58E+06 BR-85 2.31 E+07 NB-101 1.59E+08 SN-127 3.69E+06 BR-87 3.67E+07 NB-104 5.1OE+07 SN-127M 4.95E+06 BR-88 3.52E+07 NB-95 1.66E+08 SN-129 1.28E+07 BR-89 2.45E+07 NB-95M 1.89E+06 SN-129M 1.17E+07 BR-90 1.35E+07 NB-97 1.59E+08 SN-130 3.28E+07 CD-115 9.17E+05 NB-97M 1.50E+08 SN-131 2.83E+07 CD-115M 4.43E+04 NB-99 1.07E+08 SN-132 2.28E+07 25 of 205 s6v Diablo Canyon Power Plant Implementation of Alternative Source Terms Table 4.1-1 DCPP Equilibrium Core Inventory (Power Level: 3580 MWth)Dose Significant Isotopes including the Parent, Grandparent and 2nd Parent Isotopes ACTIVITY ACTIVITY ACTIVITY ACTTPE*VSOTYE*ISOTOPE* (UIS ISOTOPE* (CURIES) (CURIES) (CURIES)CD-121 7.67E+05 NB-99M 7.35E+07 SN-133 6.21E+06 CE-141 1.63E+08 ND-147 6.59E+07 SN-134 1.07E+06 CE-143 1.50E+08 ND-149 3.90E+07 SR-89 9.05E+07 CE-144 1.26E+08 ND-151 2.08E+07 SR-90 9.67E+06 CE-147 6.18E+07 NP-238 6.20E+07 SR-91 1.13E+08 CF-249 2.46E-02 NP-239 2.16E+09 SR-92 1.22E+08 CM-241 3.54E+00 NP-240 6.23E+06 SR-93 1.39E+08 CM-242 5.88E+06 PD-109 4.73E+07 SR-94 1.39E+08 CM-244 1.31E+06 PD-109M 3.12E+05 SR-95 1.25E+08 CM-245 1.26E+02 PD-1 11 7.09E+06 SR-97 4.68E+07 CO-58** 0.OOE+00 PD-112 3.14E+06 TB-160 1.87E+05 CO-60** 0.OOE+00 PD-115 7.84E+05 TC-99M 1.63E+08 CS-132 5.75E+03 PM-147 1.68E+07 TC-101 1.69E+08 CS-134 2.41 E+07 PM-148 1.88E+07 TC-103 1.65E+08 CS-134M 5.63E+06 PM-148M 2.83E+06 TC-104 1.40E+08 CS-136 7.01 E+06 PM-149 6.43E+07 TC-105 1.18E+08 CS-137 1.37E+07 PM-151 2.1OE+07 TC-106 8.80E+07 CS-138 1.85E+08 PM-153 9.77E+06 TE-127 9.03E+06 CS-139 1.72E+08 PR-142 9.47E+06 TE-127M 1.52E+06 CS-140 1.54E+08 PR-143 1.47E+08 TE-129 3.1OE+07 CS-141 1.17E+08 PR-144 1.27E+08 TE-129M 6.30E+06 CS-142 6.80E+07 PR-144M 1.76E+06 TE-131 8.28E+07 CS-143 3.41E+07 PR-147 6.52E+07 TE-131M 2.04E+07 DY-166 4.91E+02 PR-149 3.57E+07 TE-132 1.41E+08 EU-154 9.OOE+05 PR-151 1.23E+07 TE-133 1.09E+08 EU-155 3.83E+05 PU-238 5.22E+05 TE-133M 8.93E+07 EU-156 3.90E+07 PU-239 3.06E+04 TE-134 1.75E+08 EU-157 4.12E+06 PU-240 4.87E+04 TE-135 9.68E+07 EU-158 1.01 E+06 PU-241 1.36E+07 TE-136 4.29E+07 EU-159 5.15E+05 PU-242 3.34E+02 TE-137 1.45E+07 GA-72 1.71 E+03 PU-243 7.36E+07 TE-138 3.65E+06 GA-77 1.66E+05 RA-224 5.16E-01 TH-228 5.14E-01 GD-159 8.91E+05 RB-86 2.50E+05 U-239 2.17E+09 GE-77 6.48E+04 RB-86M 2.07E+04 XE-131M 1.42E+06 26 of 205 AA Diablo Canyon Power Plant Implementation ofAlternative Source Terms Table 4.1-1 DCPP Equilibrium Core Inventory (Power Level: 3580 MWth)Dose Significant Isotopes including the Parent, Grandparent and 2nd Parent Isotopes ACTIVITY ACTIVITY ACTIVITY ISOTOPE* (CURIES) ISOTOPE* (CURIES) ISOTOPE* (CURIES)GE-77M 1.70E+05 RB-88 6.60E+07 XE-133 2.01E+08 GE-83 1.24E+06 RB-89 8.57E+07 XE-1 33M 6.42E+06 H-3 6.1OE+04 RB-90 7.86E+07 XE-135 4.92E+07 HO-166 2.58E+04 RB-90M 2.53E+07 XE-135M 4.30E+07 1-129 4.00E+00 RB-91 1.05E+08 XE-137 1.84E+08 1-130 3.58E+06 RB-92 9.35E+07 XE-138 1.70E+08 1-1 30M 1.92E+06 RB-93 7.89E+07 XE-1 39 1.25E+08 1-131 9.90E+07 RB-94 4.13E+07 XE-140 8.66E+07 1-132 1.44E+08 RB-95 2.01E+07 XE-142 1.33E+07 1-133 2.01E+08 RH-103M 1.66E+08 Y-90 1.02E+07 1-134 2.22E+08 RH-l05 1.08E+08 Y-90M 7.71 E+02 I- 134M 2.07E+07 RH-105M 3.43E+07 Y-91 1.19E+08 1-135 1.92E+08 RH-106 7.53E+07 Y-91 M 6.57E+07 1-136 8.73E+07 RH-109 3.65E+07 Y-92 1.23E+08 1-137 9.40E+07 RN-220 5.16E-01 Y-93 9.41 E+07 1-138 4.80E+07 RU-103 1.66E+08 Y-94 1.50E+08 1-139 2.22E+07 RU-105 1.21 E+08 Y-95 1.57E+08 1-140 6.06E+06 RU-106 6.68E+07 Y-97 1.26E+08 IN-115M 9.17E+05 RU-109 3.16E+07 ZN-72 1.71E+03 IN-121 7.55E+04 SB-122 1.57E+05 ZR-101 9.55E+07 IN-121M 7.82E+05 SB-122M 1.57E+04 ZR-95 1.65E+08 IN-123 6.87E+05 SB-124 1.21E+05 ZR-97 1.58E+08 SB-124M 2.34E+03 ZR-99 1.66E+08 Note:* Isotopes in Bold Font are dose-significant for inhalation, submersion and direct shine. The parent, grandparent and second parent of the isotopes in Bold Font are also required to address daughter product in-growth. The group of isotopes needed to determine the "submersion and inhalation" dose in the Control Room and at the Site Boundary is typically a subset of the isotopes listed above in bold font, and represent a small group of reasonably long half-life isotopes with significant inhalation dose conversion factors which dominate the TEDE dose.To determine the TEDE resulting from inhalation and submersion following a LOCA, the DCPP LOCA dose consequence analysis uses the default group of 60 isotopes provided with computer code RADTRAD 3.03 plus 13 additional nuclides that were deemed to be dose significant (i.e., Br-82, Br-84, Rb-88, Rb-89, Te-1 33, Te-133m, Te-134, 1-130, Xe-131m, Xe-133m, Xe-138, Cs-138 and Np-238).** Co-58 / Co-60 are activation products that are developed external to the core and typically do not appear in the equilibrium core inventory. 27 of 205 A-A Diablo Canyon Power Plant Implementation ofAlternative Source Terms Table 4.2-1 Primary and Secondary Coolant Technical Specification Activity Concentrations Nuclide Primary Coolant Secondary Coolant (pCi/gm) (PCi/gm)Kr-83M 1.87E-01 Kr-85M 6.60E-01 Kr-85 5.60E+00 -----Kr-87 4.41 E-01 -----Kr-88 1.22E+00 -----Xe-131M 1.88E+00 -----Xe-133M 1.92E+00 -----Xe-133 1.29E+02 -----Xe-135M 4.07E-01 -----Xe-135 3.76E+00 1-131 7.87E-01 8.06E-02 1-132 2.99E-01 1.94E-02 1-133 1.16E+00 1.08E-01 1-134 1.67E-01 4.78E-03 1-135 6.68E-01 5.09E-02_________________ _________________________ I. ________________________ Table 4.2-2 Primary Coolant Pre-Accident Iodine Spike Concentrations & Equilibrium Iodine Appearance Rates Pre-Accident Spike Equilibrium Iodine Activity RCS Concentrations Appearance Rates into RCS (60 pCi/gm DE 1-131) (pCi/sec)Nuclide (pCi/gm)1-131 47.2 7.18E+03 1-132 17.9 7.78E+03 1-133 69.5 1.25E+04 1-134 10.0 8.91E+03 1135 40.1 9.91E+03 28 of 205 Ah, Diablo Canyon Power Plant Implementation of Alternative Source Terms Table 4.3-1 Isotopic Gap Activity Locked Rotor Accident I Control Rod Ejection Accident Fraction of Core Gap Activity Fraction of Core Gap Activity Core Activity w/o Peaking Core Activity w/o Peaking Core in Gap Factor in Gap Factor Activity LRA (Ci) CREA (Ci)Nuclide (Ci) LRA CREA KR-85 1.11E+06 0.30 3.33E+05 0.10 1.11E+05 KR-85M 2.33E+07 0.10 2.33E+06 0.10 2.33E+06 KR-87 4.65E+07 0.10 4.65E+06 0.10 4.65E+06 KR-88 6.43E+07 0.10 6.43E+06 0.10 6.43E+06 Xe-131M 1.42E+06 0.10 1.42E+05 0.10 1.42E+05 Xe-133M 6.42E+06 0.10 6.42E+05 0.10 6.42E+05 XE-133 2.01E+08 0.10 2.01E+07 0.10 2.01E+07 XE-135 4.92E+07 0.10 4.92E+06 0.10 4.92E+06 Xe-138 1.70E+08 0.10 1.70E+07 0.10 1.70E+07 1-130 3.58E+06 0.10 3.58E+05 0.10 3.58E+05 1-131 9.90E+07 0.12 1.19E+07 0.10 9.90E+06 1-132 1.44E+08 0.10 1.44E+07 0.10 1.44E+07 1-133 2.01E+08 0.10 2.01E+07 0.10 2.01E+07 1-134 2.22E+08 0.10 2.22E+07 0.10 2.22E+07 1-135 1.92E+08 0.10 1.92E+07 0.10 1.92E+07 BR-82 4.44E+05 0.10 4.44E+04 0.10 4.44E+04 BR-84 2.1OE+07 0.10 2.1OE+06 0.10 2.1OE+06 CS-134 2.41E+07 0.17 4.1OE+06 --CS-136 7.01E+06 0.17 1.19E+06 CS-137 1.37E+07 0.17 2.33E+06 --CS-138 1.85E+08 0.17 3.15E+07 --RB-86 2.50E+05 0.17 4.25E+04 --Rb-88 6.60E+07 0.17 1.12E+07 --Rb-89 8.57E+)7 0.17 1.46E+07 --Note: Values reported reflect the core isotopic gap activity assumed for the LRA and CREA. These values have to be adjusted for a) the failed fuel percentage (10%) and b) peaking factor (1.65), prior to assessing the associated dose consequences. For the isotopic gap activity associated with the FHA refer to Table 7.3-2.29 of 205 Diablo Canyon Power Plant Implementation ofAlternative Source Terms 5.0 ACCIDENT ATMOSPHERIC DISPERSION FACTORS (X/Q)The DCPP meteorological measurement program is described in DCPP UFSAR Section 2.3.3.The meteorological program was designed to meet the requirements of Safety Guide 23, February 1972 (Reference 59). The program consists of monitoring wind speed, wind direction, ambient temperature, and precipitation. Operation of the meteorological monitoring instrumentation ensures that sufficient meteorological data are available for estimating potential radiation doses to the public as a result of routine or accidental release of radioactive materials to the atmosphere. 5.1 Exclusion Area.Boundary and Low Population Zone Atmospheric Dispersion Factors Atmospheric dispersion factors (i.e., X/Qs) are calculated at the DCPP EAB and LPZ for post-accident environmental releases originating from DCPP Units 1 and 2.The applicable methodology is identified in RG 1.145 for ground level releases (Reference 5).The methodology is implemented using CB&I S&W Inc. computer program "Atmospheric Dispersion Factors" using a continuous temporally representative 5-year period of hourly meteorological data from the DCPP onsite meteorological tower (i.e., January 1, 2007 through December 31, 2011).The Regulatory Guide 1.145 methodology for ground level sources is as follows: X/Q 1 = {(u)[(n)(ay)(oy) + (A/2)11-'/Q2 = [(u)(37r)(y)(az)]-1 X/ 3 = RU) (7) (Fy) (GOV where: X/Q = relative concentration (sec/m 3);ay, a, = horizontal and vertical dispersion coefficients, respectively, based on stability class and horizontal downwind distance (m);u = wind speed at the 10-meter elevation (m/sec);A = cross-sectional building area (M2);y = (M)(Wy) for distances of 800 meters or less; and Ey = [(M-1)(c80yoom) + uy] for distances greater than 800 meters with M representing the meander factor in Reference 5.Per the guidance provided in RG 1.145, X/Q 1 and XIQ 2 values are calculated by EN-1 13 and the higher value selected. This value is then compared to the X/Q 3 value calculated by EN-1 13, and the smaller value is then selected as the appropriate value.The EAB distances for the sixteen 22.50 azimuth downwind sectors are derived from a DCPP site boundary drawing (Figure 5.1-1), taking into consideration a 45-degree azimuth sector centered on each 22.50 azimuth sector as described in RG 1.145, Regulatory Position C.1.2.The EAB X/Q values for the radiological releases from each unit are conservatively based on 30 of 205 Diablo Canyon Power Plant Implementation of Alternative Source Terms the EAB distances from the outer edge of each containment building. The release point locations are listed in Table 5.2-1.An LPZ distance of 6 miles (9,654 meters) is used in the analysis. The use of one LPZ distance in all directions from the center of the site for all release points is reasonable given the magnitude of this distance relative to the separation of the release point locations from one another.The containment building cross-sectional area along with the containment building height is used for the annual average xI.Q calculations (used as input to develop the accident X/Q values at the LPZ using Regulatory Guide 1.145 methodology). The applicable methodology is identified in RG 1.111, Regulatory Position C. I.c.(Reference

26) These annual average X/Q values are used to calculate the intermediate averaging time X/Q values for the periods of 2-8 hours, 8-24 hours, 1-4 days, and 4-30 days by logarithmic interpolation.

The following conservative assumptions are made for these calculations:

All releases are treated as point sources* All releases are treated as ground-level as there are no release conditions that are high enough to escape the aerodynamic effects of the plant buildings" The distances from the Unit 1 and Unit 2 releases are determined from the closest edge of the containment buildings to the EAB" The plume centerline from each release is transported directly over the receptor and* A terrain recirculation factor of 4 is used in the calculation of the annual average X/Q values, and* Radioactive decay or plume depletion due to deposition is not considered The highest EAB & LPZ %/Q values from among all 22.50 downwind sectors for each release/receptor combination and accident period are summarized in Table 5.1-1. EAB %IQ values are presented for releases from Unit 1 and Unit 2 while the LPZ X/Q values are applicable to both units. The 0.5% sector dependent X/Q values are presented with the worst case downwind sector indicated in parentheses.

5.2 On-Site Atmospheric Dispersion Factors Regulatory Guide 1.194, June 2003 (Reference 21), Regulatory Position C.1 through C.3, and the adjustment factor for vertically orientated energetic releases from steam relief valves and atmospheric dump valves allowed by Regulatory Position C.6 are used to determine short-term on-site %/Q values in support of design basis radiological habitability assessments. In accordance with Regulatory Position C.2 of RG 1.194, the control room and technical support center X/Q values for radiological releases from DCPP Unit 1 and Unit 2 were calculated using the NRC "Atmospheric Relative CONcentrations in Building Wakes" (ARCON96) methodology as documented in NUREG/CR-6331, Revision 1 (Reference 6). Input data consist of: hourly on-31 of 205 Diablo Canyon Power Plant Implementation ofAlternative Source Terms site meteorological data; release characteristics (e.g., release height, building area affecting the release) and various receptor parameters (e.g., the distance and direction from the release to the control room air intake and intake height).A continuous temporally representative 5-year period of hourly on-site meteorological data from the DCPP onsite meteorological tower (i.e., January 1, 2007 through December 31, 2011) was used for the ARCON96 runs. Each hour of data, at a minimum, had a validated wind speed and direction at the 10-meter level and a temperature difference between the 76- and 10-meter levels. This period of data is temporally representative and meets Safety Guide 23, (Reference

59) guidance.The ARCON96 modeling utilized to establish the Unit 1 and Unit 2 y/Q values follows the ground level release guidance of Regulatory Position C.3 of RG 1.194 (Reference

21) relative to release height (i.e., ground-level vs. elevated), release type (i.e., diffuse vs. point) and configuration of release points and receptors (iLe., building cross-sectional area, release heights, line-of-sight distance between release and receptor, initial diffusion coefficients etc.).All releases were assumed to be ground-level as none of the release points at DCPP meet the definition of an elevated release as indicated in Regulatory Position C.3.2.2 of RG 1.194 (i.e., do not meetthe requirement to be 2.5 times the height of plant buildings).

Only the containment building edge releases were treated as diffuse sources as the releases occur from the entire surface of the building. In these cases, initial values of the diffusion coefficients (ay, a,) were determined in accordance with the guidance provided in RG 1.194, June 2003, Regulatory Position C.3.2.4. Release and receptor locations are applied in accordance with the guidance provided in RG 1.194, Regulatory Position C.3.4 for building geometry and line-of-site distances (refer to Appendix A).The following recommended default values from Regulatory Guide 1.194, June 2003, Table A-2, are judged to be applicable to DCPP: " Wind direction range = 90 degrees azimuth" Wind speed assigned to calm = 0.5 m/sec" Surface roughness length = 0.20 meter and" Sector averaging constant = 4.3 (dimensionless). The following assumptions are made for these X/Q calculations: " The plume centerline from each release is transported directly over the control room or technical support center air intake/receptor (conservative)

The distances from the Unit 1 and Unit 2 containment building surfaces to the receptors are determined from the closest edge of the containment buildings and the source/receptor elevation differences are set to zero (conservative)" The applicable structure relative to quantifying building wake effects on the dispersion of the releases is based on release/receptor orientation relative-to the plant structures 32 of 205 Diablo Canyon Power Plant Implementation ofAlternative Source Terms* The releases from the Unit I and Unit 2 containment building surfaces are treated as diffuse sources* All releases are conservatively treated as ground level as there are no release conditions that merit categorization as an elevated release (i.e., 2.5 times containment building height) at this site" The X/Q value from the accident release point to the center of the control room boundary at roof level is utilized for Control Room in-leakage since the above Y/Q can be considered an average value for in-leakage locations around the Control Room envelope.

The X/Q from the accident release point to the center of the control room boundary at roof level is also utilized for Control Room ingress/egress. The outer doors to the Control Room are located at approximately the middle of a) the east side (i.e., Auxiliary building side) wall of the Control Room and b) the west side (i.e., Turbine building side) wall of the Control Room. Similarly, the X/Q from the accident release point to the center of the TSC at its roof level is utilized for TSC inleakage since the above x/Q can be considered an average value for in-leakage locations around the TSC building envelope.Summarized below are some of the other salient aspects of the DCPP control room and technical support center X/Q analyses, as applicable

Control room receptors within 10-meters of release: In accordance with RG 1.194, Regulatory Position C.3.4, the ARCON96 methodology is not recommended for use at distances less than about 10 meters. Based on engineering judgment, the ARCON96 methodology has been applied for 2 cases when the distance from the release to the receptor is slightly less than 10 meters. Use of ARCON96 methodology for release point-receptors distances less than 10 meters (i.e., 9.4 meters for Unit 1 containment building to Unit 1 control room normal intake and 7.8 meters for Unit 2 containment building to Unit 2 control room normal intake) is considered acceptable since the dominating factors in the calculation are building cross-sectional area and plume meander and not the normal atmospheric dispersion coefficients.

Note that the X/Q values for the above 2 cases were developed to establish the bounding 7IQ values; the referenced X/Q's were not the bounding values and therefore not used in the dose consequence analyses.* Control room receptors at 1.5 meters from release: Since the Unit 1 and Unit 2 MSSVs, 10% ADVs, and MSL break release points are located within 1.5 meters line-of-sight distance from the affected unit's control room normal intake, this near-field distance is considered outside of the ARCON96 application domain. Although ARCON96 is capable of estimating near-field dispersion, the 1.5-meter line-of-sight distance from the releases to the receptors is much less than the 10-meter distance recommended as the minimum applicable distance in Regulatory Position C.3.4 of RG 1.194. Thus no X/Qs are developed for the above release point / receptor combinations. The bullet below provides a discussion of the effect of atmospheric dispersion on releases from the MSSVs/1 0% ADVs with respect to the affected unit's control room normal intake.Atmospheric dispersion is not credited when determining the CR operator dose due to 33 of 205 Diablo Canyon Power Plant Implementation of Alternative Source Terms releases out of the MSL break location in the faulted steam generator. For further detail see Section 7.6.Energetic releases: The vertical velocity of the releases from the MSSVs and 10% ADVs are at least 95 times the 9 5 th percentile wind speed of 1 m/sec and approximately 5 times the highest observed 10-meter wind speed (i.e., 18.9 m/sec) over the 5-year meteorological data base. The large vertical velocities of the MSSV and 10% ADV releases, ranging from 98.9 -94.9 m/sec, preclude any down-washing of the releases by the aerodynamic effects of the containment buildings such that the control room normal intake of the same unit as the release (e.g., Unit 1 MSSV/10% ADV releases to Unit 1 CR normal intake) is not contaminated given that the horizontal distance is only 1.5 meters. Moreover, this short distance precludes the releases from reaching the control room normal intakes of the same unit given the height of the MSSV and 10% ADV releases (i.e., 27.1 and 26.5 meters, respectively) relative to the height of the normal.intakes (i.e., 22 meters). Plume rise calculations indicate that the MSSV and ADV release heights will be enhanced by 11 meters at the 95th percentile wind speed of 1 m/sec due to the large vertical velocities of the releases. Thus, for purposes of estimating dose consequences, it is appropriate to use the x/Q associated with the normal CR intake of the opposite unit for releases from the MSSVs / 10% ADVs as the worst case CR normal intake location.Vertically-oriented energetic releases: Regulatory Position C.6 of NRC Regulatory Guide 1.194 establishes the use of a deterministic reduction factor of 5 applied to ARCON96%/Q values for energetic releases from steam relief valves or atmospheric dump valves.These valves must be uncapped and vertically-oriented and the time-dependent vertical velocity must exceed the 95th-oercentile wind speed at the release point height by at least a factor of 5. Since the DCPP MSSVs and 10% ADVs are vertically oriented /uncapped and will have a vertical velocity of at least 94.9 m/sec until initiation of shutdown cooling at 10.73 hours of the accident, the reduction factor of 5 is applicable to the DCPP MSSV and 10% ADV releases. Note that since XJQ values are averaged over the identified period (i.e., 0-2 hours, 2-8 hours, 8-24 hours, etc.), and the vertical velocity has been estimated to occur for 10.73 hours, application of the factor of 5 reduction is not appropriate for X/Q values applicable to averaging periods beyond the 2-8 hours averaging period. For assessment of an environmental release between T= 8 to 10.73 hrs, continued use of the 2-8 hr 7IQ (with the factor of 5 reduction) is acceptable and conservative. Dual Intakes: The Unit 1 and Unit 2 control room emergency air intakes (also serve the technical support center), may be considered dual intakes for the purpose of providing a low contamination intake regardless of wind direction for any of the release points since the two control room emergency air intakes are never within the same wind direction window; defined as a wedge centered on the line of sight between the source and the receptor with the vertex located on the release point. The size of the wedge for each release-receptor combination is 90 degrees azimuth with the use of ARCON96, as described in Regulatory Position C.3.3.2 of RG 1.194.Redundant radiation monitors: Per RG 1.194, Regulatory Position C.3.3.2.3, based on the dual intake design of the control room pressurization intakes, and as discussed later 34 of 205 Diablo Canyon Power Plant Implementation ofAlternative Source Terms in Section 7.1, the availability of redundant PG&E Design Class I radiation monitors at each pressurization intake (which provide the capability of initial selection of the cleaner intake and support the expectation that the operator will manually make the proper intake selection throughout the event), the y/Q values applicable to the more favorable CR pressurization intake can be reduced by a factor of 4 and utilized to estimate the dose consequences. PG&E DesiQn Class II lines connectinq to PG&E Desiqn Class I Plant Vent: It has been determined that there are several PG&E Design Class It lines that connect to the PG&E Design Class I Plant Vent; specifically, a) the 40 inch containment penetration area (GE/GW) HVAC ventilation line, b) a 2 inch gaseous radwaste system line, and c)the 16 inch gland steam condenser (GSC) / steam jet air ejector (SJAE) exhaust header. In addition, it has been determined that the Plant Vent Expansion Joint may experience a tear during a seismic event, however, the plant vent will remain intact and functional. a) The 40 inch Containment Penetration Area Ventilation line and the 2 inch gaseous radwaste system line were originally designed as PG&E Design Class I, but were subsequently declassified to PG&E Design Class II. As noted in Section 2.2, the portion of these lines that connect to the Plant Vent will be re-classified as PG&E Design Class I.b) The GSC / SJAE 16 inch exhaust header connects to the Plant Vent at El 144'-6" (Centerline) on the North-East side / South-East side of the Unit I and Unit 2 containments, respectively. It has been determined that should a failure occur due to a seismic event, it would most likely occur at the interface of this line and the plant vent.c) The plant vent expansion joint is located at El 155.83' North-East side / South-East side of the Unit 1 and Unit 2 containments, respectively. As discussed earlier, the plant vent expansion joint may experience a tear during a seismic event.An assessment of the potential release locations identified in items b) and c) above indicates that the XIQ values developed for the plant vent are either conservative or representative of these potential release points (i.e., at the interface of the GSC / SJAE 16 inch line and the plant vent or at the plant vent expansion joint). In addition, and for purposes of completeness, the dose impact in the control room due to back flow from the Plant Vent out of a potential break in GSC / SJAE 16 inch exhaust header at locations other than the point of interface with the Plant Vent has been evaluated and demonstrated to be insignificant. The configuration of release point -and *receptor information used in the 7/Q calculations including all significant release and receptor location permutations, release and receptor heights above plant grade, cross-sectional building areas encountered by the release on its path to the receptor, and line-of-sight distance between release and receptor used in the ARCON96 calculations are summarized in Appendix A of this report. A drawing showing the locations of the release points and receptors is also provided in Appendix A (See Figure A-I). Also provided in Appendix A is the hourly 5-year on-site meteorological data from the DCPP onsite 35 of 205 Diablo Canyon Power Plant Implementation ofAlternative Source Terms meteorological tower (i.e., January 1, 2007 through December 31, 2011) which was used in the ARCON96 analyses.Table 5.2-1 provides the release point / receptor combinations evaluated. Tables 5.2-2 and 5.2-3 provide the control room X/Q values for the individual release point-receptor combinations for Unit I and Unit 2, respectively. Note that the specific CR X/Q values used in each of the accident analyses are presented in the accident-specific Tables presented in Chapter 7. The x/Q values selected for use in the Chapter 7 dose consequence analyses are intended to support bounding analyses for an accident that occurs at either unit. They take into consideration the various release points-receptors applicable to each accident to identify the bounding X/Q values and reflect the allowable adjustments / reductions in the values as discussed earlier in this section and summarized in the notes of Tables 5.2-2 and 5.2-3.Table 5.2-4 presents the X/Q values for the individual post-LOCA release point -TSC receptor combinations for Unit 1 and Unit 2 applicable to the TSC normal intake and the center of the TSC boundary at roof level (considered an average value for potential TSC unfiltered in-leakage locations around the envelope). The Unit I and Unit 2 control room pressurization air intakes also serve the TSC during the post-accident pressurization mode. Thus, the %/Qs presented in Tables 5.2-2 and 5.2-3 for the control room pressurization intakes inclusive of the credit for dual intake design and ability to select the more favorable intake are also applicable to the TSC.Presence of the Simulator Building in the vicinity of the DCPP Onsite Meteorological Tower PG&E conducted an analysis to determine whether the presence of the Simulator Building in the vicinity of the DCPP onsite meteorological tower is meaningfully affecting the spatial representativeness of the meteorological data used for atmospheric dispersion calculations. This was done in response to an NRC Request for Additional Information (RAI) dated July 6, 2010 concerning the distance of the location of the onsite meteorological tower being less than ten times the height of the Simulator Building (Regulatory Position C.3 of RG 1.23, Revision 1, Reference

28) and is documented here for completeness.

The distance from the Simulator Building to the meteorological tower is just over seven times the height of the Simulator Building.The assessment of the presence of the Simulator Building on the spatial representativeness of the meteorological data involved several steps.* The first step established the Simulator Building region of influence (ROI) in terms of distance and upwind fetch azimuth, on the DCPP meteorological tower.* The second step determined the source-receptor cases that are within the Simulator Building ROI which may be potentially affected by the aerodynamic effects of the Simulator Building on the local wind field.* The third step involved an analysis comparing pre-building and post-building meteorological data base sigma theta averages based on work performed by Bellinger (Reference

29) and Call (Reference 30).36 of 205 Diablo Canyon Power Plant Implementation ofAlternative Source Terms" The fourth step involved calculating and comparing the j/Q values using ARCON96 (for the control room) and Regulatory Guide 1.145 methodology (for the EAB and LPZ) for the pre-Simulator Building and post-Simulator Building meteorological data bases for the source-receptor cases within the Simulator Building ROI.* The fifth step involved calculating and comparing the X/Q values using ARCON96 (for the control room) and Regulatory Guide 1.145 methodology (for the EAB and LPZ) for the pre-Simulator Building and post-Simulator Building meteorological data bases for the source-receptor cases that were outside the Simulator Building ROI.These analyses were examined to determine whether any differences in sigma theta averages and calculated X/Q values were systematic, as well as for excessively large and outside-the range-of-expected stochastic climatic variations.

A comparison of the y/Q values within the Simulator Building ROI and relevant to the FHA and MSSV / 10% ADV / MSL Break releases, indicated that x/Q values calculated with the pre-Simulator Building onsite meteorological data (1974 to 1978) are always slightly smaller than those calculated with the 2007-2011 onsite meteorological data for all relevant time periods.A comparison of the %/Q values generated by the two different meteorological data bases for the LOCA releases, indicated that for the 0-2 hour period, all but one of the 16 cases showed a small decrease in y/Q values when calculated with the pre-Simulator Building onsite meteorological data. For the longer post-LOCA time periods, the calculated X/Qs demonstrated, for the most part, a combination of small decreases and small increases for the pre-Simulator Building onsite meteorological data versus the post-Simulator Building onsite meteorological data.In order to better determine if the differences in X/Q values developed using the pre-Simulator Building and post-Simulator Building meteorological data bases are primarily the result of expected climatological differences, the LOCA releases for those source-receptor combinations that were outside the Simulator Building ROI were also examined. The magnitude of the variation in the X/Q values were found, for the most part, to be similar to that for the %/Q values within the Simulator Building ROI, and the results reflected a similar combination of increases and decreases. It was concluded that the differences in X/Q values most likely reflect the climatological differences between the 1974-1978 and 2007-2011 data bases, as opposed to the limited aerodynamic effect of the Simulator Building on the wind flow at the meteorological tower location. If there is an effect of the Simulator Building on the meteorological data, it is overshadowed by climatological differences in the data bases.It was therefore judged that the Simulator Building has little to no influence on the DCPP meteorological tower data as it is applied to Gaussian atmospheric dispersion modeling.Moreover, the heuristic uncertainties in Gaussian modeling, which are more than adequately compensated for by conservative assumptions of point releases, centerline calculations and 95% meteorological conditions, are larger than the differences in comparative X/Q values from the meteorological databases. 37 of 205 Diablo Canyon Power Plant Implementation ofAlternative Source Terms Table 5.1-1 Site Boundary Atmospheric Dispersion Factors X/Q (sec/mz)Receptor 0 -2 hours 2 -8 hours 8 -24 hours I -4 days 4 -30 days Unit 1 EAB (NW) 2.50E-04 Unit 2 EAB (SSE) 2.30E-04 Unit 1/2 LPZ (NW) 2.12E-05 9.26E-06 6.26E-06 2.67E-06 7.86E-07 Note 1 An EAB X/Q value of 2.5E-4 sec/M 3 is used for all release points.2. The 0.5% sector dependent %/Q values are presented with the worst case downwind sector indicated in parentheses TABLE 5.2-11 DCPP On-Site Atmospheric Dispersion Factor Evaluation Post-Accident Release Point / Receptor Combinations Release Points On-Site Receptors 1. Unit 1 Containment Building Edge 1. Unit 1 Control Room Normal Intake 2. Unit 2 Containment Building Edge 2. Unit 2 Control Room Normal Intake 3. Unit I Plant Vent 3. Unit 1 Control Room Emergency Intake 4. Unit 2 Plant Vent 4. Unit 2 Control Room Emergency Intake 5. Unit 1 Refueling Water Storage Tank (RWST) Vent 2 5. Control Room Center (i.e., In-leakage)

6. Unit 2 RWST Vent 2 6. TSC Normal Intake6 7. Unit 1 Containment Penetration (GE Area) 7. TSC Center (i.e., In-leakage) 2 8. Unit 2 Containment Penetration (GE Area)9. Unit 1 Containment Penetration (GW/FW Area)10. Unit 2 Containment Penetration (GW/FW Area)11. Unit 1 Fuel Handling Building 12. Unit 2 Fuel Handling Building 13. Unit 1 Equipment Hatch 14. Unit 2 Equipment Hatch 15. Unit 1 Main Steam Safety Valves (MSSVs)16. Unit2 MSSVs 17. Unit 1 10% Atmospheric Dump Valves 4 18. Unit 1 10% Atmospheric Dump Valves 4 19. Unit 1 Main Steam Line Break Location 20. Unit 2 Main Steam Line Break Location Notes: 1. See Appendix A, Table A-1 and Figure A-1 for Release Point / Receptor Locations, including all input data used in the ARCON96 calculations.

2. X/Q s for RWST releases to the control room normal intakes are not needed for the dose calculations since the normal intakes are isolated prior to releases occurring from the RWST vent, 3. X/Q s developed only for the LOCA, i.e., release points I through 10.38 of 205 Ab Diablo Canyon Power Plant Implementation ofAlternatiie Source Terms TABLE 5.2-25 DCPP Unit 1 Control Room Atmospheric Dispersion Factors (sec/m 3)Source and Receptor 0-2 Hour 2-8 Hour 8-24 Hour 1-4 Day 4-30 Day Unit I Containment Edge to Unit I Control Room (CR) Normal Intake 1.28E-03 7.12E-04 2.87E-04 2.90E-04 2.84E-04 Unit 1 Containment Edge to Unit 2 CR Normal Intake 6.52E-04 3.51 E-04 1.51 E-04 1.49E-04 1.37E-04 Unit I Containment Edge to Unit 1 CR Emergency Intake 4 4.11E-04 2.30E-04 9.62E-05 8.69E-05 7.03E-05 Unit 1 Containment Edge to Unit 2 CR Emergency Intake 4 1.67E-04 7.95E-05 2.63E-05 2.81 E-05 2.34E-05 Unit 1 Containment Edge to CR Center 8.85E-04 4.43E-04 1.75E-04 1.77E-04 1.65E-04 Unit 1 Plant Vent to Unit 1 CR Normal Intake 1.67E-03 1.22E-03 4.90E-04 4.90E-04 4.44E-04 Unit 1 Plant Vent to Unit 2 CR Normal Intake 9.1OE-04 6.57E-04 2.68E-04 2.62E-04 2.45E-04 Unit I Plant Vent to Unit 1 CR Emergency Intake 4 5.59E-04 3.38E-04 1.32E-04 1.12E-04 8.38E-05 Unit 1 Plant Vent to Unit 2 CR Emergency Intake 4 2.26E-04 1.48E-04 5.40E-05 5.47E-05 4.45E-05 Unit 1 Plant Vent to CR Center 1.26E-03 8.96E-04 3.44E-04 3.44E-04 2.99E-04 Unit I Containment Penetration (GE Area) to Unit 1 CR Normal Intake 6.84E-03 3.08E-03 1.21 E-03 1.12E-03 8.75E-04 Unit 1 Containment Penetration (GE Area) to Unit 2 CR Normal Intake 2.24E-03 1.15E-03 3.98E-04 3.89E-04 3.20E-04 Unit _ Containment Penetration (GE Area) to Unit 1 CR Emergency Intake 4 3.75E-04 2.33E-04 9.12E-05 8.45E-05 6.62E-05 Unit 1 Containment Penetration (GE Area) to Unit 2 CR Emergency Intake 4 2.55E-04 1.25E-04 4.42E-05 4.38E-05 3.55E-05 Unit 1 Containment Penetration (GE Area) to CR Center 3.22E-03 1.42E-03 5.54E-04 5,20E-04 4.21E-04 Unit 1 Containment Penetration (GW/FW Area) to Unit 1 CR Normal Intake 4.90E-03 3.45E-03 1.37E-03 1.37E-03 1.28E-03 Unit I Containment Penetration (GW/FW Area) to Unit 2 CR Normal Intake 1.38E-03 9.83E-04 3.92E-04 3.88E-04 3.65E-04 Unit I Containment Penetration (GW/FW Area) to Unit 1 CR Emergency Intake 4 8.20E-04 5.40E-04 2.15E-04 1.87E-04 1.43E-04 Unit 1 Containment Penetration (GW/FW Area) to Unit 2 CR Emergency Intake 4 2.58E-04 1.54E-04 4.95E-05 5.26E-05 4.48E-05 Unit 1 Containment Penetration (GWJFWArea) to CR Center 2.59E-03 1.81 E-03 7.29E-04 7.15E-04 6.64E-04 Unit I RWST Vent to Unit 1 CR Emergency Intake 4 3.27E-04 1.90E-04 7.13E-05 6.99E-05 5.76E-05 Unit I RWST Vent to Unit 2 CR Emergency Intake 4 2.1OE-04 9.83E-05 3.73E-05 3.53E-05 2.86E-05 Unit I RWST Vent to CR Center 1.07E-03 4.86E-04 1.99E-04 1.75E-04 1.43E-04 39 of 205 a&Diablo Canyon Power Plant Iniplenmentation ofAlternative Source Terms TABLE 5.2-2' (Continued)

DCPP Unit I Control Room Atmospheric Dispersion Factors (sec/im 3)Source and Receptor 0-2 Hour 2-8 Hour 8-24 Hour 1-4 Day 4-30 Day Unit I MSSVs to Unit 1 CR Normal Intake 1,2 N/A N/A N/A N/A N/A Unit 1 MSSVs to Unit 2 CR Normal Intake 3 4.29E-03 2.76E-03 1.04E-03 1.06E-03 9.46E-04 Unit 1 MSSVs to Unit 1 CR Emergency Intake 3 4.66E-04 2.92E-04 1.16E-04 1.04E-04 8.08E-05 Unit 1 MSSVs to Unit 2 CR Emergency Intake 3'4 3.14E-04 1.53E-04 5.12E-05 5.29E-05 4.38E-05 Unit 1 MSSVs to CR Center 3 1.39E-02 A.40E-03 2.38E-03 2.56E-03 2.15E-03 Unit 1 10% ADVs to Unit 1 CR Normal Intake 1'2 N/A N/A N/A N/A N/A Unit 1 10% ADVs to Unit 2 CR Normal Intake 3 4.30E-03 2.79E-03 1.05E-03' 1.06E-03 9.49E-04 Unit 1 10% ADVs to Unit 1 CR Emergency Intake 3'4 4.66E-04 2.92E-04 1.16E-04 1.04E-04 8.07E-05 Unit 1 10% ADVs to Unit 2 CR Emergency Intake 3'4 3.13E-04 1.54E-04 5.13E-05 5.30E-05 4.39E-05 Unit 1 10% ADVs to CR Center" 3 1.39E-02 7.45E-03 2.39E-03 2.59E-03 2.15E-03 Unit I MSL Break Location to Unit I CR Normal Intake 1 N/A N/A N/A N/A N/A Unit 1 MSL Break Location to Unit 2 CR Normal Intake 4.23E-03 2.90E-03 1.13E-03 1.11E-03 1.02E-03 Unit I MSL Break Location to Unit 1 CR Emergency Intake 4 4.35E-04 2.94E-04 1.15E-04 1.01 E-04 7.76E-05 Unit 1 MSL Break Location to Unit 2 CR Emergency Intake 4 3.06E-04 1.54E-04 5.19E-05 5.32E-05 4.38E-05 Unit 1 MSL Break Location to CR Center 1.24E-02 7.1OE-03 2.24E-03 2.43E-03 2.07E-03 Unit 1 FHB to Unit I CR Normal Intake 6.98E-03 ---Unit 1 FHB to Unit 2 CR Normal Intake 2.93E-03 ----Unit 1 FHB to Unit 1 CR Emergency Intake4 3.31 E-04 ----Unit 1 FHB to Unit 2 CR Emergency Intake 4 2.56E-04 ----Unit I FHB to CR Center 3.78E-03 ----Unit I Equipment Hatch to Unit 1 CR Normal Intake 2.61E-02 ----Unit 1 Equipment Hatch to Unit 2 CR Normal Intake 2.88E-03 -Unit I Equipment Hatch to Unit 1 CR Emergency Intake 4 4.36E-04 -Unit 1 Equipment Hatch to Unit 2 CR Emergency Intake 4 2.64E-04 -Unit I Equipment, Hatch to CR Center 5.51E-03 -40 of 205 Diablo Canyon Power Plant'Implementation ofAlternative Source Terms Notes: i. ARCON96 based %/Q s are not applicable for these cases given that the horizontal distance from the source to the receptor is 1.5 meters (which is much less than the 10 meters required by ARCON96 methodology).

2. Due to the proximity of the release from the MSSVs/1 0% ADVs to the normal operation CR intake of the affected unit, and due 'to the high vertical velocity of the steam discharge from the MSSVs/10%

ADVs, the resultant plume from the MSSVs/10% ADVs will not contaminate the normal operation CR intake of the affected unit.3. For releases from the MSSVs and 10% ADVs (which are uncapped / vertically oriented and have a high vertical velocity discharge for the first 10.73 hours of the accident), a X/Q reduction factor of 5 is applicable to the values listed above. Since X/Q values are averaged over the identified period (i.e., 0-2 hrs, 2-8 hrs, 8-24 hrs, etc), and the vertical velocity has been estimated only up to 10.73 hrs, application of the factor of 5 reduction is not appropriate for X/Q values applicable to averaging periods beyond the 2-8 hrs averaging period. For assessment of an environmental release between T= 8 to 10.73 hrs, continued use of the 2-8 hr X/Q (with the factor of 5 reduction) is acceptable and conservative.

4. The more favorable X/Q value presented above for the CR Pressurization Intakes is further reduced by a factor of 4 to address the "dual intake" credit and the capability of initial selection of the cleaner intake and expectation that the operator will manually make the proper intake selection throughout the event.5. X/Q values for RWST releases to the control room normal intakes are not needed for the dose calculations since the normal intakes are isolated prior to releases occurring from the RWST vent.41 of 205 I;, Diablo Canyon Power Plant Implementation ofAlternative Source Termis TABLE 5.2-35 DCPP Unit 2 Control Room Atmospheric Dispersion Factors (sec/m 3)Source and Receptor 0-2 Hour 2-8 Hour 8-24 Hour 1-4 Day 4-30 Day Unit 2 Containment Edge to Unit 2 CR Normal Intake 1.96E-03 9.42E-04 4.48E-04 3.98E-04 3.18E-04 Unit 2 Containment Edge to Unit 1 CR Normal Intake 6.93E-04 3.84E-04 1.67E-04 1.42E-04 1.08E-04 Unit 2 Containment Edge to Unit 1 CR Emergency Intakel 1.70E-04 1.06E-04 4.23E-05 3.81E-05 2.95E-05 Unit 2 Containment Edge to Unit 2 CR Emergency Intake 4 3.85E-04 1.47E-04 5.94E-05 5.84E-05 4.84E-05 Unit 2 Containment Edge to CR Center 1.08E-03 5.46E-04 2.47E-04 2.12E-04 1.68E-04 Unit 2 Plant Vent to Unit I CR Normal Intake 1.51E-03 9.41E-04 3.86E-04 3.23E-04 2.23E-04 Unit 2 Plant Vent to Unit 1 CR Normal Intake 7.88E-04 4.86E-04 2.01 E-04 1.69E-04 1.17E-04 Unit 2 PlantVent to Unit I CR Emergency Intake4 2.03E-04 1.29E-04 5.13E-05 4.32E-05 3.19E-05 Unit 2 Plant Vent to Unit 2 CR Emergency Intake4 5.71E-04 2,96E-04 1.20E-04 1.04E-04 8.19E-05 Unit 2 Plant Vent to CR Center 1.13E-03 7.08E-04 2.85E-04 2.39E-04 1.70E-04 Unit 2 Containment Penetration (GE Area) to Unit 2 CR Normal Intake 6.71E-03 3.12E-03 1.21 E-03 1.22E-03 1.02E-03 Unit 2 Containment Penetration (GE Area) to Unit 1 CR Normal Intake 2.14E-03 '1.39E-03 5,72E-04 4.83E-04 3.62E-04 Unit 2 Containment Penetration (GE Area) to Unit 1 CR Emergency Intake 4 2.28E-04 1.60E-04 6.25E-05 5.52E-05 4.21 E-05 Unit 2 Containment Penetration (GE Area) to Unit 2 CR Emergency Intake 4 3.97E-04 1.76E-04 6.93E-05 6.44E-05 5.27E-05 Unit 2 Containment Penetration (GE Area) to CR Center 3.16E-03 1.85E-03 7.17E-04 6.84E-04 5.43E-04 Unit 2 Containment Penetration (GW/FW Area) to Unit 2 CR Normal Intake 3.55E-03 1.19E-03 4.82E-04 4.56E-04 3.03E-04 Unit 2 Containment Penetration (GW/FWArea) to Unit I CR Normal Intake 1.22E-03 6.26E-04 2.53E-04 2.12E-04 1.41E-04 Unit 2 Containment Penetration (GW/FWArea)to Unit CR Emergency Intake 4 2.28E-04 1.62E-04 6.58E-05 5,43E-05 3.99E-05 Unit 2 Containment Penetration (GW/FW Area)to Unit 2 CR Emergency Intake 4 8.64E-04 4.23E-04 1.50E-04 1.48E-04 1.20E-04 Unit 2 Containment Penetration (GW/FWArea)to CR Center 2.21E-03 1.17E-03 4.70E-04 3.90E-04 2.61 E-04 Unit 2 RWST Vent to Unit I CR Emergenc Intake 4 1.91E-04 1.21E-04 4.58E-05 4.39E-05 3I53E-05 Unit 2 RWST Vent to Unit 2 CR Emergency Intake4 3.29E-04 1.61 E-04 6.10E-05 5.53E-05 4.45E-05 Unit 2 RWST Vent to CR Center 1.07E-03 5.80E-04 2.18E-04 2.19E-04 1.79E-04 42 of 205 li')Diablo Canyon Power Plant Iniplenientation ofAlternative Source Terms TABLE 5.2-35 (Continued)

DCPP Unit 2 Control Room Atmospheric Dispersion Factors (seclm 3)Source and Receptor 0-2 Hour 2-8 Hour 8-24 Hour 1-4 Day 4-30 Day Unit 2 MSSVs to Unit 1 CR Normal Intake 3 3.87E-03 2.42E-03 9.89E-04 8.17E-04 6.09E-04 Unit 2 MSSVs to Unit 2 CR Normal Intake 1 2 N/A N/A N/A N/A N/A Unit 2 MSSVs to Unit I CR Emergency Intake 3'4 2.89E-04 1.91 E-04 7.45E-05 6.62E-05 5.08E-05 Unit 2 MSSVs to Unit 2 CR Emergency Intake 3'4 4.90E-04 2.29E-04 8.24E-05 8.07E-05 6.49E-05 Unit 2 MSSVs to CR Center" 1.22E-02 8.1OE-03 3.27E-03 2.76E-03 2.08E-03 Unit 2 10% ADVs to Unit 1 CR Normal Intake 3 3.88E-03 2.43E-03 9.94E-04 8.19E-04 6.1OE-04 Unit 2 10% ADVs to Unit 2 CR Normal Intake 1 2 N/A N/A N/A N/A N/A Unit 2 10% ADVs to Unit 1 CR Emergency Intake 3'4 2.87E-04 1.92E-04 7.48E-05 6.61 E-05 5.07E-05 Unit 2 10% ADVs to Unit 2 CR Emergency Intake 3 4 4.90E-04 2.29E-04 8.24E-05 8.08E-05 6.48E-05 Unit 2 10% ADVs to CR Center" 1.22E-02 8.16E-03 3.28E-03 2.78E-03 2.09E-03 Unit 2 MSL Break Location to Unit I CR Normal Intake 3.81 E-03 2.40E-03 1.01 E-03 8.09E-04 5.88E-04 Unit 2 MSL Break Location to Unit 2 CR Normal Intake 1 N/A N/A N/A N/A N/A Unit 2 MSL Break Location to Unit I CR Emergency Intake 4 2.75E-04 1.91 E-04 7.45E-05 6.53E-05 4.86E-05 Unit 2 MSL Break Location to Unit 2 CR Emergency Intake 4 4.76E-04 2.24E-04 8.14E-05 7.94E-05 6.40E-05 Unit 2 MSL Break Location to CR Center 1.09E-02 7.35E-03 3.01 E-03 2.48E-03 1.86E-03 Unit 2 FHB to Unit 1 CR Normal Intake 2.72E-03 -Unit 2 FHB to Unit 2 CR Normal Intake 6.98E-03 -Unit 2 FHB to Unit I CR Emergency Intake" 2.49E-04 -Unit 2 FHB to Unit 2 CR Emergency Intake 4 3.50E-04 -Unit 2 FHB to CR Center 3.71E-03 -.Unit 2 Equipment Hatch to Unit 1 CR Normal Intake 2.49E-03 -Unit 2 Equipment Hatch to Unit 2 CR Normal Intake 2.51 E-02 -Unit 2 Equipment Hatch to Unit 1 CR Emergency Intake 4 2.49E-04 --Unit 2 Equipment Hatch to Unit 2 CR Emergency Intake 4 4.68E-04 --Unit 2 Equipment Hatch to CR Center 5.19E-03 --43 of 205 Diablo Canyon Power Plant Implementation of Alternative Source Terms Notes: 1 ARCON96 based XIQs are not applicable for these cases given that the horizontal distance from the source to the receptor is 1.5 meters (which is much less than the 10 meters required by ARCON96 methodology). 2 Due to the proximity of the release from the MSSVs/10% ADVs to the normal operation CR intake of the affected unit, and due to the high vertical velocity of the steam discharge from the MSSVs/10% ADVs, the resultant plume from the MSSVs/10% ADVs will not contaminate the normal operation CR intake of the affected unit.3 For releases from the MSSVs and 10% ADVs (which are uncapped / vertically oriented and have a high vertical velocity discharge for the first 10.73 hours of the accident), a X/Q reduction factor of 5 is applicable to the values listed above. Since X/Q values are averaged over the identified period (i.e., 0-2 hrs, 2-8 hrs, 8-24 hrs, etc), and the vertical velocity has been estimated only up to 10.73 hrs, application of the factor of 5 reduction is not appropriate for %/Q values applicable to averaging periods beyond the 2-8 hrs averaging period. For assessment of an environmental release between T= 8 to 10.73 hrs, continued use of the 2-8 hr X/Q (with the factor of 5 reduction) is acceptable and conservative. 4 The more favorable X/Q value presented above for the CR Pressurization Intakes is further reduced by a factor of 4 to address the "dual intake" credit and the capability of initial selection of the cleaner intake and expectation that the operator will manually make the proper intake selection throughout the event.5 X/Q values for RWST releases to the control room normal intakes are not needed for the dose calculations since the normal intakes are isolated prior to releases occurring from the RWST vent.44 of 205 1Wt Diablo Canyon Power Plant Implementation ofAlternative Source Ternis TABLE 5.2-4 DCPP Units 1 and 2 Technical Support Center Atmospheric Dispersion Factors (sec/m 3)Source and Receptor 0-2 Hour 2-8 Hour 8-24 Hour 1-4 Day 4-30 Day UNIT I Unit 1 Containment Edge to TSC Normal Intake 2.57E-04 1.18E-04 4.27E-05 4.24E-05 3.50E-05 Unit 1 Containment Edge to TSC Center 2.90E-04 1.33E-04 4.98E-05 4.83E-05 4.02E-05 Unit 1 Plant Vent to TSC Normal Intake 3.12E-04 1.77E-04 6.91E-05 6.29E-05 5.21E-05 Unit 1 Plant Vent to TSC Center 3.54E-04 1.95E-04 7.71 E-05 6.70E-05 5.67E-05 Unit 1 RWST Vent to TSC Normal Intake 2.72E-04 1.27E-04 4.80E-05 4.49E-05 3.71E-05 Unit 1 RWST Vent to TSC Center 2.94E-04 1.38E-04 5.40E-05 4.89E-05 3.97E-05 Unit 1 140' Leakage to TSC Normal Intake 3.64E-04 1.74E-04 6.55E-05 6.14E-05 5.00E-05 Unit 1 140' Leakage to TSC Center 4.27E-04 1.91 E-04 7.45E-05 6.84E-05 5.62E-05 Unit 1 Penetration Leakage TSC Normal Intake 4.80E-04 2.51 E-04 8.31 E-05 8.64E-05 6.95E-05 Unit 1 Penetration Leakage to TSC Center 5.98E-04 3.03E-04 1.04E-04 1.03E-04 8.46E-05 UNIT 2 Unit 2 Containment Edge to TSC Normal Intake 5.48E-04 2.00E-04 8.52E-05 8.37E-05 6.84E-05 Unit 2 Containment Edge to TSC Center 5.57E-04 2.01E-04 8.81 E-05 8.89E-05 6.92E-05 Unit 2 Plant Vent to TSC Normal Intake 5.52E-04 2.35E-04 1.06E-04 8.71 E-05 6.95E-05 Unit 2 Plant Vent to TSC Center 5.43E-04 2.16E-04 9.97E-05 8.11E-05 6.58E-05 Unit 2 RWST Vent to TSC Normal Intake 3.63E-04 1.68E-04 6.47E-05 6.04E-05 4.91E-05 Unit 2 RWST Vent to TSC Center 3.72E-04 1.68E-04 6.64E-05 6.17E-05 5.1OE-05 Unit 2 140' Leakage to TSC Normal Intake 5.47E-04 2.41 E-04 9.36E-05 8.83E-05 7.02E-05 Unit 2 140' Leakage to TSC Center 5.72E-04 2.43E-04 9.75E-05 9.12E-05 7.52E-05 Unit 2 Penetration Leakage TSC Norrmal Intake 1.80E-03 7.72E-04 3.07E-04 2.87E-04 2.33E-04 Unit 2 Penetration Leakage to TSC Center 1.83E-03 7.49E-04 3.16E-04 2.92E-04 2.41E-04 45 of 205 ýM, Diablo Canyon Power Plant Implementation ofAlternative Source Terms Figure 5.1-1 Diablo Canyon Power Plant -Site Drawing)"I,.2 '1'V ~-~ 7<~\ -~YJ~ -? 1--.--1 -~~~L-~ .-/*,-9,-- (9~I.'K.I I'~6'I /21~II-lvI........ ]-46 of 205 Diablo Canyon Power Plant Implementation of Alternative Source Terms 6.0 DOSE CALCULATION METHODOLOGY 6.1 Inhalation and Submersion Doses from Airborne Radioactivity Computer Code RADTRAD 3.03 is used to calculate the committed effective dose equivalent (CEDE) from inhalation and the effective dose equivalent (EDE) from submersion due to airborne radioactivity at offsite locations and in the control room. The summation of CEDE and EDE is reported as the TEDE. As allowed in Section 4.1.4 of RG 1.183, since the submersion exposure is uniform to the whole body, the EDE is used in lieu of the deep dose equivalent (DDE) in determining the contribution of the submersion dose to the TEDE.The CEDE is calculated using the inhalation dose conversion factors provided 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" (Reference 19). The factors in the column headed "effective" yield doses corresponding to the CEDE and are derived based on ICRP-30.The submersion EDE is calculated using the air submersion dose coefficients provided in Table 111.1 of Federal Guidance Report 12, "External Exposure to Radionuclides in Air, Water, and Soil" (Reference 20). The dose coefficients in the column headed "effective" yield doses corresponding to the EDE and are derived based on a semi-infinite cloud model. The submersion EDE is reported as the whole body dose in the RADTRAD 3.03 output.RADTRAD 3.03 includes models for a variety of processes that can attenuate and/or transport radionuclides. It can model the effect of sprays and natural deposition that reduce the quantity of radionuclides suspended in the containment or other compartments. In addition, it can model the flow of radionuclides between compartments within a building, from buildings into the environment, and from the environment into a Control Room (CR). These flows can be through filters, piping, or simply due to air leakage. RADTRAD 3.03 can also model radioactive decay and in-growth of daughters. Ultimately the program calculates the whole body dose, the thyroid dose, and the TEDE dose (rem) to the public located offsite, and to onsite personnel located in the CR due to inhalation and submersion in airborne radioactivity based on user specified, fuel inventory, nuclear data, dispersion coefficients, and dose conversion factors. Note that the code uses a numerical solution approach to solve coupled ordinary differential equations. The basic equation for radionuclide transport and removal is the same for all compartments. The program breaks its processing into 2 parts a) radioactive transport and b) radioactive decay and daughter in-growth. CB&I S&W Inc. computer program PERC2 is used to calculate the CEDE from inhalation and the EDE from submersion due to airborne radioactivity in the Technical Support Center (TSC).PERC2 is a multiple compartment activity transport code with the dose model consistent with Regulatory Guide 1.183 guidance. The decay and daughter build-up during the activity transport among compartments and the various cleanup mechanisms are included. The CEDE is calculated using the Federal Guidance Report No.11 dose conversion factors. The EDE in the TSC is based on a finite cloud model that addresses buildup and attenuation in air. The dose equation is based on the assumption that the dose point is at the center of a hemisphere of the same volume as the TSC. The dose rate at that point is calculated as the sum of typical differential shell elements at a radius R. The equation utilizes the integrated activity in the TSC 47 of 205 Diablo Canyon Power Plant Implementation of Alternative Source Terms air space, the photon energy release rates per energy group from activity airborne in the TSC, and the ANSI/ANS 6.1.1-1991 neutron and gamma-ray fluence-to-dose factors. (Reference 62)Offsite Dose -In accordance with RG 1.183, for the first 8 hours, the breathing rate of the public located offsite is assumed to be 3.5x10-4 m 3/sec. From 8 to 24 hours following the accident, the breathing rate is assumed to be 1.8x1 0-4 m 3/sec. After that and until the end of the accident, the rate is assumed to be 2.3x10-4 m 3/sec. The maximum EAB TEDE for any two-hour period following the start of the radioactivity release is calculated and used in determining compliance with the dose criteria in IOCFR50.67. The LPZ TEDE is determined for the most limiting receptor at the outer boundary of the low population zone and is calculated for the entire accident duration.Control Room Dose -The control room inhalation CEDE is calculated assuming a breathing rate of 3.5x1 0-4 m 3/sec for the duration of the event. The following occupancy factors are credited in determining the control room TEDE: 1.0 during the first 24 hours after the event, 0.6 between 1 and 4 days, and 0.4 from 4 days to 30 days. The submersion EDE is corrected for the difference in the finite cloud geometry in the control room and the semi-infinite cloud model used in calculating the dose coefficients. The following expression obtained from RG 1.183 is used in RADTRAD 3.03 to correct the semi-infinite cloud dose, EDE., to a finite cloud dose, EDEfinite, where the control room is modeled as a hemisphere that has a volume, V, in cubic feet, equivalent to that of the control room.EDEY. VO'338 ED Ey~i,,e -1173 Technical Support Center Dose -The TSC inhalation CEDE is calculated by CB&I S&W Inc.computer code PERC2 assuming the same breathing rate and occupancy factors as those used in determining the control room dose. The submersion EDE developed by PERC2 (which computes the photon fluence at the center of TSC and utilizes the ANSI/ANS 6.1.1-1991 fluence to effective dose conversion factors), is a close approximation of the dose determined using Table 111.1 of FGR No. 12, column headed "effective" (see Section 4.1.4, RG 1.183, RO) and adjusted by the finite volume correction factor given in RG 1.183, RO, Section 4.2.7.6.2 Direct Shine Dose from External and Contained Sources CB&I S&W Inc. point kernel shielding computer program SW-QADCGGP is used to calculate the deep dose equivalent (DDE) in the control room, TSC and at the EAB due to external and contained sources. The calculated DDE is added to the inhalation (CEDE) and the submersion (EDE) dose due to airborne radioactivity to develop the final TEDE. Conservative build-up factors are used and the geometry models are prepared to ensure that un-accounted streaming/scattering paths were eliminated. The dose albedo method with conservative albedo values is used to estimate the scatter dose in situations where the scattering contributions are potentially significant. ANSI/ANS 6.1.1-1977 "neutron and gamma-ray flux-to-dose-rate factors" (Reference

31) is used to convert the gamma flux to the dose equivalent rate.48 of 205 Diablo Canyon Power Plant Implementation of Alternative Source Terms 7.0 RADIOLOGICAL CONSEQUENCES USING AST As discussed in Chapter 1, the methodology

/ scenarios used in the existing design basis accident analyses discussed in the DCPP UFSAR and listed below, are being updated to reflect AST in accordance with the guidance provided in Regulatory Guide 1.183.1. Loss of Coolant Accident 2. Fuel Handling Accident in the Fuel Handling Building 3. Fuel Handling Accident in the Containment

4. Locked Rotor Accident 5. Control Rod Ejection Accident 6. Main Steam Line Break 7. Steam Generator Tube Rupture 8. Loss-of Load Event In addition, the updated analyses reflect the results of a "licensing basis verification I design basis re-constitution" effort that was initiated by PG&E to support a total upgrade of the listed radiological post-accident dose consequence analyses.

Appendix B provides a comparison of the critical input parameter values utilized in the current licensing basis dose consequence analysis versus that used to support this AST application. Also included in this application is the use of updated atmospheric dispersion factors for the site boundary (EAB & LPZ), control room and technical support center. (Refer to Chapter 5.0 for detail)The proposed changes to the current licensing basis that are incorporated in these analyses are summarized in Section 2.1. Section 2.2 provides a summary of the proposed plant modifications. Section 2.3 identifies the key plant operating procedures that will be updated.The full set of impacted procedures will be addressed in the AST ECP.The assumptions and methodology utilized to estimate the dose consequences at the site boundary and in the control room for the listed design basis accidents are summarized in this Chapter. Parameter values are selected to ensure bounding dose consequences applicable to either unit.In accordance with the guidance provided in RG 1.183, the assumptions regarding the occurrence and timing of a Loss of Offsite Power (LOOP) during an accident are selected with the intent of maximizing the dose consequences. A LOOP is assumed for events that have the potential to cause grid perturbation. The dose consequences of the LOCA, MSLB, SGTR, LRA, CREA and LOL event are evaluated with the assumption of a LOOP concurrent with reactor trip. The assumption of a LOOP related to a postulated design basis accident which leads to a reactor trip does not directly correlate to an FHA. Specifically, a FHA does not directly cause a reactor trip and a subsequent LOOP due to grid instability; nor can a LOOP be the initiator of an FHA. Thus the FHA dose consequence analyses are evaluated without the assumption of a LOOP.The worst 2-hour period dose at the EAB, and the dose at the LPZ for the duration of the release, is calculated for each of the above events based on postulated airborne radioactivity 49 of 205 Diablo Canyon Power Plant Implementation ofAlternative Source Terms releases. This represents the post-accident dose to the public due to inhalation and submersion for each of these events. Due to distance/plant shielding, the dose contribution at the EAB / LPZ due to direct shine from contained sources is expected to be negligible forall the accidents. However, for purposes of completeness, the direct shine dose at the EAB following a LOCA was evaluated and the almost insignificant dose contribution (< 0.01 rem)included in the dose estimate.The 30-day integrated dose to an operator in the Control Room due to airborne radioactivity releases is developed for all of the listed design basis accidents. This represents the post-accident dose to the operator due to inhalation and submersion. The CR shielding design is based on the LOCA which represents the worst case DBA relative to radioactivity releases. The direct shine dose due to contained sources / the external radioactive cloud is included in the CR doses reported for the LOCA.In accordance with current licensing basis, the 30-day integrated dose to an operator in the Technical Support Center (TSC) due to immersion, inhalation and direct shine is evaluated for the DBA that has the worst case radioactivity release, i.e., the LOCA.7.1 Control Room Design I Operation I Transport Model The DCPP main control room (CR) serves both units and is located at El 140' of the Auxiliary Building. The walls facing the Unit 1 and Unit 2 containments (i.e., the north and south walls)are made of 3'-0" concrete, whereas the CR east and west walls are made up of 2'-0" concrete.The floor and ceiling thickness, / material reflect a minimum of 2'-0" and 3'-4" of concrete, respectively. The CR Mechanical Equipment and HVAC room is located adjacent to the CR (east side), at El 154'-6".The CR has a normal intake per unit (each located on opposite sides the Auxiliary Building; i.e., north and south), and a pressurization flow intake per unit (each located on either side of the Turbine Building, i.e., north and south). The DCPP CR pressurization air intakes have dual ventilation outside air intake design as defined by Regulatory Position C.3.3.2 of RG 1.194.(See Section 5.2 for additional details)During normal operation (CRVS Mode 1), both CR normal intakes are operational. Redundant PG&E Design Class I radiation monitors located at each CR normal intake (1-RE-25126, 2-RE-25/26) have the capability of isolating the CR normal intakes on detection of high radiation and switching the CRVS to Mode 4 operation (i.e., CR filtered intake and pressurization). Other signals that initiate CRVS Mode 4 operation include the safety injection signal (SIS) and Containment Isolation Phase A. The SIS does not directly initiate CRVS Mode 4, however, it initiates Containment Isolation Phase A which initiates Mode 4.CRVS Mode 4 operation utilizes redundant PG&E Design Class I radiation monitors located at each CR pressurization air intake and the provisions of acceptable control logic to automatically select the least contaminated inlet at the beginning of the accident, and manually select the least contaminated inlet during the course of the accident. Thus, during Mode 4 operation the dose consequence analyses can utilize the %/Q values for the more favorable pressurization air intake reduced by a factor of 4 to credit the "dual intake" design (See Section 5.2 for additional 50 of 205 Diablo Canyon Power Plant Implementation ofAlternative Source Terms details).During normal operations, 2100 (+/-10%) cfm of unfiltered air is drawn into the control room envelope (i.e., -170,000 ft 3 of free volume) from the Unit 1 and Unit 2 normal intakes for a total of 4200 (+/-10%) cfm. In response to a CR radiation monitor or SI signal, the CR switches to CRVS Mode 4 operation, and control logic ensures that the CRVS pressurization fan of the non-accident unit is initiated and air is taken from the less contaminated of the Unit I or Unit 2 CR pressurization air intakes. The pressurization flow at either intake ranges between 650 -900 cfm. The CR pressurization flowrate used in the dose consequence analyses is selected to maximize the estimated dose in the control room. With the exception of 100 cfm which is assumed to be unfiltered due to backdraft damper leakage, the pressurization flow is filtered.The allowable methyl iodide penetration and filter bypass for the CRVS Mode 4 Charcoal Filter is controlled by DCPP TS 5.5.11 and is <2.5% and <1%, respectively. In accordance with the NRC SER for License Amendment Nos. 163 and 165, PG&E has committed to the test methods of ASTM D3803-1989, and thus in accordance with the guidance provided in GL 99-02 (Reference 41), use is made of a safety factor of 2 in determining the charcoal filter efficiency to be used in safety analyses. Thus the CR charcoal filter efficiency for elemental and organic iodine used in the DCPP safety analyses is 100% -[(2.5% + 1%) x 2] = 93%. The acceptance criteria for the in-place test of the high efficiency particulate air (HEPA) filters in DCPP TS 5.5.11 is a "penetration plus system bypass" < 1.0%. Thus using methodology similar to the charcoal filters, the HEPA filter efficiency for particulates used in the DCPP safety analyses is 100% -[(1%) x 2] = 98%.During Mode 4 operation, the CR air is also recirculated and a portion of the recirculation flow filtered through the same filtration unit as the pressurization flow. The range of the flow through the filter banks is 1800 -2200 cfm with the minimum filtered recirculation flow being 1250 cfm.Unfiltered inleakage into the CR during Mode 1 and Mode 4 is assumed to be 70 cfm (includes 10 cfm for ingress/egress based on the guidance provided in SRP 6.4). Note that the December 2012 Control Room Tracer Gas Test recorded a maximum unfiltered inleakage of 37 cfm (i.e., 32+/-5 cfm). (Reference 50)The CRVS Mode 4 parameter values assumed in the dose consequence analyses are summarized below. These values encompass the results' of the recent control room tracer gas test.Mode 4 CR Parameters Min Flow (cfm) Max Flow (cfm)Pressurization Flow 650 900 Backdraft damper Lkg. 100 100 Filtered Intake 550 800 Charcoal Filter Flow 1800 2200 Filtered Recirc Flow 1250 1400 Unfiltered Inleakage 70 70 CR Exhaust Flow 720 970 51 of 205 aDiablo Canyon Power Plant Implementation ofAlternative Source Terms For purposes of estimating the post-accident dose consequences, the DCPP control room (CR) is modeled as a single region. When in CRVS Mode 4, the Mode 1 intakes are isolated and outside air is a) drawn into the CR through the filtered emergency intakes; b) enters the CR as infiltration, c) enters the CR during operator egress/ingress, and d) enters the CR as unfiltered leakage via the emergency intake back draft dampers. The direction of flow uncertainty on the CRVS ventilation intake flowrates (normal as well as accident), are selected to maximize the dose consequence in the CR.As discussed in Section 7.0, the dose consequence analyses for the events that can cause grid instability and credit CRVS Mode 4 operation (i.e., the LOCA, MSLB, SGTR and the CREA), assume a Loss of Offsite Power (LOOP) concurrent with reactor trip.In accordance with current licensing basis the non-accident unit is assumed unaffected by the LOOP. Thus, to address the effect of a LOOP, and taking into consideration the fact that the time of receipt of the signal to switchover from CRVS Mode 1 to Mode 4 is dependent on the time of reactor trip and is therefore accident specific:* Automatic isolation of the CR normal intake of the "non-accident" unit, is delayed by 12 seconds from receipt of the signal to switch to CRVS Mode 4. This delay takes into account a 2 second SI signal processing time and a 10 second damper closure time.* Automatic isolation of the CR normal intake of the "accident" unit, and credit for CRVS Mode 4 operation is delayed by 38.2 seconds from receipt of the signal to switch to CRVS Mode 4. This delay takes into account a) 28.2 seconds for the diesel generator to become fully operational including sequencing delays, and b) 10 seconds for the CR ventilation dampers to re-align. The 2 second Sl signal processing time occurs in parallel with diesel generator sequencing and is therefore not included as part of the delay. In addition, and as discussed earlier, the CRVS system design ensures that upon receipt of a signal to switch to Mode 4, the CR pressurization fans of the non-accident unit is initiated; thus fan ramp-up is assumed to occur well within the 38.2 seconds delay discussed above, unhampered by a LOOP.As discussed in Section 7.0, the FHA dose consequence analyses do not address the potential effects of a LOOP.The dose consequence analyses for the LRA and the LOL event assume that the CR remains in normal operation mode and do not credit CRVS Mode 4 operation. Table 7.1-1 lists key assumptions/ parameters associated with DCPP control room design.7.2 Loss of Coolant Accident (LOCA)The accidental rupture of a main coolant pipe is the event assumed to initiate a large break LOCA. Analyses of the response of the reactor system, including the emergency core cooling system (ECCS), to ruptures of various sizes are presented in DCPP UFSAR Sections 15.3.1 and 15.4.1. As demonstrated in these analyses, the ECCS, using emergency power, is designed to keep cladding temperatures well below melting and to limit zirconium-water reactions to an insignificant level. However, as a result of the increase in cladding temperature 52 of 205 Diablo Canyon Power Plant Implementation of Alternative Source Terms and the rapid depressurization of the core, some cladding failure may occur in the hottest regions of the core. Following the cladding failure, some of the core fission products would be released to the primary coolant and subsequently to the inside of the containment building.There are several passive and active fission product removal mechanisms available inside containment. Active mechanisms include radioactive particulate and iodine removal by the containment sprays inclusive of the containment air mixing provided by the containment fan coolers. DCPP UFSAR Section 6.2 describes the design and operation of the containment spray system and the containment fan coolers.RG 1.183, Appendix A, identifies the large break LOCA as the design basis case of the spectrum of break sizes for evaluating performance of release mitigation systems and the containment, and for facility siting relative to radiological consequences. AST methodology as provided by RG 1.183 presents a more credible accident scenario than the instantaneous fuel damage scenario depicted in TID-14844 with respect to fission product releases from the core following a LOCA, and the timing and chemical form of such releases.The core damage sequence of a AST LOCA scenario as defined by RG 1.183 addresses a delayed radioactivity release, i.e., a gap release starting at t=30 secs, followed by fuel melt starting at t = 30 mins and continuing on to t = 1.8 hrs. At DCPP, containment spray in the injection mode is exhausted within approximately an hour after accident initiation, or earlier if full safeguards is available. Thus in order for the containment spray to continue to be effective as a fission product removal mechanism, the sprays have to be made available beyond the injection mode and continue on in the recirculation mode.7.2.1 Use of Containment Spray in the Recirculation Mode DCPP is designed and licensed to operate using containment spray in the recirculation mode.In accordance with current licensing basis, and as documented in the NRC SER related to License Amendment No. 139 to Facility Operating License No's DPR-80 and DPR-82 (Reference 54), containment spray is not required per analyses to be actuated during recirculation, but may be actuated in accordance with the EOPs or at the discretion of the Technical Support Center. With this application, TS Bases 3.6.6 and the associated emergency operating procedures will be updated to require initiation of containment spray in the recirculation mode from the control room within 12 minutes of termination of injection spray.Minimum Core flow rate and Containment Spray flow rate when CS is operating in the recirculation mode.The minimum ECCS core flow available when containment spray is operating in the recirculation mode is 713.4 gpm in addition to the spill flow via the break. The above value is greater than the required core flow rate acceptance criteria of 709.6 gpm and is based on a single active failure of Train B to minimize the available pumps for core cooling. Thus the minimum ECCS core flow is based on operation with 1 Residual Heat Removal (RHR) Pump / I Safety Injection Pump (SI) Pump / 1 Centrifugal Charging Pump (CCP). Note that the minimum required ECCS core flow of 709.6 gpm was determined by Westinghouse as 1.2 times the core boil-off with Replacement Steam Generators. The boil-off rates are conservatively calculated based on the minimum 53 of 205 Diablo Canyon Power Plant Implementation ofAlternative Source Terms RWST drain-down time of 24 minutes, and an assumed saturation temperature of 212°F for the ECCS fluid downstream of the RHR heat exchanger. The predicted minimum ECCS core flow rate of 713.4 gpm when operating in the recirculation spray mode is based on the calculated ECCS fluid temperature downstream of the RHR heat exchanger; thus the margin between the minimum ECCS core flow rate when operating in the recirculation spray mode, and the acceptance criteria will increase if it is adjusted to 2121F.The minimum containment recirculation spray flow rate is determined to be 1211 gpm.The above value is based on Valves 8809A and B being closed during the recirculation spray mode which is in accordance with current licensing basis. A single active failure of the Train-A RHR pump is assumed to minimize the CS flow and maximize the flow to the core. In summary, the minimum CS flowrate is based on operation with 1 RHR Pump / 2 SI Pumps / 2 CCPs.7.2.2 Activity Release Pathways following a LOCA DCPP has identified six activity release paths following a LOCA 1. Release via the Containment Pressure / Vacuum Relief pathway to the environment until the containment isolation valves are closed.2. Containment leakage to the environment after containment isolation is achieved.3. Sump water leakage from ESF systems that recirculate sump water outside containment.

4. Failure of the RHR pump seal at T=24 hrs resulting in a 50 gpm leak of sump water for 30 mins.Note: DCPP design includes an ESF atmosphere filtration system, so from a regulatory standpoint per SRP 15.6.5, Appendix B (Reference 51), as well as RG 1.183, inclusion of this leakage path in dose consequences is not required.

However, the RHR pump seal failure resulting in a "filtered" release is DCPP's licensing basis with respect to passive single failure, and will be maintained for this application. Specifically,-UFSAR Section 3.1.1. 1 (Single Failure Criteria / Definitions), Item 2; discusses passive failures -"The structural failure of a static component that limits the component's effectiveness in carrying out its design function. When applied to a fluid system, this means a break in the pressure boundary resulting in abnormal leakage not exceeding 50 gpm for 30 minutes. Such leak rates are assumed for RHR pump seal failure."-UFSAR Appendix 6.3A.3.2 (discusses passive failures), indicates that -the design of the auxiliary building and related equipment is based on handling of leaks up to a maximum of 50 gpm. Means are provided to detect and isolate such leaks in the emergency core cooling pathway within 30 mins-UFSAR Section 15.5.17.2.8 indicates that -failure of an RHR pump seal at 24 hrs is assumed as the single failure that can be tolerated without loss of the required functioning of the RHR system 54 of 205 Diablo Canyon Power Plant Implementation ofAlternative Source Terms Therefore, the RHR Pump Seal Failure is retained as a release pathway for the AST dose consequence analysis.5. Releases to the environment from the MEDT which collects component leakage hard-piped to the MEDT. The collected fluid includes both post-LOCA sump water and other non-radioactive fluid.6. Releases to the environment via the refueling water storage tank (RWST) vent due to post-LOCA sump fluid back-leakage into the RWST via the mini-flow recirculation lines connecting the high head and low head safety injection pump discharge piping to the RWST.The DCPP LOCA dose consequence analysis follows the guidance provided in the pertinent sections of RG 1.183 including Appendix A. Table 7.2-1 lists the key assumptions / parameters utilized to develop the radiological consequences following a LOCA at either unit.7.2.3 Dose from Submersion and Inhalation NRC sponsored computer code RADTRAD 3.03, is used to calculate the control room and site boundary dose due to airborne radioactivity releases following a LOCA.7.2.3.1 Containment Pressure [Vacuum Relief Line Release In accordance with RG 1.183, Appendix A, Section 3.8, for containments such as DCPP that are routinely purged during normal operations, the dose consequence analysis must assume that 100% of the radionuclide inventory in the primary coolant is released to the containment at the initiation of the LOCA. The inventory of the release from containment should be based on primary coolant equilibrium activity as allowed by the Technical Specifications (see Table 4.2-1).Iodine spikes need not be considered. Thus, in accordance with the above guidance, the 12 inch containment vacuum / over pressure relief valves are assumed to be open to the extent allowed by DCPP Technical Specifications (i.e., blocked to prevent opening beyond 50 degrees) at the initiation of the LOCA, and the release via this pathway terminated as part of containment isolation. The analysis assumes that 100% of the radionuclide inventory in the primary coolant, assumed to be at Technical Specification levels, is released to the containment at T= 0 hours. It is conservatively assumed that 40% of release flashes and is instantaneously and homogeneously mixed in the containment atmosphere, and that the activity associated with the volatiles, i.e., 100% of the noble gases and 40% of the iodine in the reactor coolant, is available for release to the environment via this pathway.Containment pressurization (due to the RCS mass and energy release), combined with the relief line cross-sectional area, results in a 218 acfs release of containment air to the environment for a conservatively estimated period of 13 seconds. Credit is taken for pressure boundary integrity of the containment pressure / vacuum relief system ductwork which is classified as PG&E Design Class II, and seismically qualified; thus, environmental releases are via the Plant Vent.55 of 205 Diablo Canyon Power Plant Implementation of Alternative Source Terms Since the release is isolated within 13 seconds after LOCA, i.e., before the onset of the gap phase release, releases associated with fuel damage are not postulated. The chemical form of the iodine released from the RCS to the environment is assumed to be 97% elemental and 3%organic.7.2.3.2 Containment Leakage The inventory of fission products in the reactor core available for release into the containment following a LOCA is provided in Table 4.1-1 which represents a conservative equilibrium reactor core inventory of the dose significant isotopes assuming maximum full power operation at 1.05 times the current licensed thermal power, and taking into consideration fuel enrichment and burnup. The notes provided at the bottom of Table 4.1-1 provide information on isotopes used to estimate the inhalation and submersion doses following a DCPP LOCA, vs isotopes that are considered to estimate the post-LOCA direct shine dose.Per RG 1.183, the fission products released from the fuel are assumed to mix instantaneously and homogeneously throughout the free air volume of the primary containment as it is released from the core. Containment spray is utilized as one of the primary means of fission product cleanup following a LOCA. Mixing of the "effectively" sprayed volume of containment, with the unsprayed volume of the containment is enhanced by operation of the PG&E Design Class I containment fan coolers. In order to quantify the effectiveness of the containment spray system, both the volume fraction of containment that is sprayed, and the mixing rate between the sprayed and unsprayed volumes are quantified. The LOCA dose consequence analysis is based on an assumed worst case single failure of loss of one ESF train. A single train ESF consists of one train of ECCS, one train of CSS, and two Containment Fan Cooling Units (CFCUs). A single train scenario is selected to be consistent with the use of reduced iodine and particulate removal coefficients associated with single train operation. 7.2.3.2.1 Spray Duration Containment Spray in the injection mode is initiated at 111 seconds after the LOCA and terminated at 3798 seconds. Manual operation is credited to initiate containment recirculation spray. Thus, based on single train operation, containment spray in the recirculation mode is initiated at 4518 seconds (i.e., twelve (12) minutes after injection spray is terminated), and terminated at 22,518 seconds. In summary, containment spray operation (injection plus recirculation) is credited until 6.25 hrs post-LOCA, with a twelve minute gap after injection spray is terminated. 7.2.3.2.2 Effectively SprayedVolume Fraction of Containment The current licensing basis containment sprayed volume for the spray injection mode is calculated based on the assumption that the unoccupied containment volume above the operating floor is 100% covered by sprays. It includes the sprayed volume below the grating in the operating floor deck, and the refueling cavity volume. The percentage of the total containment free volume that is sprayed is 82.5%.56 of 205 Diablo Canyon Power Plant Implementation ofAlternative Source Terms To justify the assumption that the unoccupied containment volume above the operating floor is 100% covered by sprays, the analysis supporting the current licensing basis containment spray coverage during the spray injection mode with only one CSS train operating estimated the projected unsprayed area percentage of the containment deck area of 42% for Unit 1 and 44%for Unit 2, using actual spray flow patterns. The spray reduction factor of 0.5 used to address the spray compression effect due to elevated containment pressure was conservatively based on the containment design pressure. It was concluded that the whole volume above the operating floor can be considered well mixed and totally subjected to the sprays given the estimated spray coverage calculated above based on actual spray patterns, and the high levels of turbulence due to spray action, entrainment of the air from the volume outside the spray patterns, and induced upflow of air to balance the downflow experienced in the sprayed volume.In support of this application, an analysis was performed to evaluate the effect of the reduction in spray flow rate between the spray injection phase (i.e., 2456 gpm) and the spray recirculation phase (i.e., 1211 gpm) on the containment spray coverage. The minimum volumetric flow rate of water through the spray nozzles, the associated nozzle water pressure drop, and Figure 1 of NUREG/CR-5966 (Reference 33), were used to establish the effect of the reduction in spray flow rate on the spray pattern. The analysis determined that with the continued use of a spray reduction factor of 0.5 for spray compression, there was a 3% increase in the "unsprayed area percentage of the containment deck" for both Units. It concluded that the use of a spray pattern reduction factor of 0.5 is extremely conservative when applied to the recirculation mode (the containment pressure is substantially lower than the design pressure), and that use of a sprayed volume of 82.5% of the containment free volume, is acceptable, for both the containment spray injection as well as the containment spray recirculation mode.7.2.3.2.3 Mixing between Sprayed and Unsprayed Regions of Containment The PG&E Design Class. I containment fan cooler units support post-LOCA mixing of the sprayed and unsprayed volume of the containment at a rate higher than that justified by natural convection. The containment mixing rate between the sprayed and unsprayed regions following a LOCA is determined to be 9.13 turnovers of the unsprayed regions per hour. This mixing rate is based on the operation of two Containment Fan Coolers Units (CFCUs) with a total volumetric flow rate of 68,000 cfm, between the unsprayed regions and sprayed regions. The design flow for each CFCU is 47,000 cfm; the value used to determine the mixing rate addresses surveillance margins and uncertainty. Review of the layout and arrangement of the intake and exhaust registers of the CFCUs indicate that the air intakes are all located above the operating floor (sprayed region) and the air discharge registers are all located below the operating floor in the unsprayed region. Additional review of the containment configuration including the location of major openings, and the various active and passive mixing mechanisms, results in the conclusion that following a LOCA, credit can be taken for a) the entire flowrate provided by each operating CFCU to support mixing between the sprayed and unsprayed regions, and b) homogeneous mixing within the sprayed and unsprayed regions, of the volume of air transferred from one region to the other due to CFCU operation. CFCU operation is initiated at 86 seconds after the LOCA and operates for the duration of the accident. In accordance with RG 1.183, Appendix A, Section 3.3, prior to CFCU initiation, the dose consequence model assumes a mixing rate attributable to 57 of 205 Diablo Canyon Power Plant Implementation ofAlternative Source Terms natural convection between the sprayed and unsprayed regions of 2 turnovers of the unsprayed region per hour.7.2.3.2.4 Fission Product Removal RG 1.183, Appendix A, Section 3.3, invokes SRP 6.5.2 (Reference

42) and NUREG/ CR 5966 (Reference

33) as acceptable models for removal of iodines and particulates.

Regulatory guidance provided in*SRP 6.5.2 outlines simplified methodology to develop steady state and conservative iodine and particulate removal coefficients in the containment for post-LOCA fission products. However, since with implementation of AST, the releases from the core are assumed to be predominantly particulate in nature, refinement of the SRP 6.5.2 methodology in determining particulate removal coefficients is deemed appropriate. RG 1.183, Appendix A, Section 3.3, permits the use of time-dependent particulate aerosol removal coefficients by invoking NUREG/CR 5966, and indicates that no reduction in particulate aerosol removal coefficients is required when a DF of 50 is reached if the removal rates are based on the calculated time-dependent airborne aerosol mass.Thus the fission product removal coefficients developed for the LOCA reflect the following guidance documents: " Elemental iodine removal coefficients are calculated using guidance provided in SRP 6.5.2 which is invoked by Regulatory Guide 1.183, Appendix A, Sec 3.3* Time dependent particulate aerosol removal coefficients are estimated using guidance provided in RG 1.183 Appendix A, Sec 3.3, for alternative source terms, and use of CB&I S&W Inc. computer program SWNAUA The total elemental iodine and particulate removal coefficients in the sprayed and unsprayed region of the containment as a function of time are summarized in Table 7.2-2. The methodology utilized to develop these values is summarized below.1. Particulate Removal There are several aerosol mechanics phenomena that promote the depletion of aerosols from the containment atmosphere. These include the natural phenomena of particle growth due to agglomeration, gravitational settling of particles (also called gravitational sedimentation), diffusiophoresis (particulate removal due to steam condensation); and removal by fluid mechanical interaction with the falling droplets that enter the containment atmosphere through the spray system nozzles (i.e., containment spray).All of the above phenomena are credited for DCPP. Agglomeration of the aerosol is considered in both sprayed and unsprayed regions. In the sprayed region, the particulate removal calculation takes credit for the removal effectiveness of diffusiophoresis and sprays.Gravitational settling is considered only in the unsprayed region.58 of 205 Diablo Canyon Power Plant Implementation ofAlternative Source Terms The methodology presented below envelopes both Unit 1 and Unit 2 and addresses the development of time-dependent particulate aerosol removal coefficients.

a. Removal of Particulates by Sprays The particulate removal rate is calculated using CB&I S&W Inc. computer code SWNAUA.Computer code SWNAUA is a derivative of NAUA/MOD 4 (Reference 40). The results of SWNAUA have been accepted by the NRC for the AST applications supporting the design certification of CE System 80+, and for operating nuclear plants Beaver Valley Power Station[ML032530204]

and Fort Calhoun Station [ML013030027]. The NAUANMOD4 code does not include a model for aerosol removal by sprays. The aerosol removal model for sprays was developed and incorporated into the SWNAUA code by CB&I S&W Inc. as a conservative model suitable for design basis accident calculations. The model correlations implemented into SWNAUA conservatively underestimate the spray removal coefficient. The spray model incorporated in the SWNAUA code was originally described in Reference

34. When performing DBA calculations to determine particulate removal in the effectively sprayed region of the containment, only the conservatively developed spray removal models and conservative steam condensation rates for the diffusiophoresis calculation are utilized.

While agglomeration is considered in the calculation, its impact on the resulting particulate removal rates is negligible. In summary, the aerosol removal rates calculated by SWNAUA are conservative lower bound estimates. The spray model in SWNAUA evaluates the particulate removal efficiency for each particle size in the aerosol by the following mechanisms: inertial impaction, interception, and Brownian diffusion. The aerosol removal constant due to spray is presented in NUREG-0772 (Reference

35) as: Xspray 4 3 F m h s Vspray -Vsed sy4 Rsp w V V spray Where'\spray " Particulate removal constant for spray (sec 1)Fm = Spray mass flow rate (gm/sec)h = Spray fall height (cm)= Collision efficiency Rsp = Spray droplet radius (cm)Pw = Density of the spray droplet (gm/cm 3)V = Effectively sprayed volume of containment (cm 3)Vspry = Velocity of the spray droplets (cm/sec)Vsed = Aerosol sedimentation velocity (cm/sec)The collision efficiency is divided into three contributing mechanisms as described in BMI-2104 (Reference 36): 59 of 205 Diablo Canyon Power Plant Implementation of Alternative Source Terms Where E = Efficiency due to inertial impaction, Er Efficiency due interception and Ed Efficiency due to Brownian diffusion.

For viscous flow around the spray droplet, the inertial impaction efficiency is given in NUREG-0772 (Reference 35): 1 Stk -1.214 The critical Stokes number, Stk, for viscous flow is 1.214; for Stk below this value, the model assumes the efficiency of inertial impaction is 0.0. The Stk is calculated from BMI-2104 (Reference 36): 2 pP r 2 Cc (Vspray -Vsed)Stk -= _ _ _ _ _ _ _ _9 ýJ Rsp Where r Aerosol particle radius (cm)pp = Aerosol density (gm/cc)C = Cunningham slip correction factor, P = Gas viscosity (gm/(cm-sec)) For droplet sizes typical of nuclear plant spray systems, the data of Walton and Woolcock (Reference

37) show that collision efficiency will be closer to that predicted for potential flow around the droplet. Calvert (Reference

38) fitted this data to the expression:

EStk )2 Stk +0.7 The collision efficiency predicted by this equation is always higher than that predicted by the viscous flow expression given above. The Calvert's fit is employed in this calculation. As for the remaining constituents of the collision efficiency, the spray model employs an interception efficiency of the form:.i j 2 2 3 1RsX I Rs 60 of 205 Diablo Canyon Power Plant Implementation of Alternative Source Terms which is a conservative approximation of the expression given by BMI-2104 (Reference 36).The efficiency due to Brownian motion is also taken from this report: Sd<=3.5pe 2/3 Where Pe = Peclet number-2 vsPraYRsp/DB, DB = Aerosol diffusion coefficient (cm 2/sec)= kBoItTB (Fuchs, Reference 39, p. 181), kBoItz = Boltzmann constant= 1.3804 X 1016 (erg/°K).T = Temperature (°K)Fuchs (Reference 39, p. 27) gives the aerosol mobility, B: In most cases, the overall collision efficiency is dominated by inertial impaction, but for small aerosols, Brownian diffusion may become dominant. The collision efficiency due to inertial impaction increases as the aerosol size is increased, whereas that due to Brownian diffusion increases as the aerosol size decreases. The model has the capability of handling a distribution of up to 20 droplet radii with the spray removal efficiency being determined for each aerosol size bin.Elia and Lischer (Reference

34) investigated the use of a single spray droplet size in the analysis instead of a drop size distribution.

While Reference 34 does not specifically analyze the DCPP spray system, the parameter sensitivities for the spray model are applicable. The paper demonstrates that the droplet diameter distribution can be represented by a single diameter that is the mass mean diameter. The case 6 droplet distribution presented in the paper is for the SPRACO 1713A nozzle that is frequently used by the nuclear industry for fission product/heat removal spray systems. This diameter approximates that used in case 1, 1000p.The spray flow rate used for both case 1 and case 6 is 10,000 gpm. Table 2 in the paper indicates that the spray removal rates for these two cases are very close. The mass mean spray droplet radii for DCPP are specified in the table below.The paper also investigated the variation of particulate removal coefficient with droplet diameter.Cases 1 through 3 vary the mass mean droplet diameter from 500li to 1500p. Although Table 2 in the paper indicates that these cases assume a spray flow of 10,000 gpm, the spray removal coefficient reduction by about 67 percent is expected to be independent of spray flow rate.The bounding plant parameters for the DCPP Units are listed below.61 of 205 Diablo Canyon Power Plant Implementation of Alternative Source Terms Plant Parameters for Fission Product Cleanup Calculations Parameter Value Sprayed Containment Volume 5.960 x 1010 cm 3 Fall Height 3,536 cm Spray Flow Rate 2,456 gpm (111 -3,798 sec)0 gpm (3,798 -4,518 sec)1,211 gpm (4,518 -22,518 sec)Spray droplet radius 500 x 10-4 cm (111- 3,798 sec)*500 x 104 cm: (4,518 -22,518 sec)* Spray droplet radius during the injection phase conservatively assigned the larger droplet radius applicable to the recirculation phase.The DCPP spray coverage fraction of 82.5% is utilized for the duration of injection and recirculation spray.The DCPP containment pressure, temperature, and relative humidity transient data following the limiting DBA are presented in Table 7.2-2A.Description of Aerosol The chemical composition of the aerosol is only important as it relates to the density of aerosol utilized in the development of spray removal rates. The chemical composition during the gap release phase is assumed to be pre-dominantly CsOH. The chemical composition during the in-vessel release phase is assumed to be 20 percent CsOH, 20 percent indium, and 60 percent silver. These assumed compositions are based on a review of the SASCHA experimental results. The aerosol input data for SWNAUA are provided below.Description of Aerosol Minimum Aerosol Radius ]1.0000E-07 cm Maximum Aerosol Radius. 1.0000E-02 cm Maximum Number of Aerosol Size Bins 100 From 30 sec to 1830.0 sec Aerosol Injection Rate 10.29 (gm/sec)Mean Geometric Radius 7.50000E-06 cm Geometric Standard Deviation 1.56 Aerosol Density 3.7 gm/cc From 1830 sec to 6510.0 sec Aerosol Injection Rate 88.74 (gm/sec)Mean Geometric Radius 4.OOOOOE-05 cm Geometric Standard Deviation 1.46 Aerosol Density 4.6 gm/cc 62 of 205 MDiablo Canyon Power Plant Implementation ofAlternative Source Terms Removal of Particulates by Diffusiophoresis Particulate matter is entrained in the steam as it flows to the condensation surfaces. This phenomenon is called diffusiophoresis. Steam is assumed to condense on spray droplets, on the containment fan cooler units, and on heat sinks. The diffusiophoresis model in the SWNAUA computer code is the same as that in the NAUA/MOD4 computer code.The containment steam condensation rates used by SWNAUA are presented in Table 7.2.2B.The coefficients for removal of particulates from the effectively sprayed and unsprayed regions of the containment are plotted versus time in Figures 7.2-1 and 7.2-2, respectively. For the effectively sprayed region, the aerosol removal is due to sprays and diffusiophoresis. The particulate removal coefficient in the unsprayed region is due to gravitational settling only.2. Elemental Iodine Removal The methodology presented in Section Il1, 4.C.i, of SRP 6.5.2 (Reference

42) is used to estimate the elemental iodine removal coefficients.

The removal of elemental iodine from the containment atmosphere can be attributed due to wall deposition (AE, Wall) and due to the action of containment spray (XE, spray)a. Elemental Iodine Removal Coefficients Due to Wall Deposition (W, Wall)The elemental iodine removal coefficients due to wall deposition can be estimated using the equation provided in Reference 42.AE, Wall = Kw'A/V Where: K, = mass transfer coefficient (ft/hr)A = wetted surface area (ft 2)V = volume of the containment (ft 3)Note: A value of 4.9 m/hr (16.08 ft./hr) for Kw conservatively envelopes all available experimental data (Reference 42).The total containment surface area (435,256 ft 2) is initially available for wall deposition due to condensation on heat sink surfaces prior to spray actuation after accident. Subsequently, due to heat-up, certain portions of the heat sink surfaces become non-condensing and can no longer be considered as "wetted" surfaces. However, after spray actuation, since the heat sink surfaces in the sprayed region are continuously wetted by sprays, elemental iodine removal due to wall deposition in the sprayed region is valid over the entire period of containment spray operation. The wetted surface area within the sprayed volume is conservatively assumed to be limited to the carbon steel lined containment shell surface area (90,560 ft 2) multiplied by the spray coverage fraction of containment volume. Therefore, wall deposition elemental iodine removal coefficients are calculated during the period (1) prior to containment spray actuation, applicable to the entire containment, and (2) post containment spray actuation, applicable in the sprayed region for the duration of spray operation. 63 of 205 Diablo Canyon Power Plant Implementation of Alternative Source Terms b. Elemental iodine Removal Coefficients Due to Sprays in the Sprayed Region (AE, Spray)The elemental iodine removal coefficients due to spray actuation can be estimated using the equation provided in Reference 42.XE, Spray = (6 Kg't'F)/(V,'d) Note: This equation is valid for 10 hr- _< AE, Spray -20 hr 1 to prevent extrapolation beyond the existing data for boric acid solution with a pH of 5 (Reference 42).Where: Kg = gas phase mass transfer coefficient (ft/hr)t = time of fall of the spray droplets (= h/UT) (hr)F = volumetric spray flow rate (ft 3/hr)Vý, = effectively sprayed containment free volume (ft 3)d = mass-mean diameter of the spray droplets (ft)h = mean fall height of the spray droplet (ft)UT = terminal velocity of the spray droplet (ftlhr)The gas phase mass transfer coefficient is determined by using the equation provided on pages 418 and 441 of Reference 43.Kg = (Dg/d) x 2.0 x (1 + 0.276.Re 1/2 Sc 1 1 3)where: D 9 = diffusivity of iodine in the gas film surrounding the drop (ft 2 /hr)Re = dimensionless Reynolds number = U-pd/]p Sc = dimensionless Schmidt number = p/ (p DO)7.2.3.2.5 Sump Water pH SRP Sections 6.1.1 (Reference

45) and 6.5.2 (Reference

42) require that the pH of the sump water be controlled to maintain a minimum value of 7.0 following a LOCA. This is required to prevent re-evolution of the iodine that have been removed from the containment atmosphere by the containment spray and washed into the sump water. A neutral pH also limits material degradation, in particular, stress corrosion cracking of austenitic stainless steel components in the post LOCA environment.

Long-term retention of iodine in the sump fluid is strongly dependent on the pH. Per SRP 6.5.2, I1.1.g, long term iodine retention may be assumed only when the equilibrium sump water pH after mixing and dilution with the primary coolant is above 7. Per RG 1.183, long-term production of acids (hydrochloric acid (HCI) and nitric acid (HNO 3)) by irradiation needs to be addressed in determining whether the plant chemical addition system is adequate for long-term pH control.NUREG/CR-5732 (Reference 44), states that iodine re-evolution is not a factor if the ultimate sump water pH of > 7 is achieved prior to the time when iodine re-evolution could potentially occur. Section 3.1 of NUREG/CR-5732 notes that the following phenomena occur during the first time interval between t = 0 to t = 1000 min, a) events "leading" to the formation of 12 by 64 of 205 Diablo Canyon Power Plant Implementation ofAlternative Source Terms radiolysis, and b) all HI effects except for those related to pH. Re-evolution (i.e. vapor phase elemental iodine produced by radiolysis and partitioned between the aqueous and gas) can occur in the second time interval which is from t=1000 min to t-2 to 3 weeks. It is therefore concluded that for Light Water Reactors, a pH of 7 must be achieved within 1000 min of the initial post-LOCA release.At DCPP, RWST drain-down and chemical addition is essentially complete within -1 hour post-LOCA; thus it is expected that as a result of recirculation, the sump water will be well mixed by t=1000 minutes or- 16 hours.As part of the AST application, a conservative analysis is performed to confirm that the sump water pH at thirty (30) days following a LOCA remains greater than 7.0. The analysis assumes the minimum volume / concentration values for NaOH, in combination with the maximum volume / boration values for the water sources contributing to the sump water volume, i.e., the reactor coolant and the RWST. To establish the cable inventory inside containment, a simplified conservative upper bound approach is utilized by taking into consideration a) cable insulation data provided in NUREG/CR-5950 (Reference 46), specifically, the amounts of EPR/Hypalon cable from PWRs listed in Table 2.2 of Reference 46; and b) by examining the mass/type of cable installed in CB&I S&W constructed PWRs. A safety factor of 1.5 is applied to the largest mass of electrical cable identified (was determined to be for a 4-loop PWR with a power level slightly greater than DCPP), to estimate an upper bound value for the electrical cable installed inside the DCPP containments. Although the mass of electrical cable identified was applicable to both insulation and conductors, the DCPP analysis conservatively assumed that it was all insulation. The airborne LOCA radiation dose was conservatively assumed to be 2E+08 Rads which is commonly used for evaluating environmental qualification of electrical equipment in PWR containments, and was recommended post-TMI in IEB 79-01B (Reference

47) as a representative

/ upper bound value for the beta dose inside containment for PWRs, while the gamma dose estimate was a decade lower.Based on the above approach, the DCPP minimum ultimate sump pH was conservatively determined to be -7.8 without long-term acid production due to radiolysis, and >7.5 at T=30 days following a LOCA, inclusive of long-term acid production inside containment. Due to the availability of significant margin even with the use of conservative methodology, it is concluded that determination of the actual mass of cable insulation inside the DCPP containments is not required, and that the minimum sump pH at DCPP will be > 7.5 inclusive of acid production. Iodine volatility is analyzed for a maximum sump pH of 7.0.Based on the above assessment it is concluded that the that the DCPP sump water pH will remain greater than 7.0, and the post-LOCA dose consequence analyses need not consider iodine re-evolution from the sump fluid.7.2.3.2.6 Containment Leakage Transport Model As indicated previously, the fission products released from the fuel are assumed to mix instantaneously and homogeneously throughout the free air volume of the primary containment as it is released from the core. In accordance with RG 1.183, two fuel release phases are considered for DBA analyses: (a) the gap release, which begins 30 seconds after 65 of 205 r~t Diablo Canyon Power Plant Implementation ofAlternative Source Terms the LOCA and continues to t=30 mins and (b) the early In-Vessel release phase which begins 30 minutes into the accident and continues for 1.3 hours (i.e., t=1.8 hrs).Per RG 1.183, the core inventory release fractions, by radionuclide groups, for the gap and early in-vessel damage are as follows. Per Note 10, of Section 3.2 of RG 1.183, the release fractions listed below are acceptable for use with currently approved LWR fuel with a peak rod burnup up to 62,000 MWD/MTU.Early In-Vessel Group Gap Release Phase Release Phase Noble gas 0.05 0.95 Halogens 0.05 0.35 Alkali Metals 0.05 0.25 Tellurium Group 0.05 Ba, Sr _ 0.02 Noble Metals 0.0025 Cerium Group 0.0005 Lanthanides 0.0002 The elements in each radionuclide group released to the containment following a LOCA are assumed to be as follows (note that the groupings were expanded from that in RG 1.183 to address isotopes in the core with similar characteristics; the added isotopes are in bold font): Noble gases: Halogens: Alkali Metals: Tellurium Grp: Ba,Sr: Noble Metals: Cerium Grp: Lanthanides: Xe, Kr I, Br Cs Rb Te, Sb, Se, Sn, In, Ge, Ga, Cd, As, Ag Ba, Sr Ru, Rh, Pd, Mo, Tc, Co Ce, Pu, Np, Th La, Zr, Nd, Eu, Nb, Pm, Pr, Sm, Y, Cm, Am, Gd, Ho, Tb As discussed in Section 7.2.3.2.5, current DCPP design includes chemical addition into the containment spray system which ensures a long term sump pH equal to or greater than 7.0.Thus, the chemical form of the radioidine released from the fuel is assumed to be 95%particulate (Cesium iodide (Csl)), 4.85% elemental iodine, and 0.15% organic iodine. With the exception of noble gases, elemental and organic iodine, all fission products released are assumed to be in particulate form.The activity released from the core during each release phase is modeled as increasing in a linear fashion over the duration of the phase. The release into the containment is assumed to terminate at the end of the early in-vessel phase, approximately 1.8 hours after the LOCA.Isotopic decay, containment leakage and spray removal are credited to deplete the inventory of fission products airborne in containment. Containment Spray in the injection mode is initiated at 111 seconds after the LOCA and terminated at 3798 seconds. 'As discussed in Section 7.2.3.2.2, the sprays are estimated to cover 82.5% of the containment free volume of 2.55E+06 66 of 205 Diablo Canyon Power Plant Implementation ofAlternative Source Terms ft 3.Manual operation is credited to initiate containment recirculation sprays. Thus, based on single train operation, containment spray in the recirculation mode is initiated at 4518 seconds (i.e., twelve (12) minutes after injection spray is terminated), and terminated at 22,518 seconds.In summary, containment spray operation (injection plus recirculation) is credited until 6.25 hrs post-LOCA, with a twelve minute gap after injection spray is terminated. In the effectively sprayed region the activity transport model takes credit for aerosol removal due to steam condensation and via containment spray based on spray flowrates associated with minimum ESF. It considers mixing between the sprayed and unsprayed regions of the containment, reduction in airborne radioactivity in the containment by concentration dependent aerosol removal lambdas, and isotopic in-growth due to decay.During spray operation in the injection mode, the elemental iodine removal rate for the sprays exceeds 20 hr 1 , the maximum value permitted by SRP Section 6.5.2; thus the elemental iodine removal rate attributable to sprays is limited to 20 hr'. During recirculation spray operation, the elemental removal rate for the sprays is 19.34 hr-1.As discussed earlier, the wall deposition removal coefficient for elemental iodine has been calculated with the model provided in SRP Section 6.5.2. In sprayed and unsprayed regions, prior to spray actuation, the wall deposition removal coefficient is estimated to be 2.74 hr-1, while during spray operation, and in the sprayed region only, the wall deposition removal coefficient is estimated to be 0.57 hr 1.In the unsprayed region, the aerosol removal lambdas reflect gravitational settling. No credit is taken for elemental iodine removal in the unsprayed region.Since the spray removal coefficients are based on calculated time dependent airborne aerosol mass, there is no restriction on the DF for particulate iodine. The maximum DF for elemental iodine is based on SRP 6.5.2 and is limited to a DF of 200. The maximum allowable DF for elemental iodine is developed using methodology outlined in RG 1.183 Section 3.3.The methodology used to develop the elemental iodine and particulate removal coefficients in the sprayed and unsprayed region of the containment is discussed in Section 7.2.3.2.4. The total elemental iodine and particulate removal coefficients in the sprayed and unsprayed region of the containment as a function of time are summarized in Table 7.2-2 As discussed in Section 7.2.3.2.5, the long term sump water pH is greater than 7.0.Consequently, iodine re-evolution is not addressed. As discussed in Section 7.2.3.2.3, mixing between the sprayed and unsprayed regions of the containment is assumed for the duration of the accident. CFCU operation is initiated at 86 seconds after the LOCA and operates for the duration of the accident. The containment mixing rate between the sprayed and unsprayed regions following CFCU initiation is determined to be 9.13 turnovers of the unsprayed regions per hour with a total volumetric flow rate of 68,000 cfm between the unsprayed regions and sprayed regions. In accordance with RG 1.183, Appendix A, Section 3.3, prior to CFCU initiation, the dose consequence model assumes a mixing rate attributable to natural convection between the sprayed and unsprayed regions of 2 turnovers of the unsprayed region per hour.Radioactivity is assumed to leak from both the sprayed and unsprayed region to the environment at the containment technical specification leak rate for the first day, and half that 67 of 205 Diablo Canyon Power Plant Implementation ofAlternative Source Terms leak rate for the remaining duration of the accident (i.e., 29 days). To ensure bounding values, the atmospheric dispersion factors utilized for the containment release path reflects the worst value between the containment wall release point, the plant Vent, the Containment Penetration Area GE (EL 140') and the Containment Penetration Areas GW/FW (EL 140').7.2.3.3 ESF System leakage outside Containment In accordance with RG 1.183, with the exception of noble gases, all the fission products released from the core during the gap and early in-vessel release phases are assumed to be instantaneously and homogeneously mixed in the primary containment recirculation sump water at the time of release from the fuel. A minimum sump water volume of 480,015 gallons is utilized in this analysis.In accordance with regulatory guidance, the DCPP ESF systems that recirculate sump fluids outside containment are analyzed to leak at twice the sum of the administratively controlled total allowable leakage applicable to all components in the ESF recirculation systems. With the exception of iodine, all radioactive materials in the recirculating liquid are assumed to be retained in the liquid phase. In addition, per RG 1.183, if the temperature of the leakage exceeds 212'F, the fraction of the total iodine in the liquid that becomes airborne should be assumed equal to the fraction of the leakage that flashes to vapor. However, if the temperature of the leakage is less than 212 0 F, or the calculated flash fraction is less than 10%, the amount of iodine that becomes airborne should be assumed to be 10% of the total iodine activity in the leaked fluid unless a smaller amount can be justified based on the actual sump pH history data and area ventilation rates.ESF leakage is assumed to occur at initiation of the recirculation mode for safety injection, which at DCPP occurs as early as t=829 seconds. The maximum temperature of the recirculation fluid is 259.9'F which has a flash fraction less than 10%, thus, per RG 1.183, ten percent (10%) of the halogens associated with this leakage are assumed to be airborne and are exhausted (without mixing and without holdup) to the environment. The iodine release from the core is 95% particulate (CsI), 4.85% elemental and 0.15% organic, however after interactions with sump water the environmental release is assumed to be 97% elemental and 3% organic.At DCPP, the environmental release of ESF system leakage can occur via the 2 pathways listed below. As part of this application, DCPP is proposing to establish administrative acceptance criteria to ensure the total as-tested leakage from ESF systems that recirculate sump fluid outside containment is less than or equal to 126 cc/min, and with the following breakdown: Environmental release of ESF System leakage via the plant vent: The sum of the maximum allowable simultaneous leakage from all components in the ESF recirculation systems located in the auxiliary building is limited to 120 cc/min. Thus, and in accordance with the guidance provided in RG 1.183, the analysis addresses an ESF leakage of 240 cc/min in the Auxiliary Building (AB). The areas where these components are located are covered by the PG&E Design Class I Auxiliary Building ventilation system which discharges to the environment out of the Plant Vent. Only selected portions of the Auxiliary Building ventilation system are processed through the PG&E Design Class I AB ventilation filters. For purposes of estimating the dose consequences, it is assumed that with the exception of the RHR pump rooms (refer to 68 of 205 Diablo Canyon Power Plant Implementation ofAlternative Source Terms Section 7.2.3.4), this release pathway bypasses the PG&E Design Class I AB ventilation filters.Environmental release of ESF System leakage via Containment Penetration Area GE and Areas GW and FW: The sum of the maximum allowable simultaneous leakage from all components in the ESF recirculation systems located in the containment penetration areas is limited to 6 cc/min. Thus, and in accordance with the guidance provided in RG 1.183, the analysis addresses an ESF leakage of 12 cc/min in the containment penetration areas. The ventilation system covering this area is not PG&E Design Class 1, thus the release path to the environment is unfiltered and could occur via the Plant Vent or via the closest structural opening in the Containment Penetration Areas GE and Areas GW and FW.7.2.3.4 RHR Pump Seal Failure As discussed in Section 7.2.2, the RHR pump seal failure resulting in a filtered release via the plant vent is DCPP's licensing basis with respect to the worst case passive single failure in the RHR System. Therefore, the RHR pump Seal Failure is retained as a release pathway for the AST LOCA dose consequence analysis.The activity transport model is based on a 50 gpm leak of sump water activity for 30 minutes that occurs 24 hours after the LOCA. The temperature of the recirculation fluid is conservatively assumed to remain at the maximum temperature of 259.9°F. Thus as discussed above in Section 7.2.3.3 under ESF system leakage, the amount of iodine that becomes airborne is assumed to be 10% of the total iodine activity in the leaked fluid.The ventilation exhaust from the RHR pump rooms is covered by the PG&E Design Class I Auxiliary Building ventilation system and processed through the PG&E Design Class I AB ventilation filters. Thus, credit for filtration of the release of a RHR pump seal failure by the Auxiliary Building Ventilation system is taken in determining the dose consequences to the public at the EAB and LPZ, to the operator in the control room, and to personnel in the technical support center.The efficiency of the AB charcoal filters is determined using methodology similar to that documented in Section 7.1 for the CRVS Mode 4 ventilation filters. The allowable methyl iodide penetration / filter bypass for the Auxiliary Building Charcoal Filter is controlled by DCPP TS 5.5.11; currently the associated values are 15% and <1%, respectively. With this application, the allowable methyl iodide penetration for the AB filter will be reduced to 5%.Based on the above, an efficiency of 88% is assigned to the charcoal filters in the AB ventilation system prior to environmental release via the plant vent. Similar to the ESF system leakage, the environmental release of iodine is assumed to be 97% elemental and 3%organic.7.2.3.5 Refueling Water Storage Tank (RWS7) Back-Leakage Dose consequences associated with the potential for post-LOCA radioactive leakage to tanks vented to the atmosphere was raised in NRC Information Notice (IN) 91-56. (Reference 48)69 of 205 Diablo Canyon Power Plant Implementation ofAlternative Source Terms At DCPP, the safety injection and containment spray systems function to provide reactor core cooling and mitigate the containment pressure and temperature rise, respectively, in the event of a LOCA. Both systems initially take suction from the RWST. Once the RWST water supply is depleted, both the containment spray and safety injection systems are supplied by the RHR System. The RHR pumps take suction from the containment recirculation sump water. Under LOCA conditions, the recirculation sump water is assumed to be radioactively contaminated by fission products, of which the main contributors to airborne dose are the various isotopes of iodine.Per IN 91-56, during containment sump water recirculation, there is the potential for leakage from the mini-flow recirculation lines connecting the high head and low head safety injection pump discharge piping to the RWST. Since the RWST is vented to the atmosphere, this presents a pathway for iodine release to the atmosphere. As part of this application, DCPP is proposing to establish administrative acceptance criteria to ensure the total as-tested back leakage into the RWST from the containment recirculation sump is less than or equal to 1 gpm.The methodology discussed below to determine the post-LOCA iodine and noble gas (iodine daughters) releases via the RWST vent has been previously used and accepted by NRC for the Beaver Valley AST application (SER to License Amendment Numbers 257 and 139 for License Nos DPR-66 and NPF-73, [ML032530204]) and for the Prairie Island LOCA Re-analysis [ML091490611]. A technical paper, "Modeling Radioactive Leakage from Atmospheric Tank Vents Following a LOCA", describing this methodology was presented by CB&I S&W-Inc. in the ANS Summer Conference in 2007 and is published in Transactions of the American Nuclear Society Volume 96, Radiation Protection and Shielding Session I, pg 441. (Reference 49)Dose consequences of RWST back-leakage assumes that leakage starts at the switchover to recirculation (829 second following the LOCA) and continues for 30 days. Per regulatory guidance, a safety factor of 2 is applied to the leak rate, i.e., a 2-gpm leakage rate is assumed for the full duration of the event, which is two times the allowable leakage of 1 gpm. Also, in accordance with RG 1.183, with the exception of noble gases, all fission products released from the fuel to the containment are instantaneously and homogeneously mixed in the sump water at the time of release. However, only iodine and their daughter products are released through RWST back-leakage since the particulates would remain in the sump water.A significant portion of the iodine associated with sump water back-leakage into the RWST is retained within the RWST fluid due to the equilibrium iodine distribution balance between the RWST gas and liquid phases. The time dependent iodine partition coefficient takes into consideration the temperature and pH of the RWST liquid and sump fluid, the RWST liquid and gas volumes, and the temperature, pH and volume of the incoming leakage. The iodines that evolve into the RWST gas space as a result of the equilibrium iodine distribution balance, and the noble gas daughters of iodines, are released to the environment via the RWST vent, at a vent rate established by the temperature transient in the RWST (which includes the effect of decay heat), the increase in the liquid inventory of the RWST due to the incoming leakage, the gases evolving out of incoming leakage, and the environmental conditions outside the RWST.The average time-dependent RWST iodine release fractions along with the fractional RWST gas venting rates (may be applied to the noble gas daughters of iodines) to the atmosphere from the Units 1 and 2 RWSTs due to RWST back-leakage following switchover to the sump water recirculation mode of operation is summarized in Table 7.2-3. As discussed earlier, the release 70 of 205 Diablo Canyon Power Plant Implementation of Alternative Source Terms fractions / rates presented in Table 7.2-3 reflect a safety factor of 2 on the leak rates, i.e., are developed based on a RWST back-leakage of 2 gpm. The iodine released to the environment is assumed to be 97% elemental and 3% organic.The equilibrium iodine concentration in the RWST gas space utilized to develop Table 7.2-3 is based on the iodine mass in the sump fluid entering the RWST vapor space as back-leakage or the total iodine mass contained in the RWST liquid, whichever results in higher RWST vapor phase concentrations. The RWST maximum venting rate averaged over an interval is primarily based on RWST back-leakage entering the RWST gas space and thermally equilibrating, and is used in conjunction with the higher RWST gas space iodine concentration to calculate an iodine mass release rate as a function of time. An interval based averaging approach is utilized in preparing Table 7.2-3 to reduce the number of input values to the dose analysis while preserving the boundaries for the time periods used for atmospheric dispersion; the actual iodine release calculated in an interval is normalized to the iodine mass leaking into the RWST during that time interval.Examination of the average gas space venting rates indicate that after the first day, the noble gases formed by decay of iodine will primarily remain in the RWST during the 30 day period of evaluation and not be released. However, the dose consequence analysis conservatively releases the noble gases formed by decay of iodine, directly to the environment without taking any credit for tank holdup.7.2.3.6 Miscellaneous Equipment Drain Tank (MEDT) Leakage The DCPP Units 1 and 2 MEDT is a covered rectangular (12' x 5' x 10') stainless steel lined concrete tank located in the Auxiliary Building below El 60 ft. The MEDT tank vent is hard-piped to the Auxiliary Building ventilation ductwork; thus the airborne releases from the MEDT are ultimately discharged to the environment via the plant vent.Following a LOCA, the MEDT will receive both post-LOCA sump fluids as well as non-radioactive fluids (i.e., ESF system leakage from the accident unit, as well as non-radioactive fluids from equipment drains and RWST leakage from the non-accident unit) which are hard-piped to the MEDT. As part of this application, DCPP is proposing to establish administrative acceptance criteria to ensure the total as-tested flow hard piped to the MEDT is less than 950 cc/min of ESF system leakage and 484 cc/min of non-radioactive fluid leakage.Similar to the RWST back-leakage model, dose consequences due to releases from the MEDT assumes that leakage starts at the switchover to recirculation (829 second following the LOCA)and continues for 30 days. Per regulatory guidance, a safety factor of 2 is applied to the leak rate, i.e., 1900 cc/min of ESF system leakage and 968 cc/min of non-radioactive fluids into the MEDT is assumed for the full duration of the event, which is two times the allowable leakage.For purposes of bounding analyses, the boron concentration of the pre-existing fluid in the MEDT, as well as the incoming leakage is assumed to be at its upper bound levels. With the exception of noble gases, all fission products released from the fuel to the containment are instantaneously and homogeneously mixed in the sump water at the time of release. Only iodine and their daughter products are released through MEDT leakage since the particulates would remain in the sump water.71 of 205 Diablo Canyon Power Plant Implementation ofAlternative Source Terms The methodology used to determine the post-LOCA iodine and noble gas releases via the MEDT vent and Plant Vent has been used previously in evaluating post-LOCA leakage into the RWST. Adaptation of the methodology to address overflows/room ventilation releases is straightforward with the room ventilation rate being treated as the tank exhaust rate. As discussed earlier in Section 7.2.3.5, this equilibrium based methodology for RWST vent release paths has been accepted by NRC for the Beaver Valley AST application (SER to License Amendment Numbers 257 and 139 for License Nos DPR-66 and NPF-73, [ML032530204]) and for the Prairie Island LOCA Re-analysis [ML09149061 1].The transport model utilized to determine airborne releases from the MEDT takes into account the fact that the MEDT is a small tank with an auto-transfer capability which is PG&E Design Class II. Consequently, and for purposes of conservatism, it is assumed that a) the LOCA occurs when the MEDT water level is at the normal maximum setpoint to initiate auto transfer, b) the auto-transfer capability is not initiated because it is not a safety function, and c) the MEDT contents will spill over into the Equipment Drain Receiver Tank (EDRT) Room after the tank is full. Thus, for the post-LOCA scenario, the MEDT is conservatively assumed to overflow via its manway into the EDRT Room. The EDRT room drains into the Auxiliary Building Sump (ABS), which ultimately overflows into the U1 / U2 pipe tunnels. The ABS is also a covered rectangular (16' x 5' x 10') stainless steel lined concrete tank with a vent that is hard-piped to the Auxiliary building ductwork with a PG&E Design Class II auto transfer capability. The ABS is located adjacent to the MEDT.The bounding transient release of iodine along with the gas venting rate to the atmosphere as a result of post-LOCA leakage of radioactive and non-radioactive fluid hard-piped into the MEDT is developed in 2 parts: a) prior to MEDT overflow and b) post MEDT overflow.a) Prior to MEDT overflow -The iodines evolve into the MEDT gas space as a result of the equilibrium iodine distribution balance between the MEDT gas and liquid phases (either the MEDT liquid inventory or the incoming leakage), and are released to the environment via the plant vent, at a vent rate established by the temperature transient in the MEDT (including the effect of decay heat), the increase in the liquid inventory of the MEDT due to the incoming leakage, and the gases evolving out of the incoming leakage.b) After MEDT overflow -The equilibrium iodine distribution balance is conservatively assumed to be between the iodine concentrations in the MEDT overflow liquid and the EDRT room (or U1/U2 pipe tunnels) ventilation flow (rather than the average concentration in the EDRT room (or U1IU2 pipe tunnels) free volume). This maximizes the iodine release rate. Thus, the iodines released are a sum total of the following: i) the iodines that evolve into the EDRT room air space as a result of the equilibrium iodine distribution balance between the spilled liquid from the MEDT (at the temperature of the MEDT), and the EDRT room ventilation flow, and is released to the environment via the plant vent, at the vent rate established by the EDRT room ventilation system, and ii) the iodines that evolve into the U1/U2 Pipe Tunnel air space as a result of the equilibrium iodine distribution balance between the spilled liquid from the MEDT (at the maximum temperature of the U1/U2 Pipe Tunnel), and the U1/U2 Pipe Tunnel 72 of 205 Diablo Canyon Power Plant Implementation ofAlternative Source Terms ventilation flow, and is released to the environment via the plant vent, at the vent rate established by the U1/U2 Pipe Tunnel ventilation system The exhaust fans servicing the EDRT room and pipe tunnel are PG&E Design Class I. There is also a potential that the non-LOCA unit's ABVS will be operating with the flow being exhausted to the associated unit specific plant vent. Thus, it is conservatively assumed that the non-LOCA unit's ABVS is operating, and together with the accident units' exhaust fans, are providing the motive force to exhaust the airborne releases to the environment, unfiltered, via the respective plant vents.The average time-dependent MEDT iodine release fractions, along with the fractional MEDT gas venting rates (which may be applied to the noble gas daughters of iodines prior to MEDT overflow) to the atmosphere following switchover to the sump water recirculation mode of operation, is summarized in Table 7.2-4. As discussed earlier, the release fractions / rates presented in Table 7.2-4 reflect a safety factor of 2 on the leak rates, i.e., are developed based on an input of 1900 cc/min of ESF system leakage and 968 cc/min of non-radioactive fluids into the MEDT. Through the use of extremely conservative assumptions, the calculated iodine release fractions / gas venting rates presented in Table 7.2-4 when used in combination with the analyzed ESF system leak rate, bound the iodine releases of all combinations of radioactive and non-radioactive leakages less than or equal to the leak rates analyzed. The iodine released from the ventilation system is assumed to be 97% elemental and 3% organic, and is released to the environment via the plant vent. In addition, the dose consequence analysis conservatively releases the noble gases formed by decay of iodine, directly to the environment without taking any credit for tank holdup.7.2.4 Offsite Dose Assessment Due to the delayed post-LOCA fuel release sequence of an AST model, and the rate at which aerosols and elemental iodine are removed from the containment, the maximum 2-hour EAB dose for a PWR LOCA typically occurs between 0.5 hrs to 2.5 hrs.To establish the "worst case 2-hour release window" for the DCPP EAB dose, the integrated dose versus time for each of the six pathways discussed above was evaluated. The 0-2 hr EAB Atmospheric Dispersion Factor was utilized for all cases.The analysis demonstrated that for DCPP the maximum 2 hour EAB dose will occur, as a result of the RHR pump seal failure, between T=24 hrs to T=26 hrs, and is unrelated to the post-LOCA fuel release sequence associated with AST.For purposes of completeness, the "worst case 2-hour release window" for the DCPP EAB dose is estimated for 2 cases: a) Without the RHR pump seal failure release, and b) With the RHR pump seal failure release.The direct shine dose at the EAB due to a) the airborne activity inside containment, and b) the sump water collected in the RWST due to RWST back-leakage, was also-evaluated. Based 73 of 205 Diablo Canyon Power Plant Implementation ofAlternative Source Terms on the results of the EAB evaluation which determined that the dose contribution due to direct shine was minimal (<0.01 rem), the dose at the LPZ due to direct shine is deemed negligible. The bounding EAB and LPZ dose following a LOCA at either unit is presented in Section 8.7.2.5 Control Room Occupancy Dose 7.2.5.1 Accident Specific Control Room Model Assumptions The parameter values utilized for the control room in the accident dose transport model are discussed in Section 7.1. Provided below are the critical LOCA-specific assumptions associated with control room response and activity transport. Timing for Initiation of CRVS Mode 4 (if applicable):

An SI signal will be generated at t = 6 sec following a LOCA." The CRVS normal intake dampers of the accident unit start to close after a 28.2 second delay due to delays associated with diesel generator loading onto the 4kv buses. The CR dampers are fully closed 10 secs later, or at t=44.2 secs (i.e., 6 + 28.2 + 10). The 2 second SI signal processing time occurs in parallel with diesel generator sequencing and is therefore not included as part of the delay." In accordance with DCPP licensing basis, the CRVS normal operation dampers of the non-accident unit are not affected by the LOOP and are isolated at t=18 secs (i.e., 6 + 2 secs signal processing time + 10 sec damper closure time).Control Room Atmospheric Dispersion Factors: The bounding atmospheric dispersion factors applicable to the radioactivity release points I control room receptors applicable to a LOCA at either unit are provided in Table 7.2-5. The I/Q values presented in Table 7.2-5 take into consideration the various release points-receptors applicable to the LOCA to identify the bounding %/Q values applicable to a LOCA at either unit, and reflect the allowable adjustments

/ reductions in the values as discussed in Chapter 5 and summarized in the notes of Tables 5.2-2 and 5.2-3.7.2.5.2 Direct Shine Dose to the Control Room from External and Contained Sources The direct shine dose to an operator in the control room due to contained or external sources resulting from a postulated LOCA is calculated using CB&I S&W Inc. point kernel shielding computer program SW-QADCGGP. The post-LOCA gamma energy release rates (MeV/sec)and integrated gamma energy release (MeV-hr/sec) in the various external sources are developed with CB&I S&W Inc. computer program PERC2.The LOCA sources that could potentially impact the CR operator dose due to direct shine are identified below.74 of 205 Diablo Canyon Power Plant Implementation ofAlternative Source Terms 1. Direct shine from containment -shine from the airborne source, in the containment structure via the bulk shielding (3'-8" thick concrete walls below the bendline, 2'-6" thick concrete dome), including shine through one of the main steam line penetrations and the Personnel Hatch facing the CR 2. Direct shine from the contaminated cloud outside the control room pressure boundary resulting from containment leakage, ESF system leakage, RHR Pump seal leakage, RWST back-leakage, MEDT leakage -shine occurs through the CR walls, via wall penetrations such as CR doors to the outside, and from the airborne activity in cable spreading room below via CR floor penetrations.

3. Dose due to scattered gamma radiation through wall penetrations from the CRVS filters located in the adjacent "mechanical equipment room.4. Direct shine from the sump fluid that is postulated to collect in the RWST Cloud shine through CR doorways was found to be the most significant of all the identified contained or external post-LOCA radiation sources listed above, followed by the dose contribution through the CR floor penetrations.

Note that other radiation sources were identified and deemed insignificant due to the presence of significant shielding between the operator and the radiation sources. Examples of these dose contributors include most of the large and small electrical and pipe penetrations in the Containment outer wall that faces the CR, and the ESF system piping and components located in the Auxiliary Building.The direct shine dose estimate in the CR takes into consideration the function of Room 506 (which serves as a control room foyer adjacent to the Shift Supervisor's office), where occupancy is deemed to be minimal; i.e., conservatively estimated at less than 5% of the total time spent daily in the control room. The above "occupancy adjustment" is utilized to determine the maximum 30-day integrated dose in Control Room (i.e., the total direct shine dose in the CR includes the 30-day dose in Room 506 adjusted by the referenced occupancy factor).The bounding control room operator dose following a LOCA at either unit is presented in Section 8.7.2.6 Control Room Operator Dose during Access Diablo Canyon assumes that the dose received by the operator during routine access to the control room for the 30 day period following the LOCA is minimal. Thus, as long as some reasonable margin exists between the regulatory limit and the estimated dose to the operator during control room occupancy, the additional dose due to ingress / egress can be accommodated. This approach is consistent with the approach used by other licensees, and is reasonable since a) transit to and from the control room is only expected after the first 24 hours following the accident by which time the airborne levels inside containment has reduced significantly due to the use of active fission product removal mechanisms such as containment sprays, and radioactive decay, and b) the operator is protected from radioactive ESF fluids by the shielding provided by the buildings that house such equipment. In addition, it is expected that during a 75 of 205 Diablo Canyon Power Plant Implementation ofAlternative Source Terms postulated event, access to the control room will be controlled by Health Physics and the Emergency Plan based on real time data, with the purpose of minimizing personnel dose.It is also noted that the dose received by the operator during transit outside the control room is not a measure of the "habitability" of the control room which is defined by the radiation protection provided to the operator by the control room shielding and ventilation system design.Thus, the estimated dose to the operator during routine post-LOCA access to the control room is addressed separately from the control room occupancy dose which is used for the demonstration of control room habitability. DCPP's current licensing basis provides an estimated dose contribution to the operator during egress and ingress to the control room following a LOCA. The access dose estimate currently reported in the UFSAR has been part of DCPP's 'licensing basis since the original Safety Analysis Report. While several of the input values and the calculations for the original ingress/egress dose values are no longer available, selected inputs, the ingress/egress dose values and methodology are still in the UFSAR.To address the existing licensing basis, a TEDE dose is estimated for operator access to the control room. Because RG 1.183 does not provide guidance on determining the egress and ingress to the control room following an accident, the same inputs used to estimate the current licensing basis values for access to the control room, along with the associated dose estimate presented in the UFSAR, are used to determine the TEDE dose estimate for ingress/egress. With this application, and consistent withl the assumption made by other licensees, PG&E is proposing to demonstrate that the dose contribution due to routine ingress/egress during the accident is minimal.In accordance with DCPP original licensing basis, radiation exposures to personnel during egress and ingress (i.e., during routine access to the control room for the duration of the accident) could result from the following sources:-Airborne activity in the 6ontainment leakage plume (2) Direct gamma radiation from fission products in the containment structure Post-accident egress-ingress exposures were based on 27 outbound excursions, from the control room to the site boundary, and 26 inbound excursions, from the site boundary to the control room. It was estimated that each excursion would take 5 minutes, and no credit was taken for breathing apparatus or special whole body shielding. Egress-ingress thyroid and whole body exposures from airborne activity are functions of containment activity, containment leakage, atmospheric dispersion, and excursion time. The airborne activity concentrations were calculated and the then conventional exposure equations from Regulatory Guide 1.4, Revision 1, were used to calculate gamma, beta, and thyroid exposures (Reference 8). The exposure from betas was calculated on the basis of an infinite uniform cloud, and exposure from gammas was calculated on the basis of a semi-infinite cloud.Because of the containment shielding and short excursion time, egress-ingress containment shine exposures were estimated to be small. The shine model assumed a cylindrical radiation 76 of 205 Diablo Canyon Power Plant Implementation ofAlternative Source Terms source having the same radius and height as the containment structure with a 3.5-foot-thick concrete shield surrounding it. The receptor point was assumed to be a distance of 10 meters from the outer surface of the containment wall.The estimated egress-ingress exposures developed in support of DCPP original licensing basis are listed below.* The dose to control room personnel during egress / ingress from airborne fission products in the containment leakage pl.ume: 0.0066 rem gamma, 0.0243 rem beta, and 4.72 rem thyroid* The dose to control room personnel during egress / ingress as a result of direct radiation shine from the fission products in the containment structure: 0.022 rem.As part of this AST licensing application, DCPP has identified additional post-LOCA fission product release pathways, as discussed in Section 7.2.2. The postulated effect of these additional radioactivity release paths, as well as the implementation of AST, on the estimated dose to control room personnel during routine egress / ingress takes into consideration the following: " The transport models used to develop the dose to the control room operator during occupancy address a control room occupancy factor of 1.0 till t=24 hours after the accident. This implies that during the first 24 hours the control room operator stays in the control room. This is also reflected in the DCPP original licensing basis which addresses one more outbound trip than the inbound trips.* Routine ingress / egress to the control room during the 30 day period following a LOCA falls into the mission dose category as discussed in NUREG-0737, II. B. 2 (Reference 15).* NUREG-0737, Item II. B. 2 states that leakage of systems outside containment need not be considered as potential sources.Based on the above considerations, the dose consequences of the additional activity release paths addressed in Section 7.2.2 (and listed below), is addressed as follows:* Containment Pressure Nacuum relief release -this release occurs at accident initiation (before t=24hr), so there is no dose contribution to the control operator during routine ingress /egress during the 30 day period following the accident..Containment leakaqe: o The airborne activity in the containment after t=24 hours with an AST source term is primarily 100% of the core noble gases and 0.06% of the core iodines that were released to containment. Note: The iodine source term at t=24 hrs is essentially the organic iodines released to the containment which are not effected by sprays, and which per Regulatory Guide 1.183, 77 of 205 Diablo Canyon Power Plant Implementation ofAlternative Source Terms represent 0.06% of the core iodines (i.e., 0.15% of the 40% core iodines released to containment atmosphere at accident initiation). Also, the essentially particulate nature of the radioactivity release associated with an AST source term, and the effectiveness of particulate removal by sprays / settling makes the dose contribution from the particulate source minimal after t=24 hours.o The corresponding airborne activity in the containment after t=24 hours for a TID-14844 source term is 100% of the core noble gases and 1% of the core iodines.Note: Per Regulatory Guide 1.4, Revision 1, the organic iodines released to the containment is 4% of the 25% iodines released to containment atmosphere at accident initiation. o Based on the above it is concluded.that after t=24hrs: a. The dose consequences due to containment leakage based on a TID-14844 based scenario will bound the dose consequences based on an AST scenario.b. Since the thyroid dose is primarily due to iodines, the associated dose to the operator will vary proportionately to the amount of iodine airborne in containment. Thus the thyroid dose to the operator during ingress /egress for an AST scenario may be estimated by adjusting the TID-14844 based dose by the ratio of the iodine estimated to be airborne in containment for each of the scenarios. As noted earlier, the current licensing basis thyroid dose to the operator during ingress / egress is 4.72 rem. The corresponding thyroid dose based on an AST scenario is estimated to be 4.72 x 0.06 = 0.28 rem thyroid.The RHR Pump Seal Failure, ESF System Leakage, RWST back leakage and MEDT leakage -All of these releases are based on leakage of systems outside containment. In accordance with NUREG-0737 II. B. 2, the dose contribution due to these sources need not be considered for access calculations. To address the TEDE dose acceptance criteria applicable to the use of AST, the original licensing basis egress-ingress exposures have been updated as noted below using the guidance provided in 1OCFR20.1003 (Reference 61). The referenced federal regulation defines TEDE as the sum of the deep dose equivalent for external exposures (i.e., external whole body exposure) and the committed effective dose equivalent for internal exposures (i.e., sum of the product of the weighting factor applicable to each organ irradiated and the dose to that organ).Per 10CFR20.1003, the weighting factor for the whole body is 1.0 and for the thyroid is 0.03.While the weighting factor for beta radiation is undefined, the contribution of the beta dose to the total effective dose equivalent is expected to be insignificant. Therefore," Radiation from airborne fission products in the containment leakage plume to the control room personnel during egress ingress is approximately 0.0066 rem + 0.28 x 0.03 rem, i.e., 0.015 rem TEDE* Direct radiation from the fission products in the containment structure to control room personnel during egress ingress is 0.022 rem TEDE.78 of 205 Diablo Canyon Power Plant Implementation ofAlternative Source Terms Thus the total dose to the control room operator during access is estimated to be 0.037 rem TEDE; i.e., 0.015 + 0.022. This value is 1% of the estimated operator dose due to control room occupancy following a LOCA (Refer to Table 8.1-1) and is therefore considered to be minimal.7.2.7 Technical Support Center Dose In accordance with current licensing basis, the Technical Support Center (TSC) design has been evaluated for the LOCA.CB&I S&W Inc. computer code PERC2 is used to calculate the dose to TSC personnel due to airborne radioactivity releases following a LOCA. The direct shine dose to an operator in the TSC due to contained or external sources resulting from a postulated LOCA is calculated using CB&I S&W Inc. point kernel shielding computer program SW-QADCGGP. The post-LOCA gamma energy release rates (MeV/sec) and integrated gamma energy release (MeV-hr/sec) in the various external sources are developed with computer program PERC2.The TSC serves both units and is located at El 104' on the south-west side of the Unit 2 turbine building and is shared between Unit 1 and Unit 2. The north and south walls are made up of 2'-2" of concrete, whereas the east and west walls are made of 1'-4" and 1'-6" of concrete, respectively. The floor and ceiling thickness / material reflect a minimum of 1V-0" and 1'-8" of concrete, respectively. As part of this application, DCPP proposes to process the TSC normal ventilation intake flow through a HEPA filter. In addition, the nominal air intake flowrate during normal operations will be 500 cfm. The above air intake is filtered through the referenced HEPA filter and drawn into the TSC envelope which has a free volume of 51,250 ft 3.The TSC normal intake is isolated and the TSC ventilation placed into filtered I pressurized Mode 4 operation by manual operator action within 2 hours of the LOCA.The post-accident pressurization flow to the TSC is provided via the CRVS Mode 4 pressurization intakes (i.e., 1 per unit, each located on either side of the Turbine Building). As noted in Section 7.1, the DCPP CR pressurization air intakes have dual ventilation outside air intake design. The nominal air intake flowrate during the TSC pressurization mode is 500 cfm.Mode 4 Operation of the TSC utilizes the CRVS. As discussed in Section 7.1, CRVS Mode 4 operation (which also serves the TSC) utilizes redundant PG&E Design Class I radiation monitors located at each pressurization air intake and has the provisions of acceptable control logic to automatically select the least contaminated inlet at the beginning of the accident, and manually select the least contaminated inlet during the course of the accident. Thus, during Mode 4 operation the TSC dose consequence analysis can utilize the X/Q values for the more favorable pressurization air intake reduced by a factor of 4 to credit the "dual intake" design (See Section 5.2 for additional details).The allowable methyl iodide penetration and filter bypass for the TSC Mode 4 Charcoal Filter is<2.5% and <1%, respectively. Thus in accordance with GL 99-02 (Reference 41), the CR charcoal filter efficiency for elemental and organic iodine used in the TSC dose analysis is 100%-[(2.5% + 1 %) x 2] = 93%. The acceptance criteria for the TSC normal operation and Mode 4 HEPA filters is "penetration plus system bypass" < 1.0%. Thus, using methodology similar to 79 of 205 Diablo Canyon Power Plant Implementation ofAlternative Source Terms the charcoal filters, the HEPA filter efficiency for particulates used in the TSC dose analysis is 100% -[(1%) x 2] = 98%.During TSC Mode 4 operation, the TSC air is also recirculated through the same filtration unit as the pressurization flow. The air flow allowable through the pressurization charcoal / HEPA filter during TSC Mode 4 is 1000 cfm.. This flow is comprised of a minimum filtered recirculation flow of 500 cfm and the previously discussed pressurization flow of 500 cfm, for a total flow through the filtration unit of 1000 cfm (pressurization plus recirculation). Unfiltered inleakage into the TSC during Mode 1 and Mode 4 is assumed to be 60 cfm (includes 10 cfm for ingress/egress based on the guidance provided in SRP 6.4).For purposes of estimating the post-LOCA dose consequences, the DCPP TSC is modeled as a single region. When in TSC Mode 4, the Mode I intakes are isolated and outside air is a) drawn into the TSC through the filtered emergency intakes; b) enters the TSC as infiltration, and c)enters the TSC during operator egress/ingress. The dose assessment model utilizes nominal values for the ventilation intake flowrates since the intake pathways (normal as well as accident) are filtered, thus the controlling dose contributor is the unfiltered inleakage. The effect of intake flow uncertainty on the TSC dose is expected to be insignificant. The bounding atmospheric dispersion factors applicable to the radioactivity release points / TSC receptors applicable to a LOCA at either unit are provided in Table 7.2-6. The %/Q values presented take into consideration the various release points-receptors applicable to the LOCA to identify the bounding X/Q values applicable to an accident at either unit, and reflect the allowable adjustments / reductions in the values as discussed in Chapter 5.The direct shine dose into the TSC due to the external cloud and contained sources is calculated in a manner similar to that described for the control room in Section 7.2.5.2. The LOCA sources that could potentially impact the TSC operator dose due to direct shine are identified below.1. Direct shine from containment -shine from the airborne source in the containment structure via the bulk shielding (3'-8" thick concrete walls below the bendline, 2'-6" thick concrete dome), including shine through the Personnel Hatch facing the TSC 2. Direct shine from the contaminated cloud outside the TSC pressure boundary resulting from containment leakage, ESF system leakage, RHR Pump seal leakage, RWST back-leakage, MEDT leakage -shine occurs through the TSC walls and via wall penetrations such as TSC doors to the outside.3. Dose due to scattered gamma radiation through wall penetrations from the TSC filters located in the adjacent mechanical equipment room, and scatter past labyrinths provided for selected doors.Note that other radiation sources were identified and deemed insignificant due to the presence of significant shielding between the operator in the TSC and the radiation sources.80 of 205 Diablo Canyon Power Plant Implementation ofAlternative Source Terms Table 7.1-2 lists key assumptions / parameters associated with DCPP TSC design.The bounding TSC operator dose following a LOCA at either unit is presented in Section 8.7.3 Fuel Handling Accident (FHA)NRC sponsored computer code RADTRAD 3.03, is used to calculate the control room and site boundary dose due to airborne radioactivity releases following a FHA. In accordance with current licensing basis, an FHA is assumed to occur in the Spent Fuel Pool located in the Fuel Handling building (FHB), or in the Containment. This event -postulates that a spent fuel assembly is dropped during refueling in the Spent Fuel Pool (SFP) located in the FHB, or in the reactor cavity located in the Containment. In accordance with current licensing basis, all of the fuel rods (264 rods) in the dropped fuel assembly are assumed to be damaged; thus all of the activity in the fuel gap of the dropped assembly is assumed to be instantaneously released into the SFP or into the reactor cavity. As documented in the NRC SER for License Amendments 8 and 6 to DCPP Facility Operating License Nos. DPR-80 and DPR-82, respectively (Reference 58), the assumption that all fuel rods in one assembly rupture is conservative because the kinetic energy available for causing damage to a fuel assembly dropped through water is fixed by the drop distance. The kinetic energy associated with the maximum drop height for a fuel handling accident is not considered sufficient to rupture the equivalent number of fuel rods of one assembly in both the dropped assembly and the impacted assembly.This assessment follows the guidance provided for the FHA in pertinent sections of RG 1.183 including Appendix B. As discussed in Section 4.3, and as part of the change in licensing basis requested with this application, the core gap activity is assumed to be comprised of 12% of the core 1-131 inventory, 30% of the core Kr-85 activity, 10% of the remaining noble gas and halogen isotopes, and 17% of the core alkali metals (Cesium and Rubidium). Table 7.3-1 lists the key assumptions / parameters utilized to develop the radiological consequences following an FHA at either location and at either unit.Current DCPP procedures prohibit movement of recently irradiated fuel which is defined as fuel that has occupied part of a critical reactor core within the previous 100 hours. As a part of this application, it is proposed that the definition of recently irradiated fuel at DCPP be updated to reflect fuel that has occupied part of a critical reactor core within the previous 72 hours. Table 7.3-2 provides the gap activity inventory of the noble gases, iodines and alkali metals in a single fuel assembly at t=72 hrs post reactor shutdown.DCPP TS 3.7.15 requires the SFP water level to be >23 feet over the top of irradiated fuel assemblies seated in the storage racks. TS 3.9.7 requires the refueling cavity water level to be maintained >23 feet above the top of the reactor vessel flange. Additional margin is provided through operating procedures. The fission product inventory in the fuel rod gap of all the rods in the damaged assembly are assumed to be instantaneously released into the spent fuel pool or reactor cavity, both of which 81 of 205 Diablo Canyon Power Plant Implementation ofAlternative Source Terms have a minimum of 23 ft of water above the damaged fuel assembly. A radial peaking factor of 1.65 is applied to the activity release.Per RG 1.183, the radioiodine released from the fuel gap is assumed to be 95% particulate (Csl), 4.85% elemental, and 0.15% organic. Due to the acidic nature of the water in the fuel pool (pH less than 7), the CsI is assumed to immediately disassociate and re-evolve as elemental iodine, thus changing the chemical form of iodine to 99.85% elemental and 0.15%organic. In addition, and per RG 1.183, an iodine decontamination factor of 200 is assumed for the SFP / reactor cavity. Noble gases and unscrubbed iodines rise to the water surface where they are mixed in the available air space. All of the alkali metals released from the gap are retained in the pool. In accordance with RG 1.183, the chemical form of the iodines above the pool is 57% elemental and 43% organic.Per RG 1.183, the activity released due to an FHA is assumed to be discharged to the environment in a period of 2 hrs (or less if the ventilation system promotes a faster release rate).FHA in the FHB The radioactivity release pathways following an FHA in the FHB are established taking into consideration the following Administration Controls: During fuel movement in the FHB: o The movable wall is put in place and secured o No exit door is propped open o One FHBVS exhaust fan is operating (The supply fan flow (if operating) has been confirmed by design to have less flow than the exhaust fan)Operation of the Fuel Handling Building Ventilation system (FHBVS) with a minimum of 1 exhaust fan operating and all significant openings administratively closed will ensure negative pressure in the FHB which will result in post-accident environmental release of radioactivity occurring via the Plant Vent. The activity release due to the FHA in the FHB is assumed to be discharged to the environment as follows: o A maximum release rate of 46,000 cfm via the Plant Vent due to operation of the FHBVS with a closed FHB configuration. o A maximum conservatively assumed outleakage of 500 cfm occurring from the closest edge of the FHB to the control room normal intake (i.e., 30 cfm outleakage is assumed for ingress/egress; 470 cfm is assumed for outleakage from miscellaneous gaps/openings in the FHB structure). It has been determined that for the FHA in the FHB, the actual release rate lambda based on the FHBVS exhaust (i.e., 8.7 hr-') is larger than the release rate applicable to "a 2-hr release" per regulatory guidance (i.e., 3.45 hr-). Thus the larger exhaust rate lambda associated with FHBVS operation plus the exhaust rate lambda for the 500 cfm outleakage is utilized in the analysis.82 of 205 Diablo Canyon Power Plant Implementation of Alternative Source Terms FHA in the Containment The potential radioactivity release pathways following a FHA in the containment are established taking into consideration o Operation of the containment purge system which would result in radioactivity release via the plant vent o Plant Technical Specification Section 3.9.4 that allows for an "open containment" during fuel movement in containment during offload or reload. The most significant containment opening closest to the Control room normal operation intake is the equipment hatch. The equipment hatch is an approximately 20-ft wide circular opening in containment. In the event the containment purge system ceased to operate (a viable scenario since it is single train and has non-vital power), the density driven convective flow out of the equipment hatch (due to the thermal gradient between inside and outside containment conditions), could be significant. It has been determined that for the FHA in the Containment, the release rate assuming a regulatory based 2 hr release is larger than that dictated by the containment purge ventilation system, or convective flow out of the equipment hatch. Thus the regulatory based release rate (i.e., 3.45 hr'), is utilized for this analysis. Review of the atmospheric dispersion factors associated with the plant vent vs the equipment hatch indicates that dose consequences due to releases via the equipment hatch will be bounding.EAB 2 hr Worst Case Window AST methodology requires that the worst case dose to an individual located at any point on the boundary at the EAB, for any 2-hr period following the onset of the accident be reported as the EAB dose. Since the FHA is based on a 2-hour release, the worst 2-hour period for the EAB is the 0 to 2-hour period.The bounding EAB and LPZ dose following a FHA at either location and at either unit is presented in Section 8.Accident Specific Control Room Model Assumptions The parameter values utilized for the control room in the accident dose transport model are discussed in Section 7.1. Provided below are the critical FHA-specific assumptions associated with control room response and activity transport. Design Basis FHA (occurs at t=72 hours after reactor shutdown)The current licensing basis analyses supporting the DBA FHA in the FHB and the FHA in the Containment assume that the CR remains in normal operation mode for the duration of the accident.As part of this application, credit is taken for PG&E Design Class I area radiation monitors located at the Control Room (CR) normal intakes (1-RE-25/26, 2-RE-25/26) to initiate CRVS 83 of 205 Diablo Canyon Power Plant Implementation ofAlternative Source Terms Mode 4 (filtered / pressurized accident ventilation) upon detection of high radiation levels at the CR normal intakes as a result of an FHA.An analytical safety limit of 1 mR/hr for the gamma radiation environment at the CR normal operation air intakes has been used in the FHA analyses to initiate CRVS Mode 4. As discussed in Section 2.2, the actual monitor trip setpoint will be lower to include the instrument loop uncertainty. The radiation monitor response time is primarily dependent on the type of monitor, the setpoint, the background radiation levels and the magnitude of increase in the radiation environment at the detector location.For a monitor with an instrument time constant of "-c" (2 seconds) and a background of 0.05 mR/hr, the response time "t" to a high alarm Setpoint (HASP < 1 mr/hr), for a step increase of radiation level DR (mR/hr) is determined by solving the following equation that represents the monitor reading approaching the final reading exponentially. t HASP = 0.0S + DR(l -e-T)It is determined that a DBA FHA (i.e., occurs at 72 hrs post shutdown) will result in a radiation environment at the CR normal operation intakes that greatly exceed the analytical limit of 1 mR/hr for initiating CRVS Mode 4. This will result in an almost instantaneous generation of a radiation monitor signal to initiate CRVS Mode 4 (radiation monitor response time is estimated to be < 1 sec). For purposes of conservatism, and since the delay in isolation of the normal intake has a significant impact on the estimated dose consequences, the analysis conservatively assumes a monitor response time to the HASP of 10 secs.As discussed in Section 7.1, when crediting CRVS Mode 4, the FHA dose consequence analyses does not address the potential effects of a LOOP. Thus delays associated with diesel generator sequencing are not required.Therefore, the time delay between the arrival of radioactivity released due to a DBA FHA at both the CR normal Intakes (assumed to be instantaneous) and CRVS Mode 4 operation is estimated to be the sum total of the monitor response time (10 secs), the signal processing time (2 secs) and the damper closure time (10 secs) for a total delay of 22 seconds.Delayed FHA: It is recognized that the response time for radiation monitors are dependent on the magnitude of the radiation level / energy spectrum of the airborne cloud at the location of the detectors, which in turn are dependent on the fuel assembly decay time. Thus an additional case is considered for each of the two FHA scenarios described above (i.e., a FHA in the FHB and a FHA in Containment) when determining the dose to the OR operator; i.e., a case that reflects a delayed FHA at Fuel Offload or a FHA during Reload, occurring at a time when the fuel has decayed to such an extent that the radiation environment at the CR normal intake radiation monitors is just below the setpoint; thus the CR remains in normal operation mode and CRVS Mode 4 is not initiated. 84 of 205 Diablo Canyon Power Plant Implementation of Alternative Source Terms The analyses determined that the dose consequences of a DBA FHA bound that associated with the delayed FHA for both the FHA in the FHB and the FHA in the containment. The bounding atmospheric dispersion factors applicable to the radioactivity release points I control room receptors applicable to an FHA at either location, and at either unit, are provided in Table 7.3-3. The X/Q values presented in Table 7.3-3 take into consideration the various release points-receptors applicable to the FHA to identify the bounding X/Q values applicable to a FHA at either unit and at either location, and reflect the allowable adjustments I reductions in the values as discussed in Chapter 5 and summarized in the notes of Tables 5.2-2 and 5.2-3.The bounding Control Room dose following a FHA at either location and at either unit is presented in Section 8.7.4 Locked Rotor Accident (LRA)NRC sponsored computer code RADTRAD 3.03, is used to calculate the control room and site boundary dose due to airborne radioactivity releases following a LRA.This event is caused by an instantaneous seizure of a primary reactor coolant pump (RCP)rotor. Flow through the affected loop is rapidly reduced, causing a reactor trip due to a low primary loop flow signal. Fuel damage is predicted to occur as a result of this accident. Due to the pressure differential between the primary and secondary systems and assumed Steam Generator (SG) tube leakage, fission products are discharged from the primary into the secondary system. A portion of this radioactivity is released to the outside atmosphere from the secondary coolant system via the 10% ADVs and MSSVs. Following reactor trip, and based on an assumption of a Loss of Offsite Power (LOOP) coincident with reactor trip, the condenser is assumed to be unavailable and reactor cooldown is achieved using steam releases from the SG MSSVs and 10% ADVs until initiation of shutdown cooling. DCPP has established that the LOL event generates the maximum primary to secondary heat transfer and the LRA assumes these same conservatively bounding secondary steam releases.Regulatory guidance provided for the LRA in pertinent sections of RG 1.183 including Appendix G is used to develop the dose consequence model. The fuel gap fractions used for non-LOCA events are discussed in Section 4.3. Table 7.4-1 lists the key assumptions I parameters utilized to develop the radiological consequences following a LRA.Consistent with current licensing basis, the LRA is postulated to result in 10% fuel failure resulting in the release of the associated gap activity. As discussed in Section 4.3, and as part of the licensing basis change requested with this application, the core gap activity is assumed to be comprised of 12% of the core 1-131 inventory, 30% of the core Kr-85 activity, 10% of the remaining noble gas and halogen isotopes, and 17% of the core alkali metals (Cesium and Rubidium). In accordance with RG 1.183, the activity released from the fuel is assumed to be released instantaneously and mixed homogenously through the primary coolant mass and transmitted to the secondary side via primary to secondary SG tube leakage. A radial peaking factor of 1.65 is applied to the activity release from the fuel gap. The activity associated with the 85 of 205 Diablo Canyon Power Plant Implementation of Alternative Source Terms release of the primary to secondary leakage of normal operation RCS, (at Technical Specification levels) via the MSSVs/10% ADVs are insignificant compared to the failed fuel release and are therefore not included in this assessment. DCPP Plant Technical Specification 3.4.13d limits primary to secondary SG tube leakage to 150 gpd per steam generator for a total of 600 gpd in all 4 SGs. To accommodate any potential accident induced leakage, the LRA dose consequence analysis addresses a limit of 0.75 gpm from all 4 SGs (or a total of 1080 gpd).The chemical form of the iodines in the gap are assumed to be 95% particulate (Csi), 4.85%elemental and 0.15% organic. The effect of SG tube uncovery in intact SGs (for SGTR and non-SGTR events), has been evaluated for potential impact on dose consequences as part of a Westinghouse Owners Group (WOG) Program and demonstrated to be insignificant; therefore, the gap iodines are assumed to have a partition coefficient of 100 in the SG. The iodine releases to the environment from the SG are assumed to be 97% elemental and 3%organic. The gap noble gases are released freely to the environment without retention in the SG whereas the particulates are assumed to be carried over in accordance with the design basis SG moisture carryover fraction.The condenser is assumed unavailable due to the loss of offsite power. Consequently, the radioactivity release resulting from a LRA is discharged to the environment from all steam generators via the MSSVs and the 10% ADVs. The SG releases continue for 10.73 hours, at which time shutdown cooling is initiated via operation of the RHR system, and environmental releases are terminated. EAB 2 hr Worst Case Window AST methodology requires that the worst case dose to an individual located at any point on the boundary at the EAB, for any 2-hr period following the onset of the accident be reported as the EAB dose. For the LRA, the worst two hour period can occur either during the 0-2 hr period when the noble gas release rate is the highest, or during the t=8.73 hr to 10.73 hr period when the iodine and particulate level in the SG liquid peaks (SG releases are terminated at T=10.73 hrs). Regardless of the starting point of the worst 2 hr window, the 0-2 hr EAB y/Q is utilized.The bounding EAB and LPZ dose following a LRA at either unit is presented in Section 8.Accident Specific Control Room Model Assumptions The parameter values utilized for the control room in the accident dose transport model are discussed in Section 7.1. Provided below are the critical LRA-specific assumptions associated with control room response and activity transport. Timing for Initiation of CRVS Mode 4 (if applicable): The LRA does not initiate any signal which could automatically start the control room emergency ventilation. Thus the dose consequence analysis for the LRA assumes that the CR remains in normal operation mode.86 of 205 Diablo Canyon Power Plant Implementation of Alternative Source Terms Control Room Atmospheric Dispersion Factors As noted in Section 5.0, because of the proximity of the MSSV/10% ADVs to the CR normal intake of the affected unit (- 15 ft above the CR intake, horizontal distance is -1.5 meters), and because the releases from the MSSVs/10% ADVs have a vertically upward discharge, it is expected that the concentrations near the normal operation CR intake of the faulted unit (closest to the release point) will be insignificant. Therefore, only the unaffected unit's CR normal intake is assumed to be contaminated by a release from the MSSVs/1 0% ADVs.The bounding atmospheric dispersion factors applicable to the radioactivity release points /control room receptors applicable to an LRA at either unit are provided in Table 7.4-2. The ,/Q values presented in Table 7.4-2 take into consideration the various release points-receptors applicable to the LRA to identify the bounding.X/Q values applicable to a LRA at either unit, and reflect the allowable adjustments / reductions in the values as discussed in Chapter 5 and summarized in the notes of Tables 5.2-2 and 5.2-3.The bounding Control Room dose following a LRA at either unit is presented in Section 8.7.5 Control Rod Ejection Accident (CREA)NRC sponsored computer code RADTRAD 3.03, is used to calculate the control room and site boundary dose due to airborne radioactivity releases following a CREA.This event consists of an uncontrolled withdrawal of a control rod from the reactor core. The CREA results in reactivity insertion that leads to a core power level increase, and under adverse combinations of circumstances fuel failure, and a subsequent reactor trip. In this case, some of the activity in the fuel rod gaps would be released to the coolant and in turn to the inside of the containment building. As a result of pressurization of the containment, some of this activity could leak to the environment. Following reactor trip, and based on an assumption of a Loss of Offsite Power coincident with reactor trip, the condenser is assumed to be unavailable and reactor cooldown is achieved using steam releases from the SG MSSVs and 10% ADVs until initiation of shutdown cooling.DCPP has established that the LOL event generates the maximum primary to secondary heat transfer and the CREA assumes these same conservatively bounding secondary steam releases.Regulatory guidance provided for the CREA in pertinent sections of RG 1.183 including Appendix H is used to develop the dose consequence model. Table 7.5-1 lists the key assumptions / parameters utilized to develop the radiological consequences following a CREA.Consistent with current licensing basis, the CREA is postulated to result in 10% fuel failure resulting in the release of the. associated gap activity. Per RG 1.183, the core gap activity is assumed to be comprised of 10% of the core noble gases and halogens. A radial peaking factor of 1.65 is applied to the activity release from the fuel gap.87 of 205 Diablo Canyon Power Plant Implementation ofAlternative Source Terms In accordance with guidance provided in RG 1.183, two independent release paths to the environment are analyzed: first, via containment leakage of the fission products released due to the event from the primary system to containment, assuming that the containment pathway is the only one available; and second, via releases from the secondary system, outside containment, following primary-to-secondary leakage in the steam generators, assuming that the latter pathway is the only one available. The actual doses resulting from a postulated CREA would be a composite of doses resulting from portions of the release going out via the containment building and, portions via the secondary system. If regulatory compliance to dose limits can be demonstrated for each of the scenarios, the dose consequence of a scenario that is a combination of the two will be encompassed by the more restrictive of the two analyzed scenarios. The DCPP CREA dose consequence analysis evaluates the following two scenarios. Scenario 1: The failed fuel resulting from a postulated CREA is released into the RCS, which is released in its entirety into the containment via the ruptured control rod drive mechanism housing, is mixed in the free volume of the containment, and then released to the environment at the containment technical specification leak rate for the first 24 hrs and at half that value for the remaining 29 days.Scenario 2: The failed fuel resulting from a postulated CREA is released into the RCS which is then transmitted to the secondary side via steam generator tube leakage. The condenser is assumed to be unavailable due to a loss of offsite power. Environmental releases occur from the steam generators via the MSSVs and 10% ADVs.The chemical composition of the iodine in the gap is assumed to be 95% particulate (Csl), 4.85% elemental and 0.15% organic. However, because the sump pH is not controlled following a CREA, it is conservatively assumed that the iodine released via the containment leakage pathway has the same composition as the iodine released via the secondary system release pathway; i.e.; it is assumed that for both scenarios, 97% of all halogens available for release to the environment are elemental, while the remaining 3% is organic.Scenario 1: Transport From Containment The failed fuel activity released due to a CREA into the RCS is assumed to be instantaneously released into the containment where it mixes homogeneously in the containment free volume.The containment is assumed to leak at the technical specification leak rate of 0.001 per day for the first 24 hours and at half that value for the remaining 29 days after the event. Except for decay, no credit is taken for depleting the halogen or noble gas concentrations airborne in the containment. Per RG 1.183, the chemical composition of the iodine in the gap fuel is 95%particulate (CsI), 4.85% elemental and 0.15% organic. However, since no credit is taken for the actuation of sprays or pH control, the iodine released via containment leakage pathway is assumed to have the same composition as iodine activity released to the environment from the secondary coolant; i.e.; 97% elemental and 3% organic. Environmental releases due to containment leakage can occur unfiltered as a diffuse source from the containment wall, and as a point source via the containment penetration areas or the Plant Vent. The dose consequences are estimated based on the worst case atmospheric dispersion factors, i.e., an 88 of 205 Diablo Canyon Power Plant Implementation of Alternative Source Terms assumed environmental release via the containment penetration areas.Scenario 2: Transport from Secondary System The failed fuel activity released due to a CREA into the RCS is assumed to be instantaneously and homogeneously mixed in the reactor coolant system and transmitted to the secondary side via primary to secondary SG tube leakage. The activity associated with the release of the initial inventory in secondary steam/liquid, and primary to secondary leakage of normal operation RCS, (both at Technical Specification levels) via the MSSVs/10% ADVs are insignificant compared to the failed fuel release, and are therefore not included in this assessment. DCPP Plant Technical Specification 3.4.13d limits primary to secondary SG tube leakage to 150 gpd per steam generator for a total of 600 gpd in all 4 SGs. To accommodate any potential accident induced leakage, the CREA dose consequence analysis addresses a limit of 0.75 gpm from all 4 SGs (or a total of 1080 gpd).The effect of SG tube uncovery in intact SGs (for SGTR and non-SGTR, events), has been evaluated for potential impact on dose consequences as part of a WOG Program and demonstrated to be insignificant; therefore, the gap iodines have a partition coefficient of 100 in the SG. The gap noble gases are released freely to the environment without retention in the SG.The condenser is assumed unavailable due to the loss of offsite power. Consequently, the radioactivity release resulting from a CREA is discharged to the environment from steam generators via the MSSVs and the 10% ADVs. Per RG 1.183, 97% of all halogens available for release to the environment via the Secondary System are elemental, while the remaining 3% are organic. The SG releases continue until shutdown cooling is initiated via operation of the RHR system (10.73 hours after the accident) and environmental releases are terminated. EAB 2 hr Worst Case Window AST methodology requires that the worst case dose to an individual located at any point on the boundary at the EAB, for any 2-hr period following the onset of the accident be reported as the EAB dose. For Scenario 1 (release via Containment leakage), the worst case 2-hour period occurs during the first 2 hours). For Scenario 2 (release via secondary side), the worst two hour period can occur either during the 0-2 hr period when the noble gas release rate is the highest, or during the t=8.73 hr to 10.73 hr period when the iodine and particulate level in the SG liquid peaks (SG releases are terminated at T=1 0.73 hrs). Regardless of the starting point of the worst 2 hr window, the 0-2 hr EAB %/Q is utilized.The bounding EAB and LPZ dose following a CREA at either unit for both scenarios are presented in Section 8.89 of 205 Diablo Canyon Power Plant Implementation of Alternative Source Terms Accident Specific Control Room Model Assumptions The parameter values utilized for the control room in the accident dose transport model are discussed in Section 7.1. Provided below are the critical CREA-specific assumptions associated with control room response and activity transport. Timinq for Initiation of CRVS Mode 4: The time to generate a signal to switch CRVS operation from Mode 1 to Mode 4 is based on the containment pressure response following a 2 inch small-break LOCA (SBLOCA), and the.fact that at DCPP, a Containment High Pressure signal will initiate a SI signal which will automatically initiate CRVS Mode 4 pressurization. The containment pressure response analysis for a 2 inch SBLOCA shows that the 5 psig setpoint for Containment High Pressure is reached in -150 seconds after the SBLOCA. As indicated earlier, releases to the containment following a CREA are through a ruptured control rod drive mechanism housing.The control rod shaft diameter is 1.840 inches and the RCCA housing penetration opening is 4 inches in diameter. Based on the above and for the purposes of conservatism, the time to generate the Containment High Pressure SI signal following a CREA is assumed to be double the value applicable to the 2 inch SBLOCA, or 300 seconds.Based on the above, following a CREA,* An SI signal will be generated at t = 300 sec following a CREA.* The CRVS normal intake dampers of the accident unit start to close after a 28.2 second delay due to delays associated with diesel generator loading onto the 4kv buses. The CR dampers are fully closed 10 secs later, or at t=338.2 secs (i.e., 300 + 28.2 + 10).The 2 second SI signal processing time occurs in parallel with diesel generator sequencing and is therefore not included as part of the delay.* In accordance with DCPP licensing basis, the CRVS normal operation dampers of the non-accident unit are not affected by the LOOP and are isolated at t=312 secs (i.e., 300+ 2 secs signal processing time + 10 sec damper closure time).Control Room Atmospheric Dispersion Factors: As noted in Section 5.0, because of the proximity of the MSSV/10% ADVs to the CR normal intake of the affected unit (- 15 ft above the CR intake, horizontal distance is -1.5 meters), and because the releases from the MSSVs/10% ADVs have a vertically upward discharge, it is expected that the concentrations near the normal operation CR intake of the faulted unit (closest to the release point) will be insignificant. Therefore, prior to switchover to CRVS Mode 4 pressurization, only the unaffected unit's CR normal intake is assumed to be contaminated by a release from the MSSVs/1 0% ADVs.The bounding atmospheric dispersion factors applicable to the radioactivity release points /control room receptors applicable to a CREA at either unit are provided in Table 7.5-2. The XIQ values presented in Table 7.5-2 take into consideration the various release points-receptors applicable to the CREA to identify the bounding XfQ values applicable to a CREA at 90 of 205 Diablo Canyon Power Plant Implementation ofAfternative Source Terms either unit, and reflect the allowable adjustments / reductions in the values as discussed in Chapter 5 and summarized in the notes of Tables 5.2-2 and 5.2-3.The bounding Control Room dose following a CREA at either unit is presented in Section 8.7.6 Main Steam Line Break (MSLB)NRC sponsored computer code RADTRAD 3.03, is used to calculate the control room and site boundary dose due to airborne radioactivity releases following a MSLB.This event consists of a double-ended break of one main steam line. The analysis focusses on a MSLB outside the containment since a MSLB inside containment will clearly result in a lesser dose to a control room operator or to the offsite public due to hold-up of activity in the containment. Following a MSLB, the affected SG rapidly depressurizes and releases the initial contents to the environment via the break. Based on an assumption of a Loss of Offsite Power coincident with reactor trip, the condenser is assumed to be unavailable, and environmental steam releases via the MSSVs / 10% ADVs of the intact steam generators are used to cool down the reactor until initiation of shutdown cooling. The activity in the ROS leaks into the faulted and intact steam generators via SG tube leakage and is released to the environment from the break point, anrd from the MSSVs / 10% ADVs, respectively. Regulatory guidance provided for the MSLB in pertinent sections of RG 1.183 including Appendix E is used to develop the dose consequence model. Table 7.6-1 lists the key assumptions / parameters utilized to develop the radiological consequences following a MSLB.No melt or clad breach is postulated for the DCPP MSLB event. Thus, and in accordance with RG 1.183, Appendix E, item 2, the activity released is based on the maximum coolant activity allowed by the plant technical specifications. The plant technical specifications focus on the noble gases and iodines. In addition, and per RG 1.183, two scenarios are addressed, i.e., a)a pre-accident iodine spike and b) an accident-initiated iodine spike.a. Pre-accident Iodine Spike -the initial primary coolant iodine activity is assumed to be 60 j.Ci/gm of DE 1-131 which is the transient Technical Specification limit for full power operation. The initial primary coolant noble gas activity is assumed to be at Technical Specification levels.b. Accident-Initiated Iodine Spike -the initial primary coolant iodine activity is assumed to be at Technical Specification of 1 l.iCi/gm DE 1-131 (equilibrium Technical Specification limit for full power operation). Immediately following the accident the iodine appearance rate from the fuel to the primary coolant is assumed to increase to 500 times the equilibrium appearance rate corresponding to the 1 .Ci/gm DE 1-131 coolant concentration. The duration of the assumed spike is 8 hours. The initial primary coolant noble gas activity is assumed to be at Technical Specification levels.91 of 205 Diablo Canyon Power Plant Implementation ofAlternative Source Terms The initial secondary coolant iodine activity is assumed to be at the Technical Specification limit of 0.1 ltCi/gm DE 1-131.DCPP Plant Technical Specification 3.4.13d limits primary to secondary SG tube leakage to 150 gpd per steam generator for a total of 600 gpd in all 4 SGs. To accommodate any potential accident induced leakage, the MSLB dose consequence analysis addresses a limit of 0.75 gpm from all 4 SGs (or a total of 1080 gpd).Following a MSLB, the primary and secondary reactor coolant activity is released to the environment via two pathways.Faulted Steam Generator The release from the faulted SG occurs via the postulated break point of the main-steam line.The faulted SG is estimated to dry-out almost instantaneously following the MSLB (in -10 seconds), releasing all of the iodine in the secondary coolant (at Technical Specification concentrations) that was initially contained in the steam generator. The EAB and LPZ dose to the public is calculated using an instantaneous release of the iodine inventory (Ci) in the SG liquid in the faulted SG. The secondary steam activity initially contained in the faulted steam generator is also released; however, the associated dose contribution is not included in this analysis since it is considered insignificant. To maximize the control room and offsite doses following a MSLB, the maximum allowable primary to secondary SG tube leakage for all SGs (0.75 gpm or 1080 gpd at Standard Temperature and Pressure (STP) conditions), is conservatively assumed to occur in the faulted SG. All iodine and noble gas activities in the referenced tube leakage are released directly to the environment without hold-up or decontamination. The primary to secondary SG tube leakage is assumed to go on until the RCS reaches 212 0 F, which based on minimum heat transfer rates, is conservatively estimated to occur 30 hours after the event.Intact Steam Generators The initial iodine activities in the secondary coolant at Technical Specification levels are released to the environment in proportion to the steaming rate and the inverse of the partition coefficient (limited to 100) defined in RG 1.183. The noble gases are released freely to the environment without retention in the steam generators. However, there is no primary to secondary leakage into the intact SG as all primary to secondary leakage (1080 gpd or 0.75 gpm) is assumed to be occurring in the faulted SG.The iodine releases to the environment from the SG are assumed to be 97% elemental and 3% organic. The condenser is assumed unavailable due to the loss of offsite power. The SG releases continue for 10.73 hours, at which time shutdown cooling is initiated via operation of the RHR system and environmental releases are terminated. 92 of 205 Diablo Canyon Power Plant Implementation ofAlternative Source Terms EAB 2 hr Worst Case Window AST methodology requires that the worst case dose to an individual located at any point on the boundary at the EAB, for any 2-hr period following the onset of the accident be reported as the EAB dose.The Source/Release for the Pre-incident Spike Case is at its maximum levels between 0 and 2 hours.The Source/Release for the Accident-Initiated Spike Case is at its maximum levels towards the end of the spiking period.Regardless of the starting point of the "Worst 2-hr Window," the 0-2 hrs X/Q is utilized.The bounding EAB and LPZ dose following a MSLB at either unit for both scenarios are presented in Section 8.Accident Specific Control Room Model Assumptions The parameter values utilized for the control room in the accident dose transport model are discussed in Section 7.1. Provided below are the critical MSLB-specific assumptions associated with control room response and activity transport. Timing for Initiation of CRVS Mode 4:* An Sl signal will be generated at t = 0.6 sec following a MSLB.* The CRVS normal intake dampers of the accident unit start to close after a 28.2 second delay due to delays associated with diesel generator loading onto the 4kv buses. The CR dampers are fully closed 10 secs later, or at t=38.8 secs (i.e., 0.6 + 28.2 + 10). The 2 second SI signal processing time occurs in parallel with diesel generator sequencing and is therefore not included as part of the delay." In accordance with DCPP licensing basis, the CRVS normal operation dampers of the non-accident unit are not affected by the LOOP and are isolated at t=12.6 secs (i.e., 0.6+ 2 secs signal processing time + 10 sec damper closure time).Transport of radioactivity from the Break Location Since the normal operation (CRVS Mode 1) control room intake of the faulted unit is in such close proximity to the break point, an atmospheric dispersion factor (XJQ) cannot be accurately determined. Thus, atmospheric dispersion is not credited when determining the control room operator dose from the secondary coolant discharge or the primary to secondary SG tube leakage released from the faulted SG via the break point.Secondary Coolant Discharge: The radioactivity release due to the almost immediate dry-out of the faulted SG following a MSLB is based on a) the radioactivity concentration of the iodine in a finite cloud created by the secondary coolant liquid flash at the break point; b) conservation of total iodine activity in the SG liquid. The 93 of 205 6Diablo Canyon Power Plant Implementation ofAlternative Source Terms activity concentration at the release point is conservatively based on saturated steam at a density of 5.98E-04 gm/cm 3 , (i.e., at 1 atmosphere and 212 0 F). The activity concentration entering the CR is assumed to be the same as the concentration at the break point until the CR normal ventilation is isolated and the CRVS re-aligned to Mode 4 Pressurization. Primary to Secondary SG Tube Leakage: Due to the close proximity of the normal operation CR intake of the faulted unit and MSL break release point and consequent unavailability of viable atmospheric dispersion factors, the primary to secondary SG tube leakage into the faulted SG is conservatively assumed to be piped directly into the control room. This model is reasonable since the relatively small plume of steam created by the -0.485 gallon {i.e., (0.75 gallon/min)(38.8 s) / 60 s/man) of reactor coolant released due to SG tube leakage via the MSL break point could easily be swept into the CR due to the close proximity of the CR normal intake to the break point.Control Room Atmospheric Dispersion Factors As noted in Section 5.0, because of the proximity of the MSSVs/10% ADVs to the CR normal intake of the affected unit (- 15 ft above the CR intake, horizontal distance is -1.5 meters), and because the releases from the MSSVs/10% ADVs have a vertically upward discharge, it is expected that the concentrations near the normal operation CR intake of the affected unit (closest to the release point) will be insignificant. Therefore, prior to switchover to CRVS Mode 4 pressurization, only the unaffected unit's CR normal intake is assumed to be contaminated by releases from the MSSVs/10% ADVs.The bounding atmospheric dispersion factors applicable to the radioactivity release points /control room receptors applicable to a MSLB at either unit are provided in Table 7.6-2. The X/Q values presented in Table 7.6-2 take into consideration the various release points-receptors applicable to the MSLB to identify the bounding X/Q values applicable to a MSLB at either unit, and reflect the allowable adjustments / reductions in the values as discussed in Chapter 5 and summarized in the notes of Tables 5.2-2 and 5.2-3.The bounding Control Room dose following a MSLB at either unit is presented in Section 8.7.7 Steam Generator Tube Rupture (SGTR)NRC sponsored computer code RADTRAD 3.03, is used to calculate the control room and site boundary dose due to airborne radioactivity releases following a SGTR.This event is caused by the instantaneous rupture of a SG tube with a resultant release of primary coolant into the lower pressure secondary system. No melt or clad breach is postulated for the SGTR event. The calculation assumes a stuck-open 10% ADV of the ruptured steam generator for 30 minutes. Based on an assumption of a Loss of Offsite Power coincident with reactor trip, the condenser is assumed to be unavailable, and environmental steam releases via the MSSVs / 10% ADVs of the intact steam generators are used to cool down the reactor until initiation of shutdown cooling. A portion of the primary coolant break flow in the ruptured SG flashes and is released a) to the condenser before reactor trip and b)94 of 205 Diablo Canyon Power Plant Implementation of Alternative Source Terms directly to the environment after reactor trip, via the MSSVs and 10% ADVs. The remaining break flow mixes with the secondary side liquid, and is released to the environment via steam releases through MSSVs and 10% ADVs. The activity in the RCS also leaks into the intact steam generators via SG tube leakage and is released to the environment from the MSSVs /10% ADVs.Regulatory guidance provided for the SGTR in pertinent sections of RG 1.183 including Appendix F is used to develop the dose consequence model. Table 7.7-1 lists the key assumptions / parameters utilized to develop the radiological consequences following a SGTR. Table 7.7-2 provides the time dependent steam flow from the ruptured and intact SGs and the flashed and unflashed break flow in the ruptured SG.No melt or clad breach is postulated for the DCPP SGTR event. Thus, and in accordance with RG 1.183, Appendix F, item 2, the activity released is based on the maximum coolant activity allowed by the plant technical specifications. The plant technical specifications focus on the noble gases and iodines. In addition, and per RG 1.183, two scenarios are addressed, i.e., a) a pre-accident iodine spike and b) an accident-initiated iodine spike.a. Pre-accident Iodine Spike -the initial primary coolant iodine activity is assumed to be 60 ytCi/gm of DE 1-131 which is the transient Technical Specification limit for full power operation. The initial primary coolant noble gas activity is assumed to be at Technical Specification levels.b. Accident-Initiated Iodine Spike -the initial primary coolant iodine activity is assumed to be at Technical Specification of 1 pCi/gm DE 1-131 (equilibrium Technical Specification limit for full power operation). Immediately following the accident the iodine appearance rate from the fuel to the primary coolant is assumed to increase to 335 times the equilibrium appearance rate corresponding to the I YtCi/gm DE 1-131 coolant concentration. The duration of the assumed spike is 8 hours. The initial primary coolant noble gas activity is assumed to be at Technical Specification levels.The initial secondary coolant iodine activity is assumed to be at the Technical Specification limit of 0.1 pCi/gm DE 1-131.DCPP Plant Technical Specification 3.4.13d limits primary to secondary SG tube leakage to 150 gpd per steam generator for a total of 600 gpd in all 4 SGs. To accommodate any potential accident induced leakage, the SGTR.dose consequence analysis addresses a limit of 0.75 gpm from all 4 SGs (or a total of 1080 gpd). To maximize the dose consequences, the analysis conservatively assumes that all of the 0.75 gpm SG tube leakage occurs in the intact SGs.Following a SGTR, the primary and secondary reactor coolant activity is released to the environment via two pathways.Ruptured Steam Generator A SGTR will result in a large amount of primary coolant being released to the rutured steam generator via the break location with a significant portion of it flashed to the steam space.95 of 205 Diablo Canyon Power Plant implementation ofAlternative Source Terms In accordance with the guidance provided in RG 1.183, the noble gases in the entire break flow and the iodine in the flashed portion of the break flow are assumed to be immediately available for release from the steam generator. The iodine in the non-flashed portion of the break flow mixes uniformly with the steam generator liquid mass and is released into the steam space in proportion to the steaming rate and the inverse of the allowable partition coefficient of 100. The iodine releases from the SGs are assumed to be 97% elemental and 3% organic.Before the reactor trip at t=179 seconds, the radioactivity in the steam is released to the environment from the air ejector which discharges into the plant vent. All noble gases and organic iodines in the steam are released directly to the environment. Only a portion of the elemental iodine carried with the steam is partitioned to the air ejector and released to the environment. The rest is partitioned to the condensate, returns to both the intact steam generators and the ruptured steam generator and will be available for future steaming releases.After the reactor trip, the radioactivity in the steam is released to the environment from the MSSVs/10% ADVs, due to the assumption of loss of offsite power (LOOP). To isolate the ruptured steam loop, the auxiliary feed water to the ruptured SG is secured. The calculation assumes the 10% ADV of the ruptured SG fails open for 30 minutes. The fail-open 10% ADV is isolated at t = 2653 seconds at which time the ruptured 'steam loop is isolated. The break flow continues until the primary system is in equilibrium with the secondary side of the ruptured SG at t = 5872 seconds. The iodines in the flashed break flow and the noble gases in the entire break flow is bottled up in the steam space of the ruptured SG and released to the environment during the manual depressurization of the ruptured SG after t = 2 hours.Intact Steam Generators The radioactivity released from the intact steam generators includes two components: (a) a portion of the break flow activity that is transferred to the intact steam generators via the condenser before reactor trip, and (b) due to SG tube leakage.Approximately 75% (3 intact SGs vs 1 ruptured SG) of the flashed break flow activity that is transported and retained in the condenser before reactor trip will be transferred to the intact steam generators and released to the environment during the cool-down phase.The total primary-to-secondary tube leak rate in the 3 intact SGs is conservatively assumed to be 0.75 gpm. The effect of SG tube uncovery in intact SGs (for SGTR and non-SGTR events)has been evaluated for potential impact on dose consequences as part of a WOG Program and demonstrated to be insignificant. Thus all leaked primary coolant iodine activities are assumed to mix uniformly with the steam generator liquid and are released in proportion to the steaming rate and the inverse of the partition coefficient. Before the reactor trip, the activity in the main steam is released from the plant vent via the air ejector/ condenser. After the reactor trip, the steam is released from the MSSVs/10% ADVs. The reactor coolant noble gases that enter the intact steam generator are released directly to the environment without holdup. The iodine releases from the SGs are assumed to be 97% elemental and 3% organic. The intact SG steam release continues until shutdown cooling (SDC) is initiated at t = 10.73 hours 96 of 205 Diablo Canyon Power Plant Implementation ofAlternative Source Terms Initial Secondary Coolant Activity Release The initial iodine activities in the secondary coolant are released to the environment in proportion to the steaming rate and the inverse of the partition coefficient from the ruptured and intact SGs. Twenty five percent of the initial secondary coolant iodine inventory is in the ruptured SG and 75% of the initial secondary coolant iodine inventory is in the 3 intact SGs.EAB 2 hr Worst Case Window AST methodology requires that the worst case dose to an individual located at any point on the boundary at the EAB, for any 2-hr period following the onset of the accident be reported as the EAB dose.For the SGTR, the EAB dose is controlled by the release of the flashed break flow in the ruptured SG which stops at 3402 seconds. The break flow stops at 5872 seconds and the ruptured SG is manually depressurized 2 hours after the accident. Therefore the maximum EAB dose occurs during the 0-2hr period for both the pre-accident and accident initiated iodine spike cases.Regardless of the starting point of the "Worst 2-hr Window," the 0-2 hrs X/Q is utilized.The bounding EAB and LPZ dose following a SGTR at either unit for both scenarios are presented in Section 8.Accident Specific Control Room Model Assumptions The parameter values utilized for the control room in the accident dose transport model are discussed in Section 7.1. Provided below are the critical SGTR-specific assumptions associated with control room response and activity transport. Timing for Initiation of CRVS Mode 4:* An SI signal will be generated at t = 219 sec following a SGTR." The CRVS normal intake dampers of the accident unit start to close after a 28.2 second delay due to delays associated with diesel generator loading onto the 4kv buses. The CR dampers are fully closed 10 secs later, or at t=257.2 secs (i.e., 219 + 28.2 + 10).The 2 second SI signal processing time occurs in parallel with diesel generator sequencing and is therefore not included as part of the delay.* In accordance with DCPP licensing basis, the CRVS normal operation dampers of the non-accident unit are not affected by the LOOP and are isolated at t=231 secs (i.e., 219+ 2 secs signal processing time + 10 sec damper closure time).Control Room Atmospheric Dispersion Factors As noted in Section 5.0, because of the proximity of the MSSVs/10% ADVs to the CR normal intake of the affected unit (- 15 ft above the CR intake, horizontal distance is -1.5 meters), 97 of 205 Diablo Canyon Power Plant Implementation of Alternative Source Terms and because the releases from the MSSVs/10% ADVs have a vertically upward discharge, it is expected that the concentrations near the normal operation CR intake of the affected unit (closest to the release point) will be insignificant. Therefore, prior to switchover to CRVS Mode 4 pressurization, only the unaffected unit's CR normal intake is assumed to be contaminated by releases from the MSSVs/10% ADVs.The bounding atmospheric dispersion factors applicable to the radioactivity release points I control room receptors applicable to a SGTR at either unit are provided in Table 7.7-3. The z/Q values presented in Table 7.7-3 take into consideration the various release points-receptors applicable to the SGTR to identify the bounding X/Q values applicable to a SGTR at either unit, and reflect the allowable adjustments / reductions in the values as discussed in Chapter 5 and summarized in the notes of Tables 5.2-2 and 5.2-3.The bounding Control Room dose following a SGTR at either unit is presented in Section 8.7.8 Loss of Load Event (LOL)NRC sponsored computer code RADTRAD 3.03, is used to calculate the control room and site boundary dose due to airborne radioactivity releases following a LOL event.In accordance with DCPP current licensing basis documented in UFSAR Chapter 15.5.10, Condition II events that are expected to result in atmospheric steam releases are:* Loss of electrical load and/or turbine trip (UFSAR 15.2.7)* Loss of normal Feedwater (UFSAR 15.2.8)" Loss of offsite power to the station auxiliaries (UFSAR 15.2.9)* Accidental depressurization of the main steam system (UFSAR 15.2.14)At DCPP, the mass of environmental steam releases for the Loss of Load Event bound all Condition II events and encompass the LRA and CREA.SRP 15.2.1 to 15.2.5 (Reference

52) indicates that a Loss of Load event is different from the Loss of Alternating Current (AC) power condition (discussed in SRP 15.2.6, Reference 53), in that offsite AC power remains available to support station auxiliaries (e.g., reactor coolant pumps). The Loss of AC power condition results in the condenser being unavailable and reactor cooldown being achieved using steam releases from the SG MSSVs and 10% ADVs until initiation of shutdown cooling.In keeping with the concept of developing steam releases that bound all Condition II events and encompass the LRA and CREA, the analysis performed to determine the mass of steam released following a Loss of Load event incorporates the assumption of Loss of offsite power to the station auxiliaries.

Neither RG 1.183 nor NUREG-0800 provides specific guidance with respect to scenarios to be assumed to determine radiological dose consequences from Condition 11 events. Thus the scenario outlined below for the bounding Condition II event that results in environmental releases is based on the conservative assumptions outlined in RG 1.183 for the MSLB.98 of 205 Diablo Canyon Power Plant Implementation ofAlternative Source Terms Table 7.8-1 lists the key assumptions / parameters utilized to develop the radiological consequences following a LOL event. The conservative assumptions utilized to assess the dose consequences ensure that it represents the Limiting Condition I1 event.As noted in DCPP UFSAR Section 15.2, no melt or clad breach is postulated for the DCPP LOL- Limiting Condition II event. Thus, and in accordance with RG 1.183, Appendix E, item 2, the activity released is based on the maximum coolant activity allowed by the plant technical specifications. The plant technical specifications focus on the noble gases and iodines. In addition, and per RG 1.183, two scenarios are addressed, i.e., a) a pre-accident iodine spike and b) an accident-initiated iodine spike.a. Pre-accident Iodine Spike -the initial primary coolant iodine activity is assumed to be 60 ptCi/gm of DE 1-131 which is the transient Technical Specification limit for full power operation. The initial primary coolant noble gas activity is assumed to be at Technical Specification levels.b. Accident-initiated Iodine Spike -the initial primary coolant iodine activity is assumed to be at Technical Specification of 1 ýtCi/gm DE 1-131 (equilibrium Technical Specification limit for full power operation). Immediately following the accident the iodine appearance rate from the fuel to the primary coolant is assumed to increase to 500 times the equilibrium appearance rate corresponding to the I ýtCi/gm DE 1-131 coolant concentration. The duration of the assumed spike is 8 hours. The initial primary coolant noble gas activity is assumed to be at Technical Specification levels.The initial secondary coolant iodine activity is the Technical Specification limit of 0.1 ýtCi/gm DE 1-131.DCPP Plant Technical Specification 3.4.13d limits primary to secondary SG tube leakage to 150 gpd per steam generator for a total of 600 gpd in all 4 SGs. To accommodate any potential accident induced leakage, the LOL dose consequence analysis addresses a limit of 0.75 gpm from all 4 SGs (or a total of 1080 gpd).The entire primary-to-secondary tube leakage of 0.75 gpm (maximum leak rate at STP conditions; total for all 4 SGs) is leaked into an effective SG. In accordance with RG 1.183, the pre-existing iodine activity in the secondary coolant and iodine activity due to reactor coolant leakage into the 4 SGs is assumed to be homogeneously mixed in the bulk secondary coolant. The effect of SG tube uncovery in intact SGs (for SGTR and non-SGTR events) has been evaluated for potential impact on dose consequences as part of a WOG Program and demonstrated to be insignificant. Therefore, per RG 1.183, the iodines are released to the environment via the MSSVs I 10% ADVs in proportion to the steaming rate and the inverse of a partition coefficient of 100. The iodine releases to the environment from the SG are assumed to be 97% elemental and 3% organic. The noble gases are released freely to the environment without retention in the SG.The condenser is assumed unavailable due to a loss of offsite power coincident with reactor trip. Consequently, the radioactivity release resulting from a LOL-Limiting Condition II event is discharged to the environment from the steam generators via the MSSVs and the 10% ADVs.99 of 205 Diablo Canyon Power Plant Implementation ofAlternative Source Terms The SG releases continue for 10.73 hours, at which time shutdown cooling is initiated via operation of the RHR system and environmental releases are terminated. EAB 2 hr Worst Case Window AST methodology requires that the worst case dose to an individual located at any point on the boundary at the EAB, for any 2-hr period following the onset of the accident be reported as the EAB dose. For the LOL-Limiting Condition 11 event, the worst two hour period can occur either during the 0-2 hr period when the noble gas release rate is the highest, or during the t=8.73 hr to 10.73 hr period when the iodine level in the SG liquid peaks (SG releases are terminated at T=10.73 hrs). Regardless of the starting point of the worst 2 hr window, the 0-2 hr EAB %/Q is utilized.The bounding EAB and LPZ dose following a LOL-Limiting Condition 11 event at either unit is presented in Section 8.Accident Specific Control Room Model Assumptions The parameter values utilized for the control room in the accident dose transport model are discussed in Section 7.1. Provided below are the critical LOL event-specific assumptions associated with control room response and activity transport. Timinq for Initiation of CRVS Mode 4 (if applicable): The LOL-Limiting Condition II event does not initiate any signal which could automatically start the control room pressurization air ventilation. Thus the dose consequence analysis for the LOL-Limiting Condition II event assumes that the CR remains in normal operation mode.Control Room Atmospheric Dispersion Factors As noted in Section 5.0, because of the proximity of the MSSVs/10% ADVs to the CR normal intake of the affected unit (- 15 ft above the CR intake, horizontal distance is -1.5 meters), and because the releases from the MSSVs/10% ADVs have a vertically upward discharge, it is expected that the concentrations near the normal operation CR intake of the affected unit (closest to the release point) will be insignificant. Therefore, only the unaffected unit's CR normal intake is assumed to be contaminated by releases from the MSSVs/10% ADVs.The bounding atmospheric dispersion factors applicable to the radioactivity release points /control room receptors applicable to an LOL-Limiting Condition II event at either unit are provided in Table 7.8-2. The %IQ values presented in Table 7.8-2 take into consideration the various release points-receptors applicable to the LOL to identify the bounding 7/Q values applicable to a LOL-Limiting Condition II event at either unit, and reflect the allowable adjustments / reductions in the values as discussed in Chapter 5 and summarized in the notes of Tables 5.2-2 and 5.2-3.The bounding Control Room dose following a LOL-Limiting Condition II event at either unit is presented in Section 8.100 of 205 A" 419 Diablo Canyon Power Plant Implementation ofAlternative Source Terms Table 7.1-1 Control Room Analysis Assumptions & Key Parameter Values Parameter Value Free Volume 170,000 ftl Unfiltered Normal Operation Intake Unit 1: 2100 cfm +/- 10%Unit 2: 2100 cfm +/- 10%Emergency Pressurization Flow Rate 650 -900 cfm Maximum Unfiltered Backdraft Damper Leakage 100 cfm during CR Pressurization Operation Carbon / HEPA Filter Flow during CR Pressurization 1800 -2200 cfm Operation Emergency Filtered Recirculation Rate 1250 cfm (minimum)Pregsurization Intake and Recirculation Carbon/HEPA 93% (iodine)Filter Efficiency (includes filter bypass) 98% (particulates) Unfiltered Inleakage 70 cfm (maximum)(Normal and Pressurization Mode) Includes 10 cfm ingress / egress Occupancy Factors 0-24 hr (1.0)1- 4 d (0.6)4-30 d (0.4)Operator Breathing Rate 0-30 d (3.50E-04 md/sec)101 of 205 4M., Diablo Canyon Power Plant Implementation ofAlternative Source Terms I Table 7.1-2 Technical Support Center Analysis Assumptions & Key Parameter Values Parameter Value Free Volume 51,250 ft3 Filtered (HEPA only) Normal Operation Intake Flow 500 cfm Rate Normal Intake HEPA Filter Efficiency 98% (particulates)(includes filter bypass)Filtered (Carbon / HEPA) Pressurization Flow Rate 500 cfm Flow through Carbon / HEPA Filter during 1000 cfm Pressurization mode Filtered Recirculation flow rate during Pressurization 500 cfm (minimum)mode Pressurization Intake and Recirculation Carbon/HEPA 93% (iodine)Filter Efficiency (includes filter bypass) 98% (particulates) Unfiltered Inleakage 60 cfm (maximum)Includes 10 cfm ingress/egress Occupancy Factors 0-24 hr (1.0)1 -4 d (0.6)4-30 d (0.4)Operator Breathing Rate 0-30 d (3.50E-04 m"Isec)102 of 205 46 Diablo Canyon Power Plant Implementation of Alternative Source Terms Table 7.2-1 Loss of Coolant Accident Assumptions & Key Parameter Values Parameter Value Core Power Level (105% of the rated power of 3411 MWth) 3580 MWt Fuel Activity Release Fractions Per Reg. Guide 1.183 (See Section 7.2.3.2.6) Fuel Release Timing (gap) Onset: 30 sec Duration: 0.5 hr Fuel Release Timing (Early-In-Vessel) Onset: 0.5 hr Duration: 1.3 hr Core Activity Table 4.1-1 Chemical Form of Iodine released from fuel to containment 4.85% elemental atmosphere 95% particulate 0.15% organic Chemical Form of Iodine Released from RCS and sump 97% elemental water 3% organic Containment Vacuum/Pressure Relief Parameters Minimum Containment Free Volume: 2.550E+06 Primary Coolant Tech Spec Activity Table 4.2-1 Chemical Form of Iodine Released 97% elemental; 3% organic Maximum RCS flash fraction after LOCA Noble Gases 100%Halogens 40%Maximum containment pressure relief line air flow rate 218 actual cubic feet per second (acfs)Maximum duration of release via containment pressure relief 13 sec line Release Point Plant Vent Containment Leakage Parameters Containment Spray Coverage -Injection Spray and 82.5% (sprayed fraction)Recirculation Spray Modes: Sprayed Volume 2.103E+06 ft 3 Unsprayed Volume 4.470E+05 ft 3 Minimum mixing flow rate from unsprayed to sprayed region: Before actuation of CFCUs 2 unsprayed regions/hr After actuation of CFCUs 9.13 unsprayed regions/hr Minimum duration of mixing via CFCUs Start = 86 sec End = 30 days Containment spray in injection mode Initiation time 111 sec Termination time 3798 sec 103 of 205 A&,ý6 1P. -&1D Diablo Canyon Power Plant Implementation of Alternative Source Terms Table 7.2-1 Loss of Coolant Accident Assumptions & Key Parameter Values Parameter Value Maximum delay between end of injection spray and 12 min (based on manual operator action)initiation of recirculation spray Containment spray in recirculation mode Initiation time 4518 sec Termination time 22,518 sec Long-term Sump Water pH >_ 7.5 Maximum allowable DF for fission product removal Elemental Iodine: 200 Others: not applicable Elemental iodine and particulate spray removal coefficients See Table 7.2-2 in sprayed region during both injection spray and recirculation spray modes Elemental iodine removal coefficients due to wall deposition See Table 7.2-2 Particulate removal coefficients in unsprayed region due to See Table 7.2-2 gravitational settling Containment Leak rate (0-24 hr) 0.1% weight fraction per day Containment Leak rate (1-30 day) 0.05% weight fraction per day Containment Leakage Release Point (Unfiltered) From the worst case release point of the following: Diffuse source via the containment wall Via Plant Vent Via Containment Pen Area GE Via Containment Pen Areas GW & FW ESF System Environmental Leakage Parameters Minimum post-LOCA containment water volume sources 480,015 gal.Minimum time after LOCA when recirculation is initiated 829 sec Duration of leakage 30 days Maximum ECCS fluid temperature after initiation of 259.9 'F recirculation Maximum ECCS leak rate (including safety factor of 2) Unfiltered via plant vent = 240 cc/min Unfiltered via Containment Penetration Areas GE or GW & FW = 12 cc/min RHR pump seal failure Filtered') via plant vent 50 gpm starting at t =24 hrs for 30 min Iodine Airborne Release Fraction 10%Auxiliary Building ESF Ventilation System filter efficiency Elemental iodine: 88%Organic iodine: 88%104 of 205 Ah Diablo Canyon Power Plant Implementation ofAlternative Source Terms Table 7.2-1 Loss of Coolant Accident Assumptions & Key Parameter Values Parameter I I Value Refueling Water Storage Tank (RWST) Back-Leakage Parameters Earliest initiation time of RWST back-leakage 829 sec Maximum ECCS I sump water back-leakage rate to RWST 2 gpm (includes safety factor of 2)RWST back-leakage iodine release fractions See Table 7.2-3 RWST back-leakage noble gas, as iodine daughters, See Table 7.2-3 release rate from the RWST vent Miscellaneous Equipment Drain Tank (MEDT) Leakage Parameters MEDT inflow rate (includes safety factor of 2) 1900 cc/min MEDT leakage-Iodine release fractions See Table 7.2-4 MEDT leakage noble gas, as iodine daughters release rate See Table 7.2-4 from plant vent CR Emergency Ventilation: Initiation Signal/Timing Initiation time (signal) SI signal generated: 6 sec Non-Affected Unit NOP Intake Isolated: 18 sec Affected Unit NOP Intake Isolated and CRVS Mode 4 in full Operation: 44.2 sec Bounding Control Room Atmospheric Dispersion Table 7.2-5 Factors for LOCA Note: (1) Releases from the RHR Pump Seal failure are filtered for CR dose evaluation and filtered for Site Boundary Dose Evaluation 105 of 205 Ah Diablo Canyon Power Plant Implementation ofAlternative Source Terms Table 7.2-2 Loss of Coolant Accident Total Elemental Iodine & Particulate Removal Coefficients Elemental Iodine Removal Coefficient Particulate Removal Coefficient (hr 1) -Note 1 (hr" 1)From Time To Time Sprayed Unsprayed Sprayed Unsprayed (sec) (sec) Region Region Region Region 0 30 NIANA 2.74 2.74 NIA 30 111 5.89 0.0062 111 1,800 2.24 0.0071 1,800 3,798 2057 (Note 2) 9.35 0.1144 3,798 4,518 0.00 (Note 3) 1.02 0.1229 4,518 5,030 7.50 0.1239 5,030 6,480 6.40 0.1237 6,480 7,200 19.91 (Note 2) 0.00 4.74 0.1236 7,200 8,004 3.39 0.1222 8,004 22,152 1.53 0.1040 22,152 22,518 0.00 22,518 720 hrs 0.00 0.00 (Note 4)Notes: 1. Per RG 1.183 and SRP 6.5.2, removal credit for elemental iodine by sprays is eliminated after a DF=200 is reached in the containment atmosphere.

2. Wall deposition removal coefficient (0.57 hr-) is included.3. Time period without spray.4. For purposes of conservatism, no credit is taken for particulate removal in the sprayed region after termination of recirculation spray 106 of 205 Ab Diablo Canyon Power Plant Implementation ofAlternative Source Terms Figure 7.2-1 Aerosol Removal Rates Within Sprayed Region Post-LOCA Containment Particle Removal Rate Profile in Sprayed Volume 12 --Diffusiophoresis Patti j8.0 ..- -Sprays Particle Remoa-Total Particle Remov'6.0J I 4.0 2.0'0.0 0 5,000 10,O00 15,0DO cle Removal val 20,000 post-LOCA time (sec)107 of 205 I.Diablo Canyon Power Plant Implementation ofAlternative Source Terms Figure 7.2-2 Aerosol Removal Rates Within Unsprayed Region Post-LOCA Containment Particle Removal Rate Profile in Unsprayed Volume 0.14 0.10 oJo-G.08-Sedimentation Particle Removal E CIJ U 0,.0.04 0 5,000 10,000 15,000 20,ODO post-LOCA time (sec)108 of 205 9,63 Diablo Canyon Power Plant Implementation of Alternative Source Terms Table 7.2-2A Containment Pressure/Temperaturel Relative Humidity Data -LOCA Post-LOCA Time Containment Containment Containment RH Pressure Temperature Seconds psia °F %0.00 16.00 120 18 0.52 18.81 145.9 72.3 1.04 21.45 163.8 88.2 1.54 23.76 176.78 93.5 2.04 25.88 186.8 96 2.54 27.80 194.83 97.4 3.04 29.44 201.03 98.1 3.54 30.86 205.97 98.5 4.04 32.10 210.01 98.9 4.54 33.21 213.43 99.1 5.04 34.24 216.47 99.2 7.04 38.04 226.75 99.5 7.54 38.96 229.03 99.5 8.54 40.68 233.06 99.6 10.04 43.06 238.34 99.7 10.54 43.81 239.9 99.7 11.54 45.21 242.76 99.8 13.04 47.12 246.47 99.8 14.54 48.82 249.65 99.8 16.04 50.36 252.4 99,9 17.54 51.70 254.7 99.9 19.04 52.87 256.67 99.9 20.54 53.88 258.34 99.9 21.54 54.24 258.9 99.9 22.04 54.36 259.08 100 23.54 54.48 259.25 100 25.04 54.40 259.12 100 29.54 53.87 258.24 100 32.54 53.62 257.85 100 48.54 53.14 257.02 100 54.54 53.37 257.03 100 68.04 53.70 256.79 100 86.54 53.18 255.93 100 144.18 50.88 251.95 100 158.18 50.44 251.15 100 109 of 205 Ab Diablo Canyon Power Plant Implementation ofAlternative Source Terms Table 7.2-2A Containment Pressure/Temperature/

Relative Humidity Data -LOCA Post-LOCA Time Containment Containment Containment RH Pressure Temperature Seconds psia -F %188.18 49.70 249.82 100 200.18 49.48 249.41 100 212.18 49.33 249.15 100 266.18 49.21 248.92 100 333.18 49.37 249.2 100 400.18 49.70 249.82 99.9 534.18 50.56 251.76 99.1 668.18 51.49 254.34 97.4 803.18 52.60 256.32 97.3 816.18 52.43 255.22 98.7 857.18 52.08 254.17 99.5 912.18 51.73 253.6 99.4 1021.19 51.21 252.76 99.3 1131.19 50.83 252.11 99.2 1240.19 50.55 251.61 99.1 1458.19 50.16 250.91 99 1677.19 49.94 250.49 98.9 1730.19 50.43 251.61 98.5 1746.19 50.26 250.56 99.9 1859.19 49.41 248.91 100 1988.19 48.57 247.32 100 2247.19 47.07 244.4 100 2505.19 45.75 241.75 100 2764.19 44.56 239.29 100 3022.19 43.47 236.96 99.9 3281.19 42.45 234.71 99.8 3604.24 41.26 231.94 99.9 3798.24 40.10 229.1 100 3888.29 40.52 230.99 98.2 3978.29 40.92 233.84 94.6 4068.29 41.24 235.73 92.5 4158.29 41.48 237.00 91.3 4338.29 41.86 238.43 90.3 4518.29 42.13 239.07 90.3 4536.29 42.05 237.81 92.2 110 of 205 M6 Diablo Canyon Power Plant Implementation of Alternative Source Terms Table 7.2-2A Containment Pressure/Temperature/ Relative Humidity Data -LOCA Post-LOCA Time Containment Containment Containment RH Pressure Temperature Seconds psia °F %4555.29 41.94 235.91 95.2 4573.29 41.87 234.69 97.2 4592.29 41.82 233.97 98.4 4666.29 41.74 233.23 99.5 5110.73 41.37 232.3 99.6 5700.73 40.79 230.94 99.6 6890.73 39.55 227.94 99.6 8080.73 38.36 224.96 99.5 10000.80 36.64 220.36 99.4 11001.50 35.82 218.08 99.4 12001.50 35.08 215.93 99.4 13001.50 34.39 213.85 99.4 14001.50 33.74 211.85 99.4 15001.50 33.13 209.91 99.4 16001.50 32.57 208.07 99.3 18001.50 31.59 204.69 99.3 20001.50 30.71 201.53 99.3 21001.50 30.31 200.06 99.3 22518.00 29.78 198.01 99.4 111 of 205 Diablo Canyon Power Plant Implementation ofAlternative Source Terms Table 7.2-2B Containment Steam Condensation Data -Loss of Coolant Accident Post-LOCA Steam Condensation Rate Time Thermal Containment Injection Recirculation Total Steam Condensation Rate Conductor Fan Coolers Spray Spray Seconds Ibm/sec Ibm/sec Ibm/sec Ibm/sec Ibm/sec g/sec 0.00 0.00 0.00 0.00 0.00 0.00 0.00 0.52 8.37 0.00 0.00 0.00 8.37 3796.57 1.04 49.71 0.00 0.00 0.00 49.71 22548.08 2.54 204.71 0.00 0.00 0.00 204.71 92854.90 3.04 250.47 0.00 0.00 0.00 250.47 113611.29 3.54 290.16 0.00 0.00 0.00 290.16 131614.37 4.04 325.28 0.00 0.00 0.00 325.28 147544.54 5.04 384.15 0.00 0.00 0.00 384.15 174247.52 7.54 509.97 0.00 0.00 0.00 509.97 231318.52 10.04 611.01 0.00 0.00 0.00 611.01 277149.49 12.54 684.86 0.00 0.00 0.00 684.86 310647.29 15.04 734.83 0.00 0.00 0.00 734.83 333313.30 17.54 766.16 0.00 0.00 0.00 766.16 347524.35 20.54 783.18 0.00 0.00 0.00 783.18 355244.50 24.54 742.58 0.00 0.00 0.00 742.58 336828.64 28.54 684.40 0.00 0.00 0.00 684.40 310438.64 32.54 644.28 0.00 0.00 0.00 644.28 292240.51 37.04 623.09 0.00 0.00 0.00 623.09 282628.89 53.54 538.96 0.00 0.00 0.00 538.96 244468.16 70.04 469.30 0.00 0.00 0.00 469.30 212870.91 87.04 412.48 0.00 0.00 0.00 412.48 187097.79 87.57 410.75 44.28 0.00 0.00 455.03 206398.15 88.07 409.15 45.29 0.00 0.00 454.44 206130.53 107.14 354.70 44.87 13.61 0.00 413.18 187415.31 124.18 312.19 44.28 72.76 0.00 429.23 194695.47 146.18 267.97 43.51 70.37 0.00 381.85 173204.26 169.18 231.50 42.78 68.46 0.00 342.74 155464.26 197.18 197.26 41.97 66.66 0.00 305.89 138749.38 234.18 166.95 41.32 49.25 0.00 257.52 116809.11 262.18 149.95 41.00 48.48 0.00 239.43 108603.63 327.18 121.98 40.50 47.28 0.00 209.76 95145.54 403.18 100.43 40.23 46.58 0.00 187.24 84930.64 449.18 91.02 40.17 46.66 0.00 177.85 80671.41 502.18 82.32 40.11 47.29 0.00 169.72 76983.70 112 of 205 Diablo Canyon Power Plant Implementation of Alternative Source Terms Table 7.2-2B Containment Steam Condensation Data -Loss of CoolantAccident Post-LOCA Steam Condensation Rate Time Thermal Containment Injection Recirculation Total Steam Condensation Rate Conductor Fan Coolers Spray Spray Seconds Ibm/sec Ibm/sec Ibm/sec Ibm/sec Ibm/sec g/sec 558.18 74.79 40.09 48.64 0.00 163.52 74171.43 560.18 74.55 40.08 48.66 0.00 163.29 74067.10 629.18 67.45 40.12 47.87 0.00 155.44 70506.40 683.18 63.43 40.22 47.23 0.00 150.88 68438.02 754.18 59.41 40.45 47.01 0.00 146.87 66619.12 802.18 57.28 40.64 46.99 0.00 144.91 65730.07 832.18 49.92 40.37 57.62 0.00 147.91 67090.85 876.18 43.40 40.00 58.41 0.00 141.81 64323.94 937.18 37.29 39.57 57.66 0.00 134.52 61017.25 1013.19 32.26 39.16 56.94 0.00 128.36 58223.12 1094.19 28.58 38.82 56.39 0.00 123.79 56150.20 1148.19 26.76 38.64 56.09 0.00 121.49 55106.94 1243.19 24.16 38.39 55.68 0.00 118.23 53628.23 1341.19 22.03 38.19 55.34 0.00 115.56 52417.14 1423.19 20.55 38.07 55.11 0.00 113.73 51587.06 1492.19 19.48 37.98 54.94 0.00 112.40 50983.79 1564.19 18.50 37.90 54.79 0.00 111.19 50434.94 1607.19 17.97 37.86 54.72 0.00 110.55 50144.64 1644.19 17.54 37.83 54.66 0.00 110.03 49908.77 1672.19 17.23 37.81 54.61 0.00 109.65 49736.41 1678.19 17.19 37.71 54.73 0.00 109.63 49727.33 1730.19 19.93 36.37 52.72 0.00 109.02 49450.64 1794.19 16.03 35.05 60.34 0.00 111.42 50539.27 1859.19 14.01 33.89 60.13 0.00 108.03 49001.59 1985.19 11.88 32.19 59.74 0.00 103.81 47087.43 2052.19 11.05 31.50 59.53 0.00 102.08 46302.71 2116.19 10.33 30.96 59.34 0.00 100.63 45645.00 2244.19 9.09 30.09 58.96 0.00 98.14 44515.56 2311.19 8.50 29.72 58.76 0.00 96.98 43989.39 2439.19 7.47 29.12 58.40 0.00 94.99 43086.74 2567.19. 6.53 28.62 57.93 0.00 93.08 42220.38 2695.19 5.66 28.19 57.37 0.00 91.22 41376.70 2763.19 5.23 27.94 57.08 0.00 90.25 40936.71 2890.19 4.47 27.51 56.53 0.00 88.51 40147.46 113 df 205 Ab Diablo Canyon Power Plant Implementation ofAlternative Source Terms Table 7.2-2B Containment Steam Condensation Data -Loss of Coolant Accident Post-LOCA Steam Condensation Rate Time Thermal Containment Injection Recirculation Total Steam Condensation Rate Conductor Fan Coolers Spray Spray Seconds Ibm/sec Ibm/sec Ibm/sec Ibm/sec Ibm/sec g/sec 3018.19 3.76 27.11 56.00 0.00 86.87 39403.57 3082.19 3.43 26.92 55.74 0.00 86.09 39049.77 3210.19 2.80 26.56 55.23 0.00 84.59 38369.38 3338.19 2.21 26.22 54.74 0.00 83.17 37725.28 3466.19 1.64 25.89 54.25 0.00 81.78 37094.79 3594.19 1.12 25.58 53.78 0.00 80.48 36505.12 3722.24 0.24 25.02 54.13 0.00 79.39 36010.70 3796.24 0.02 24.72 53.61 0.00 78.35 35538.96 3843.29 2.24 24.91 0.00 0.00 27.15 12315.03 3901.29 4.39 25.13 0.00 0.00 29.52 13390.05 3995.29 6.72 25,42 0.00 0.00 32.14 14578.46 4105.29 8.33 25.70 0.00 0.00 34.03 15435.75 4189.29 9.21 25.87 0.00 0.00 35.08 15912.02 4291.29 10.03 26.06 0.00 0.00 36.09 16370.15 4383.29 10.63 26.21 0.00 0.00 36.84 16710.34 4463.29 11.06 26.33 0.00 0.00 37.39 16959.82 4515.29 11.25 26.41 0.00 0.00 37.66 17082.29 4518.29 11.26 26.41 0.00 0.92 38.59 17504.13 4584.29 10.17 26.48 0.00 8.26 44.91 20370.83 4592.29 10.15 26.49 0.00 9.02 45.66 20711.03 4654.29 10.18 26.55 0.00 11.12 47.85 21704.40 4698.29 10.19 26.59 0.00 11.33 48.11 21822.33 4734.29 10.21 26.60 0.00 11.39 48.20 21863.15 4785.29 10.19 26.62 0.00 11.43 48.24 21881.30 4807.29 10.17 26.63 0.00 11.44 48.24 21881.30 4843.29 10.14 26.63 0.00 11.46 48.23 21876.76 4851.29 10.13 26.64 0.00 11.46 48.23 21876.76 4895.29 10.09 26.64 0.00 11.48 48.21 21867.69 4926.29 10.05 26.64 0.00 11.49 48.18 21854.08 4932.29 10.06 26.63 0.00 11.49 48.18 21854.08 4988.29 9.99 26.63 0.00 11.51 48.13 21831.40 6120.73 8.70 25.98 0.00 11.43 46.11 20915.15 7400.73 7.52 25.02 0.00 11.13 43.67 19808.38 8680.73 6.51 24.04 0.00 10.86 41.41 18783.26 114 of 205 -A-6 Diablo Canyon Power Plant Implementation ofAlternative Source Terms Table 7.2-26 Containment Steam Condensation Data -Loss of Coolant Accident Post-LOCA Steam Condensation Rate Time Thermal Containment Injection Recirculation Total Steam Condensation Rate Conductor Fan Coolers Spray Spray Seconds Ibm/sec Ibm/sec Ibm/sec Ibm/sec Ibm/sec g/sec 9510.73 6.06 23.47 0.00 10.70 40.23 18248.02 14001.50 4.42 20.91 0.00 9.98 35.31 16016.35 18001.50 3.69 18.95 0.00 9.59 32.23 14619.28 22518.00 3.23 17.41 0.00 9.27 29.91 13566.95 115 of 205 &3 Diablo Canyon Power Plant Implementation ofAlternative Source Terms Table 7.2-3 Loss of Coolant Accident RWST Iodine Release Fraction and Gas Venting Rate to Atmosphere Average Interval Time To Time Iodine Release Fraction Weighted Gas Space to Atmosphere Venting Rate to Atmosphere Sec Sec Fraction lrleased/ lentering Fraction Vr.st / day 829 7200 9.451E-05 2.610E+00 7200 28,800 6.357E-05 7.291 E-01 28,800 86,400 8.796E-06 7.375E-02 86,400 345,600 4.560E-07 9.955E-03 345,600 471,600 6.347E-07 1.311E-02 471,600 1,011,600 8.231 E-07 1.489E-02 1,011,600 2,048,400 1.114E-06 1.547E-02 2,048,400 2,592,000 1.483E-06 1.702E-02 Where: Ireleased = Total Iodine mass released to atmosphere during specified time interval, gm Ientering = Total Iodine mass entering to the RWST during specified time interval, gm Frac. Vrt = Rate of Fractional RWST gas volume vented during specified time interval 116 of 205 a al.b Diablo Canyon Power Plant Implementation ofAlternative Source Terms Table 7.2-4 Loss of Coolant Accident MEDT Iodine Release Fraction and Gaseous Venting Rate to Atmosphere Iodine Release Fraction to Average Interval Weighted From Time To Time Atmosphere Gas Space Venting Rate to Atmosphere Sec Sec Fraction Ireleased/ ]entering Fraction VMEDT / day 829 7,200 4.521 E-07 5.024E+00 7,200 28,800 1.386E-08 3.024E-02 28,800 86,400 2.362E-07 3.324E-01 86,400 183,289 3.950E-07 6.497E+00 183,289 345,600 1.236E-02 (Note 2) (Note 1)345,600 752,400 2.028E-02 (Note 2) (Note 1)752,400 1,530,000 2.390E-02 (Note 2) (Note 1)1,530,000 2,592,000 2.166E-02 (Note 2) (Note 1)Where:[released = Total Iodine mass released to atmosphere during specified time interval, gm lentering -Total Iodine mass entering to the MEDT during specified time interval, gm Frac VMEDT = Rate of Fractional MEDT gas volume vented during specified time interval Note 1: After the MEDT overflows at t = 183,289 sec, the gas venting rates are 2640 cfm from the EDRT room, and 1760 cfm from the UI/U2 Pipe Tunnels (i.e., the exhaust ventilation rate from the respective rooms + 10%). To be consistent with the methodology used to determine the iodine release fractions after spillover, the noble gases generated by decay of iodines in the tank and spilled liquid after overflow occurs, should also be released instantaneously to the environment without hold-up.Note 2: The room ventilation flows addressed in Note 1 (utilized as clean in-coming air) are incorporated into the determination of the iodine equilibrium concentration in the EDRT room and U1/U2 Pipe Tunnels air space, respectively. The bounding iodine release fractions presented above after spillover assume instantaneous release of iodines to the environment without hold-up in the room.117 of 205 A 6 Diablo Canyon Power Plant Implementation ofAlternative Source Terms Table 7.2-5 Loss of Coolant Accident Control Room Limiting Atmospheric Dispersion Factors (sec/m 3)Release Location I Receptor 0-2 hr 2-8 hr 8-24 hr 24-96 hr 96-720 hr Control Room Normal Intakes Plant Vent Release-Affected Unit Intake 1.67E-03-Non-Affected Unit Intake 9.10E-04 Containment Penetration Areas-Affected U nit Intake 6.84E-03 -----...... -Non-Affected Unit Intake 2.24E-03 .................... Control Room Infiltration Plant Vent 1.26E-03 8.96E-04 3.44E-04 3.44E-04 2.99E-04 Containment Penetration Areas 3.22E-03 1.85E-03 7.29E-04 7.15E-04 6.64E-04 RWST Vent 1.07E-03 5.80E-04 2.18E-04 2.19E-04 1.79E-04 Control Room Pressurization Intake Plant Vent 5.65E-05 3.70E-05 1.35E-05 1.37E-05 1.11 E-05 Containment Penetration Areas 6.45E-05 4.05E-05 1.65E-05 1.38E-05 1.1 2E-05 RWST Vent 5.25E-05 3.03E-05 1.15E-05 1.10E-05 8.83E-06 Note 1: Release from the Containment penetration areas (i.e., areas GE or GW & FVV): applicable to containment leakage and ESF system leakage that occurs in the Containment Penetration Area Note 2: Release from Plant Vent: applicable to ESF system leakage that occurs in the Auxiliary building, MEDT releases, RHR Pump Seal Failure Release and Containment Vacuum/Pressure Relief Line Release 118 of 205 A h Diablo Canyon Power Plant Implementation of Alternative Source Terms Table 7.2-6 Loss of Coolant Accident TSC Limiting Atmospheric Dispersion Factors (seclm 3)Release Location / Receptor 0-2 hr 2-8 hr 8-24 hr 24-96 hr 96-720 hr TSC Normal Intakes Plant Vent Release 5.52E-04 ............... C o n t a i n m e n t P e n e t r a t i o n A r e a s 1 .8 0 E -0 3 ...... ..... ....RWST Vent 3.63E-04 .............. TSC Infiltration Plant Vent 5.43E-04 2.16E-04 9.97E-05 8.11E-05 6.58E-05 Containment Penetration Areas 1.83E-03 7.49E-04 3.16E-04 2.92E-04 2.41 E-04 RWST Vent 3.72E-04 1.68E-04 6.64E-05 6.17E-05 5.10E-05 CRITSC Pressurization Intake Plant Vent -3.70E-05 1.35E-05 1.37E-05 1.11 E-05 Containment Penetration Areas ----- 4.05E-05 1.65E-05 1.38E-05 1.12E-05 RWST Vent .3.03E-05 1.15E-05 1.10E-05 8.83E-06 Note 1: Release from the Containment penetration areas (i.e., areas GE or GW & FW): applicable to containment leakage and ESF system leakage that occurs in the Containment Penetration Area Note 2: Release from Plant Vent: applicable to ESF system leakage that occurs in the Auxiliary building, MEDT releases, RHR Pump Seal Failure Release and Containment Vacuum/Pressure Relief Line Release 119 of 205 Diablo Canyon Power Plant Implementation ofAlternative Source Terms Table 7.3-1 Fuel Handling Accident in Fuel Handling Building or Containment Analysis Assumptions & Key Parameter Values Parameter Value Power Level 3580 MWt Number of Damaged Fuel Assemblies 1 Total Number of Fuel Assemblies 264 Decay Time Prior to Fuel Movement 72 hours Radial Peaking Factor 1.65 Fraction of Core Inventory in gap 1-131 (12%)Kr-85 (30%)Other Noble Gases (10%)-Other Halides (10%)Alkali Metals (17%)Isotopic Inventory in Fuel Gap (Decayed 72 hours) Table 7.3-2 Iodine form of gap release before scrubbing 99.85% elemental 0.15% Organic Iodine form of gap release after scrubbing 57% elemental 43% Organic Scrubbing Decontamination Factors Iodine (200, effective) Noble Gas (1)Particulates (co)Rate of Release from Fuel Puff Environmental Release Rate All airborne activity released within a 2 hour period (or less if the ventilation system promotes a faster release rate)Environmental Release Points and Rates Accident in SFP in the FHB -Release flow rates -Plant Vent -46,000 cfm FHB Outleakage-Ingress/Egress locations -30 cfm-Miscellaneous gaps/openings -470 cfm Minimum free volume in FHB above SFP 317,000 ftW Accident in Containment -Release flow rates -Open Equipment Hatch -All airborne activity released in 2 hrs Minimum Free Volume in Containment above Operating 2,013,000 ft3 Floor CR Emergency Ventilation: Initiation Signal/Timing Signal(s) available to switch the CRVS from normal Radiation signals from gamma sensitive operation (NOP) Ventilation (Mode 1) to Pressurized Filtered intake monitors that initiate closure of the Ventilation (Mode 4) following a FHA CR normal intake dampers and switch the CRVS from normal operation Ventilation Mode 1 to Pressurized Filtered Ventilation Mode 4.120 of 205 'p Diablo Canyon Power Plant Implementation ofAlternative Source Terms Table 7.3-1 Fuel Handling Accident in Fuel Handling Building or Containment Analysis Assumptions & Key Parameter Values Parameter Value Radiation Monitor Analytical Safety Limit 1 mR/hr Delay time for CRVS Mode 4 operation, including monitor 22 seconds (see below)response, signal processing, and damper closure time Radiation Monitor Response Time 10 seconds (conservative assumption) -(Refer to Section 7.3)Radiation monitor signal processing time 2 seconds NOP Ventilation Damper Closure Time 10 seconds Bounding Control Room Atmospheric Dispersion Table 7.3-3 Factors for FHA 121 of 205 P Diablo Canyon Power Plant Implementation ofAlternative Source Terms Table 7.3-2 Isotopic Gap Activity -Fuel Handling Accident Single Fuel Assembly (Decayed 72 hours)Activity Gap Fraction Gap Activity per Per Assembly Assembly (w/o Peaking Factor)Nuclide (Ci)1-129 2.07E-02 0.10 2.07E-03 1-130 3.29E+02 0.10 3.29E+01 1-131 4.09E+05 0.12 4.91 E+04 1-132 3.99E+05 0.10 3.99E+04 1-133 9.73E+04 0.10 9.73E+03 1-135 5.01E+02 0.10 5.01 E+01 KR-83M 2.51E-04 0.10 2.51 E-05 KR-85 5.75E+03 0.30 1.73E+03 KR-85M 1.77E+00 0.10 1.77E-01 KR-88 7.73E-03 0.10 7.73E-04 XE-127 9.64E-02 0.10 9.64E-03 XE-129M 5.28E+01 0.10 5.28E+00 XE-131M 6.96E+03 0.10 6.96E+02 XE-133 8.31E+05 0.10 8.31 E+04 XE-133M 1.88E+04 0.10 1.88E+03 XE-135 1.07E+04 0.10 1.07E+03 XE-135M 8.18E+01 0.10 8.18E+00 CS-132 2.16E+01 0.17 3.67E+00 CS-134 1.25E+05 0.17 2.13E+04 CS-134M 1.04E-03 0.17 1.77E-04 CS-135 3.01E-01 0.17 5.12E-02 CS-136 3.1OE+04 0.17 5.27E+03 0S-137 7.1OE+04 0.17 1.21E+04 RB-86 1.16E+03 0.17 1.97E+02 RB-87 1.37E-05 0.17 2.33E-06 RB-88 8.63E-03 0.17 1.47E-03 122 of 205 a 6t Diablo Canyon Power Plant Implementation of Alternative Source Terms Table 7.3-3 Fuel Handling Accident Control Room Limiting Atmospheric Dispersion Factors (sec/m 3)Release Location / Receptor 0-22 sec 22 sec -2 hr 2-8 hr 8-24 hr 1-4 d 4-30 d Control Room Normal Intakes Containment Hatch Release-Affected Unit Intake 2.61 E-02 ............... ..... .....-Non-Affected Unit Intake 2.88E-03 -----......... ...Plant Vent Release-A ff e c t e d U n it I n t a k e 1 .6 7 E -0 3 -----. ... ... ... ..-Non-Affected Unit Intake 9.10E-04 FHB Out-leakage points-Affected Unit Intake 6.98E-03 ............... ..... .....-Non-Affected Unit Intake 2.93E-03 -----......... ..... ....Control Room Infiltration C on tainm en t H atch R ele a se 5 .5 1 E -03 5 .5 1 E ----............ Plant Vent 1.26E-03 1.26E-03 .......... ..... .....F H B O u t-le a k a g e p o in ts 3 .7 8 E -0 3 3 .7 8 E -0 3 -----............-Control Room Pressurization Intake Containm ent Hatch Release 6.60E-05 ........ ----- ....P la n t V e n t -5 .6 5 E -0 5 ...... .-..-. ....----FH B O ut-leakage points ----- 6.40E-05 -----............ Note 1: Release from the Containment Hatch: applicable to FHA in Containment Note 2: Release from Plant Vent / FHB Out-leakage: applicable to FHA in FHB 123 of 205 A 6 Diablo Canyon Power Plant Implementation ofAlternative Source Terms Table 7.4-1 Locked Rotor Accident Analysis Assumptions & Key Parameter Values Parameter Value Power Level 3580 MWt Reactor Coolant Mass 446,486 Ibm Primary to Secondary SG tube leakage 0.75 qpm (total for all 4 SGs); leakage density 62.4 Ibm/ft)Melted Fuel Percentage 0%Failed Fuel Percentage 10%Equilibrium Core Activity Table 4.1-1 Radial Peaking Factor 1.65 Fraction of Core Inventory in Fuel Gap 1-131: 12%Kr-85: 30%Other Noble Gases: 10%Other Halogens: 10%Alkali Metals: 17%Isotopic Inventory in Fuel Gap Table 4.3-1 Iodine Chemical Form in Gap 4.85% elemental 95% Particulate 0.15% organic Secondary Side Parameters Initial and Minimum SG Liquid Mass 92,301 lbm/SG Iodine Species Released to Environment 97% elemental; 3% organic Time period when tubes not totally submerged insignificant Steam Releases 0-2 hrs: 651,000 Ibm 2-8 hrs: 1,023,000 Ibm 8-10.73 hrs: same release rate as that for 2-8 hrs iodine Partition Coefficient in SGs 100 Particulate Carry-Over Fraction in SGs 0.0005 by weight Fraction of Noble Gas Released 1.0 (Released without holdup)Termination of releases from SGs 10.73 hours Environmental Release Point MSSVs/10% ADVs CR emergency Ventilation

Initiation SignallTiming Control Room is assumed to remain on normal ventilation (CRVS Mode 1) for duration of the accident.Control Room Atmospheric Dispersion Factors Table 7.4-2 124 of 205 96 Diablo Canyon Power Plant Implementation ofAlternative Source Terms Table 7.4-2 Locked Rotor Accident Control Room Limiting Atmospheric Dispersion Factors (sec/m 3)Release point and receptor 0-2hr 2-8 hr 8-10.73 hr MSSVs/10%

ADVs to CR NOP Intake (Note 1) 8.60E-04 5.58E-04 5.58E-04 MSSVs/1 0% ADVs to CR In-leakage (CR Centerline) 2.78E-03 1.63E-03 1.63E-03 Note 1: Due to the proximity of the release from the MSSVs/10% ADVs-to the normal operation CR intake of the affected unit, and due to the high vertical velocity of the steam discharge from the MSSVs/10% ADVs, the resultant plume from the MSSVs/1 0% ADVs will not contaminate the normal operation CR intake of the affected unit. Thus the X/Q s presented reflect those applicable to the CR intake of the unaffected unit.125 of 205 A6 Diablo Canyon Power Plant Implementation ofAlternative Source Terms Table 7.5-1 Control Rod Ejection Accident Analysis Assumptions & Key Parameter Values Parameters Value Containment Leakage Pathway Power Level 3580 MWt Free Volume 2.550E+06 Containment leak rate (0-24 hr) 0.1% vol. fraction per day Containment leak rate(I-30 day) 0.05% vol. fraction per day Failed Fuel Percentage 10%Percentage of Core Inventory in Fuel Gap 10% (noble gases & halogens)Melted Fuel Percentage 0%Chemical Form of Iodine in Failed fuel 4.85% elemental 95% particulate 0.15% organic Radial Peaking Factor 1.65 Core Activity Release Timing Puff Form of Failed Iodine in the Containment Atmosphere 97% elemental 3% organic Equilibrium Core Activity Table 4.1-1 Termination of Containment Release 30 days Environmental Release Point Same as LOCA Containment Leakage pathway Secondary Side Pathway Reactor Coolant Mass 446,486 Ibm Primary-to-Secondary Leak rate 0.75 qpm (total for all 4 SGs); leakage density 62.4 Ibm/ft Failed Fuel Percentage Same as containment leakage pathway Percentage of Core Inventory in Fuel Gap Same as containment leakage pathway Minimum Post-Accident SG Liquid Mass 92,301 Ibm / SG Iodine Species released to Environment 97% elemental 3% organic Time period when tubes not totally submerged Insignificant Steam Releases 0-2 hrs: 651,000 Ibm 2-8 hrs: 1,023,000 Ibm 8-10.73 hrs: same release rate as that for 2-8 hrs.Iodine Partition Coefficient in SGs 100 126 of 205 Ah Diablo Canyon Power Plant Implementation ofAlternative Source Terms Table 7.5-1 Control Rod Ejection Accident Analysis Assumptions & Key Parameter Values Parameters Value Fraction of Noble Gas Released 1.0 (Released without holdup)Termination of Release from SGs 10.73 hours Envircnmental Release Point MSSVs/10% ADVs CR emergency Ventilation: Initiation Signal/Timing Initiation time (signal) 300 sec (SIS Generated) 312 sec (Non-Affected Unit NOP Intake fully Closed)338.2 sec (Affected Unit NOP Intake fully Closed with full Mode 4 Emergency Ventilation Operation). Control Room Atmospheric Dispersion Factors Table 7.5-2 127 of 205 Ab Diablo Canyon Power Plant Implementation ofAlternative Source Terms Table 7.5-2 Control Rod Ejection Accident Control Room Limiting Atmospheric Dispersion Factors (sec/m 3)Release Location / Receptor 0-2 hr 2-8 hr 8-10.73 10.73-24 hr 24-96 hr 96-720 hr Control Room Normal Intakes Containment leakage-Affected Unit Intake 6.84E-03 ........ ........ --Non-Affected Unit Intake 2.24E-03 .................. MSSVs/1 0% ADVs-Affected Unit Intake Note 3 ......................... -Non-Affected Unit Intake 8.60E-04 -----.................. Control Room Infiltration Containment leakage 3.22E-03 1.85E-03 7.29E-04 7.29E-04 7.15E-04 6.64E-04 MSSVs/1 0% ADVs 2.78E-03 1.63E-03 1.63E-03 .......Control Room Pressurization Intake Containment leakage 6.45E-05 4.05E-05 1.65E-05 1.65E-05 1.38E-05 1.12E-05 MSSVs/10% ADVs 1.57E-05 9.60E-06 9.60E-06 ............... Note 1: Containment leakage: Used for Containment release scenario; based on Containment penetration area release point.Note 2: MSSV /10% ADVs: Used for Secondary System Release Scenario;Note 3: Due to the proximity of the release from the MSSVs/10% ADVs to the normal operation CR intake of the affected unit, and due to the high vertical velocity of the steam discharge from the MSSVs/1 0% ADVs, the resultant plume from the MSSVs/10% ADVs will not contaminate the normal operation CR intake of the affected unit.128 of 205 Diablo Canyon Power Plant Implementation ofAlternative Source Terms TABLE 7.6-1 Main Steam Line Break Analysis Assumptions & Key Parameter Values Parameter Value Power Level 3580 MWt Reactor Coolant Mass 446,486 Ibm Leak rate to Faulted Steam Generator 0.75 gpm (3 conservative assumption)

leakage density 62.4 Ibm/ft Leak rate to Intact Steam Generators 0 gpm (all leakage assumed into faulted SG)Failed/Melted Fuel Percentage 0%RCS Tech Spec Iodine Conc. Table 4.2-1 (1 pCi/gm DE 1-131)RCS Tech Spec Noble Gas Conc. Table 4.2-1 (270 pCi/gm DE Xe-133)RCS Equilibrium.

Iodine Appearance Rates Table 4.2-2 (1 pCi/gm DE 1-131)Pre-Accident Iodine Spike Concentrations Table 4.2-2 (60 pCi/gm DE 1-131)Accident-Initiated Iodine Spike Appearance Rate 500 times equilibrium appearance rate Duration of Accident-Initiated Iodine Spike 8 hours Initial Secondary Coolant Iodine Concentrations Table 4.2-1 1-131)Secondary System Release Parameters Iodine Species released to Environment 97% elemental; 3% organic Fraction of Iodine Released form Faulted SG 1.0 (Released to Environ without holdup)Fraction of Noble Gas Released from Faulted SG 1.0 (Released to Environ without holdup)Liquid mass in each SG Faulted: 182,544 Ibm (max.)Intact: 92,301 Ibm (min. and initial)Release Rate of SG liquid activity from Faulted SG Dryout within1 0 seconds Time period when tubes not totally submerged (intact SG) Insignificant Steam Releases from intact SGs 0-2 hrs: 384,000 Ibm 2-8 hrs: 893,000 Ibm 8-10.73 hrs: Same release rate as that for 2-8 hrs Iodine Partition Coefficient in Intact SG 100 (SGs fully covered)Termination of release (0.75 gpm leak): Faulted SG 30 hrs when RCS reaches 212 'F Termination of release from Intact SG 10.73 hours Release Point: Faulted SG Outside containment, at the steam line break location 129 of 205 MAI Diablo Canyon Power Plant Implementation ofAlternative Source Terms TABLE 7.6-1 Main Steam Line Break Analysis Assumptions & Key Parameter Values Parameter Value Release Point: Intact SG MSSVs/1 0% ADVs CR Emergency Ventilation

Initiation Signal/Timing Initiation (signal) SIS Unaffected Unit CRVS inlet damper fully closed Within 12.6 seconds Affected Unit CRVS inlet dampers fully closed Within 38.8 seconds Control Room Atmospheric Dispersion Factors Table 7.6-2 Table 7.6-2 Main Steam Line Break Accident Control Room Limiting Atmospheric Dispersion Factors (sec/m 3)Receptor -Release Point 0-2hr 2-8 hr 8-10.73 hr 10.73-30 hr CR NOP Intake -Faulted SG (Break Location)

Note 1 CR NOP Intake -Intact SG (MSSVs/10% ADVs) -Note 2 8.60E-04 CR Inleakage -Faulted SG (Break Location) 1.24E-02 7.35E-03 3.01E-03 3.01 E-03 CR Inleakage -Intact SG (MSSVs/10% ADVs) 2.78E-03 1.63E-03 1.63E-03 -----CR Emergency Intake & Bypass -Faulted SG (Break 7.65E-05 4.78E-05 1.86E-05 1.86E-05 Location) I CR Emergency Intake & Bypass -Intact SG (MSSVs/1 0% 1.57E-05 9.60E-06 9.60E-06 ADVs)Notes: 1. ARCON96 based X/Q s are not applicable for these cases given that the horizontal distance from the source to the receptor is 1.5 meters (which is much less than the 10 meters required by ARCON96 methodology).

2. Due to the proximity of the release from the MSSVs/10%

ADVs to the normal operation CR intake of the affected unit, and due to the high vertical velocity of the steam discharge from the MSSVs/10% ADVs, the resultant plume from the MSSVs/10% ADVs will not contaminate the normal operation CR intake of the affected unit. Thus the X/Q s presented reflect those applicable to the CR intake of the unaffected unit.130 of 205 Ab Diablo Canyon Power Plant Implementation ofAlternative Source Terms Table 7.7-1 Steam Generator Tube Rupture Analysis Assumptions & Key Parameter Values Parameter Value Power Level 3580 MWt Reactor Coolant Mass 446,486 Ibm Time of Reactor Trip 179.0 sec Time of isolation of stuck-open 10% ADV on the Ruptured 2653 sec SG Termination of Break Flow from Ruptured SG that flashes 3402 sec Termination of Break Flow from Ruptured SG 5872 sec Time of manual depressurization of the Ruptured SG 2 hours Break Flow to Ruptured Steam Generator that flashes Table 7.7-2, Column "A" Break Flow to Ruptured Steam Generator that does not Table 7.7-2, Column "B" flash Tube Leakage rate to Intact Steam Generators 0.75 gpm (total for all 4 SGs; conservatively assumed for 3 intact SGs); leakage density 62.4 Ibm/ft 3 Failed/Melted Fuel Percentage 0%RCS Tech Spec Iodine Concentration 1 pCi/gm DE 1-131 (Table 4.2-1)RCS Tech Spec Noble Gas Concentration 270 pCi/gm DE Xe-133 (Table 4.2-1)RCS Equilibrium Iodine Appearance Rates Table 4.2-2 (1 pCi/gm DE 1-131)Pre-Accident Iodine Spike Concentration 60 pCi/gm DE 1-131 (Table 4.2-2)Accident-Initiated Iodine Spike Appearance Rate 335 times TS equilibrium appearance rate Duration of Accident-initiated Iodine Spike 8 hours Initial Secondary Coolant Iodine Concentrations 0.1 pCi/gm DE 1-131 (Table 4.2-1)Secondary System Release Parameters Initial SG liquid mass 89,707 Ibm /SG Iodine Species released to Environment 97% elemental; 3% organic Steam flow rate to condenser from Ruptured SG before trip 63,000 Ibm/min Steam flow rate to condenser from intact SGs before trip 189,000 Ibm/min Partition Factor in Main Condenser 0.01 (elemental iodine)1 (organic iodine and noble gases)Steam Releases from Ruptured SG Table 7.7-2, Column "C" Steam Releases from intact SG Table 7.7-2, Column "D" Post-accident minimum SG liquid mass for Ruptured SG 89,707 Ibm Post-accident minimum SG liquid mass for intact SGs 89,707 Ibm per SG 131 of 205 &Z Diablo Canyon Power Plant Implementation ofAlternative Source Terms Table 7.7-1 Steam Generator Tube Rupture Analysis Assumptions & Key Parameter Values Parameter Value Time period when tubes not totally submerged (intact SG) insignificant Fraction of Iodine Released (flashed portion) 1.0 (Released without holdup)Fraction of Noble Gas Released from all SGs 1.0 (Released without holdup)Iodine Partition Coefficient 100 Termination of Release from intact SG 10.73 hrs Environmental Release Points Plant Vent: 0 -179 sec MSSVs/10% ADVs:179 sec- 10.73 hr CR emergency Ventilation: Initiation Signal/Timing Initiation time (signal) SIS: 219 sec Unaffected Unit inlet damper closed: 231 sec Affected Unit inlet damper closed: 257.2 sec Control Room Atmospheric Dispersion Factors Table 7.7-3 132 of 205 iF Diablo Canyon Power Plant Implementation ofAlternative Source Terms Table 7.7-2 Steam Generator Tube Rupture Break Flows and Steam Releases Break Flow and Steam Release within each Time Interval A B C D Time Flashed Un-flashed Ruptured SG Intact SGs from Break Break Flow Break Flow Steam Releases Steam Releases (sec) (Ibm) (Ibm) (Ibm) (Ibm)0 1678 8422 187822 563100 179 2217 30003 10527 42565 853 12121 90754 113657 118 2653 1355 15906 0 146 2953 779 23177 0 85467 3402 0 45026 0 97164 4324 0 16870 0 9237 4739 0 23892 0 29103 5872 0 0 0 103300 7200 0 0 270000 1,342,400 38628 0 0 0 0 Note: Data in row for T=0 is applicable to time interval between T=0 sec to T=1 79 sec (typ)TABLE 7.7-3 Steam Generator Tube Rupture Accident Control Room Limiting Atmospheric Dispersion Factors (sec/m 3)Release Location I Receptor 0-179s 179-257.2 s 257.2 s- 2 h 2-8 hr 8-10.73 hr Control Room Normal Intakes-Plant Vent 1.29E-03 ................. -MSSVs/10% ADVs (Note 1) ..... 8.60E-04 Control Room Infiltration -Plant Vent 1.26E-03 .......-MSSVs/1 0% ADVs ..... 2.78E-03 2.78E-03 1.49E-03 1.49E-03 Control Room Pressurization Intake-MSSVs/10% ADVs ......... 1.57E-05 7.65E-06 7.65E-06 Note 1: Due to the proximity of the release from the MSSVs/1 0% ADVs to the normal operation CR intake of the affected unit, and due to the high vertical velocity of the steam discharge from the MSSVs/10% ADVs, the resultant plume from the MSSVs/10% ADVs will not contaminate the normal operation CR intake of the affected unit. Thus the X/Q s presented reflect those applicable to the CR intake of the unaffected unit.133 of 205 Diablo Canyon Power Plant Implementation ofAlternative Source Terms Table 7.8-1 Loss of Load Analysis Assumptions & Key Parameter Values Parameter Value Power Level 3580 MWt Reactor Coolant Mass 446,486 Ibm Primary to Secondary SG tube leakage 0.75 ,qpm (total for all 4 SGs); leakage density 62.4 Ibm/ft Failed/Melted Fuel Percentage 0%RCS Technical Specification Iodine Levels Table 4.2-1 (1 pCi/gm DE 1-131)RCS Technical Specification Noble Gas Levels Table 4.2-1 (270 pCi/gm DE Xe-133)RCS Equilibrium Iodine Appearance Rates Table 4.2-2 (1 pCi/gm DE 1-131)Pre-Accident Iodine Spike Concentration Table 4.2-2 (60 pCi/gm DE 1-131)Accident-initiated Iodine Spike Appearance Rate 500 times TS equilibrium appearance rate Duration of Accident-Initiated Iodine Spike 8 hours Initial Secondary Coolant Iodine Concentrations 0.1 pCi/gm DE 1-131 (Table 4.2-1)Initial and Minimum SG Liquid Mass 92,301 lbm/SG Time period of tubes uncovered insignificant Steam Releases 0-2 hrs: 651,000 Ibm 2-8 hrs: 1,023,000 Ibm 8-10.73 hrs: same release rate as that for 2-8 hrs Iodine Partition Coefficient in SGs 100 Iodine Species Released to Environment 97% elemental; 3% organic Fraction of Noble Gas Released 1.0 (Released without holdup)Termination of releases from SGs 10.73 hours Environmental Release Point MSSVs/1 0% ADVs CR emergency Ventilation

Initiation Signal/Timing Control Room is assumed to remain on normal ventilation for duration of the accident.Control Room Atmospheric Dispersion Factors Table 7.8-2 134 of 205 AIKýýM91111 Diablo Canyon Power Plant Implementation ofAlternative Source Terms Table 7.8-2 Loss of Load Accident Control Room Limiting Atmospheric Dispersion Factors (seclm 3)Release point and receptor 0-2hr 2-8 hr 8-10.73 hr MSSVs/10%

ADVs to CR NOP Intake (Note 1) 8.60E-04 5.58E-04 5.58E-04 MSSVs/10% ADVs to CR Inleakage (CR Centerline) 2.78E-03 1.63E-03 1.63E-03 Note 1: Due to the proximity of the release from the MSSVs/100/% ADVs to the normal operation CR intake of the affected unit, and due to the high vertical velocity of the steam discharge from the MSSVs/10% ADVs, the resultant plume from the MSSVs/10% ADVs will not contaminate the normal operation CR intake of the affected unit. Thus the x/Q s presented reflect those applicable to the CR intake of the unaffected unit.135 of 205 Diablo Canyon Power Plant Implementation of Alternative Source Terms 8.0

SUMMARY

OF RESULTS: CONTROL ROOM I SITE BOUNDARY DOSES The accidents listed below have been analyzed for dose consequences at the site boundary and control room.1. Loss of Coolant Accident 2. Fuel Handling Accident in the Fuel Handling Building 3. Fuel Handling Accident in the Containment

4. Locked Rotor Accident 5. Control Rod Ejection Accident 6. Main Steam Line Break 7. Steam Generator Tube Rupture 8. Loss-of Load Event In accordance with RG 1.183, the "worst 2-hour period" dose at the EAB, and the dose at the LPZ "for the duration of the release" is presented in Table 8.1-1. These dose values represent the post-accident dose to the public due to inhalation and submersion for each of these events. Due to distance/plant shielding, the dose contribution at the EAB/LPZ due to direct shine from contained sources is considered negligible for all the accidents.

The associated regulatory limit as discussed in Section 2.4 is also presented. Per regulatory guidance, the CR dose is integrated over 30 days. The calculated doses address the fact that for events with a duration less than 30 days, the CR dose needs to include the remnant radioactivity within the CR envelope after the event has terminated. The 30-day integrated dose to the control room operator, due to inhalation and submersion, is presented in Table 8.1-1 for all of the referenced design basis accidents. No credit is taken for use of personal protective equipment or prophylactic drugs.The CR shielding design is based on the LOCA which represents the worst case DBA relative to radioactivity releases. The dose contribution due to direct shine from post LOCA contained sources/external cloud is identified and included in the CR doses reported for the LOCA in Table 8.1-1.The dose received by the operator during transit outside the control room is not a measure of the "habitability" of the control room which is defined by the radiation protection provided to the operator by the control room shielding and ventilation system design. Thus, the estimated dose to the operator during routine post-LOCA access to the control room is addressed separately from the control room occupancy dose and is not included with the control room occupancy dose for the demonstration of control room habitability. As demonstrated in Section 7.2.6, the dose contribution to the operator during routine access to control room for the duration of the LOCA is minimal (- 1% of the occupancy dose).In accordance with current licensing basis, the TSC design has been evaluated for the worst case DBA, i.e., the LOCA. The 30-day integrated dose to the TSC operator due to inhalation, submersion, and direct shine from the post LOCA contained sources/external cloud is estimated to be 4.1 rem TEDE (note: the dose contribution of direct shine to this total is -1.3 rem TEDE).136 of 205 Ah Diablo Canyon Power Plant Implementation ofAlternative Source Terms Table 8.1-1 AST Site Boundary and Control Room Dose (TEDE, rem)Regulatory Control Regulatory Accident EAB (1) (4) LPZ (2) Limit Room Limit LOCA 5.6 1 25 3.7 (0.7) (3) 5 Fuel Handling Accident in Fuel Handling Building 1.5 0.2 6.3 1.1 5 Fuel Handling Accident in Containment 1.5 0.2 6.3 4.7 5 Locked Rotor Accident 0.8 0.2 2.5 2:4 5 Control Rod Ejection Accident Containment Release 0.7 0.3 6.3 3.4 5 Secondary Release 0.7 0.2 0.5 Main Steam Line Break Pre-incident iodine Spike 0.1 <0.1 25 2.0 5 Accident-initiated Iodine Spike 0.7 0.2 2.5 4.1 Steam Generator Tube Rupture Pre-incident iodine Spike 1.3 0.1 25 0.6 5 Accident-Initiated Iodine Spike 0.7 <0.1 2.5 0.3 Loss of Load Pre-incident iodine Spike <0.1 <0.1 2.5 <0.1 5 Accident-Initiated Iodine Spike <0.1 <0.1 2.5 <0.1 Notes (1) EAB doses are based on worst 2-hour period following onset of accident. Except as noted, the maximum 2-hr dose period for the EAB dose for each of the accidents is the 0 to 2 hrs time period.S a LOCA: 24-26 hrs (based on RHR Pump Seal Failure; see note 4 below for additional information) LRA: 8.73 to 10.73 hrs MSLB (accident initiated spike model): 7.6 to 9.6 hrs LOL (accident initiated spike model): 8.73 to 10.73 hrs.(2) LPZ Doses are based on the duration of the release.(3) The dose presented represents the operator dose due to occupancy. Value shown in parenthesis represents that portion of the total dose reported that is the contribution of direct shine from contained sources/external cloud. The dose to the CR operator during routine access for the 30 day duration of the accident is discussed in Section 7.2.6 and summarized in the text of Section 8.0.(4) The maximum 2 hr EAB dose is based on the assumed RHR pump seal failure resulting in a 50 gpm leak of sump water occurring at t=24 hr for 30 mins. This release pathway is considered a part of DCPP licensing basis with respect to passive system failure. If this assumed release pathway were not included, the maximum 2 hr dose at the EAB would occur between t=0.5 hrs to t=2.5 hrs (i.e., during the post-LOCA ex-vessel release phase and would be 3.4 rem.137 of 205 Diablo Canyon Power Plant Implementation ofAlternative Source Terms

9.0 CONCLUSION

S The Alternative Source Term as defined in Regulatory Guide 1.183 has been incorporated into the DCPP site boundary and control room dose re-analyses discussed herein. In accordance with current licensing basis, the dose to the Technical Support Center has been evaluated for the DBA that has the worst case radioactivity release, i.e., the LOCA. The estimated DCPP dose consequences for all design basis events meet the acceptance criteria specified in 10CFR50.67 and RG 1.183. This represents a full implementation of the Alternative Source Terms in which the RG 1.183 source term will become the licensing basis for DCPP.138 of 205 Diablo Canyon Power Plant Implementation ofAlternative Source Terms

10.0 REFERENCES

1. Code of Federal Regulations 10CFR50.67, "Accident Source Term".2. NUREG-0800, Standard Review Plan 15.0.1, "Radiological Consequence Analyses using Alternative Source Terms," Revision 0.3. Regulatory Guide 1.183, Revision 0, "Alternative Radiological Source Terms for Evaluating Design Basis Accidents at Nuclear Power Reactors", July 2000.4. NRC Safety Evaluation Report Related to License Amendment No. 163 to Facility Operating License No. DPR-80 and License Amendment No 165 to Facility Operating License No. DPR-82, PG&E, Diablo Canyon Power Station, Units I and 2, Docket Nos 50-275 and 50-323, dated February 27, 2004.5. Regulatory Guide 1.145, "Atmospheric Dispersion Models for Potential Accident Consequence Assessments at Nuclear Power Plants", Revision 1.6. Ramsdell, J. V. Jr. and C. A. Simonen, "Atmospheric Relative Concentrations in Building Wakes". Prepared by Pacific Northwest Laboratory for the U.S. Nuclear Regulatory Commission, PNL-10521, NUREG/CR-6331, Revision 1, May 1997.7. TID-14844, "Calculation of Distance Factors for Power and Test Reactor Sites", 1962.8. Regulatory Guide 1.4, Revision 1, "Assumptions Used for Evaluating the Potential Radiological Consequences of a Loss of Coolant Accident for Pressurized Water Reactors".

9. Code of Federal Regulations 1 OCFR50.44, "Combustible Gas Control for Nuclear Power Reactors".

10. NRC Safety Evaluation Report Related to License Amendment No. 168 to Facility Operating License No. DPR-80 and License Amendment No 169 to Facility Operating License No. DPR-82, PG&E, Diablo Canyon Power Station, Units 1 and 2, Docket Nos 50-275 and 50-323, dated May 4, 2004.11. Code of Federal Regulations, IOCFR100.11, "Determination of Exclusion Area, Low Population Zone, and Population Center Distance." 12. Code of Federal Regulations, 1OCFR50, Appendix A, GDC 19, "Control Room".13. NUREG-0800, SRP 6.4, Revision 3, "Control Room Habitability System".14. Code of Federal Regulations, 10CFR50.49, "Environmental Qualification of Electric Equipment Important to Safety for Nuclear Power Plants." 15. NUREG-0737, "Clarification of TMI Action Plan Requirements," Nov. 1980.139 of 205 Diablo Canyon Power Plant Implementation ofAlternative Source Terms 16. NUREG-1465, "Accident Source Terms for Light-Water Nuclear Power Plants", February 1995.17. SECY-98-154, "Results of the Revised (NUREG-1465)

Source Term Rebaselining for Operating Reactors," June 30, 1998.18. NRC Safety Evaluation Report Related to License Amendment No. 201 to Facility Operating License no. DPR-40, OPPD, Fort Calhoun Station Unit No. 1, Docket No. 50-285, dated December 5, 2001.19. Federal Guidance Report No. 11, 1988, "Limiting Values of Radionuclide Intake and Air Concentration and Dose Conversion factors for Inhalation, Submersion, and Ingestion" 20. Federal Guidance Report No. 12, 1993, "External Exposure to Radionuclides in Air, Water, and Soil" 21. Regulatory Guide 1.194, June 2003, Atmospheric Relative Concentrations for Control Room Radiological Habitability Assessments at Nuclear Power plants.22. NUREG 0800, Standard Review Plan 6.2.4, Revision 2, "Containment Isolation System".23. Safety Guide 25, March 23, 1972, "Assumptions Used for Evaluating the Potential Radiological Consequences of a Fuel Handling Accident in the Fuel Handling and Storage Facility for Boiling and Pressurized Water Reactors.24. NUREG/CR 5009, Assessment of the Use of Extended Burn Fuel in LWRs, Jan 1988.25. NUREG-0017, "Calculation of Releases of Radioactive Materials in Gaseous and Liquid Effluents from Pressurized Water Reactors", Revision 1.26. Regulatory Guide 1.111, "Methods for Estimating Atmospheric Transport and Dispersion of Gaseous Effluents in Routine Releases from Light Water Cooled Reactors", Revision 1.27. NUREG-0737, Supplement 1, Clarification of TMI Action Plan Requirements, January 1983.28. Regulatory Guide 1.23, "Meteorological Monitoring Programs for Nuclear Power Plants", Revision 1, March, 2007.29. Bellinger, Thomas E., "The Impact of Nearby Structures and Trees on Sigma Theta Measurements", Illinois Department of Nuclear Safety, presented at the May 2002 NUMUG meeting, St. Charles, Illinois.30. Call, Jennifer, "Evaluation of Obstruction Impacts on Wind Flow at Clinch River Nuclear Plant", February 2013.31. ANSI/ANS 6.1.1-1977, "Neutron and Gamma-ray Flux-to-Dose-Rate Factors" 140 of 205

32.33.34.35.36.37.38.39.40.41.42.43.44.45.46.47.Diablo Canyon Power Plant Implementation of Alternative Source Terms NUREG-0800, Standard Review Plan 15.2.8, Revision 2, "Feedwater System Pipe Break Inside and Outside Containment (PWR)".NUREG/CR-5966, "A Simplified Model of Aerosol Removal by Containment Sprays", June 1993.Elia, Frank A. Jr. and Lischer, D. Jeffrey, Advanced Method for Calculating the Removal of Airborne Particles with Sprays, 1993, ASME paper no. 93-WA/SERA-5.

NUREG-0772, June 1981, "Technical Bases for Estimating Fission Product Behavior During LWR Accidents." Battelle Columbus Laboratories, BMI-2104, Vol. III, draft report, 1984, "Radionuclide Release Under Specific LWR Accident Conditions." Walton, W. H., and Woolcock, A., 1960, "The Suppression of Airborne Dust by Water Spray," Interm. J. Air Pollution 3, 129-153.Calvert, S., 1970, "Venturi and Other Atomizing Scrubbers Efficiency and Pressure Drop," AIChE Journal 16, 392-396.Fuchs, N.A., 1964, "The Mechanics of Aerosols," revised and enlarged edition, Dover Publications, Inc.Bunz, H., Kayro, M., Sch6ck, W., 1982, NAUA/MOD4 -A Code for Calculating Aerosol Behaviour in LWR Core Melt Accidents, Code Description and User Manual, KfK.NRC Generic Letter No. 99-02, Laboratory Testing of Nuclear Grade Activated Charcoal, June 3, 1999.NUREG-0800, 1988, Standard Review Plan, "Containment Spray as a Fission Product Cleanup System", Section 6.5.2, Revision 4.Prentice-Hall W. M. Rohsenow and Harry Choi, "Heat, Mass and Momentum Transfer", 1961 NUREG/CR-5732, "Iodine Chemical Forms in LWR Severe Accidents -Final Report," April 1992.NUREG-0800, Standard Review Plan, Section 6.1.1, "Engineered Safety Features Materials", Revision 2 NUREG/CR-5950, "Iodine Evolution and pH Control", December 1992.IE Bulletin No. 79-01B, Environmental Qualification of Class IE Equipment, January 14, 1980, including Enclosure 4, Guidelines for Evaluating Environmental Qualification of Class IE Electrical Equipment in Operating Reactors.141 of 205 Diablo Canyon Power Plant Implementation of Alternative Source Terms 48. Information Notice 91-56, September 19, 1991, "Potential Radioactive Leakage to Tank Vented to Atmosphere".

49. Joon Cho, Joseph Baron, Keith Ferguson, ""Modeling Radioactive Leakage from Atmospheric Tank Vents Following a LOCA", ANS Summer Conference in 2007, published in Transactions of the American Nuclear Society Volume 96, Radiation Protection and Shielding Session I, pg 441.50. NUCON International Inc., "Control room Habitability Tracer Gas Leak Testing at Diablo Canyon Power Plant", December 2012.51. NUREG-0800, Section 15.6.5, Appendix B, Revision 1, "Radiological Consequences of a Design Basis LOCA: Leakage from engineered Safety Feature Components outside Containment".

52. NUREG-0800, Standard Review Plan (SRP) Sections 15.2.1-15.2.5, "Loss of External Load; Turbine Trip; Loss of Condenser Vacuum; Closure of Main Steam Isolation Valve (BWR); and Steam Pressure Regulator Failure (Closed)", Revision 1.53. NUREG-0800, SRP Section 15.2.6, "Loss of Non-Emergency AC Power to the Station Auxiliaries", Revision 1.54. NRC Safety Evaluation Report Related to License Amendment No. 139 to Facility Operating License No. DPR-80 and No. DPR-82, PG&E, Diablo Canyon Power Station, Units 1 and 2, Docket Nos 50-275 and 50-323, dated February 9, 2000.55. NUREG-0933, "Resolution of Generic Safety Issues", Item 187, closed June 30, 2000 56. Code of Federal Regulations, 10CFR20.1003, "Definitions".

57. ANSI/ANS 6.1.1-1991, "Neutron and Gamma-ray Fluence-to-dose Factors".58. NRC SER Related to License Amendment No. 8 and 6 to Facility Operating License No.DPR-80 and. DPR-82, PG&E, Diablo Canyon Power Station, Units 1 and 2, dated May 30, 1986.59. Safety Guide 23, "Onsite Meteorological Programs", February 17, 1972 60. SCALE 4.3, "Modular Code System for Performing Standardized Computer Analyses for Licensing Evaluation for Workstations And Personal Computers," Control Module SAS2-CB&I S&W Inc. QA Category I Computer Code NU-230, V04, L03.61. ACTIVITY2, "Fission Products in a Nuclear Reactor" -CB&I S&W Inc. Proprietary QA Category I Computer Code NU-014, V01, L03.62. IONEXCHANGER, -CB&I S&W Inc. Proprietary QA Category I Computer Code NU-009, Ver. 01, Lev. 03.142 of 205 Diablo Canyon Power Plant Implementation ofAlternative Source Terms 63. EN-1 13, "Atmospheric Dispersion Factors" -CB&I S&W Inc. Proprietary QA Category I Computer Code EN-1 13, V06, L08.64. ARCON96, "Atmospheric Relative Concentrations in Building Wakes" -CB&I S&W Inc.QA Category I Computer Code EN-292, V00, LOD.65. SWNAUA, "Aerosol Behavior in Condensing Atmosphere", CB&I S&W Inc. Proprietary QA Category I Computer Code NU-185, V02, LO.66. RADTRAD 3.03 "A Simplified Model for RADionuclide Transport and Removal And Dose Estimates" -CB&I S&W Inc. QA Category I computer code No. NU-232, Version 3.03, Level (NA).67. PERC2, "Passive Evolutionary Regulatory Consequence Code" -CB&I S&W Inc.Proprietary QA Category I Computer Code, NU-226, VOO, L02.68. SW-QADCGGP, "A Combinatorial Geometry Version of QAD-5A" -CB&I S&W Proprietary QA Category I Computer Code, NU-222, V00, L02.69. GOTHIC, "Generation of Thermal-Hydraulic Information for Containments", CB&I S&W QA Category I computer code No. ME-376, Version 8.0, Lev (NA).70. NRC Regulatory Issue Summary 2006-04, "Experience with Implementation of Alternative Source Terms", March 7, 2006 143 of 205 96 Diablo Canyon Power Plant Implementation ofAlternative Source Terms APPENDIX A: DCPP ARCON96 ATMOSPHERIC DISPERSION FACTOR INPUTS Appendix A provides the DCPP Unit 1 and Unit 2 release point and receptor configuration information (e.g., height, velocity, distances, direction with respect to true north, etc.), release mode (e.g., ground, elevated, surface), and meteorological sensor configuration, used as input into ARCON96.Also included as Figure A-i, is a site building layout and arrangement drawing that depicts the locations of the postulated release points and receptors.

Figure A-1 shows plant north. The directions provided in the Unit 1 and Unit 2 release point and receptor configuration Tables presented herein account for the 23-degree azimuth clockwise offset of true north from plant north.The on-site meteorological data input to ARCON96 (January 1, 2007 through December 31, 2011), in the ARCON96 input data format, is embedded below Attachment DCPP ARCON96 Atmospheric Dispersion Factor Inputs On-site meteorological data in to ARCON96 (1/1/2007 -12/31/2007) On-site meteorological data in to ARCON96 (1/1/2008 -12/31/2008) On-site meteorological data in to ARCON96 (1/1/2009 -12/31/2009) On-site meteorological data in to ARCON96 (1/1/2010 -12/31/2010) On-site meteorological data in to ARCON96 (1/1/2011 -12/31/2011) 144 of 205 ýft, Diablo Canyon Power Plant Implementation ofAlternative Source Terms Table A-1 On-Site Atmospheric Dispersion Factor Evaluation Post-accident Release Point and Receptor Description / Location ID Release/Receptor Description (See Figure A-1)Note 1 Release Point Unit 1 Containment Building (CB) edge Note 1 Release Point Unit 2 Containment Building (CB) edge 1 Release Point U1 Plant Vent 2 Release Point U2 Plant Vent 3 Receptor U1 Control Room Normal Intake 4 Receptor U2 Control Room Normal Intake 5 Receptor U1 Control Room Emergency Intake 6 Receptor U2 Control Room Emergency Intake 7 Release Point UI RWST Vent 8 Release Point U2 RWST Vent 9 Receptor Control Room Center (location assigned for unfiltered inleakage) 10 Release Point Unit 1 Containment Penetration Area, GE 11 Release Point Unit 2 Containment Penetration Area, GE 12 Release Point Unit I Containment Penetration Area, FW/GW 13 Release Point Unit 2 Containment Penetration Area, FW/GW 14 Release Point U1 Fuel Handling Building 15 Release Point U2 Fuel Handling Building 16 Release Point U1 Equipment Hatch 17 Release Point U2 Equipment Hatch 18 Release Point U1 MSSV 19 Release Point U2 MSSV 20 Release Point U1 10% ADVs 21 Release Point U2 10% ADVs 22 Release Point U1 MSL Break location 23 Release Point U2 MSL Break location 24 Receptor TSC Normal Intake 25 Receptor TSC Center (location assigned for unfiltered inleakage) Note 1: Though not depicted in Figure A-i, atmospheric dispersion factors were also calculated from the closest edge of the containment building to the various receptors; this release point was treated as a diffuse source.145 of 205 "133 Diablo Canyon Power Plant Implementation of Alternative Source Terms Figure A-1 Diablo Canyon Power Plant -Site Layout and Arrangement Post-Accident Environmental Release Point I Receptor Locations ii A ( II Ii I 9 9 i 9 99 99 9 Y 400-350-3W j--------------- rM IMINIMM OL--------------

jFE F G---------------

-3.2-- --- -- 22------------- 1FW---------------L 2 70'H 1 12 1 dvi----------------------- T ; R it 9 1 14U MooII 1 CM0 W" tJJR -r- W-----------Q.s-.-, -- -----91~ P mu r ( i i 1 1 1 1 1 1I I -I I i I-I I I 1 -1 11 11 F 1 1 I" AU F i F I- V, aw- I I I 11 1 B C BA A I I I 2L b I I I F)mY s~o 150 4. u*, Z, a. --$1 01 n K; Ei 11 7";.4 ii 9~I;I I N Vi i. t 1 1 :3 1NO" T---.-- -.-- -----------------------)20 250 390 350 400 ~4ý0 b00 sk, COOR*INAUI-COLUWN L94 CROSS WW"MNC 6600-A 76o 80 8&0......... 146 of 205 ýffi Diablo Canyon Power Plant Implementation ofAlternative Source Terms Unit I LOCA Releases to CR Receptors Release Point / Receptor ARCON96 Parameter11 U1 CB / U1 CR U1 CB / U2 CR 1U1 CB/U1 CR U1 CB/U2CR Normal Intake Normal Intake Emergency Intake Emergency Intake Case No. 1 39 2 3 Meteorological Information: Period of Meteorological Data 2007 -2011 2007 -2011 2007 -2011 2007 -2011 Lower Measurement Height (m) 10 10 10 10 Upper Measurement Height (m) 76 76 76 76 Wind Speed Units m/sec m/sec m/sec m/sec Meteorological Data File Names: dcpp2007.met-dcpp2007.met-dcpp2007.met-dcpp2007.met-dcpp2011.met dcpp2011 .met dcpp2011 .met dcpp2011 .met Source Information: Release Type ground -diffuse ground -diffuse ground -diffuse ground -diffuse Release Height (m) 22.0 22.0 32.0 32.0 Building Area (m') 2,744.5 2,744.5 2,744.5 2,744.5 Vertical Velocity (m/sec) 0.0 0.0 0.0 0.0 Stack Flow (ms/sec) 0.0 0.0 0.0 0.0 Stack Radius (m) 0.0 0.0 0.0 0.0 Receptor Information: Distance to Receptor (m) 9.4 44.6 71.8 151.4 Intake Height (m) 22.0 22.0 32.0 32.0.Elevation Difference (m) 0.0 0.0 0.0 0.0 Direction to Source (degrees azimuth) 345 339 115 003 Default Information: Surface Roughness Length (m) 0.20 0.20 0.20 0.20 Wind Direction Window (degrees azimuth) 90 90 90 90 Minimum Wind Speed (m/sec) 0.5 0.5 0.5 0.5 Averaging Sector Width Constant 4.3 4.3 4.3 4.3 Initial Diffusion Coefficients (m) 7.42, 11.07 7.42, 11.07 7.42, 11.07 7.42,11.07 Report No. 14078101 -RADR-004-1 147 of 205 Revision 1 Diablo Canyon Power Plant Implementation of Alternative Source Terms Unit I LOCA Releases to CR Receptors Receptor ARCON96 Parameter UI CB I CR Center U1 PVIUt CR U1 PVIU2 CR U1 PV/U1 CR Normal Intake Normal Intake Emergency Intake Case No. 4 37 41 5 Meteorological Information: Period of Meteorological Data 2007 -2011 2007 -2011 2007 -2011 2007 -2011 Lower Measurement Height (m) 10 10 10 10 Upper Measurement Height (m) 76 76 76 76 Wind Speed Units m/sec m/sec m/sec m/sec Meteorological Data File Names: dcpp2007.met-dcpp2007.met-dcpp2007.met-dcpp2007.met-dcpp2011.met dcpp20l11.met dcpp20l11.met dcpp2011.met Source Information: Release Type ground -diffuse ground -point ground -point ground -point Release Height (m) 24.4 74.1 74.1 74.1 Building Area (min) 2,744.5 2,744.5 2,744.5 2,744.5 Vertical Velocity (m/sec) 0.0 0.0 0.0 0.0 Stack Flow (m 3/sec) 0.0 0.0 0.0 0.0 Stack Radius (m) 0.0 0.0 0.0 0.0 Receptor Information: Distance to Receptor (m) 27.4 31.7 66.8 93.6 Intake Height (m) 24.4 22.0 22.0 32.0 Elevation Difference (m) 0.0 0.0 0.0 0.0 Direction to Source (degrees azimuth) 347 345 339 115 Default Information: Surface Roughness Length (m) 0.20 0.20 0.20 0.20 Wind Direction Window (degrees azimuth) 90 90 90 90 Minimum Wind Speed (m/sec) 0.5 0.5 0.5 0.5 Averaging Sector Width Constant 4.3 4.3 4.3 4.3 Initial Diffusion Coefficients (m) 7.42; 11.07 0.0, 0.0 0.0, 0.0 0.0, 0.0 Report No. 14078101-RADR-004-1 148 of 205 Revision I "it Diablo Canyon Power Plant Implementation ofAlternative Source Terms Unit I LOCA Releases to CR Receptors Receptor ARCON96 Parameter A UI PV / U2 CR UI PV / CR Center U1 RWSTI UI CR U1 RWST I U2 CR Emergency Intake Emergency Intake Emergency Intake Case No. 6 7 8 9 Meteorological Information: Period of Meteorological Data 2007 -2011 2007 -2011 2007 -2011 2007 -2011 Lower Measurement Height (m) 10 10 10 10 Upper Measurement Height (m) 76 76 76 76 Wind Speed Units m/sec m/sec m/sec m/sec Meteorological Data File Names: dcpp2007.met-dcpp2007.met-dcpp2007.met-dcpp2007.met-dcpp2011.met dcpp2011.met dcpp2011.met dcpp20ll.met Source Information: Release Type ground -point ground -point ground -point ground -point Release Height (m) 74.1 74.1 26.8 26.8 Building Area (m 2) 2,744.5 2,744.5 215.5 215.5 Vertical Velocity (m/sec) 0.0 0.0 0.0 0.0 Stack Flow (m'/sec) 0.0 0.0 0.0 0.0 Stack Radius (m) 0.0 0.0 0.0 0.0 Receptor Information: Distance to Receptor (m) 173.2 50.7 141.2 189.6 Intake Height (m) 32.0 24.4 32.0 32.0 Elevation Difference (m) 0.0 0.0 0.01 0.01 Direction to Source (degrees azimuth) 003 354 100 019 Default Information: Surface Roughness Length (m) 0.20 0.20 0.20 0.20 Wind Direction Window (degrees azimuth) 90 90 90 90 Minimum Wind Speed (m/sec) 0.5 0.5 0.5 0.5 Averaging Sector Width Constant 4.3 4.3 4.3 4.3 Initial Diffusion Coefficients (m) 0.0, 0.0 0.0, 0.0 0.0, 0.0 0.0, 0.0 1. Difference in local grade elevation between the RWST (115 ft) and the CR (85 ft) is accounted for in the release height such that source/receptor elevation difference is set to zero.Report No. 14078101 -RADR-004-1 149 of 205 Revision 1 Diablo Canyon Power Plant Implementation ofAlternative Source Terms Unit I LOCA Releases to CR Receptors Receptor ARCON96 Parameter UI RWST / CR U1 Cont. Pen GE! UI CR UI Cont. Pen GE I U2 UI Cont. Pen GE / UI Center Normal Intake CR Normal Intake CR Emergency Intake Case No. 10 21 43 22 Meteorological Information: Period of Meteorological Data 2007 -2011 2007 -2011 2007 -2011 2007 -2011 Lower Measurement Height (m) 10 10 10 10 Upper Measurement Height (m) 76 76 76 76 Wind Speed Units m/sec m/sec m/sec m/sec Meteorological Data File Names: dcpp2007.met-dcpp2007.met-dcpp2007.met-dcpp2007.met-dcpp2011,nmet dcpp2011. met dcpp2011.met dcpp2011.rmet Source Information:__________ Release Type ground -point ground -point ground -point ground -point Release Height (m) 26.8 16.8 16.8 16.8 Building Area (min) 215.5 530.4 530.4 2,744.5" Vertical Velocity (m/sec) 0.0_ 0.0 0.0 0.0 Stack Flow (mr/sec) 0.0 0.0 0.0 0.0 Stack Radius (m) 0.0 0.0 0.0 0.0 Receptor Information: Distance to Receptor (m) 74.9 26.7 51.2 128.0 Intake Height (m) 24.4 22.0 22.0 32.0 Elevation Difference (m) 0.01 0.0 0.0 0.0 Direction to Source (degrees azimuth) 037 047 007 112 Default Information: Surface Roughness Length (m) 0.20 0.20 0.20 0.20 Wind Direction Window (degrees azimuth) 90 90 90 90 Minimum Wind Speed (m/sec) 0.5 0.5 0.5 0.5 Averaging Sector Width Constant 4.3 4.3 4.3 4.3 Initial Diffusion Coefficients (m) 0.0, 0.0 0.0, 0.0 0.0, 0.0 0.0, 0.0 1. Difference in local grade elevation between the RWST (115 ft) and the CR (85 ft) is accounted for in the release height such that source/receptor elevation difference is set to zero.2. Release/receptor path goes around the Unit I CB.Report No. 14078101 -RADR-004-1 150 of 205 Revision I Diablo Canyon Power Plant Implementation ofAlternative Source Terms Unit 1 LOCA Releases to CR Receptors Receptor ARCON96 Parameter U1 Cont. Pen GE I U2 U1 Cont. Pen GE I U1 PL GW/FW / Unit U1 PL GWIFW I U2 CR Emergency Intake CR Center I CR Normal Intake CR Normal Intake Case No. 23 24 25 44 Meteorological information: Period of Meteorological Data 2007 -2011 2007 -2011 2007 -2011 2007 -2011 Lower Measurement Height (m) 10 10 10 10 Upper Measurement Height (m) 76 76 76 76 Wind Speed Units m/sec m/sec m/sec m/sec Meteorological Data File Names: dcpp2007.met -dcpp2007. met -dcpp2007. met -dcpp2007. met -dcpp20l1 .met dcpp20l1 .met dcpp20l l.met dcpp2011 .met Source Information: Release Type ground -point ground -point ground -point ground -point Release Height (m) 16.8 16.8 16.8 16.8 Building Area (m-) 530.4 530.4 2,744.5 2,744.5 Vertical Velocity (m/sec) 0.0 0.0 0.0 0.0 Stack Flow (ma/sec) 0.0 0.0 0.0 0.0 Stack Radius (m) 0.0 0.0 0.0 Receptor Information: Distance to Receptor (m) 165.0 41.0 34.0 67.1 Intake Height (m) 32.0 24.4 22.0 22.0 Elevation Difference (m) 0.0 0.0 0.0 0.0 Direction to Source (degrees azimuth) 012 027 300 319 Default Information: Surface Roughness Length (m) 0.20 0.20 0.20 0.20 Wind Direction Window (degrees azimuth) 90 90 90 90'Minimum Wind Speed (m/sec) 0.5 0.5 0.5 0.5 Averaging Sector Width Constant 4.3 4.3 4.3 4.3 Initial Diffusion Coefficients (m) 0.0, 0.0 0.0, 0.0 0.0, 0.0 0.0, 0.0 Report No. 14078101-RADR-004-1 151 of 205 Revision I Diablo Canyon Power Plant Implementation ofAlternative Source Terms Unit I LOCA Releases to CR Receptors Receptor ARCON96 Parameter U1 PL GWIFW / U1 Ul PL GW/FW / U2 CR U1 PL GWIFW /CR Emergency Intake Emergency Intake CR Center Case No. 26 27 28 Meteorological Information: Period of Meteorological Data 2007-2011 2007-2011 2007 -2011 Lower Measurement Height (m) 10 10 10 Upper Measurement Height (m) 76 76 76 Wind Speed Units m/sec m/sec m/sec Meteorological Data File Names: dcpp2007.met-dcpp2007.met-dcpp2007.met-dcpp2011 .met dcpp20l1 .met dcpp20l1 .met Source Information: Release Type ground -point ground -point ground -point Release Height (m) 16.8 16.8 16.8 Building Area (m'_) 2,744.5 2,744.5 .2,744.5 Vertical Velocity (m/sec) 0.0 0.0 0.0 Stack Flow (me/sec) 0.0 0.0 0.0 Stack Radius (m) 0.0 0.0 0.0 Receptor Information: Distance to Receptor (m) 83.0 166.0 48.0 Intake Height (m) 32.0 32.0 24.4 Elevation Difference (m) 0.0 0.0 0.0 Direction to Source (degrees azimuth) 124 354 317 Default Information: Surface Roughness Length (m) 0.20 0.20 0.20 Wind Direction Window (degrees azimuth) 90 90 90 Minimum Wind Speed (m/sec) 0.5 0.5 0.5 Averaging Sector Width Constant 4.3 4.3 4.3 Initial Diffusion Coefficients (m) 0.0, 0.0 0.0, 0.0 0.0, 0.0 Report No. 14078101-RADR-004-1 162 of 205 Revision I ýft Diablo Canyon Power Plant Implementation ofAlternative Source Terms Unit I LOCA Releases to TSC Receptors Receptor ARCON96 Parameter CB / TSC Normal UI CB I TSC Center U1PV/TSC Normal UI PV / TSC Center Intake Intake Case No. 1 2 3 4 Meteorological Information: Period of Meteorological Data 2007 -2011 2007 -2011 2007 -2011 2007 -2011 Lower Measurement Height (m) 10 10 10 10 Upper Measurement Height (m) 76 76 76 76 Wind Speed Units m/sec m/sec m/sec m/sec Meteorological Data File Names: dcpp2007.met-dcpp2007.met-dcpp2007.met-dcpp2007.met-dcpp20ll.met dcpp20ll.met dcpp2011.met dcpp21ll.met Source Information: Release Type ground -diffuse ground -diffuse ground -point ground -point Release Height (m) 11.0 10.7 74.1 74.1 Building Area (me) 2,744.5 2,744.5 2,744.5 2,744.5 Vertical Velocit (m/sec) 0.0 0.0 0.0 0.0 Stack Flow (m /sec) 0.0 0.0 0.0 0.0 Stack Radius (m) 0.0 0.0 0.0 0.0 Receptor Information: Distance to Receptor (m) 109.5 97.4 131.7 119.7 Intake Height (m) 11.0 10.7 11.0 10.7 _Elevation Difference (m) 0.0 0.0 0.0 0.0 Direction to Source (degrees azimuth) 013 017 013 017 Default Information: Surface Roughness Length (m) 0.20 0.20 0.20 0.20 Wind Direction Window (degrees azimuth) 90 90 90 90 Minimum Wind Speed (m/sec) 0.5 0. 5 0.5 Averaging Sector Width Constant 4.3 4.3 4.3 4.3 Initial Diffusion Coefficients (m) 7.42, 11.07 7.42, 11.07 0.0, 0.0 0.0, 0.0 Report No. 14078101 -RADR-004-1 153 of 205 Revision I 46; %Diablo Canyon Power Plant Implementation of Alternative Source Terms Unit I LOCA Releases to TSC Receptors Receptor ARCON96 ParameterII UI RWSTI TSC U1 RWST/ TSC U1 Cont. Pen GE / U1 Cont. Pen GE I Normal Intake Center TSC Normal Intake TSC Center Case No. 5 6 7 8 Meteorological Information: Period of Meteorological Data 2007 -2011 2007 -2011 2007 -2011 2007 -2011 Lower Measurement Height (m) 10 10 10 10 Upper Measurement Height (m) 76 76 76 76 Wind Speed Units m/sec m/sec m/sec m/sec Meteorological Data File Names: dcpp2007.met-dcpp2007.met-dcpp2007.met-dcpp2007.met-dcpp2011.met dcpp2011.met dcpp20ll.met dcpp2011.met Source Information: Release Type ground -point ground -point ground -point ground -point Release Height (m) 26.8 26.8 16.8 16.8 Building Area (mi) 215.5 215.5 530.4 530.4 Vertical Velocity (m/sec) 0.0 0.0 0.0 0.0 Stack Flow (ms/sec) 0.0 0.0 0.0 0.0 Stack Radius (m) 0.0 0.0 0.0 0.0 Receptor Information: Distance to Receptor (m) 159.4 150.8 130.6 121.3 Intake Height (m) 11.0 10.7 11.0 10.7 Elevation Difference (m) 0.0 0.0 0.0 0.0 Direction to Source (degrees azimuth) 030 035 027 032 Default Information: Surface Roughness Length (m) 0.20 0.20 0.20 0.20 Wind Direction Window (degrees azimuth) 90 90 90 90 Minimum Wind Speed (m/sec) 0.5 0.5 0.5 0.5 Averaging Sector Width Constant 4.3 4.3 4.3 4.3 Initial Diffusion Coefficients (m) 0.0, 0.0 0.0, 0.0 0.0, 0.0 0.0, 0.0 Report No. 14078101 -RADR-004-1 154 of 205 Revision I Diablo Canyon Power Plant Implementation ofAlternative Source Terms Unit 1 LOCA Releases to TSC Receptors Receptor ARCON96 Parameter1 U1 PL GW/FW / TSC U1 PL GW/FW I TSC Normal Intake Center Case No. 9 10 Meteorological Information: Period of Meteorological Data 2007 -2011 2007 -2011 Lower Measurement Height (m) 10 10 Upper Measurement Height (m) 76 76 Wind Speed Units m/sec m/sec Meteorological Data File Names: dcpp2007.met-dcpp2007.met-dcpp2011 .met dcpp2011 .met Source Information: Release Type ground -point ground -point Release Height (m) 16.8 16.8 Building Area (min) 2,744.5 2,744.5 Vertical Velocity (m/sec) 0.0 0.0 Stack Flow (ms/sec) 0.0 0.0 Stack Radius (m) 0.0 0.0 Receptor Information: Distance to Receptor (m) 116.7 103.3 Intake Height (m) 11.0 10.7 Elevation Difference (m) 0.0 0.0 Direction to Source (degrees azimuth) 004 008 Default Information: Surface Roughness Length (m) 0.20 0.20 Wind Direction Window (degrees azimuth) 90 90 Minimum Wind Speed (m/sec) 0.5 0.5 Averaging Sector Width Constant 4.3 4.3 Initial Diffusion Coefficients (m) 0.0, 0.0 0.0, 0.0 Report No. 14078101 -RADR-004-1 155 of 205 Revision I Ah Diablo Canyon Power Plant Implementation of Alternative Source Terms Unit 2 LOCA Releases to CR Receptors Receptor ARCON96 ParameterI U2 CBIU1 CR U2 CBIU2 CR U2 CBIUI CR U2 CBIU2 CR Normal Intake Normal Intake Emergency Intake Emergency Intake Case No. 40 11 12 13 Meteorological Information: Period of Meteorological Data 2007-2011 2007-2011 2007-2011 2007-2011 Lower Measurement Height (m) 10 10 10 10 Upper Measurement Height (m) 76 76 76 76 Wind Speed Units m/sec m/sec m/sec m/sec Meteorological Data File Names: dcpp2007.met-dcpp2007.met-dcpp2007.met-dcpp2007.met-dcpp2011 .met dcpp2011.met dcpp2011 .met dcpp2011.met Source Information: Release Type ground -diffuse ground -diffuse ground -diffuse ground -diffuse Release Height (m) 22.0 22.0 32.0 32.0 Building Area (mi) 2,744.5 2-744.5 2,744.5 2,744.5 Vertical Velocit (m/sec) 0.0 0.0 0.0 0.0 Stack Flow (m /sec) 0.0 0.0 0.0 0.0 Stack Radius (m) 0.0 0.0 0.0 0.0 Receptor Information: Distance to Receptor (m) 44.9 7.8 150.6 71.0 Intake Height (m) 22.0 22.0 32.0 32.0 Elevation Difference (m) 0.0 0.0 0.0 0.0 Direction to Source (degrees azimuth) 154 150 132 027 Default Information: Surface Roughness Length (m) 0.20 0.20 0.20 0.20 Wind Direction Window (degrees azimuth) 90 90 90 90 Minimum Wind Speed (m/sec) 0.5 0.5 0.5 0.5 Averaging Sector Width Constant 4.3 4.3 4.3 4.3 Initial Diffusion Coefficients (m) 7.42, 11.07 7.42, 11.07 7.42, 11.07 7.42, 11.07 Report No. 14078101-RADR-004-1 156 of 205 Revision I Diablo Canyon Power Plant Implementation ofAlternative Source Termts Unit 2 LOCA Releases to CR Receptors Receptor ARCON96 Parameter U2 CB I CR Center U2 PVI U1 CR U2 PV/U2 CR U2 PVIU1 CR Normal Intake Normal Intake Emergency Intake Case No. 14 38 42 15 Meteorological Information: Period of Meteorological Data 2007-2011 2007-2011 2007-2011 2007-2011 Lower Measurement Height (m) 10 10 10 10 Upper Measurement Height (m) 76 76 76 76 Wind Speed Units m/sec m/sec m/sec m/sec Meteorological Data File Names: dcpp2007.met-dcpp2007.met-dcpp2007.met-dcpp2007.met-dcpp20ll.met dcpp2011.met dcpp2011.met dcpp2011.met Source Information: Release Type ground -diffuse ground -point ground -point ground -point Release Height (m) 24.4 74.1 74.1 74.1 Building Area (min) 2,744.5 2,744.5 2,744.5 2,744.5 Vertical Velocity (m/sec) 0.0 0.0 0.0 0.0 Stack Flow (m6/sec) 0.0 0.0 0.0 0.0 Stack Radius (.m) 0.0 0.0 0.0 0.0 Receptor Information: Distance to Receptor (m) 27.3 67.1 30.1 173.2 Intake Height (m) 24.4 22.0 22.0 32.0 Elevation Difference (m) 0.0 0.0 0.0 0.0 Direction to Source (degrees azimuth) 140 154 150 132 Default Information: Surface Roughness Length (m) 0.20 0.20 0.20 0.20 Wind Direction Window (degrees azimuth) 90 90 90 90-Minimum Wind Speed (m/sec) 0.5 0.5 0.5 0.5 Averaging Sector Width Constant 4.3 4.3 4.3 4.3 Initial Diffusion Coefficients (m) 7.42, 11.07 0.0, 0.0 0.0, 0.0 0.0, 0.0 Report No. 14078101 -RADR-004-1 157 of 205 Revision 1 a&Diablo Canyon Power Plant Implementation ofAlternative Source Terms Unit 2 LOCA Releases to CR Receptors Receptor ARCON96 Parameterr U2 PVI U2 CR U2 PV(CR Center U2 RWSTIUI CR U2 RWST(U2 CR Emergency Intake Emergency Intake Emergency Intake Case No. 16 17 18 19 Meteorological Information: Period of Meteorological Data 2007 -2011 2007 -2011 2007 -2011 2007 -2011 Lower Measurement Height (m) 10 10 10 10 Upper Measurement Height (m) 76 76 76 76 Wind Speed Units m/sec m/sec m/sec m/sec Meteorological Data File Names: dcpp2007.met-dcpp2007.met-dcpp2007.met-dcpp2007.met-dcpp2O 1 .met dcpp201 I.met dcpp2011 .met dcpp2O 11 .met Source Information: Release Type ground -point ground -point ground -point ground -point Release Height (m) 74.1 74.1 26.8 26.8 Building Area (mi) 2,744.5 2,744.5 215.5 215.5 Vertical Velocity (m/sec) 0.0 0.0 0.0 0.0 Stack Flow (m /sec) 0.0 0.0 0.0 0.0 Stack Radius (m) 0.0 0.0 0.0 0.0 Receptor Information: Distance to Receptor (m) 93.7 49.9 190.4 140.5 Intake Height (m) 32.0 24.4 32.0 32.0 Elevation Difference (m) 0.0 0.0 0.01 0.01 Direction to Source (degrees azimuth) 027 140 105 035 Default Information: Surface Roughness Length (m) 0.20 0.20 0.20 0.20 Wind Direction Window (degrees azimuth) 90 90 90 90 Minimum Wind Speed (m/sec) 0.5 0.5 0.5 0.5 Averaging Sector Width Constant 4.3 4.3 4.3 4.3 Initial Diffusion Coefficients (m) 0.0, 0.0 0.0, 0.0 0.0, 0.0 0.0, 0.0.Difference in locai grade elevation between the iv3 I kI -15 T) and tne difference is set to zero.UI k85 TD) IS acco.unted Tor in me release neignt sucn that source/receptor elevation Report No. 14078101-RADR-004-1 158 of 205 Revision I Report No. 14078101 -RADR-004-1 158 of 205 Revision I ýft Diablo Canyon Power Plant Implementation ofAlternative Source Terms Unit 2 LOCA Releases to CR Receptors Receptor ARCON96 Parameter U2 RWST/CR Center U2 Cont. Pen GE / UI U2 Cont. Pen GE / U2 U2 Cont. Pen GE I UI CR Normal Intake CR Normal Intake CR Emergency Intake Case No. 20 45 29 30 Meteorological Information: Period of Meteorological Data 2007 -2011 2007 -2011 2007 -2011 2007 -2011 Lower Measurement Height (m) 10 10 10 10 Upper Measurement Height (m) 76 76 76 76 Wind Speed Units m/sec m/sec m/sec m/sec Meteorological Data File Names: dcpp2007.met-dcpp2007.met-dcpp2007.met-dcpp2007.met-dcpp2011.met dcpp2011.met dcpp2011.met dcpp2011.met Source Information: Release Type ground -point ground -point ground -point ground -point Release Height (m) 26.8 16.8 16.8 16.8 Building Area (min) 215.5 530.4 530.4 530.4 Vertical Velocity m/sec) 0.0 0.0 0.0 0.0 Stack Flow (m_/sec) 0.0 0.0 0.0 0.0 Stack Radius (m) 0.0 0.0 0.0 0.0 Receptor Information: Distance to Receptor (m) 74.9 50.9 26.7 168.0 Intake Height (m) 24.4 22.0 22.0 32.0 Elevation Difference (m) 0.01 0.0 0.0 0.0 Direction to Source (degrees azimuth) 099 127 084 124 Default Information: Surface Roughness Length (m) 0.20 0.20 0.20 0.20 Wind Direction Window (degrees azimuth) 90 90 90 90 Minimum Wind Speed (m/sec) 0.5 0.5 0.5 0.5 Averaging Sector Width Constant 4.3 4.3 4.3 4.3 Initial Diffusion Coefficients (m) 0.0, 0.0 0.0, 0.0 0.0, 0.0 0.0, 0.0 1. UDirerence in local grade elevation difference is set to zero.between the RWVVST (,11 ift) and the UK (86 ft) IS accounted tor In the release height such that source/receptor elevation Report No. 140781 01-RADR-004-1 159 of 205 Revision I Report No. 14078101 -RADR-004-1 159 of 205 Revision I Diablo Canyon Power Plant Implementation of Alternati'e Source Terms Unit 2 LOCA Releases to CR Receptors Receptor ARCON96 Parameter U2 Cont. Pen GE / U2 U2 Cont. Pen GE I U2 PL GW/FW I U1 U2 PL GW/FW I U2 CR Emergency Intake CR Center CR Normal Intake CR Normal Intake Case No. 31 32 46 33 Meteorological Information: Period of Meteorological Data 2007 -2011 2007 -2011 2007 -2011 2007 -2011 Lower Measurement Height (m) 10 10 10 10 Upper Measurement Height (m) 76 76 76 76 Wind Speed Units m/sec m/sec m/sec m/sec Meteorological Data File Names: dcpp2007.met-dcpp2007. met- dcpp2007.met-dcpp2007.met-dcpp2011 .met dcpp20ll.met dcpp2011. met dcpp2011 .met Source Information: Release Type ground -point ground -point ground -point ground -point Release Height (m) 16.8 16.8 16.8 16.8 Building Area (m') 2,744.52 530.4 2,744.5 2,744.5 Vertical Velocity (m/sec) 0.0 0.0 0.0 0.0 Stack Flow (m6/sec) 0.0 0.0 0.0 0.0 Stack Radius (m) 0.0 0.0 0.0 0.0 Receptor Information: Distance to Receptor (m) 125.5 41.0 67.1 34.0 Intake Height (m) 32. D 24.4 22.0 22.0 Elevation Difference (m) 0.0 0.0 0.0 0.0 Direction to Source (degrees azimuth) 025 107 175 194 Default information: Surface Roughness Length (m) 0.20 0.20 0.20 0.20 Wind Direction Window (degrees azimuth) 90 90 90 90 Minimum Wind Speed (m/sec) 0.5 0.5 0.5 0.5 Averaging Sector Width Constant 4.3 4.3 4.3 4.3 Initial Diffusion Coefficients (m) 0.0, 0.0 0.0, 0.0 0.0, 0.0 0.0, 0.0 1. Release/receptor path goes around the Unit 2 CB.Report No. 14078101 -RADR-004-1 160 of 205 Revision I Diablo Canyon Power Plant Implementation ofAlternative Source Terms Unit 2 LOCA Releases to CR Receptors Receptor ARCQN96 Parameter U2 PL GW/FW / U1 U2 PL GW/FW / U2 CR 1U2 PL GW/FW CR Emergency Intake Emergency Intake I CR Center Case No. 34 35 36 Meteorological Information: Period of Meteorological Data 2007 -2011 2007 -2011 2007 -2011 Lower Measurement Height (m) 10 10 10 Upper Measurement Height (m) 76 76 76 Wind Speed Units m/sec m/sec m/sec Meteorological Data File Names: dcpp2007.met -dcpp2007.met, dcpp2007.met dcpp2011.met dcpp2011.met dcpp2011 .met Source Information: 'Release Type ground -point ground -point ground -point Release Height (m) 16.8 16.8 16.8 Building Area (m 2) 2,744.5 2,744.5 2,744.5 Vertical Veloci rm/sec) 0.0 0.0 0.0 Stack Flow (m'/sec) 0.0 0.0 0.0 Stack Radius (m) 0.0 0.0 0.0 Receptor Information: Distance to Receptor (m) 165.5 83.0 48.0 Intake Height (m) 32.0 32.0 24.4 Elevation Difference (m) 0.0 0.0 0.0 Direction to Source (degrees azimuth) 140 012 175 Default Information: Surface Roughness Length (m) 0.20 0.20 0.20 Wind Direction Window (degrees azimuth) 90 90 90 Minimum Wind Speed (m/sec) 0.5 0.5 0.5 Averaging Sector Width Constant 4.3 4.3 4.3 Initial Diffusion Coefficients (m) 0.0, 0.0 0.0, 0.0 0.0, 0.0 Repot N. 1478101-RDR 161of 05 Rvison Report No. 14078101 -RADR-004-1 161 of 205 Revision 1 Diablo Canyon Power Plant Implementation ofAlternative Source Ternms Unit 2 LOCA Releases to TSC Receptors Receptor ARCON96 ParameterI U2 CBI TSC Normal U2 CB/ TSC Center U2 PVITSC Normal U2 PV/ TSC Center Intake Intake Case No. 11 12 13 14 Meteorological Information: Period of Meteorological Data 2007 -2011 2007 -2011 2007 -2011 2007 -2011 Lower Measurement Height (m) 10 10 10 10 Upper Measurement Height (m) 76 76 76 76 Wind Speed Units m/sec m/sec m/sec m/sec Meteorological Data File Names: dcpp2007.met-dcpp2007.met-dcpp2007.met-dcpp2007.met-dcpp2011.met dcpp2011.met dcpp2011.met dcpp2011.met Source Information: Release Type ground -diffuse ground -diffuse ground -point ground -point Release Height (m) 11.0 10.7 74.1 74.1 Building Area (m^) 2,744.5 2,744.5 2,744.5 2,744.5 Vertical Velocity (m/sec) 0.0 0.0 0.0 0.0 Stack Flow (m /sec) 0.0 0.0 0.0 0.0 Stack Radius (m) 0.0 0.0 0.0 0.0 Receptor Information: Distance to Receptor (m) 55.1 54.8 77.4 77.1 Intake Height (m) 11.0 10.7 11.0 10.7 Elevation Difference (m) 0.0 0.0 0.0 0.0 Direction to Source (degrees azimuth) 060 072 060 072 Default Information: Surface Roughness Length (m) 0.20 0.20 0.20 0.20 Wind Direction Window (degrees azimuth) 90 90 90 90 Minimum Wind Speed (m/sec) 0.5 0.5 0.5 0.5 Averaging Sector Width Constant 4.3 4.3 4.3 4.3 Initial Diffusion Coefficients (i) 7.42, 11.07 .7.42, 11.07 0.0, 0.0 0.0, 0.0 Report No. 14078101 -RADR-004-1 162 of 205 Revision 1 I~.Diablo Canyon Power Plant Implementation of Alternative Source Ternms Unit 2 LOCA Releases to TSC Receptors Receptor ARCON96 Parameter U2 RWST/ TSC U2 RWSTITSC U2 Cont. Pen GE I U2 Cont. Pen GE I Normal Intake Center TSC Normal Intake TSC Center Case No. 15 16 17 18 Meteorological Information: Period of Meteorological Data 2007 -2011 2007 -2011 2007 -2011 2007 -2011 Lower Measurement Height (m) 10 10 10 10 Upper Measurement Height (m) 76 76 76 76 Wind Speed Units m/sec m/sec m/sec m/sec Meteorological Data File Names: dcpp2007.met-dcpp2007.met-dcpp2007.met-dcpp2007.met-dcpp2 11 .met dcpp201 1 met dcpp2011.met dcpp2011. met Source Information: Release Type ground -point ground -point ground -point ground -point Release Height (m) 26.8 26.8 16.8 16.8 Building Area (m2) 215.5 215.5. 530.4 530.4 Vertical Velocity (m/sec) 0.0 0.0 0.0 0.0 Stack Flow (m /sec) 0.0 0.0 0.0 0.0 Stack Radius (m) 0.0 0.0 0.0 0.0 Receptor Information: Distance to Receptor (m) 129.4 127.8 104.4 100.8 Intake Height (m) 11.0 10.7 11.0 10.7 Elevation Difference (m) 0.0 0.0 0.0 0.0 Direction to Source (degrees azimuth) 058 064 049 057 Default Information: Surface Roughness Length (m) 0.20 0.20 0.20 D.20 Wind Direction Window (degrees azimuth) 90 90 90 90 Minimum Wind Speed (m/sec) 0.5 0.5 0.5 0.5 Averaging Sector Width Constant 4.3 4.3 4.3 4.3 Initial Diffusion Coefficients (m) 0.0, 0.0 0.0, 0.0 0.0, 0.0 0.0, 0.0 Report No. 14078101 -RADR-004-1 163 of 205 Revision 1 Pýftol Diablo Canyon Power Plant Implementation ofAlternative Source Terms Unit 2 LOCA Releases to TSC Receptors Receptor ARCON96 Parameter U2 PL GWIFW I TSC U2 PL GWIFW I TSC Normal Intake Center Case No. 19 20 Meteorological Information: Period of Meteorological Data 2007 -2011 2007 -2011 Lower Measurement Height (m) 10 10 Upper Measurement Height (m) 76 76 Wind Speed Units m/sec m/sec Meteorological Data File Names: dcpp2007. met -dcpp2007. met -dcpp2011 .met dcpp2011 .met Source Information: Release Type ground -point ground -point Release Height (m) 16.8 16.8 Building Area (inm) 2,744.5 2,744.5 Vertical Velocity (m/sec) 0.0 0.0 Stack Flow (me/sec) 0.0 0.0 Stack Radius (m) 0.0 0.0 Receptor Information: Distance to Receptor (m) 54.5 53.1 Intake Height.(m) 11.0 10.7 Elevation Difference (m) 0.0 0.0 Direction to Source (degrees azimuth) 054 070 Default Information: Surface Roughness Length (m) 0.20 0.20 Wind Direction Window (degrees azimuth) 90 90 Minimum Wind Speed (m/sec) 0.5 0.5 Averaging Sector Width Constant 4.3 4.3 Initial Diffusion Coefficients (m) 0.0, 0.0 0.0, 0.0 Report No. 14078101 -RADR-004-1 164 of 205 Revision I Diablo Canyon Power Plant Implementation ofAlternative Source Terms Unit I FHA Releases to CR Receptors Receptor ARCON96 Parameter UI FHBI UI CR U1 FHBI U2 CR UI FHBI UI CR U1 FHBI U2 CR Normal Intake Normal Intake Emergency Intake Emergency Intake Case No. 1 2. 17 3 Meteorological Information:

Period of Meteorological Data 2007 -2011 2007 -2011 2007 -2011 2007 -2011 Lower Measurement Height (m) 10 10 10 10 Upper Measurement Height (m) 76 76 76 76 Wind Speed Units rm/sec m/sec m/sec m/sec Meteorological Data File Names: dcpp2007.met

-dcpp2007. met -dcpp2007.met-dcpp2007.met-dcpp20ll.met dcpp2011.met dcpp2011.met dcpp2011 .met Source Information: Release Type ground -point ground -point ground -point ground -point Release Height (m) 18.3 18.3 18.3 18.3 Building Area (min) 530.4 530.4 2,744.51 530.4 Vertical Velocity (m/sec) 0.0 0.0 0.0 0.0 Stack Flow (mi/sec) 0.0 0.0 0.0 0.0 Stack Radius (m) 0.0 0.0 0.0 0.0 Receptor Information: Distance to Receptor (m) 26.5 44.7 135.1 165.0 Intake Height (m) 22.0 22.0 32.0 32.0 Elevation Difference (m) 0.0 0.0 0.0 0.0 Direction to Source (degrees azimuth) 067 013 112 013 Default Information: Surface Roughness Length (m) 0.20 0.20 0.20 0.20 Wind Direction Window (degrees azimuth) 90 90 90 90 Minimum Wind Speed (m/sec) 0.5 0.5 0.5 0.5 Averaging Sector Width Constant 4.3 4.3 4.3 4.3 Initial Diffusion Coefficients (m) 0.0, 0.0 0.0, 0.0 0.0, 0.0 0.0, 0.0 1. Release/receptor path goes around the Unit 1 CB.Report No. 14078101-RADR-004-1 165 of 205 Revision I ýft Diablo Canyon Power Plant Implementation of Alternative Source Terms Unit 1 FHA Releases to CR Receptors Receptor ARCON96 Parameter U1 FHB/CR Center UI EHI UI CR UI EHI U2 CR U1 EHI UI CR_ Normal Intake Normal Intake Emergency Intake Case No. 4 5 6 19 Meteorological Information: Period of Meteorological Data 2007 -2011 2007 -2011 2007 -2011 2007 -2011 Lower Measurement Height (m) 10 10 10 10 Upper Measurement Height (m) 76 76 76 76 Wind Speed Units m/sec m/sec m/sec m/sec Meteorological Data File Names: dcpp2007.met-dcpp2007.met-dcpp2007.met -dcpp2007.met-dcpp20l1 .met dcpp2011 .met dcpp201 I.met dcpp20ll .met Source Information: Release Type ground -point ground -point ground -point ground -point Release Height (m) 18.3 20.1 20.1 20.1 Building Area (mi) 530.4 2,744.5 2,744.5 2,744.5 Vertical Velocity (m/sec) 0.0 0.0 0.0 0.0 Stack Flow (mi/sec) 0.0 0.0 0.0 0.0 Stack Radius (m) 0.0 0.0 0.0 0.0 Receptor Information: Distance to Receptor (m) 37.5 14.0 46.8 116.7 Intake Height (m) 24.4 22.0 22.0 32.0 Elevation Difference (m) 0.0 0.0 0.0 0.0 Direction to Source (degrees azimuth) 038 023 349 114 Default Information: Surface Roughness Length (m) 0.20 0.20 0.20 0.20 Wind Direction Window (degrees azimuth) 90 90 90 90 Minimum Wind Speed (misec) 0.5 0.5 0.5 0.5 Averaging Sector Width Constant 4.3 4.3 4.3 4.3 Initial Diffusion Coefficients (m) 0.0, 0.0 0.0, 0.0 0.0, 0.0 0.0, 0.0 Report No. 14078101 -RADR-004-1 166 of 205 Revision I Diablo Canyon Power Plant Implementation ofAlternative Source Terms Unit I FHA Releases to CR Receptors Receptor ARCON96 ParameterI U1 EHI U2 CR U1 EHICR Center__ Emergency Intake Case No. 7 8 Meteorological Information: Period of Meteorological Data 2007 -2011 2007 -2011 Lower Measurement Height (m) 10 10 Upper Measurement Height (m) 76 76 Wind Speed Units m/sec m/sec Meteorological Data File Names: dcpp2007.met-dcpp2007.met-dcpp2011 .met dcpp2011 .met-Source Information: Release Type ground -point ground -point Release Height (m) 20.1 20.1 Building Area (m') 2,744.5 2,744.5 Vertical Velocity (m/sec) 0.0 0.0 Stack Flow (md/sec) 0.0 0.0 Stack Radius (m) 0.0 0.0 Receptor Information: Distance to Receptor (m) 161.9 32.3 Intake Height (m) 32.0 24.4 Elevation Difference (m) 0.0 0.0 Direction to Source (degrees azimuth) 006 008 Default Information: Surface Roughness Length (m) 0.20 0.20-Wind Direction Window (degrees azimuth) 90 90 Minimum Wind Speed (m/sec) 0.5 0.5 Averaging Sector Width Constant 4.3 4.3 Initial Diffusion Coefficients (m) 0.0, 0.0 0.0, 0.0 Report No. 14078101 -RADR-004-1 167 of 205 Revision 1 A -6l, Diablo Canyon Power Plant Implementation of Alternative Source Terms Unit 2 FHA Releases to CR Receptors Receptor ARCON96 Parameter U2 FHB/UI CR U2 FHBI U2 CR U2 FHB[ UI CR U2 FHBI U2 CR Normal Intake Normal Intake Emergency Intake Emergency Intake Case No. 9 10 11 18 Meteorological Information: Period of Meteorological Data 2007-2011 2007-2011 2007-2011 2007-2011 Lower Measurement Height (m) 10 10 10 10 Upper Measurement Height (m) 76 76 76 76 Wind Speed Units m/sec m/sec m/sec m/sec Meteorological Data File Names: dcpp2007.met-dcpp2007.met-dcpp2007.met-dcpp2007.met-dcpp2011 .met dcpp20l1 .met dcpp20l l.met dcpp20l1 .met Source, Information: Release Type ground -point ground -point ground -point ground -point Release Height (m) 18.3 18.3 18.3 18.3 Building Area (m') 530.4 530.4 530.4 2,744.51 Vertical Velocity (m/sec) 0.0 0.0 0.0 0.0 Stack Flow (m /sec) 0.0 0.0 0.0 0.0 Stack Radius (m) 0.0 0.0 0.0 0.0 Receptor Information: Distance to Receptor (m) 44.7 26.5 162.5 135.1 Intake Height (m) 22.0 22.0 32.0 32.0 Elevation Difference (m) 0.0 0.0 0.0 0.0 Direction to Source (degrees azimuth) 121 067 112 022 Default Information: Surface Roughness Length (m) 0.20 0.20 0.20 0.20 Wind Direction Window (degrees azimuth) 90 90 90 90 Minimum Wind Speed (m/sec) 0.5 0.5 0.5 0.5 Averaging Sector Width Constant 4.3 4.3 4.3 4.3 Initial Diffusion Coefficients (m) 0.0, 0.0 0.0, 0.0 0.0, 0.0 0.0, 0.0 1. Release/receptor path goes around the Unit 2 CB.Repot No 1408101RADR004-168of 25 Reisio Report No. 14078101 -RADR-004-1 168 of 205 Revision 1 Diablo Canyon Power Plant Implementation ofAlternative Source Terms Unit 2 FHA Releases to CR Receptors Receptor ARCON96 Parameter U2 FHB/CR Center U2 EHI UI CR U2 EH/ U2 CR U2 EH/ UI CR Normal Intake Normal Intake Emergency Intake Case No. 12 13 14 15 Meteorological Information: Period of Meteorological Data 2007-2011 2007-2011 2007-2011 2007-2011 Lower Measurement Height (m) 10 10 10 10 Upper Measurement Height (m) 76 76 76 76 Wind Speed Units m/sec m/sec m/sec m/sec Meteorological Data File Names: dcpp2007.met-dcpp2007.met-dcpp2007.met-dcpp2007.met-dcpp2011 .met dcpp2011.met dcpp2011.met dcpp2011 .met Source Information: Release Type ground -point ground -point ground -point ground -point Release Height (mn) 18.3 20.1 20.1 20.1 Building Area (m 2) 530.4 2,744.5 2,744.5 2,744.5 Vertical Velocity (m/sec) 0.0 0.0 0.0 0.0 Stack Flow (m /sec) 0.0 0.0 0.0 0.0 Stack Radius (m) 0.0 0.0 0.0 0.0 Receptor Information: Distance to Receptor (m) 37.5 46.5 13.8 161.6 Intake Height (m) 24.4 22.0 22.0 32.0 Elevation Difference (m) 0.0 0.0 0.0 0.0 Direction to Source (degrees azimuth) 097 145 110 128 Default Information: Surface Roughness Length (m) 0.20 0.20 0.20 0.20 Wind Direction Window (degrees azimuth) 90 90 90 90 Minimum Wind Speed (m/sec) 0.5 0.5 0.5 0.5 Averaging Sector Width Constant 4.3 4.3 4.3 4.3 Initial Diffusion Coefficients (m) 0.0, 0.0 0.0, 0.0 0.0, 0.0 0.0, 0.0 Report No. 14D781 01 -RADR-004-1 169 of 205 Revision I Diablo Canyon Power Plant Implementation ofAlternative Source Terms Unit 2 FHA Releases Receptor ARCON96 Parameter U2 EHI U2 CR U2 EHICR Center Emergency Intake Case No. 20 16 Meteorologicall Information: Period of Meteorological Data 2007 -2011 2007 -2011 Lower Measurement Height (m) 10 10 Upper Measurement Height (m) 76 76 Wind Speed Units m/sec m/sec Meteorological Data File Names: dcpp2007.met-dcpp2007.met-dcpp2011.met dcpp2011.met Source Information: Release Type ground -point ground -point Release Height (m) 20.1 20.1 Building Area (mi) 2,744.5 2,744.5 Vertical Velocity (m/sec) 0.0 0.0 Stack Flow (m'fsec) 0.0 0.0 Stack Radius (m) 0.0 0.0 Receptor Information: Distance to Receptor (m) 116.9 32.0 Intake Height (m) 32.0 24.4 Elevation Difference (m) 0.0 0.0 Direction to Source (degrees azimuth) 020 126 Default Information: Surface Roughness Length (m) 0.20 0.20 Wind Direction Window (degrees azimuth) 90 90 Minimum Wind Speed (m/sec) 0.5 0.5 Averaging Sector Width Constant 4.3 4.3 Initial Diffusion Coefficients (m) 0.0, 0.0 0.0, 0.0 Report No. 14078101 -RADR-004-1 170 of 205 Revision I Diablo Canyon Power Plant Implementation ofAlternative Source Terms Unit I MSSVs 110% ADVs / MSL Break Location Releases to CR Receptors Receptor ARCON96 Parameter UI MSSV/U1 CR Ul MSSV / U2 CR U1 MSSV / U1 CR Ul MSSV / U2 CR Normal Intake Normal Intake Emergency Intake Emergency Intake Case No. 1 2 25 3 Meteorological Information: Period of Meteorological Data 2007 -2011 2007 -2011 2007 -2011 2007 -2011 Lower Measurement Height (m) 10 10 10 10 Upper Measurement Height (m) 76 76 76 76 Wind Speed Units m/sec m/sec m/sec m/sec Meteorological Data File Names: dcpp2007.met-dcpp2007.met-dcpp2007.met-dcpp2007.met-dcpp2011 .met dcpp2011 .met dcpp2011 .met dcpp2011 .met Source Information: Release Type ground -point ground -point ground -point ground -point Release Height (m) 27.1 27.1 27.1 27.1 Building Area (m') 0.0 2,744.5 2,744.5 2,744.5 Vertical Velocity (m/sec) 0.0 0.0 0.0 0.0 Stack Flow (m3/sec) 0.0 0.0 0.0 0.0 Stack Radius (m) 0.0 0.0 0.0 0.0 Receptor Information: Distance to Receptor (m) 1.5 37.5 116.6 149.8 Intake Height (m) 22.0 22.0 32.0 32.0 Elevation Difference (m) 0.0 0.0 0.0 0.0 Direction to Source (degrees azimuth) 247 337 121 005 Default Information: Surface Roughness Length (m) 0.20 0.20 0.20 0.20 Wind Direction Window (degrees azimuth) 90 90 90 90 Minimum Wind Speed (m/sec) 0.5 0.5 0.5 0.5 Averaging Sector Width Constant 4.3 4.3 4.3 4.3 Initial Diffusion Coefficients (m) 0.0, 0.0 0.0, 0.0 0.0, 0.0 0.0, 0.0 Report No. 14078101 -RADR-004-1 171 of 205 Revision 1 Diablo Canyon Power Plant Implementation ofAlternative Source Terms Unit I MSSVs 10% ADVs I MSL Break Location Releases to CR Receptors Receptor ARCON96 Parameter U1 MSSV/CR U1 10%ADVs/ Unit U1 10% ADVs / U2 U1 10% ADVs / U1 CR Center 1 CR Normal Intake CR Normal Intake Emergency Intake Case No. 4 9 10 27 Meteorological Information: Period of Meteorological Data 2007- 2011 2007 -2011 2007 -2011 2007 -2011 Lower Measurement Height (m) 10 10 10 10 Upper Measurement Height (m), 76 76 76 76 Wind Speed Units m/sec m/sec m/sec m/sec Meteorological Data File Names: dcpp2007.met-dcpp2007.rmet-dcpp2007.met-dcpp2007.met-dcpp2011. met dcpp2011.met dcpp2011.met dcpp2011.met Source Information: Release Type ground -point ground -point ground -point ground -point Release Height (m) 27.1 26.5 26.5 26.5 Building Area (mi) 2,744.5 2,744.5 2,744.5 2,744.5 Vertical Velocity (mfsec) 0.0 0.0 0.0 0.0 Stack Flow (m"/sec) 0.0 0.0 0.0 0.0 Stack Radius (m) 0.0 0.0 0.0 0.0 Receptor Information: Distance to Receptor (m) 20.5 1.5 37.5 116.6 Intake Height (m) 24.4 22.0 22.0 32.0 Elevation Difference (m) 0.0 0.0 0.0 0.0 Direction to Source (degrees azimuth) 355 247 337 121 Default Information: Surface Roughness Length (m) 0.20 0.20 0.20 0.20 Wind Direction Window (degrees azimuth) 90 90 90 90 Minimum Wind Speed (m/sec) 0.5 0.5 0.5 0.5 Averaging Sector Width Constant 4.3 4.3 4.3 4.3 Initial Diffusion Coefficients in 0.0, 0.0 0.0, 0.0 0.0, 0.0 0.0, 0.0 Report No. 14078101 -RADR-004-1 172 of 205 Revision I II, Diablo Canyon Power Plant Implementation ofAlternative Source Terms Unit I MSSVs /10% ADVs / MSL Break Location Releases to CR Receptors Receptor ARCON96 ParameterII UIIO%ADVs/U2CR U1I10%ADVs/CR U1 MSLB/ U1 CR UI MSLB/ U2 CR Emergency Intake Center Normal Intake Normal Intake Case No. 11 12 17 18 Meteorological Information: Period of Meteorological Data 2007 -2011 2007 -2011 2007 -2011 2007 -2011 Lower Measurement Height (m) 10 10 10 10 Upper Measurement Height (m) 76 76 76 76 Wind Speed Units m/sec m/sec m/sec m/sec Meteorological Data File Names: dcpp2007.met-dcpp2007.met-dcpp2007.met-dcpp2007.met-dcpp2011 .met dcpp20ll.met dcpp2011 .met dcpp20ll.met Source Information: Release Type ground -point ground -point ground -point ground -point Release Height (m) 26.5 26.5 17.7 17.7 Building Area (m n) 2,744.5 2,744.5 0.0 2,744.5 Vertical Velocity (m/sec) 0.0 0.0 0.0 0.0 Stack Flow (me/sec) 0.0 0.0 0.0 0.0 Stack Radius (m) 0.0 0.0 0.0 0.0 Receptor Information: Distance to Receptor (m) 149.8 20.5 1.5 37.5 Intake Height (m) 32.0 24.4 22.0 22.0 Elevation Difference (m) 0.0 0.0 0.0 0.0 Direction to Source (degrees azimuth) 005 355 247 337 Default Information: Surface Roughness Length (m) 0.20 0.20 0.20 0.20 Wind Direction Window (degrees azimuth) 90 90 90 90 Minimum Wind Speed (m/sec) 0.5 0.5 0.5 0.5 Averaging Sector Width Constant 4.3 4.3 4.3 4.3 Initial Diffusion Coefficients (m) 0.0, 0.0 0.0, 0.0 0.0, 0.0 0.0, 0.0 Report No. 14078101 -RADR-OD4-1 173 of 205 Revision 1 Ah Diablo Canyon Power Plant Implementation ofAlternative Source Terms Unit I MSSVs /10% ADVs / MSL Break Location Releases to CR Receptors Receptor ARCON96 Parameter U1 MSLB I UI CR UlI MSLB I U2 CR UI MSLB/CR Emergency Intake Emergency Intake Center Case No. 29 19 20 Meteorological Information: Period of Meteorological Data 2007 -2011 2007 -2011 2007 -2011 Lower Measurement Height (m) 10 10 10 Upper Measurement Height (m) 76 76 76 Wind Speed Units m/sec m/sec m/sec Meteorological Data File Names: dcpp2007.met-dcpp2007.met-dcpp2007.met-dcpp201I .met dcpp2011.met dcpp201I.met Source Information: Release Type ground -point ground -point ground -point Release Height (m) 17.7 17.7 17.7 Building Area (m") 2,744.5 2,744.5 2,744.5 Vertical Velocity (m/sec) 0.0 0.0 0.0 Stack Flow (ms/sec) 0.0 0.0 0.0 Stack Radius (m) 0.0 0.0 0.0 Receptor Information: Distance to Receptor (m) 116.6 149.8 20.5 Intake Height (m) 32.0 32.0 24.4 Elevation Difference (m) 0.0 0.0 0.0 Direction to Source (degrees azimuth) 121 005 355 Default Information: Surface Roughness Length (m) 0.20 0.20 0.20 Wind Direction Window (degrees azimuth) 90 90 90 Minimum Wind Speed (m/sec) 0.5 0.5 0.5 Averaging Sector Width Constant 4.3 4.3 4.3 Initial Diffusion Coefficients (m) 0.0, 0.0 0.0, 0.0 0.0, 0.0 *ReotN.1080-AR00-7 f25Rvso Report No. 14078101 -RADR-OD4-1 174 of 205 Revision I A Diablo Canyon Power Plant Implementation ofAlternative Source Termis Unit 2 MSSVs / 10% ADVs I MSL Break Location Releases to CR Receptors Receptor ARCON96 Parameter U2 MSSVI UI CR U2 MSSV / U2 CR U2 MSSVIUI CR U2 MSSV/U2 CR Normal Intake Normal Intake Emergency Intake Emergency Intake Case No. 5 6 7 26 Meteorological Information: Period of Meteorological Data 2007 -2011 2007 -2011 2007 -2011 2007 -2011 Lower Measurement Height (m) 10 10 10 10 Upper Measurement Height (m) 76 76 76 76 Wind Speed Units m/sec m/sec m/sec m/sec Meteorological Data File Names: dcpp2007.met-dcpp2007.met-dcpp2007.met-dcpp2007.met-dcpp2011 .met dcpp201 I.met dcpp2011 .met dcpp2011 .met Source Information: Release Type ground -point ground -point ground -point ground -point Release Height (m) 27.1 27.1 27.1 27.1 Building Area (m') 2,744.5 0.0 2,744.5 2,744.5 Vertical Velocity (m/sec) 0.0 0.0 0.0 0.0 Stack Flow (m /sec) 0.0 0.0 0.0 0.0 Stack Radius (m) 0.0 0.0 0.0 0.0 Receptor Information: Distance to Receptor (m) 37.5 1.5 149.8 116.6 Intake Height (m) 22.0 22.0 32.0 32.0 Elevation Difference (m) 0.0 0.0 0.0 0.0 Direction to Source (degrees azimuth) 157 247 130 013 Default Information: Surface Roughness Length (m) 0.20 0.20 0.20 0.20 Wind Direction Window (degrees azimuth) 90 90 90 90 Minimum Wind Speed (m/sec) 0.5 0.5 0.5 0.5 Averaging Sector Width Constant 4.3 4.3 4.3 4.3 Initial Diffusion Coefficients (m) 0.0, 0.0 0.0, 0.0 0.0, 0.0 0.0, 0.0 Report No. 14078101 -RADR-004-1 175 of 205 Revision 1 Diablo Canyon Power Plant Implementation of Alternative Source Terms Unit 2 MSSVs 110% ADVs I MSL Break Location Releases to CR Receptors Receptor ARCON96 Parameter U2 MSSV /CR U2 10% ADVs I U1 U2 10% ADVs I U2 U2 10% ADVs I U1 CR Center CR Normal Intake CR Normal Intake Emergency Intake Case No. 8 13 14 15 Meteorological.Information: Period of Meteorological Data 2007 -2011 2007 -2011 2007 -2011 2007 -2011 Lower Measurement Height (m) 10 10 10 10 Upper Measurement Height (m) 76 76 76 76 Wind Speed Units m/sec m/sec m/sec m/sec Meteorological Data File Names: dcpp2007.met-dcpp2007.met-dcpp2007.met -dcpp2007.met-dcpp20ll.met dcpp2011.met dcpp2011.met dcpp2011.met Source Information: Release Type ground -point ground -point ground -point ground -point Release Height (m) 27.1 26.5 26.5 26.5 Building Area (m') 2,744.5 2,744.5 0.0 2,744.5 Vertical Velocity (m/sec) 0.0 0.0 0.0 0.0 Stack Flow (m /sec) 0.0 0.0 0.0 0.0 Stack Radius (m) 0.0 0.0 0.0 0.0 Receptor Information: Distance to Receptor (m) 20.5 37.5 1.5 149.8 Intake Height (m) 24.4 22.0 22.0 32.0 Elevation Difference (m) 0.0 0.0 0.0 0.0 Direction to Source (degrees azimuth) 139 157 247 130 Default Information: Surface Roughness Length (m) 0.20 0.20 0.20 0.20 Wind Direction Window (degrees azimuth) 90 90 90 90 Minimum Wind Speed (m/sec) 0.5 0.5 0.5 0.5 Averaging Sector Width Constant 4.3 4.3 4.3 4.3 Initial Diffusion Coefficients (m) 0.0, 0.0 0.0, 0.0 0.0, 0.0 0.0, 0.0 Report No. 14078101 -RADR-004-1 176 of 205 Revision I1 'pt Diablo Canyon Power Plant Implementation ofAlternative Source Terms Unit 2 MSSVs /10% ADVs I MSL Break Location Releases to CR Receptors Receptor ARCON96 Parameter U2 10% ADVs / U2 CR U2 10% ADVs/CR U2 MSLB/ U1 CR U2 MSLB/ U2 CR Emergency Intake Center Normal Intake Normal Intake Case No. 28 16 21 22 Meteorological Information: Period of Meteorological Data 2007 -2011 2007 -2011 2007 -2011 2007 -2011 Lower Measurement Height (m) 10 10 10 10 Upper Measurement Height (m) 76 76 76 76 Wind Speed Units m/sec m/sec m/sec m/sec Meteorological Data File Names: dcpp2007.met-dcpp2007.met-dcpp2007.met-dcpp2007.met-dcpp2011 .met dcpp201 I.met dcpp2011 .met dcpp2011 .met Source Information: Release Type ground -point ground -point ground -point ground -point Release Height (m) 26.5 26.5 17.7 17.7 Building Area (m') 2,744.5 2,744.5 2,744.5 0.0 Vertical Velocity (m/sec) 0.0 0.0 0.0 0.0 Stack Flow (m-/sec) 0.0 0.0 0.0 0.0 Stack Radius (m) 0.0 0.0 0.0 0.0 Receptor Information: Distance to Receptor (m) 116.6 20.5 37.5 1.5 Intake Height (m) 32.0 24.4 22.0 22.0 Elevation Difference (m) 0.0 0.0 0.0 0.0 Direction to Source (degrees azimuth) 013 139 157 247 Default Information: Surface Roughness Length (m) 0.20 0.20 0.20 0.20 Wind Direction Window (degrees azimuth) 90 90 90 90 Minimum Wind Speed (m/sec) 0.5 0.5 0.5 0.5 Averaging Sector Width Constant 4.3 4.3 4.3 4.3 Initial Diffusion Coefficients (m) 0.0, 0.0 0.0, 0.0 0.0, 0.0 0.0, 0.0 Report No. 14078101 -RADR-004-1 177 of 205 Revision 1 Diablo Canyon Power Plant Implementation of Alternative Source Terins Unit 2 MSSVs /10% ADVs / MSL Break Location Releases to CR Receptors Receptor ARCON96 Parameter I MLB / U1l CR U2 MSLB I U2 CR U2 MSLBICR Emergency Intake Emergency Intake Center Case No. 23 30 24 Meteorological Information: Period of Meteorological Data 2007 -2011 2007 -2011 2007 -2011 Lower Measurement Height (m) 10 10 10 Upper Measurement Height (m) 76 76 76 Wind Speed Units m/sec m/sec m/sec Meteorological Data File Names: dcpp2007.met-dcpp2007.met-dcpp2007.met-dcpp2011. met dcpp2011 met dcpp2011. met Source Information: Release Type ground -point ground -point ground -point Release Height (m) 17.7 17.7 17.7 Building Area (m') 2,744.5 2,744.5 2,744.5 Vertical Velocity (m/sec) 0.0 0.0 0.0 Stack Flow (mrIsec) 0.0 0.0 0.0 Stack Radius (m) 0.0 0.0 0.0 Receptor Information: Distance to Receptor (m) 149.8 116.6 20.5 Intake Height (m) 32.0 32.0 24.4 Elevation Difference (m) 0.0 0.0 0.0 Direction to Source (degrees azimuth) 130 013 139 Default Information: _Surface Roughness Length (m) 0.20 0.20 0.20 Wind Direction Window (degrees azimuth) 90 90 90 Minimum Wind Speed (m/sec) 0.5 0.5 0.5 Averaging Sector Width Constant 4.3 4.3 4.3 Initial Diffusion Coefficients (m) 0.0, 0.0 0.0, 0.0 0.0, 0.0 Report No. 14078101 -RADR-004-1 178 of 205 Revision 1 LI'Diablo Canyon Power Plant Implementation of Alternative Source Terms APPENDIX B: CHANGES TO KEY DESIGN INPUT VALUES (BY ACCIDENT): CLB VS AST As noted in Section 1.0, with this application, and in the interest of evaluating DCPP design against a more realistic accident sequence, as well as in gaining dose analysis margin, the methodology / scenarios used in the following design basis accident (DBA) analyses discussed in the DCPP UFSAR are being updated to reflect the AST guidance provided in RG 1.183.1.2.3.4.5.6.7.Loss of Coolant Accident (LOCA)FHA in the Containment or in the Fuel Handling Building (FHA)Locked Rotor Accident (LRA)Control Rod Ejection Accident (CREA)Main Steam Line Break (MSLB)Steam Generator Tube Rupture (SGTR)Loss-of Load (LOL) Event Appendix B provides a comparison between the design input values used in the current licensing basis (CLB) dose consequence analyses supporting DCPP Units 1 and 2, to those utilized in the AST analyses supporting this application. The basis for the change from the CLB value is also included, as applicable. It is noted that the DCPP CLB assesses Control Room habitability for the LOCA, MSLB, SGTR and FHA. The methodology used to assess the CLB analyses supporting the CREA, LRA and LOL event are DCPP-specific with pre-NUREG-0800 assumptions. In addition, the CLB analyses for the CREA, LRA and LOL only address offsite dose consequences. The AST vs CLB information is provided by accident and in tabular format. The control room parameter values and the offsite atmospheric dispersion factors, both of which are generally applicable to all accidents, are summarized separately. Table No. Subject Table B.1-1 Control Room (CR)Table B. 1-2 Site Boundary Atmospheric Dispersion Factors (y/Q)Table B.2-1 Loss of Coolant Accident (LOCA)Table B.2-2A Limiting CR y/Q for LOCA -AST Values Table B.2-2B CR X/Q's for LOCA -CLB Values Table B.3-1 Fuel Handling Accident (FHA)Table B.3-2A Limiting CR X/Q's for FHA -AST Values Table B.3-2B CR X/Q's for FHA -CLB Values Table B.4-1 Locked Rotor Accident (LRA)Table B.5-1 Control Rod Ejection Accident (CREA)Table B.6-1 Main Steam Line Break (MSLB)Table B.6-2A Limiting CR -/Q's for MSLB -AST Values Table B.6-2B CR I/Q's for MSLB -CLB Values Table B.7-1 Steam Generator Tube Rupture (SGTR)Table B.7-2A Limiting CR X/Q's for SGTR -AST Values Table B.7-2B CR X/Q's for SGTR -CLB Values Table B.8-1 Loss of Load (LOL) Event 179 of 205 IF Diablo Canyon Power Plant Implementation of Alternative Source Terms Table B.1-1 Control Room 1 Changes to Key Input Parameter Values: AST vs CLB Parameter AST Value CLB Value Remark Free Volume 170,000 ft 3 4200 cfm total AST: addresses flow uncertainty Unit 1: 2100 cfm +/- (SGTR)Unfiltered Normal Operation 10%Intake Unit 2: 2100 cfm +/- 2100 cfm 10% (LOCA / MSLB/ FHA)AST: Based on results of 2012 Emergency Pressurization 650 -900 cfm 2100 cfm Tracer Gas.-Test Flow Rate Maximum Unfiltered AST: Based on results of 2012 Backdraft Damper Leakage 100 cfm None Tracer Gas Test during CR Pressurization CLB: Analysis predates modification Operation Carbon / HEPA Filter Flow AST: Based on results of 2012 during CR Pressurization 1800 -2200 cfm 2100 cfm Tracer Gas Test Operation AST: Based on results of 2012 Emergency Filtered 1250 cfm (minimum) 2100 cfm Tracer Gas Test Recirculation Rate Pressurization Intake and 93% (elemental and AST: Based on test acceptance Recirculation Carbon/HEPA 93% (elem 95% (particulates & all criteria (see Section 7.1)Filter Efficiency (includes organic iodine)filter bypass) 98% (particulates) AST: Includes 10 cfm for ingress /Unfiltered Inleakage 70 cfm (m.aximum) 10 cfm egress; Conservative value based on results of 2012 Tracer Gas Test of 37 cfm.0-24 hr (1.0)Occupancy Factors 4 d (0.6)4-30 d (0.4)0-30 d (3.5E-04 0-30 d (3.47E-04 Operator Breathing Rate m3/sec) m 3/sec)Inhalation Dose Conversion Federal Guidance Various: ICRP-30 (FGR Factors Report No.11 11), RG 1.109, ICRP 2 AST: For full list of codes used see Computer Codes used for RADTRAD 3.03, RADTRAD 3.03 -SGTR Section 3.0.Compuser Calculaiod s sARA 3LOCADOSE -LOCA / CLB -the site boundary dose for FHA / MSLB other accidents was developed using EMERALD Note 1: CLB: Control Room dose only assessed for LOCA / MSLB I SGTR / FHA; the FHA dose consequence analysis does not take credit for CRVS Mode 4 operation AST: Control Room dose estimated for all analyzed accidents; i.e., LOCA / MSLB I SGTR I FHA / LRA / CREA / LOL event 180 of 205 Diablo Canyon Power Plant Implementation ofAlternative Source Terms Table B.1-2 Site Boundary Atmospheric Dispersion Factors (X/Q)Changes to Key Input Parameter Values: AST vs CLB_/Q (sec/m 3)Receptor 0 -2 hours 2 -8 hours 8 -24 hours 1 -4da 4-30 days AST Unit 1 EAB (NW) 2.50E-04 Unit 2 EAB (SSE) 2.30E-04 Unit 1/2 LPZ (NW) 2.12E-05 9.26E-06 6.26E-06 2.67E-06 7.86E-07 CLB EAB 5.29E-04 F LPZ 2.20E-05 2.20E-05 4.75E-06 1.54E-06 3.40E-07 Note: AST: Based on RG 1.145, Revision 1 methodology and use of hourly meteorological data; see Section 5.1 for detail.CLB: Based on RG 1.4, Revision I methodology. 181 of 205 46 Diablo Canyon Power Plant Implementation ofAlternative Source Terms Table B.2-1 Loss of Coolant Accident (LOCA)Changes to Key Input Parameter Values: AST vs CLB Parameter AST Value CLB Value Remark Core Power Level (105% of the rated power of 3411 MWth) 3580 MWt Fuel Activity Release Fractions Per Reg. Guide 1.183 Per Reg. Guide 1.4: into containment (see Section 7.2.3.2.6) 100% core noble gases, 25% of core iodine.Fuel Release Timing (gap) Onset: 30 sec Instantaneous Duration: 0.5 hr release Fuel Release Timing (Early-In-Onset: 0.5 hr Vessel) Duration: 1.3 hr Fuel Activity Release Fractions Per Reg. Guide 1.183: 10% of core iodine into sump liquid Same as that released (i.e., 100% of fuel to containment with gap activity) are the exception of the released to the noble gases which sump water escape into the containment atmosphere Core Activity Composite core Composite core Comparison of some core source calculated by source calculated for isotopes (Ci)SCALE4.3 SAS2/ 3.5% and 4.5% AST CLB ORIGEN-S based on a enrichments, and a range of 4.2% to 5.0% burnup range from 1-131 9.90E7 9.760E7 U-235 enrichment, a 0.1 to 50 1-133 2.01E8 1.992E8 19-month averagefuel GWD/MTU, by Kr-85 1.11E6 1.226E6 cycle, and a maximum ORIGEN2 code; Kr-88 6.43E7 1.045E8 core average burnup Activities of some Xe-133 2.01 E8 1.929E8 of 50 GWD/MTU. isotopes are listed in Xe-1 35 4.92E7 5.570E7 See Table 4.1-1 Remark column for comparison. Chemical Form of Iodine released 4.85% elemental 91% elemental AST: Per RG 1.183 from fuel to containment 95% particulate 5% particulate CLB: Per RG 1.4, RI atmosphere 0.15% organic 4% organic Chemical Form of Iodine 97% elemental 99.75% elemental AST: Based on RG 1.183 Released from RCS and sump 3% organic 0.25% organic CLB: Based on SG 25 water Containment Vacuum/Pressure Relief Parameters Minimum Containment Free 2.550E+06 ftý Not Evaluated Volume: Primary Coolant Tech Spec Isotopic activity Activity concentrations corresponding to 1 pCi/gm DE 1-131 and 270 pCi/gm DE Xe-133 182 of 205 A6 Diablo Canyon Power Plant Implementation ofAlternative Source Terms Table B.2-1 Loss of Coolant Accident (LOCA)Changes to Key Input Parameter Values: AST vs CLB Parameter AST Value CLB Value Remark Chemical Form of Iodine 97% elemental; 3%Released organic Maximum RCS flash fraction after LOCA Noble Gases 100%Halogens 40%Maximum containment pressure 218 acfs relief line air flow rate Maximum duration of release via 13 sec containment pressure relief line Release Point Plant Vent Containment Leakage Parameters Containment Spray Coverage -: Iniection and Iniection Spray Recirculation Spray Mode Modes 82.5% (sprayed 82.5% (sprayed fraction) of Total fraction) of volume of 2.55E+06 ft 3 Total volume: 2.55E+06 ft 3 Sprayed Volume Unsprayed Volume 2.103E+06 ft 3 2.103E+06 ft 3 4.470E+05 ft 3 4.470E+05 ft 3 Minimum mixing flow rate from 94,000 cfrn between AST: CFCU flowrate consistent unsprayed to sprayed region: sprayed and with current-containment analysis unsprayed regions Before actuation of 2 unsprayed CFCUs regions/hr (based on natural convection) 9.13 unsprayed After actuation of regions/hr (based on CFCUs 68,000cfm CFCU flow to address surveillance margins and uncertainty) Minimum duration of mixing via Start = 86 sec Start = 0 sec AST: CFCU start and termination CFCUs End = 30 days End = 30 days time consistent with current containment analysis for one train operation. 183 of 205 Diablo Canyon Power Plant Implementation ofAlternative Source Terms Table B.2-1 Loss of Coolant Accident (LOCA)Changes to Key Input Parameter Values: AST vs CLB Parameter AST Value CLB Value Remark Containment spray in injection AST: Injection spray start and mode 111 sec 86.5 sec termination times consistent with Initiation time 3798 sec Not used current containment analysis for Termination time one train operation. Spray removal credit Spray removal credit ends for elemental ends when the iodine when the maximum DF for maximum DF is elemental iodine is reached reached Maximum delay between end of 12 min (based on Not Applicable injection spray and initiation of manual operator recirculation spray action)Containment spray in recirculation Not applicable AST: Spray initiation / termination mode in recirculation mode consistent Initiation time 4518 sec with analysis performed to obtain Termination time 22,518 sec containment conditions if recirculation CS is initiated 12 minutes after injection spray is terminated (1 train operation). Long-term Sump Water pH > 7.5 (includes acid No iodine re-production), thus no evolution iodine revolution Maximum allowable DF for fission Elemental Iodine: 200 Elemental Iodine: CLB: The value of 100 is applied product removal Others: not applicable 100 to the initial release of 25% core Others: not iodine to the containment applicable assuming animmediate 50%plateout. The effective DF is 200 if the initial release fraction of 50%was addressed. Elemental iodine and particulate See Table 7.2-2 Constant spray spray removal coefficients in removal coefficients sprayed region during both as given below injection spray and recirculation spray modes Elemental iodine:31 hr 1 Particulate iodine: 0 hr 1 Elemental iodine removal See Table 7.2-2 None coefficients due to wall deposition Particulate removal coefficients in See Table 7.2-2 None unsprayed region due~to gravitational settling Containment Leak rate (0-24 hr)0.1% weight fraction per day 184 of 205 Diablo Canyon Power Plant Implementation of Alternative Source Terms Table B.2-1 Loss of Coolant Accident (LOCA)Changes to Key Input Parameter Values: AST vs CLB Parameter AST Value CLB Value Remark Containment Leak rate (1-30 day)0.05% weight fraction per day Containment Leakage Release From the worst case Top of Containment. Point (Unfiltered) release point of the following: " Diffuse source via the containment wall" Via Plant Vent" Via Containment Pen Area GE" Via Containment Pen Areas GW &FW ESF System Environmental Leakage Parameters Minimum post-LOCA containment 480,015 gal. 469,000 gal for the AST: Based on current analysis water volume sources ESF leakage source (Min volume of RWST, RCS, SAT and Accumulators) 373,220 gal for the RHR pump seal CLB: Based on RWST, RCS and failure source Accumulators; volumes used are different from current analysis.Minimum time after LOCA when 829 sec 0 sec AST: Based on current recirculation is initiated containment analysis Duration of leakage 30 days Maximum ESF fluid temperature 259.9 OF See DF values AST: Based on current after initiation of recirculation below, containment analysis assuming (used to establish iodine airborne operation of CS in the recirculation fraction) mode.Maximum ESF leak Unfiltered via plant Unfiltered via plant AST: Based on current operational vent = 240 cc/min vent = 1910 cc/hr (- data with margin. Listed values Unfiltered via 32 cc/min) include a factor of 2.Containment Penetration Areas GE A maximum CLB: The maximum additional or GW & FW = 12 additional filtered / filtered / unfiltered leak rates of cc/min unfiltered leak via 1.85 gpm (7003 cc/min) / 0.186 the plant vent of gpm (or 704 cc/min) are derived in 1.85 gpm (7003 the CLB analysis and are each cc/min) / 0.186 gpm based on using all of the dose (or 704 cc/min) margin to the regulatory limit. In addition, none of the CLB leakage values have a "safety factor of 2." 185 of 205 A6 Diablo Canyon Power Plant Implementation ofrAlternative Source Terms Table B.2-1 Loss of Coolant Accident (LOCA)Changes to Key Input Parameter Values: AST vs CLB Parameter AST Value CLB Value Remark RHR pump seal failure Filtered via plant vent 50 gpm starting at t = 24 AST: Releases from the RHR hrs for 30 min Pump Seal failure are filtered for CR dose evaluation and filtered for Site Boundary Dose Evaluation Iodine Airborne Release Fraction 10% (for both RHR DF for RHR pump L.B: Iodine partition credited for pump seal and ESF seal failure release: RHR pump seal failure which is a leakage) 132 large short-term release.DF for ESF leakage: Iodine partition not credited for 1.0 ESF leakage which is a small long-term release.Auxiliary Building ESF Ventilation Elemental iodine: 88% Elemental iodine: AST: AB Ventilation filter credited System filter efficiency Organic iodine: 88% 90% for RHR Pump Seal Failure.(See Section 7.2.3.4) Organic iodine: 70%Particulate iodine: CLB: AB Ventilation filter credited 90% for RHR Pump Seal Failure Refueling Water Storage Tank (RWST) Back-Leakage Parameters Earliest initiation time of RWST 829 sec Not Evaluated back-leakage Maximum ECCS I sump water 2 gpm back-leakage rate to RWST (includes safety factor of 2)RWST back-leakage iodine See Table 7.2-3 release fractions RWST back-leakage'noble gas, See Table 7.2-3 as iodine daughters, release rate from the RWST vent Miscellaneous Equipment Drain Tank (MEDT) Leakage Parameters MEDT inflow rate (includes safety 1900 cc/min Not Evaluated factor of 2)MEDT leakage Iodine release See Table 7.2-4 fractions MEDT leakage noble gas, as See Table 7.2-4 iodine daughters release rate from plant vent CR Emergency Ventilation: Initiation Signal/Timing 186 of 205 M6 Diablo Canyon Power Plant Implementation ofAlternative Source Terms Table B.2-1 Loss of Coolant Accident (LOCA)Changes to Key Input Parameter Values: AST vs CLB Parameter AST Value CLB Value Remark Initiation time (signal) SI signal generated: 6 CLB assumes a 10-sec second closure time Non-Affected Unit for the CRVS NOP Intake Isolated: outside air isolation 18 sec dampers.Affected Unit NOP Intake Isolated and CRVS Mode 4 in full Operation: 44.2 sec Bounding Control Room Table B.2-2A Table B.2-2B Atmospheric Dispersion (Same as Table 7.2-5, Based on a modified Factors for LOCA Refer to Section 5.2 Halitsky for detail) methodology 187 of 205 Diablo Canyon Power Plant Implementation ofAlternative Source Terms Table B.2-2A Loss of Coolant Accident (LOCA)AST Values: Limiting Control Room Atmospheric Dispersion Factors (sec/m 3)Release Location / Receptor 0-2 hr 2-8 hr 8-24 hr 24-96 hr 96-720 hr Control Room Normal Intakes Plant Vent Release-Affected Unit Intake 1.67E-03-Non-Affected Unit Intake 9.10E-04 .................... Containment Penetration Areas-A ff e c t e d U n it I n t a k e 6 .8 4 E -0 3 -----. .... .... ...-N o n -A ff e c te d U n it In ta k e 2 .2 4 E -0 3 -----.. ..........Control Room Infiltration Plant Vent 1.26E-03 8.96E-04 3.44E-04 3.44E-04 2.99E-04 Containment Penetration Areas 3.22E-03 1.85E-03 7.29E-04 7.15E-04 6.64E-04 RWST Vent 1.07E-03 5.80E-04 2.18E-04 2.19E-04 1.79E-04 Control Room Pressurization Intake Plant Vent 5.65E-05 3.70E-05 1.35E-05 1.37E-05 1.11 E-05 Containment Penetration Areas 6.45E-05 4.05E-05 1.65E-05 1.38E-05 1.12E&05 RWST Vent 5.25E-05 3.03E-05 1.15E-05 1.10E-05 8.83E-06 Note 1: Release from the Containment penetration areas (i.e., areas GE or GW & FW): applicable to containment leakage and ESF system leakage that occurs in the Containment Penetration Area Note 2: Release from Plant Vent: applicable to ESF system leakage that occurs in the Auxiliary building, MEDT releases, RHR Pump Seal Failure Release and Containment Vacuum/Pressure Relief Line Release Table B.2-2B Loss of Coolant Accident (LOCA)CLB Values 1: Control Room Atmospheric Dispersion Factors (sec/m 3)Release Location I Receptor 0-8 hr 8-24 -hr 1-4 days 4-30 days Unfiltered inleakage/intake 1.96E-04 1.49E-04 1.08E-04 6.29E-05 Filtered pressurization intake 7.05E-05 5.38E-05 3.91 E-05 2.27E-05 Note 1: The above control room X/Q values are used for all postulated releases and are based on a release point on the top of containment. 188 of 205 Diablo Canyon Power Plant Implementation ofAlternative Source Terms Table B.3-1 Fuel Handling Accident in Fuel Handling Building or Containment (FHA)Changes to Key Input Parameter Values: AST vs CLB Parameter AST Value CLB Value Remark Power Level 3580 MWt Number of Damaged Fuel 1 Assemblies Total Number of Fuel Rods 264 (i.e., all rods in damaged fuel assembly)damaged Decay Time Prior to Fuel 72 hours 100 hrs Movement Radial Peaking Factor 1.65 Fraction of Core Inventory in Kr-85 (30%) Kr-85 (30%)gap Other Noble Gases (10%) Other Noble 1-131 (12%) Gases (10%)Other Halides (10%) All Halides (10%)Alkali Metals (17%)Based on SG 25 Refer to Section 4.3 Equilibrium Fuel Assembly Based on 72 hour decay. Based on 100 Comparison of a single fuel assembly activity (Ci)Activity Refer to Table 7.3-2 hour decay. AST(72 hr decay) CLB(100 hr decay)Activities of some isotopes after 1-131 4.09E5 3.625E5 100 hr decay are 1-133 9.73E4 3.783E4 listed in Remark Kr-85 5.75E3 6.350E3 column for Xe-133 8.31E5 6.914E5 comparison. Iodine form of gap release 99.85% elemental 99.75% AST: Based on RG 1.183 before scrubbing in pool/reactor 0.15% Organic elemental CLB: Based on SG 25 cavity 0.25% Organic Iodine form of gap release after 57% elemental 75% elemental AST: Based on RG 1.183 scrubbing in pool/reactor cavity 43% Organic 25% Organic CLB: Based on SG 25 Pool I reactor cavity scrubbing Iodine (200, effective) AST/CLB: Based on RG 1.183 Decontamination Factors Noble Gas (1)Particulates (oo)Rate of Release from Fuel Puff Environmental Release Points and Rates Environmental Release Rate All airborne activity FHA in FHB: All AST FHA: released within a 2 hour airborne activity period (or less if the released'within a In FHB: Analysis uses the actual release rate ventilation system 2 hour period lambda based on the FHBVS exhaust (i.e., 8.7 hr-1)promotes a faster release since it is larger than the release rate applicable to rate) FHA in 'a 2-hr release" per regulatory guidance (i.e., 3.45 Containment: hr-l).Puff release In Containment: Analysis uses a release rate applicable to "a 2-hr release" FHA in the FHB -Release flow -Plant Vent -46,000 cfm Plant vent -AST: Per Fan curve, 46,000 cfm is maximum flow;rates 40,000 cfm other out leakages are conservative assumptions, FHB Outleakage see Section 7.3-Ingress/Egress -30 cfm-Miscellaneous gaps /189 of 205 Diablo Canyon Power Plant Implementation of Alternative Source Terms Table B.3-1 Fuel Handling Accident in Fuel Handling Building or Containment (FHA)Changes to Key Input Parameter Values: AST vs CLB Parameter AST Value CLB Value Remark openings -470 cfm Minimum free volume in FHB 317,000 435,000 ft" AST: Based on updated assessment of free volume above SFP above the SFP FHA in Containment -Release -Open Equipment Hatch -Open flow rates All airborne activity containment -All released in 2 hrs airborne activity released in 1 second (puff release)Minimum Free Volume in 2,013,000 ftW 33,600 ft6 AST: Based on updated assessment of volume Containment above Operating containment available for dilution of FHA releases Floor volume above fuel pool CLB: Conservative value based on a small rectangular parallelepiped of air space above the 25 ft x 70 ft square feet pool surface in reactor cavity.CR Emergency Ventilation: Initiation Signal/Timing Signal(s) available to switch the Radiation signals from Assumes CR is CRVS from normal operation gamma sensitive intake in normal (NOP) Ventilation (Mode 1) to monitors that initiate ventilation mode, Pressurized Filtered Ventilation closure of the CR normal with unfiltered (Mode 4) following a FHA intake dampers and switch inlet/inleakage the CRVS from normal and exhaust flow operation Ventilation Mode rate of 2110 cfm 1 to Pressurized Filtered for the duration Ventilation Mode 4. of the accident.(Refer to Section 7.3)No LOOP No LOOP (Refer to Section 7.1)Delay time for CRVS Mode 4 22 seconds (see below) Not applicable operation, including monitor response, signal processing, and damper closure time Radiation Monitor Response 10 seconds (conservative Time assumption) -(Refer to Section 7.3)Radiation monitor signal 2 seconds processing time Damper Closure Time 10 seconds Bounding Control Room Table B.3-2A (Same as Table B.3-2B Atmospheric Dispersion Table 7.3-3; Refer to Based on a Factors for FHA Section 5.2 for detail) modified Halitsky methodology 190 of 205 Diablo Canyon Power Plant Implementation ofAlternative Source Terms Table B.3-2A Fuel Handling Accident (FHA)AST Values: Limiting Control Room Atmospheric Dispersion Factors (sec/m 3)Release Location I Receptor 0-22 sec 22 sec -2 hr 2-8 hr 8-24 hr 1-4 d 4-30 d Control Room Normal Intakes Containment Hatch Release-Affected Unit Intake 2.61 E-02 -----........ ..... ....-Non-Affected Unit Intake 2.88E-03 Plant Vent Release-A ffected U nit Intake 1.67E-03 -----.. ... .-Non-Affected Unit Intake 9.1OE-04 -----......... ..... ....FHB Out-leakage points-Affected Unit Intake 6.98E-03 ............... -Non-Affected Unit Intake 2.93E-03 ...Control Room Infiltration C ontainm en t H atch R elease 5 .5 1 E -03 5 .5 1 E --............ Plant Vent 1.26E-03 1.26E-03 .......... ..... .....FHB Out-leakage points 3.78E-03 3.78E-03 -----Control Room Pressurization Intake Containm ent Hatch Release ----- 6.60E-05 -----......... ...P la n t V e n t ----- 5 .6 5 E -0 5 -----.......... ..FHB Out-leakage points 6.40E-05 -----...... Table B.3-2B Fuel Handling Accident (FHA)CLB Values 1: Control Room Atmospheric Dispersion Factors (sec/m 3)Release Location / Receptor 0-8 hr 8-24 hr 1-4 days 4-30 days Unfiltered inleakage/intake 1.96E-04 " -Note 1: The above control room X/Q values are used for all postulated releases and are based on a release point on the top of containment. 191 of 205 ri, Diablo Canyon Power Plant Implementation of Alternative Source Terms Table B.4-1 Locked Rotor Accident (LRA)Changes to Key Input Parameter Values: AST vs CLB Parameter AST Value CLB Value Remark Power Level 3580 MWt 3568 MWt Reactor Coolant Mass 446,486 Ibm Not Available Primary to Secondary SG tube 0.75 gpm at STP (total 1.0 gpm at STP (total leakage for all 4 SGs) for all 4 SGs)Melted Fuel Percentage 0%Failed Fuel Percentage 10%Equilibrium Core Activity Composite core source Calculated by calculated by EMERALD NORMAL SCALE4.3 SAS2I computer program ORIGEN-S based on based on 3568 MWt current enrichment and core power, 3.18%burnup. enrichment, and 12 month fuel cycle.(Refer to Section 4.1 &Table 4.1-1) (Refer to UFSAR Table 11.1-2 & 11.1-4)Radial Peaking Factor 1.65 Not used Fraction of Core Inventory in Fuel 1-131: 12% DCPP specific: gap Kr-85: 30% Gap fractions are Other Noble Gases: isotope dependent. 10% Provided below are Other Halogens: 10% some of the values Alkali Metals: 17% (based on hot channel factor of 1.70): 1-131- 0.822%Kr-85 -16.7%Xe-133 -0.667%(Refer to UFSAR Table 11.1-7)Iodine Chemical Form in Gap 4.85% elemental 100% elemental 95% Particulate 0.15% organic Secondary Side Parameters Initial and Minimum SG Liquid Mass 92,301 Ibm/SG Not Available Iodine Species Released to 97% elemental; 3% 100% elemental Environment organic Time period of tubes being insignificant uncovered 192 of 205 Ah Diablo Canyon Power Plant Implementation of Alternative Source Terms Table B.4-1 Locked Rotor Accident (LRA)Changes to Key Input Parameter Values: AST vs CLB Parameter AST Value CLB Value Remark Steam Releases 0-2 hrs: 651,000 Ibm 0-2 hrs: 656,000 Ibm AST: Based on RSG and current allowable Tavg and Tfeed range 2-8 hrs: 1,023,000 Ibm 2-8 hrs: 1,035,000 Ibm 8-10.73 hrs: same release rate as that for 2-8 hrs Iodine Partition Coefficient in SGs 100 Particulate Carry-Over Fraction in 0.0005 by weight Not Applicable SGs Fraction of Noble Gas Released 1.0 (Released without holdup)Termination of releases from SGs 10.73 hours 8 hrs Environmental Release Point MSSVs/10% ADVs Initial and Minimum SG Liquid Mass 92,301 lbm/SG Not Available Note: No comparison is provided for Control Room parameters since the CLB does not include a dose assessment in the Control room following a LRA.193 of 205 Diablo Canyon Power Plant Implementation ofAlternative Source Terms Table B.5-1 Control Rod Ejection Accident (CREA)Changes to Key Input Parameter Values: AST vs CLB Parameters AST Value CLB Value Remark Containment Leakage Pathway Power Level 3580 MWt 3568 MWt Equilibrium Core Activity Composite core source The core activity is calculated by SCALE4.3 calculated by SAS2/ ORIGEN-S based EMERALD NORMAL on current enrichment and computer program burnup. based on 3568 MWt core power, 3.18%(Refer to Section 4.1 & enrichment, and 12 Table 4.1-1) month fuel cycle.(Refer to UFSAR Table 11.1-2 & 11.1-4)Free Volume 2.55E+06 ft6 Containment leak rate (0 -24 hr) 0.1% vol. fraction per day Containment leak rate(1-30 day) 0.05% vol. fraction per day Failed Fuel Percentage 10%Percentage of Core Inventory in Fuel 10% core noble gases DCPP specific: AST: Per RG 1.183 Gap 10% core halogens Gap fractions are CLB: Plant specific isotopic dependent. Provided below are some of the values are (based on hot channel factor of 1.70): 1-131- 0.822%Kr 16.7%Xe-133 -0.667%(Refer to UFSAR Table 11.1-7)Percentage of fission products released 100% of the noble gases 100% of the noble to coolant that are released to the 100% of the iodines gases containment atmosphere 10% of the halogens Melted Fuel Percentage 0%Chemical Form of Iodine in Failed fuel 4.85% elemental; 95% Not stated AST: Per RG 1.183 Particulate; 0.15% organic Radial Peaking Factor 1.65 Not used Core Activity Release Timing Puff Form of Iodine from failed fuel in the 97% elemental; 3% 100% elemental AST: Per RG 1.183 Containment Atmosphere organic 194 of 205 fi~Diablo Canyon Power Plant Implementation ofAlternative Source Terms Table B.5-1 Control Rod Ejection Accident (CREA)Changes to Key Input Parameter Values: AST vs CLB Parameters AST Value CLB Value Remark Credit taken for containment sprays No Yes Iodine removal coefficients by sprays Not applicable. Same as LOCA Elemental iodine:31 hr-1 Termination of Containment Release 30 days Environmental Release Point Same as LOCA Containment Leakage pathway Secondary Side Pathway Reactor Coolant Mass 446,486 Ibm The assumptions and inputs for the Primary-to-Secondary Leak rate 0.75 gpm for all 4 SGs secondary side Failed Fuel Percentage Same as containment pathway are the leakage pathway same as that described in Locked Percentage of Core Inventory in Fuel Same as containment Rotor Accident (Table Gap leakage pathway B.4-1) for CLB Minimum Post-Accident SG Liquid 92,301 Ibm / SG Mass Iodine Species released to Environment 97% elemental; 3%organic Time period when tubes not totally insignificant submerged Steam Releases 0-2 hrs: 651,000 Ibm 2-8 hrs: 1,023,000 lbm-8-10.73 hrs: same release rate as that for 2-8 hrs. _Iodine Partition Coefficient in SGs 100 Fraction of Noble Gas Released 1.0 (Released without holdup)Termination of Release from SGs 10.73 hours Environmental Release Point MSSVsI10% ADVs Note: No comparison is provided for Control Room parameters since the CLB does not calculate a dose assessment in the Control room following a CREA. A qualitative statement made in the current UFSAR indicates that since the activity releases following a CREA will be less than a LOCA, the CR dose will be below GDC 19.195 of 205 195 of 205 169L;hý%0'MY Diablo Canyon Power Plant Implementation of Alternative Source Terms TABLE B.6-1 Main Steam Line Break (MSLB)Changes to Key Input Parameter Values: AST vs CLB Parameter AST Value CLB Value Remark Power Level 3580 MWt Reactor Coolant Mass 446,486 Ibm 566,000 Ibm AST: A smaller RCS mass is more conservative for the accident-initiated iodine spike (AIS) case. RCS mass is not sensitive to the dose consequences for the pre-accident iodine spike (PIS) case because the amount of RCS loss due to SG leakage is small compared to the total inventory. Leak rate to Faulted Steam 0.75 gpm at STP The calculated Generator (conservative maximum allowable assumption) accident induced primary to secondary leak rate of 10.5 gpm based on Alternate Repair Criteria.Leak rate to Intact Steam 0 gpm (all leakage 0.3125 gpm (total for 3 CLB: TS 3.4.13d allows a maximum Generators assumed into faulted SG) intact SGs) operational leakage of 150 gpd per SG, which is equivalent to 0.3125 gpm for 3 SGs.Failed/Melted Fuel Percentage 0%RCS Tech Spec Iodine Conc. 1 pCi/gm DE 1-131 RCS Tech Spec Noble Gas 270 pCi/gm DE Xe-133 RCS NG activity is AST: The AST NG values correspond to Conc. based on 1% fuel 0.5% fuel defects.defects.RCS Equilibrium Iodine Fuel to RCS appearance Appearance rate that Comparison of 1-131 appearance rate: Appearance Rates rate that results in 1 results in 1 pCi/gm DE pCi/gm DE 1-131, based 1-131, based on 132 AST-4.31E-01 Ci/min on 132-gpm letdown flow gpm letdown flow rate, CLBB- 4.29E-01 Ci/min rate, 100% ion-exchanger 100% ion-exchanger efficiency and 11 gpm efficiency, 11 gpm RCS leakage RCS leakage, I gpm boron control shim bleed, and 1 gpm tube leakage Pre-Accident iodine Spike 60 pCi/gm DE 1-131 Concentrations Accident-Initiated Iodine Spike 500 times equilibrium appearance rate Appearance Rate Duration of Accident-Initiated 8 hours Iodine Spike Initial Secondary Coolant Iodine 0.1 pCi/gm DE 1-131 Concentrations 196 of 205 Diablo Canyon Power Plant Implementation ofAlternative Source Terms TABLE B.6-1 Main Steam Line Break (MSLB)Changes to Key Input Parameter Values: AST vs CLB Parameter AST Value CLB Value Remark Secondary System Release Parameters Iodine Species released to 97% elemental; 3% Not specified Since the CR filter efficiency is the same for Environment organic the elemental iodine and the organic iodine, this parameter value is inconsequential for CR dose.Fraction of Iodine Released from Faulted SG 1.0 (Released to Environ without holdup)Fraction of Noble Gas Released from Faulted SG 1.0 (Released to Environ without holdup)Liquid mass in each SG Faulted: 182,544 Ibm Faulted: 162,784 Ibm AST: Uses maximum mass for Faulted SG (max.) (Hot Zero Power value, RSG / current allowable Tavg and Tfeed range) to maximize Intact: 92,301 Ibm (min. Intact: 81,500 Ibm I dose consequences of release from faulted and initial) SG SG.The initial SG liquid mass for the intact SGs based on RSG I current allowable Tvg and Tfeed range. The initial SG liquid mass is used to determine the total iodine inventory in the SG liquid prior to the accident. The SG liquid mass increases following a reactor trip, so the minimum SG liquid mass post-accident is also the SG initial liquid mass.The minimum post-accident SG liquid mass is used to determine the iodine activity release rate during the accident.Release Rate of SG liquid from Dryout of SG liquid Instantaneous dryout AST: Dryout time based on RSG / current Faulted SG withinl0 seconds allowable T.,, and Tfeed range.Termination of 0.75 gpm leak 30 hrs (when RCS 8 hrs (assumption) primary to secondary leak from reaches 212 'F)Faulted SG Steam Releases from intact 0-2 hrs: 384,000 Ibm 0-2 hrs: 393,464 Ibm AST: Based on RSG I current allowable T,,g SGs 2-8 hrs: 893,000 Ibm 2-8 hrs: 915,000 Ibm and Tfeed range 8-10.73 hrs: Same release rate as that for 2-8 hrs Iodine Partition Coefficient in 100 (SGs fully covered) No iodine partition Intact SG factor is credited.Termination of release from 10.73 hours (calculated 8 hrs (assumption) Intact SG time for initiation of shutdown cooling)Release Point: Faulted SG Outside containment, at the steam line break location 197 of 205 A-1 9- *ORE.)Diablo Canyon Power Plant Implementation ofAlternative Source Terms TABLE B.6-1 Main Steam Line Break (MSLB)Changes to Key Input Parameter Values: AST vs CLB Parameter AST Value CLB Value Remark Release Point: Intact SG MSSVs/10% ADVs CR Emergency Ventilation: Initiation Signal/Timing Initiation (signal) SIS / Phase A Unaffected Unit CRVS inlet Within 12.6 seconds Conservatively damper fully closed assumes that the Affected Unit CRVS inlet Within 38.8 seconds control room is dampers fully closed isolated in 2 minutes Control Room Atmospheric Table B.6-2A (Same as Table B.6-2B (based Dispersion Factors Table 7.6-2; Refer to on a modified Halitsky Section 5.2-for detail) methodology) 198 of 205 Ah Diablo Canyon Power Plant Implementation ofAlternative Source Terms Table B.6-2A Main Steam Line Break (MSLB)AST Values: Limiting Control Room Atmospheric Dispersion Factors (sec/m 3)Receptor -Release Point 0-2hr 2-8 hr 8-10.73 hr 10.73-30 hr CR NOP Intake -Faulted SG (Break Location) Note 1 CR NOP Intake -Intact SG (MSSVs/10% ADVs) 8.60E-04 CR Inleakage -Faulted SG (Break Location) 1.24E-02 7.35E-03 3.01E-03 3.01 E-03 CR Inleakage -Intact SG (MSSVs/1 0% ADVs) 2.78E-03 1.63E-03 1.63E-03 -----CR Emergency Intake & Bypass -Faulted SG (Break 7.65E-05 4.78E-05 1.86E-05 1.86E-05 Location) I CR Emergency Intake & Bypass -Intact SG (MSSVs/10% 1.57E-05 9.60E-06 9.60E-06 ADVs)Note 1: ARCON96 based X/Q s are not applicable for this case Table B.6-2B Main Steam Line Break (MSLB)CLB Values': Control Room Atmospheric Dispersion Factors (sec/m 3)Receptor 0-8 hr 8-24 hr 1-4 days 4-30 days Unfiltered inleakage/intake 1.96E-04 1.49E-04 1.08E-04 6.29E-05 Filtered pressurization intake 7.05E-05 5.38E-05 3.91 E-05 2.27E-05 Note 1: The above control room X/Q values are used for all postulated releases and are based on a release point on the top of containment 199 of 205 Diablo Canyon Power Plant *Implementation ofAlternative Source Terms Table B.7-1 Steam Generator Tube Rupture (SGTR)Changes to Key Input Parameter Values AST vs CLB Parameter AST Value CLB Value Remark Power Level 3580 MWt Reactor Coolant Mass 446,486 Ibm 499,500 Ibm AST: A smaller RCS mass is more conservative for the accident-initiated iodine spike (AIS) case. A larger RCS mass is more conservative for the pre-accident iodine spike (PIS) case because the integrated RCS break flow is significant compared to the initial inventory. AIS case is the more limiting case relative to the regulatory limits.Time of Reactor Trip 179.0 sec Time of isolation of stuck-open 10% 2653 sec ADV on the Ruptured SG Termination of Break Flow from 3402 sec Ruptured SG that flashes Termination of Break Flow from 5872 sec Ruptured SG Time of manual depressurization of 2 hours the Ruptured SG Break Flow to Ruptured Steam Table 7.7-2, Column "A" Generator that flashes Break Flow to Ruptured Steam Table 7.7-2, Column "B" Generator that does not flash Tube Leakage rate to Intact Steam 0.75 gpm at STP (for 1.0 gpm at STP (for 3 Generators 3 intact SGs) intact SGs)Failed/Melted Fuel.Percentage 0%.RCS Tech Spec Iodine 1 pCi/gm DE 1-131 Concentration RCS Tech Spec Noble Gas 270 pCi/gm DE Xe- RCS NG activity is AST: The AST NG values correspond to Concentration 133 based on 1% fuel 0.5% fuel defects.defects.RCS Equilibrium Iodine Fuel to RCS Appearance rate that Comparison of 1-131 appearance rate: Appearance Rates appearance rate that results in 1 pCi/gm DE results in 1 pCi/gm 1-131, based on 132 AST-4.31E-01 Ci/min DE 1-131, based on gpm letdown flow rate, CLB -4.41 E-01 Ci/min 132 gpm letdown flow 100% ion-exchanger rate, 100% ion- efficiency and 11 gpm exchanger efficiency RCS leakage and 11 gpm RCS leakage Pre-Accident Iodine Spike 60 pCi/gm DE 1-131 Concentration 200 of 205 MEW Diablo Canyon Power Plant Implementation ofAlternative Source Terms Table B.7-1 Steam Generator Tube Rupture (SGTR)Changes to Key Input Parameter Values AST vs CLB Parameter AST Value CLB Value Remark Accident-Initiated Iodine Spike 335 times TS equilibrium appearance rate Appearance Rate Duration of Accident-initiated Iodine 8 hours_Spike Initial Secondary Coolant Iodine 0.1 pCi/gm DE 1-131 Concentrations Secondary System Release Parameters Initial SG liquid mass 89,707 Ibm/SG 106,000 Ibm for CLB: Values reflect average SG mass ruptured SG during the transient. 118,500 Ibm for each AST: Value based on current analysis of the 3 intact SGs supporting RSG i current allowable T.,v and Tfeed range. The initial SG liquid mass is used to determine the total iodine inventory in the SG liquid prior to the.accident Iodine Species released to 97% elemental; 3% Not specified Since the CR filter efficiency is the same Environment organic for the elemental iodine and the organic iodine, this parameter value is inconsequential for CR dose.Steam flow rate to condenser from 63,000 Ibm/min Ruptured SG before trip Steam flow rate to condenser from 189,000 Ibm/min intact SGs before trip Partition Factor in Main Condenser 0.01 (elemental iodine)1 (organic iodine and noble gases)Steam Releases from Ruptured SG Table .7.7-2, Column "C" Steam Releases from intact SG Table 7.7-2, Column "D" Post-accident minimum SG liquid 89,707 Ibm 106,000 Ibm CLB value is the average of the initial SG mass for Ruptured SG mass (89,707 Ibm) and the minimum mass (122,500 Ibm) during the stuck open 10% ADV phase of the transient. AST: The minimum post-accident SG liquid mass is used to determine the iodine activity release rate during the accident. The SG liquid mass increases following a reactor trip, so the minimum SG liquid mass post-accident is also the SG initial liquid mass 201 of 205 Diablo Canyon Power Plant Implementation ofAlternative Source Terms Table B.7-1 Steam Generator Tube Rupture (SGTR)Changes to Key Input Parameter Values AST vs CLB Parameter AST Value CLB Value Remark Post-accident minimum SG liquid 89,707 Ibm per SG 118,500 Ibm per SG CLB: value is the average of the initial SG mass for intact SGs mass (89,707 Ibm) and the minimum mass (147,400 Ibm) during the transient at the end of the cooldown.AST: The minimum post-accident SG liquid mass is used to determine the iodine activity release rate during the accident. The SG liquid mass increases following a reactor trip, so the minimum SG liquid mass post-accident is also the SG initial liquid mass Time period of tube uncovery for insignificant intact SG Fraction of Iodine Released 1.0 (Released without holdup)(flashed portion)Fraction of Noble Gas Released 1.0 (Released without holdup)from all SGs Iodine Partition Coefficient 100 Termination of Release from intact 10.73 hrs (calculated 8 hrs (assumed)SG time for initiation of shutdown cooling)Environmental Release Points Plant Vent : 0 -179 Condenser exhaust: sec 0 -179 sec MSSVs/10% ADVs: MSSVs/10%179 sec- 10.73 hr ADVs:179 sec- 8 hr CR emergency Ventilation Initiation Signal/Timing Initiation time (signal) SIS: 219 sec SIS: 215 sec Unaffected Unit inlet Unaffected Unit &damper closed: 231 Affected Unit inlet sec damper closed: 250 Affected Unit inlet sec damper closed: 257.2 sec Control Room Atmospheric Table B.7-2A (Same Table B.7-2B (based Dispersion Factors as Table 7.7-3; Refer on a modified Halitsky to Section 5.2 for methodology) detail)202 of 205 0@0 Diablo Canyon Power Plant Implementation of Alternative Source Terms TABLE B.7-2A Steam Generator Tube Rupture (SGTR)AST Values: Limiting Control Room Atmospheric Dispersion Factors (sec/m 3)Release Location / Receptor 0-179 s 179-257.2s 257.2 s- 2 h 2-8 hr 8-10.73 hr Control Room Normal Intakes Plant Vent 1.29E-03 .................... MSSVs/10% ADVs (Note 1) ----- 8.60E-04 ---Control Room Infiltration Plant Vent 1.26E-03 -----MSSVs/10% ADVs .... 2.78E-03 2.78E-03 1.49E-03 1.49E-03 Control Room Pressurization Intake MSSVs/10% ADVs ----- 1.57E-05 7.65E-06 7.65E-06 Note 1: Due to the proximity of the release from the MSSVs/10% ADVs, to the normal operation CR intake of the affected unit, and due to the high vertical velocity of the steam discharge from the MSSVs/10% ADVs, the resultant plume from the MSSVs/10% ADVs will not contaminate the normal operation CR intake of the affected unit. Thus the X/Q s presented reflect those applicable to the CR intake of the unaffected unit.Table B.7-2B Steam Generator Tube Rupture (SGTR)CLB Values 1: Control Room Atmospheric Dispersion Factors (sec/m 3)Receptor 0-8 hr 8-24 hr 1-4 days 4-30 days Unfiltered inleakage/intake 1.96E-04 1.49E-04 1.08E-04 6.29E-05 Filtered pressurization intake 7.05E-05 5.38E-05 3.91E-05 2.27E-05 Note 1: The above control room X/Q values are used for all postulated releases and are based on a release point on the top of containment. 203 of 205 Diablo Canyon Power Plant Implementation ofAlternative Source Terms Table B.8-1 Loss of Load (LOL) Event Changes to Key Input Parameter Values: AST vs CLB Parameter AST Value CLB Value Remark Power Level 3580 MWt 3568 MWt Reactor Coolant Mass 446,486 Ibm Not Available Primary to Secondary SG tube 0.75 gpm at STP (Four 1.0 gpm at STP (Four leakage SGs) SGs)Failed/Melted Fuel Percentage 0%RCS Technical Specification Iodine 1 pCi/gm DE 1-131 Design basis RCS (1%Levels fuel cladding defects)RCS Technical Specification Noble 270 pCi/gm DE Xe-133 noble gas and iodine Gas Levels activity RCS Equilibrium Iodine Fuel to RCS appearance DCPP specific Appearance Rates rate that results in 1 pCi/gm DE 1-131, based on 132 gpm letdown flow rate, 100% ion-exchanger efficiency and 11 gpm RCS leakage Accident-Initiated Iodine Spike 500 times TS equilibrium 30 times the Appearance Rate appearance rate appearance rate of normal operation Pre-Accident Iodine Spike 60 pCi/gm DE 1-131 Not evaluated Concentration Duration of Accident-Initiated 8 hrs Iodine Spike Initial Secondary Coolant Iodine 0.1 pCi/gm DE 1-131 Based on .1% fuel Concentrations defects and 1 gpm Primary to secondary Leakage Initial and Minimum SG Liquid 92,301 Ibm/SG Not Available Mass Time period of tubes uncovered insignificant Steam Releases 0-2 hrs: 651,000 Ibm 0-2 hrs: 656,000 Ibm AST: Based on RSG and current allowable 2-8 hrs: 1,023,000 Ibm Tavg and Tfeed range 8-10.73 hrs: same 2-8 hrs: 1,035,000 Ibm release rate as that for 2-8 hrs Iodine Partition Coefficient in SGs 100 Iodine Species Released to 97% elemental; 3% 100% elemental Environment organic Fraction of Noble Gas Released 1.0 (Released without holdup)204 of 205 ri~Diablo Canyon Power Plant Implementation of Alternative Source Terms Table B.8-1 Loss of Load (LOL) Event Changes to Key Input Parameter Values: AST vs CLB Parameter AST Value CLB Value Remark Termination of releases from SGs 10.73 hours 8 hrs Environmental Release Point MSSVs/1 0% ADVs Note: No comparison is provided for Control Room parameters since the CLB does not include a dose assessment in the Control room following a Loss of Load Event 205 of 205 Enclosure Attachment 5 PG&E Letter DCL-15-069 Attachment 5 Regulatory Guide 1.183 Conformance Tables In Table A5 -A through A5 -H, the text shown in "RG Position" columns is taken from RG 1.183; therefore, references to footnotes, tables, and numbered references, may be found in .RG 1.183. Only Pressurized Water Reactor items are addressed. References in the "Comments" columns are specific to this License Amendment Request.NOTE: Table A5 -A Table A5 -B Table A5 -C Table A5 -D Table A5 -E Table A5 -F Table A5 -G Table A5 -H Conformance with Regulatory Guide 1.183 Main Sections Conformance with Regulatory Guide 1.183 Appendix A (Loss-of-Coolant Accident)Conformance with Regulatory Guide 1.183 Appendix B (Fuel Handling Accident)Conformance with Regulatory Guide 1.183 Appendix E (PWR Main Steam Line Break)Conformance with Regulatory Guide 1.183 Appendix F (PWR Steam Generator Tube Rupture Accident)Conformance with Regulatory Guide 1.183 Appendix G (PWR Locked Rotor Accident)Conformance with Regulatory Guide 1.183 Appendix H (PWR Rod Ejection Accident)Conformance with Regulatory Guide 1.183 Appendix I (Equipment Qualification) -TbeA5-A: Conformance -with Regulatory Guide 1.183,MaimiSdections SetiriRG,!Position. ~Analysjs f Com~ments, -3. -Accident Source Term 3.1 The inventory of fission products in the reactor core and available Conforms The licensed power level for both units is¶1 for release to the containment should be based on the maximum 3411 MWt (DPR-80, DPR-82). Analyzed full power operation of the core with, as a minimum, current power level is 3580 MWt, approximately licensed values for fuel enrichment, fuel burnup, and an assumed 105% of 3411 MWt. The maximum core core power equal to the current licensed rated thermal power average burnup used is 50 GWD/MTU, times the ECCS evaluation uncertainty. 8 The period of irradiation which is the maximum core average should be of sufficient duration to allow the activity of dose- burnup used in previous consequence significant radionuclides to reach equilibrium or to reach analyses. (Att. 4, Section 4.0, and Table maximum values.9 The core inventory should be determined using 4.1-1, Table B.2-1)an appropriate isotope generation and depletion computer code such as ORIGEN 2 (Ref. 17) or ORIGEN-ARP (Ref. 18). Core ORIGEN-S is used to calculate the DCPP inventory factors (Ci/MWt) provided in TID 14844 and used in core inventory. The ORIGEN-S calculation some analysis computer codes were derived for low burnup, low is performed for over 800 isotopes by enrichment fuel and should not be used with higher burnup and utilizing the Control Module SAS2 of the higher enrichment fuels. SCALE 4.3 computer code package.SAS2/ORIGEN-S has been used in prior 8 The uncertainty factory used in determining the core inventory AST licensing applications. (Att. 4, should be that value provided in Appendix K to 10 CFR Part 50, Section 3.0)Typically 1.02.9 Note that for some radionuclides, such as Cs-I 37, equilibrium will not be reached prior to fuel offload. Thus, the maximum inventory at the end of life should be used.A5-1 Trable A5-A: Conformance with Regulatory Guide,1483 Main Sections RG. -.DCPP Section -RG Po-sition A~nAlysis~ Comments 3.1 For the DBA LOCA, all fuel assemblies in the core are assumed Conforms For the DBA LOCA, all fuel assemblies in¶2 to be affected and the core average inventory should be used. For the core are assumed to be affected and DBA events that do not involve the entire core, the fission product the core average inventory is used for inventory of each of the damaged fuel rods is determined by dose consequences. dividing the total core inventory by the number of fuel rods in the core. To account for differences in power level across the core, For DBA events that do not involve the radial peaking factors from the facility's core operating limits entire core, the fission product inventory of report (COLR) or technical specifications should be applied in each of the damaged fuel rods is determining the inventory of the damaged rods. determined by: 1) dividing the total core inventory by the number of fuel rods in the core, 2) multiplying by the resultant core average inventory per rod by the total number of damaged rods, and 3)multiplying the resultant total damaged rod inventory by a core radial peaking factor 1.65 from the COLR.(Att. 4 Sections 7.3, 7.4, and 7.5)3.1 No adjustment to the fission product inventory should be made for Conforms No adjustments for less than full power¶3 events postulated to occur during power operations at less than operation are made in any analysis.full rated power or those postulated to occur at the beginning of core life. For events postulated to occur while the facility is shutdown, e.g., a fuel handling accident, radioactive decay from the time of shutdown may be modeled.A5-2 T~able- A5"A, Q-onformanlce-with Regulat6ry Guide 1.18 3MainSections -CinRG Pqftion, $ ';~KT 3JjU sis.y 'Co~mments, 3.2 The core inventory release fractions, by radionuclide groups, for Conforms The release fractions from Regulatory ¶1 the gap release and early in-vessel damage phases for DBA Position 3.2, Table 2 are used. Footnote LOCAs are listed in Table 1 for BWRs and Table 2 for PWRs. 10 criterion is met in that peak fuel rod These fractions are applied to the equilibrium core inventory burnup is limited to 62,000 MWD/MTU.described in Regulatory Position 3.1. (Footnote 10 applies to The equilibrium core average isotopic entire RG Section 3.2.) inventory that meets regulatory Position 3.1 was used for LOCA.RG 1.183, Table 2 PWR Core Inventory Fraction Released Into Containment (Att. 4 Section 7.2.3.2.6, Table 7.2-1)Early Gap In-Release Vessel Group Phase Phase Total Noble Gases 0.05 0.95 1.0 Halogens 0.05 0.35 0.4 Alkali Metals 0.05 0.25 0.3 Tellurium Metals 0.00 0.05 0.05 Ba, Sr 0.00 0.02 0.02 Noble Metals 0.00 0.0025 0.0025 Ceruim Group 0.00 0.0005 0.0005 Lanthanides 0.00 0.0002 0.0002 1"The release fractions listed here have been determined to be acceptable for use with currently approved LWR fuel with a peak burnup up to 62,000 MWD/MTU. The data in this section may not be applicable to cores containing mixed oxide (MOX) fuel.A5-3 ______Table A5-A: Conformance withRegulatory.' Guide 1A183 Main Sections RG 1DP Secqtion RG Po~sitio-n -Ana~lysis [Comm~ents_____ 3.2¶2 For non-LOCA events, the fractions of the core inventory assumed to be in the gap for the various radionuclides are given in Table 3. The release fractions from Table 3 are used in conjunction with the fission product inventory calculated with the maximum core radial peaking factor.RG 1.183, Table 31 Non-LOCA Fraction of Fission Product Inventory in Gap Group Fraction 1-131 0.08 Kr-85 0.10 Other Noble Gases 0.05 Other Halogens 0.05 Alkali Metals 0.12"The release fractions listed here have been determined to be acceptable for use with currently approved LWR fuel with a peak burnup up to 62,000 MWD/MTU provided that the maximum linear heat generation rate does not exceed 6.3 kw/ft peak rod average power for burnups exceeding 54 GWD/MTU. As an alternative, fission gas release calculations performed using NRC-approved methodologies may be considered on a case-by-case basis. To be acceptable, these calculations must use a projected power history that will bound the limiting projected plant-specific power history for the specific fuel load. For the BWR rod drop accident and the PWR rod ejection accident, the gap fractions are assumed to be 10% for iodines and noble gases.Exceeds/Conforms To support flexibility in future DCPP fuel management schemes, the gap fraction used for the non-LOCA events (with the exception of the CREA) are based on the bounding (higher) values provided per isotope/isotope class in the following: RG 1.25, NUREG/CR 5009, Section 3.2.2, RG 1.183, Table 3.The CREA gap fractions are assumed to be 10% for iodines and noble gases, as provided in RG 1.183, Appendix H.(Att. 4, Section 2.1 and Section 4.3)A5-4 Table A5--A: Conformance with Regulatory' v~ide 1.1 88,Main Sections RG. -IDCIPIP Section. RG 'Position IAnalyi cm ets 3.3 Table 4 tabulates the onset and duration of each sequential Conforms The core inventory release timing for gap¶1 release phase for DBA LOCAs at PWRs and BWRs. The releases and early in-vessel releases from specified onset is the time following the initiation of the accident Regulatory Position 3.3, Table 4 are used (i.e., time = 0). The early in-vessel phase immediately follows the in the DBA LOCA. The activity released gap release phase.1 2 The activity released from the core during from the core during each release phase is each release phase should be modeled as increasing in a linear modeled as increasing in a linear fashion fashion over the duration of the phase. For non-LOCA DBAs in over the duration of the phase.which fuel damage is projected, the release from the fuel gapand the fuel pellet should be assumed to occur instantaneously with (Att. 4 Section 7.2 and Table 7.2-1)the onset of the projected damage.For non-LOCA events in which fuel Table 4 damage is projected (FHA, LRA, and LOCA Release Phases CREA), the release from the fuel gap is PWR assumed to occur instantaneously. Phase Onset Duration Gap Release 30 sec 0.5 hr (Att. 4 Sections 7.3, 7.4, and 7.5)Early In-Vessel 0.5 hr 1.3 hr 1 2 1n lieu of treating the release in a linear ramp manner, the activity for each phase can be modeled as being released instantaneously at the start of that release phase, i.e., in step increases 3.3 For facilities licensed with leak-before-break methodology, the Conforms DCPP does not take credit for the leak-¶2 onset of the gap release phase may be assumed to be 10 before-break delay in the accident minutes. A licensee may propose an alternative time for the sequence and the values from RG 1.183 onset of the gap release phase, based on facility-specific Table 4 are used.calculations using suitable analysis codes or on an accepted topical report shown to be applicable to the specific facility. In the (Att. 4 Section 7.2 and Table 7.2-1)absence of approved alternatives, the gap release phase onsets in Table 4 should be used.A5-5 Table ~A5-A: confobrmanhce with'Rogulatorly vide 1.183 Main Sections.RG DCPP~Secqtion .P, Pos~itipm .Analysis, Conimeonts> 3.4 Table 5 lists the elements in each radionuclide group that should Conforms The elements in each radionuclide group be considered in design basis analyses. from Regulatory Position 3.4, Table 5, are included the DCPP Equilibrium Core Table 5 Inventory. Radionuclide Groups Group Elements To determine the total effective dose Nobel Gases Xe, Kr equivalent (TEDE) resulting from Halogens I, Br inhalation and submersion following a Alkali Metals Cs, Rb LOCA, the DCPP LOCA dose Tellurium Te, Sb, Se, Ba, Sr consequence analysis uses the default Group group of 60 isotopes provided with Nobel Metals Ru, Rh, Pd, Mo Tc, Co computer code RADTRAD 3.03 plus 13 Lanthanides La, Zr, Nd, Eu, Nb, Pm, Pr, Sm, Y, Cm, additional nuclides that were deemed to be Am dose significant (i.e., Br-82, Br-84, Rb-88, Cerium Ce, Pu, Np Rb-89, Te-133, Te-133m, Te-134, 1-130, Xe-131m, Xe-133m, Xe-138, Cs-138 and Np-238).(Att. 4 Section 4.0 and Table 4.1-1)3.5 Of the radioiodine released from the reactor coolant system Conforms The assumed chemical form of iodine (RCS) to the containment in a postulated accident, 95 percent of released to containment following a DBA the iodine released should be assumed to be cesium iodide (Csl), LOCA is 95% psarticulate in the form of 4.85 percent elemental iodine, and 0.15 percent organic iodide. cesium iodide (Csl), 4.85% elemental This includes releases from the gap and the fuel pellets. With the iodine, and 0.15% organic iodide. (Att. 4 exception of elemental and organic iodine and noble gases, Section 7.2 and Table 7.2-1)fission products should be assumed to be in particulate form. The same chemical form is assumed in releases from fuel pins in See details for each event.FHAs and from releases from the fuel pins through the RCS in DBAs other than FHAs or LOCAs. However, the transport of these iodine species following release from the fuel may affect these assumed fractions. The accident-specific appendices to this regulatory guide provide additional details.A5-6 Table A5-ýA: Conformance with, Regulatory Guide-A1.8.3 Main Sections, RG,,. >. DCPP.-.Sectioin RdGPositi~on h Ana~lysis Comments 3.6 The amount of fuel damage caused by non-LOCA design basis Conforms The amount of fuel damage caused by events should be analyzed to determine, for the case resulting in non-LOCA design basis events has the highest radioactivity release, the fraction of the fuel that previously been determined, as described reaches or exceeds the initiation temperature of fuel melt and the in UFSAR Chapter 15. The amount of fuel fraction of fuel elements for which the fuel clad is breached, damage evaluated is consistent with Although the NRC staff has traditionally relied upon the departure current licensing basis.from nucleate boiling ration (DNBR) as a fuel damage criterion, licensees, my propose other methods to the NRC staff, such as See individual event conformance tables in those based upon enthalpy deposition, for estimating fuel damage this Attachment. for the purpose of establishing radioactivity releases.The amount of fuel damage caused by a FHA is addressed in Appendix B of this guide.4. -Dose Calculation Methodology 4.1.1 The dose calculations should determine the TEDE. TEDE is the Conforms The dose calculations determine the TEDE sum of the committed effective dose equivalent (CEDE) from dose, with all significant progeny included, inhalation and the deep dose equivalent (DDE) from external as the sum of the CEDE and DDE.exposure. The calculation of these two components of the TEDE should consider all radionuclides, including progeny from the As allowed in Section 4.1.4 of RG 1.183, decay of parent radionuclides, that are significant with regard to since the submersion exposure is uniform dose consequences and the released radioactivity. 1 3 to the whole body, the EDE is used in lieu of the deep dose equivalent (DDE) in 1 3 The prior practice of basing inhalation exposure on only determining the contribution of the radioiodine and not inc.uding radioiodine in external exposure submersion dose to the TEDE.calculations is not consistent with the definition of TEDE and the characteristics of the revised source term. (Att. 4 Section 6.1)A5-7 ______ Table A-5-A: Conformance with Repuat Gude 1.1'8-3 Main Sections-RG>[ DCIPIPýSe ction jRGPOsition j Ai~nalý.is .1Comments 4.1.2 The exposure-to-CEDE factors for inhalation of radioactive material should be derived from the data provided in ICRP Publication 30, "Limits for Intakes of Radionuclides by Workers" (Ref. 19). 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" (Ref. 20), provides tables of conversion factors acceptable to the NRC staff. The factors in the column headed "effective" yield doses corresponding to the CEDE.Conforms The CEDE is calculated using the inhalation dose conversion factors provided 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". The factors in the column headed"effective" yield doses corresponding to the CEDE and are derived based on ICRP-30.(Att.-4 Section 6.1)For the first 8 hours, the breathing rate of persons offsite should Conforms The assumed offsite breathing rates are be assumed to be 3.5 x 10.4 cubic meters per second. From 8 to those specified in Section 4.1.3 of RG 24 hours following the accident, the breathing rate should be 1.183.assumed to be 1.8 x 10-4 cubic meters per second. After that and until the end of the accident, the rate should be assumed to be (Att. 4 Section 6.1)2.3 x 1 0-4 cubic meters per second.4.1.4 The DDE should be calculated assuming submergence in semi-infinite cloud assumptions with appropriate credit for attenuation by body tissue. The DDE is nominally equivalent to the effective dose equivalent (EDE) from external exposure if the whole body is irradiated uniformly. Since this is a reasonable assumption for submergence exposure situations, EDE may be used in lieu of DDE in determining the contribution of external dose to the TEDE.Table 111.1 of Federal Guidance Report 12, "External Exposure to Radionuclides in Air, Water, and Soil" (Ref. 21), provides external EDE conversion factors acceptable to the NRC staff. The factors in the column headed "effective" yield doses corresponding to the EDE.Conforms The submersion EDE is calculated using the air submersion dose coefficients provided in Table 111.1 of Federal Guidance Report 12, "External Exposure to Radionuclides in Air, Water, and Soil." (Att. 4 Section 6.1)A5-8 Table A5-A: Conformance with Regulatory Guide 1.1.83 Main Sections=RRG Poiin ~DCPP~The An~alysis 5 Comments 4.1.5 The TEDE should be determined for the most limiting person at Conforms The Maximum EAB TEDE for any two-hour the EAB. The maximum EAB TEDE for any two-hour period period is determined and documented in following the start of the radioactivity release should be each analysis. See individual events in determined and used in determining compliance with the dose the Att. 4.criteria in 10 CFR 50.67.14 The maximum two-hour TEDE should be determined by calculating the postulated dose for a series of small time increments and performing a "sliding" sum over the increments for successive two-hour periods. The maximum TEDE obtained is submitted. The time increments should appropriately reflect the progression of the accident to capture the peak dose interval between the start of the event and the end of radioactivity release (see also Table 6).1 4 With regard to the EAB TEDE, the maximum two-hour value is the basis for screening and evaluation under 10 CFR 50.59.Changes to doses outside of the two-hour window are only considered in the context of their impact on the maximum two-hour EAB TEDE.4.1.6 TEDE should be determined for the most limiting receptor at the Conforms The TEDE is determined for the most outer boundary of the low population zone (LPZ) and should be limiting person at the LPZ.used in determining compliance with the dose criteria in 10 CFR 50.67. (Att. 4 Section 6.1)4.1.7 No correction should be made for depletion of the effluent plume Conforms No plume depletion due to ground by deposition on the ground. deposition is credited.(Att. 4 Section 5.1)A5-9 Table A5-A: Conformfance with lRegulatory Guid'e 1A.I3 Main.iSeetions .Secqtion, RG,,Pos ition Analysýis Cohiinents, 4.2.1 The TEDE analysis should consider all sources of radiation that Conforms The radiation dose to personnel within the will cause exposure to control room personnel. The applicable control room envelope includes inhalation sources will vary from facility to facility, but typically will include: and immersion doses due to releases as a" Contamination of the control room atmosphere by the result of each event. The control room intake or infiltration of the radioactive material contained in shielding design is based on the LOCA, the radioactive plume released from the facility, which represents the worst case DBA" Contamination of the control room atmosphere by the relative to radioactivity releases.intake or infiltration of airborne radioactive material from Therefore, only the LOCA addresses shine areas and structures adjacent to the control room dose. Direct shine doses from contained envelope, sources and the external plume are also* Radiation shine from the external radioactive plume evaluated. released from the facility," Radiation shine from radioactive material in the reactor (Att. 4, Section 7.2)containment," Radiation shine from radioactive material in systems and See individual events for details.components inside or external to the control room envelope, e.g., radioactive material buildup in recirculation filters.4.2.2 The radioactive material releases and radiation levels used in the Conforms The control room doses are determined control room dose analysis should be determined using the same using the same source term, transport, source term, transport, and release assumptions used for and release assumptions used for determining the EAB and the LPZ TEDE values, unless these determining the EAB and the LPZ TEDE assumptions would result in non-conservative results for the values, resulting in conservative results for control room. the control room. See individual events for details.A5-10 Table A5-A: Conformance wiffiReguifatoiry.G de 183..Main etos 4.23 nRG osiionAnalysis: Comtments-4.2.3 The models used to transport radioactive material into and Conforms The models used to transport radioactive through the control room, 1 5 and. the shielding models used to material from the fuel to the control room determine radiation dose rates from external sources, should be and the shielding models used to structured to provide suitably conservative estimates of the determine radiation dose rates from exposure to control room personnel. external sources (SW-QADCGGP and PERC2), are structured to provide suitably 1 5 The iodine protection factor (IFP) methodology of Reference 22 conservative estimates of the exposure to may not be adequately conservative for all DBAs and control control room personnel. room arrangements since it models a steady-state control room condition. Since many analysis parameters change over the (Att. 4 Section 3.0)duration of the event, the IPF methodology should only be used with caution. The NRC computer codes HABIT (Ref. 23) and RADTRAD (Ref. 24) incorporate suitable methodologies. A5-11 STable Conformance with, Regulatory Guide 1.1083 Main Secti.osns RG. .Io"pp7 -S~etion4 RGPsto S Analyss .omments 4.2.4 Credit for engineered safety features that mitigate airborne radioactive material within the control room may be assumed.Such features may include control room isolation or pressurization, or intake or recirculation filtration. Refer to Section 6.5.1, "ESF Atmospheric Cleanup System," of the SRP (Ref. 3)and Regulatory Guide 1.52, "Design, Testing, and Maintenance Criteria for Post-accident Engineered-Safety-Feature Atmosphere Cleanup System Air Filtration and Adsorption Units of Light-Water-Cooled Nuclear Power Plants" (Ref. 25), for guidance. The control room design is often optimized for the DBA LOCA and the protection afforded for other accident sequences may not be as advantageous. In most designs, control room isolation is actuated by engineered safeguards feature (ESF) signals or radiation monitors (RMs). In some cases, the ESF signal is effective only for selected accidents, placing reliance on the RMs for the remaining accidents. Several aspects of RMs can delay the control room isolation, including the delay for activity to build up to concentrations equivalent to the alarm setpoint and the effects of different radionuclide accident isotopic mixes on monitor response.Conforms Credit is taken for automatic initiation of CRVS Mode 4 (filtration and pressurization) during all analyzed events except the LRA and the Loss of Load Limiting Condition II event. Signals that initiate CRVS Mode 4 include radiation monitors located at the CR normal air intakes, safety injection signal (SIS), and Containment Isolation Phase A. The SIS does not directly initiate CRVS Mode 4, however, it initiates Containment Isolation Phase A, which initiates Mode 4. (Att. 4 Section 7.1)CR radiation monitors (1/2 -RE-25/26) located at the CR normal intakes have the capability of isolating the CR normal intakes on high radiation and switching to CRVS Mode 4.Setpoint changes to these monitors will accommodate CR isolation during an FHA.(Att. 4 Sections 2.2, 7.1, and 7.3)Credit is taken for the dual ventilation intake design of the CR pressurization air intakes per RG 1.194, June 2003. (Att. 4 Sections 2.1, 5.2, and 7.1)Filters credited for offsite and CR dose are qualified and acceptable per the DCPP Ventilation Filter Testing Program (VFTP) (TS 5.5.11), which states that the VFTP is in accordance with RG 1.52, Revision 2, ANSI N510 1980, and ASTM D3803-1989. A5-12 Table A5-A: Conformance wvith-Regula~tory Guwide 1.183 Main SectionsýRG --fCIPP Section, RG Position -Aaysis~ Commeants 4.2.5 Credit should generally not be taken for the use of personal Conforms No credit is taken for the use of personal protective equipment or prophylactic drugs. Deviations may be protective equipment or prophylactic considered on a case-by-case basis. drugs.(Att. 4 Section 8.0)4.2.6 The dose receptor for these analyses is the hypothetical Conforms The assumed breathing rates and maximum exposed individual who is present in the control room occupancy factors used for DCPP control for 100% of the time during the first 24 hours after the event, 60% room operator dose are those specified in of the time between 1 and 4 days, and 40% of the time from 4 Section 4.2.6 of RG 1.183.days to 30 days.1 6 For the duration of the event, the breathing rate of this individual should be assumed to be 3.5 x 10 4 cubic meters (Att. 4 Section 6.1)per second.ARCON96 was used for determining CR 1 The occupancy is modeled in the X/Q values determined in X/Q values. Occupancy assumptions Reference 22 (Murphy-Campe) and should not be credited twice. were an input in RADTRAD.The ARCON96 Code (Ref. 26) does not incorporate these occupancy assumptions, making it necessary to apply this (Att. 4 Section 5)correction in the dose calculations. 4.2.7 Control room doses should be calculated using dose conversion Conforms Control room doses are calculated using factors identified in Regulatory Position 4.1 above for use in dose conversion factors identified in offsite dose analyses. The DDE from photons may be corrected Regulatory Position 4.1.for the difference between finite cloud geometry in the control room and the semi-infinite cloud assumption used in calculating Equation I given in RG 1.183, Regulatory the dose conversion factors. The following expression may be Position 4.2.7, is used for finite cloud used to correct the semi-infinite cloud dose, DDE, to a finite cloud correction when calculating immersion dose, DDEtinte, where the control room is modeled as a doses due to the airborne activity inside hemisphere that has a volume, V, in cubic feet, equivalent to that the control room.of the control room (Ref. 22).(Att. 4 Section 6.1)DDEfinite = DDEoOV 0 , 3 3 8 Equation 1 1173 A5-13 Table A5-A: Conformance with Regullatory-G uide 1.1983M~ain~section Uh , -Section ,RG'Position ,AnAlys~isi Comments 4.3 The guidance provided in Regulatory Positions 4.1 and 4.2 should Conforms DCPP is applying for full implementation be used, as applicable, in re-assessing the radiological analyses of AST as described in LAR Enclosure identified in Regulatory Position 1.3.1, such as those in NUREG- Section 1. Regulatory Positions 4.1 and 0737 (Ref. 2). Design envelope source terms provided in 4.2 have been used in re-assessing the NUREG-0737 should be updated for consistency with the AST. In applicable radiological analyses identified general, radiation exposures to plant personnel identified in in LAR Enclosure Section .1.Regulatory Position 1.3.1 should be expressed in terms of TEDE.Integrated radiation exposure of plant equipment should be determined using the guidance of Appendix I of this guide.4.4 The radiological criteria for the EAB, the outer boundary of the Conforms The EAB and LPZ acceptance criteria LPZ, and for the control room are in 10 CFR 50.67. These criteria used are those of Table 6 of RG 1.183.are stated for evaluating reactor accidents of exceedingly low The control room acceptance criterion is 5 probability of occurrence and low risk of public exposure to rem TEDE. Updates to applicable criteria radiation, e.g., a large-break LOCA. The control room criterion are included in the LAR to be consistent applies to all accidents. For events with a higher probability of with the TEDE criterion in 10 CFR occurrence, postulated EAB and LPZ doses should not exceed 50.67(b)(2)(iii), including updating to GDC the criteria tabulated in Table 6. 19, 1999, for dose only, upon implementation of AST. The dose'The acceptance criteria for the various NUREG-0737 (Ref. 2) acceptance criterion for the TSC, which items generally reference General Design Criteria 19 (GDC 19) was based on Section 8.2.1, Item f of from Appendix A to 10 CFR Part 50 or specify criteria derived NUREG-0737, Supplement 1, will be 5 from GDC-19. These criteria are generally specified in terms of rem TEDE. See LAR Enclosure Section whole body dose, or its equivalent to any body organ. For facilities 2.1, items 4, 5, and 6.applying for, or having received, approval for the use of an AST, the applicable criteria should be updated for consistency with the TEDE criterion in 10 CFR 50.67(b)(2)(iii).

5. -Analysis Assumptions and Methodology A5-14

.Table A5-A: CQnformance with"Reguiatoryý Guide 1.11*8Z3 Main Sections, RG. DP Se ction, RG, Position. AnlCjomyM- _ts:-5.1.1 The evaluations required by 10 CFR 50.67 are re-analyses of the Conforms The analyses are prepared, reviewed, and¶1 design basis safety analyses and evaluations required by 10 CFR maintained in accordance with quality 50.34; they are considered to be a significant input to the assurance programs that comply with 10 evaluations required by 10 CFR 50.92 or 10 CFR 50.59. These CFR Part 50, Appendix B, 'Quality analyses should be prepared, reviewed, and maintained in Assurance Criteria for Nuclear Power accordance with quality assurance programs that comply with Plants and Fuel Reprocessing Plants." Appendix B, "Quality Assurance Criteria for Nuclear Power Plants and Fuel Reprocessing Plants," to 10 CFR Part 50.5.1.1 These design basis analyses were structured to provide a Conforms These analyses have been performed as¶2 conservative set of assumptions to test the performance of one or specified in the guidance. See more aspects of the facility design. Many physical processes and conformance tables for the individual phenomena are represented by conservative, bounding analyses.assumptions rather than being modeled directly. The staff has selected assumptions and models that provide an appropriate and prudent safety margin against unpredicted events in the course of an accident and compensate for large uncertainties in facility parameters, accident progression, radioactive material transport, and atmospheric dispersion. Licensees should exercise caution in proposing deviations based upon data from a specific accident sequence since the DBAs were never intended to represent any specific accident sequence -- the proposed deviation may not be conservative for other accident sequences. A5-15 _____ 1jjj~Table A5-A: tonfoermaice with Regulatory Guide' 1.183 Main Sections, RG IDCIPIP ______ _____S.ection R-G RIosition -,Analysis 'Comm nents, 5.1.2 Credit may be taken for accident mitigation features that are classified as safety-related, are required to be operable by technical specifications, are powered by emergency power sources, and are either automatically actuated or, in limited cases, have actuation requirements explicitly addressed in emergency operating procedures. The single active component failure that results in the most limiting radiological consequences should be assumed. Assumptions regarding the occurrence and timing of a loss of offsite power should be selected with the objective of maximizing the postulated radiological consequences. Conforms Credit is taken for the ESF equipment discussed in RG Position 4.2.4. Credit is also taken for the PG&E Design Class I ABVS and filters, which are also controlled by TS 3.7.12 and TS 5.5.11. Credit is taken for operating Containment Spray during recirculation (Att. 4, Section 2.1, item 14). A Time Critical Operator Action will be implemented to ensure that the realignment from injection to recirculation is performed within 12 minutes of termination of injection spray (Att. 4, Section 2.5). No new system is credited in the analyses; therefore, all ESF systems have been previously reviewed by NRC.Assumptions regarding the occurrence and timing of a loss of offsite power (LOOP) are selected with the intent of maximizing doses. A LOOP is assumed for events that have the potential to cause grid perturbation (LOCA, LRA, CREA, MSLB, SGTR, and LOL). (Att. 4 Sections 7.0)A FHA cannot cause grid instability, nor can a LOOP cause a FHA. Thus the FHA is evaluated without the assumption of a LOOP. (Att. 4 Section 7.0)A5-16 ______Table A5-A: Coniformance Wilth Reglulato6ry Guide 1.183 MaIn -Sections RG~ 1 K DCPP Section R&Iosition JAnalysisj C'O mments 5.1.3 The numeric values that are chosen as inputs to the analyses required by 10 CFR 50.67 should be selected with the objective of determining a conservative postulated dose. In some instances, a particular parameter may be conservative in one portion of an analysis but be non-conservative in another portion of the same analysis. For example, assuming minimum containment system spray flow is usually conservative for estimating iodine scrubbing, but in many cases may be nonconservative when determining sump pH. Sensitivity analyses may be needed to determine the appropriate value to use. As a conservative alternative, the limiting value applicable to each portion of the analysis may be used in the evaluation of that portion. A single value may not be applicable for a parameter for the duration of the event, particularly for parameters affected by changes in density. For parameters addressed by technical specifications, the value used in the analysis should be that specified in the technical specifications. 1 8 If a range of values or a tolerance band is specified, the value that would result in a conservative postulated dose should be used. If the parameter is based on the results of less frequent surveillance testing, e.g., steam generator nondestructive testing (NDT), consideration should be given to the degradation that may occur between periodic tests in establishing the analysis value.1 8 Note that for some parameters, the technical specification value may be adjusted for analysis purposes by factors provided in other regulatory guidance. For example, ESF filter efficiencies are based on the guidance in Regulatory Guide 1.52 (Ref. 25)and in Generic Letter 99-02 (Ref. 27) rather than the surveillance test criteria in the technical specifications. Generally, these adjustments address potential changes in the parameter between scheduled surveillance tests.Conforms Conservative parameters are assumed when calculating each contributor in the dose analyses. See individual events for more information. A5-17 Tab~le A5-A-: Conformance with Re'g~uatory Guide 1.1:83 Main SeGcti~ons RG. WDCPP K-Scin R~stoni. Anallysis Copmments 5.1.4 The NRC staff considers the implementation of an AST to be a Conforms The analyses assumptions and methods significant change to the design basis of the facility that is are compatible with the ASTs and the voluntarily initiated by the licensee. In order to issue a license TEDE criteria per RG 1.183 guidance.amendment authorizing the use of an AST and the TEDE dose criteria, the NRC staff must make a current finding of compliance with regulations applicable to the amendment. The characteristics of the ASTs and the revised dose calculational methodology may be incompatible with many of the analysis assumptions and methods currently reflected in the facility's design basis analyses.The NRC staff may find that new or unreviewed issues are created by a particular site-specific implementation of the AST, warranting review of staff positions approved subsequent to the initial issuance of the license. This is not considered a backfit as defined by 10 CFR 50.109, "Backfitting." However, prior design bases that are unrelated to the use of the AST, or are unaffected by the AST, may continue as the facility's design basis. Licensees should ensure that analysis assumptions and methods are compatible with the ASTs and the TEDE criteria.5.2 The appendices to this regulatory guide provide accident-specific Conforms The postulated accident radiological ¶1 assumptions that are acceptable to the staff for performing consequence analyses have been updated analyses that are required by 10 CFR 50.67. The DBAs for AST. The DBA LOCA, FHA, MSLB, addressed in these attachments were selected from accidents SGTR, LRA, and CREA have been that may involve damage to irradiated fuel. This guide does not analyzed. In addition, the Loss of Load address DBAs with radiological consequences based on technical event, which is the limiting Condition II specification reactor or secondary coolant-specific activities only. event, is also updated for AST. See The inclusion or exclusion of a particular DBA in this guide should conformance tables for individual events.not be interpreted as indicating that an analysis of that DBA is The dose consequences for other events required or not required. Licensees should analyze the DBAs that that have an accident source term and'are are affected by the specific proposed applications of an AST. part of the current DCPP licensing basis are addressed by qualitative comparisons to the above analyzed accidents as allowed by RG 1.183, Position 1.3.3.A5-18 Table A5-A: Conformance with Regulatory Guide 1.183 Main Seotions, .r-.RIG -,DCPPJ< -Setion RPo~sition Analysis Comm.ents 5.2 The NRC staff has determined that the analysis assumptions in Conforms Assumptions for each analysis have been¶2 the appendices to this guide provide an integrated approach to addressed, as shown in the conformance performing the individual analyses and generally expects tables for the individual events.licensees to address each assumption or propose acceptable alternatives. Such alternatives may be justifiable on the basis of plant-specific considerations, updated technical analyses, or, in some cases, a previously approved licensing basis consideration. The assumptions in the appendices.are deemed consistent with the AST identified in Regulatory Position 3 and internally consistent with each other. Although licensees are free to propose alternatives to these assumptions for consideration by the NRC staff, licensees should avoid use of previously approved staff positions that would adversely affect this consistency. 5.2 The NRC is committed to using probabilistic risk analysis (PRA) Conforms No changes have been made to analysis¶3 insights in its regulatory activities and will consider licensee assumptions based upon risk insights.proposals for changes in analysis assumptions based upon risk insights. The staff will not approve proposals that would reduce the defense in depth deemed necessary to provide adequate protection for public health and safety. In some cases, this defense in depth compensates for uncertainties in the PRA analyses and addresses accident considerations not adequately addressed by the core damage frequency (CDF) and large early release frequency (LERF) surrogate indicators of overall risk.A5-19 T~able A5-A: Conformance with Reg~ullatory Gu-ide 1.183 Main Sections.RýG O': ~~'P PJ:<;Se.ctipn RG Position Aays omet 5.3 Atmospheric dispersion values (X/Q) for the EAB, the LPZ, and Conforms New atmospheric dispersion values (X/Q)%1 the control room that were approved by the staff during initial for the EAB, LPZ, control room, and the facility licensing or in subsequent licensing proceedings may be TSC were developed. used in performing the radiological analyses identified by this guide. Methodologies that have been used for determining X/Q (Att. 4 Section 5)values are documented in Regulatory Guides 1.3 and 1.4, Regulatory Guide 1.145, "Atmospheric Dispersion Models for Potential Accident Consequence Assessments at Nuclear Power Plants," and the paper, "Nuclear Power Plant Control Room Ventilation System Design for Meeting General Criterion 19" (Refs. 6, 7, 22, and 28).A5-20 ____ Table A5-A-:'Conformance with; Regulatory Luide 1. 183,Main Sections locpp DCP -S Seqtiotvpn RG Positionm A ~ i_____ nCmhts 5.3¶2 References 22 [Murphy -Campe] and 28 [RG 1.145] should be used if the FSAR X/Q values are to be- revised or if values are to be determined for new release points or receptor distances. Fumigation should be considered where applicable for the EAB and LPZ. For the EAB, the assumed fumigation period should be timed to be included in the worst 2-hour exposure period. The NRC computer code PAVAN (Ref. 29) implements Regulatory Guide 1.145 (Ref. 28) and its use is acceptable to the NRC staff.The methodology of the NRC computer code ARCON96 1 9 (Ref.26) is generally acceptable to the NRC staff for use in determining control room X/Q values. Meteorological data collected in accordance with the site-specific meteorological measurements program described in the facility FSAR should be used in generating accident X/ Q values. Additional guidance is provided in Regulatory Guide 1.23, "Onsite Meteorological Programs" (Ref.30). All changes in X/Q analysis methodology should be reviewed by the NRC staff.1 9 The ARCON 96 computer code contains processing options that may yield X/Q values that are not sufficiently conservative for use in accident consequence assessments or may be incompatible with release point and ventilation intake configurations at particular sites. The applicability of these options and associated input parameters should be evaluated on a case-by-case basis.The assumptions made in the examples in the ARCON96 documentation are illustrative only and do not imply NRC staff acceptance of the methods or data used in the example.Conforms New atmospheric dispersion values (X/Q)for the EAB, LPZ, control room, and the TSC were developed. Meteorological data acquired in accordance with the DCPP meteorological measurement program for the five-year period from 2007 to 2011 are used to calculate X/Q values. The DCPP meteorological measurement program is described in DCPP UFSAR Section 2.3.3 and was designed to meet the requirements of Safety Guide 23, February 1972.The onsite X/Q methodology (CR and TSC) is based upon the methods in RG 1.194 using the computer code ARCON96.The recommended default values from RG 1.194, Table A-2 were used to develop onsite X/Q values. (Att. 4 Sections 5.1 and 5.2)The offsite X/Q methodology (EAB and LPZ) is based upon the methodology in RG 1.145 for ground level releases using CBI S&W Proprietary code EN-113. Per RG 1.145, fumigation is applicable to stack releases. Releases for DCPP are treated as ground level releases, therefore fumigation is not considered. (Att. 4 Sections 3 and 5)A5-21 .Jalble A5-A: Conformanceewith,,Regulatory.Guide' 1.18'3 VaMin Sections RGDP Secio RG Position <Analysis ýComments, 6. -Assumptions for Evaluating the radiation Doses for Equipment Qualification The assumptions in Appendix I to this guide are acceptable to the Conforms Generic Safety Issue (GSI) 187, "The NRC staff for performing radiological assessments associated with Potential Impact of Postulated Cesium equipment qualification. The assumptions in Appendix I will Concentration on Equipment supersede Regulatory Positions 2.c(1) and 2.c(2) and Appendix D Qbalification," has been resolved. The of Revision 1 of Regulatory Guide 1.89, "Environmental NRC staff concluded that there is no clear Qualification of Certain Electric Equipment Important to Safety for basis for a requirement to modify the Nuclear Power Plants" (Ref. 11), for operating reactors that have design basis for equipment qualification to amended their licensing basis to use an alternative source term. adopt AST since there would be no Except as stated in Appendix I, all other assumptions, methods, and discernible risk reduction associated with provisions of Revision I of Regulatory Guide 1.89 remain effective, such a requirement. Therefore, this LAR does not propose to modify the EQ design The NRC staff is assessing the effect of increased cesium releases basisto adopt AST. The DCPP EQ on EQ doses to determine whether licensee action is warranted, analysis will continue to be based upon Until such time as this generic issue is resolved, licensees may use TID-1 4844 assumptions at this time.either the AST or the TID14844 assumptions for performing the required EQ analyses. However, no plant modifications are required to address the impact of the difference in source term characteristics (i.e., AST vs TID14844) on EQ doses pending the outcome of the evaluation of the generic issue.A5-22 L~-- " b~ A -B4 '- nomac &thRg [-~~ ui 1 ~ fr A (oso~~~nA'cdn S~1Vf R biti"OnP-I_ -npv -: Source Term AssumDtions

1. Acceptable assumptions regarding core inventory and the Conforms See responses to Regulatory Positon 3 release of radionuclides from the fuel are provided in'Regulatory located in Table A5-A.Position 3 of this guide.2. If the sump or suppression pool pH is controlled at values of 7 or Conforms The sump pH is controlled at a value greater, the chemical form of radioiodine released to the greater than 7.0. Evaluation of pH takes containment should be assumed to be 95% cesium iodide (CsI), into consideration acid production due to 4.85 percent elemental iodine, and 0.15 percent organic iodide, the radiation environment associated Iodine species, including those from iodine re-evolution, for sump with the accident.

The assumed or suppression pool pH values less than 7 will be evaluated on a chemical form of iodine released to case-by-case basis. Evaluations of pH should consider the effect containment following a DBA LOCA is of acids and bases created during the LOCA event, e.g., 95% particulate in the form of cesium radiolysis products. With the exception of elemental and organic iodide (CsI), 4.85% elemental iodine, iodine and noble gases, fission products should be assumed to and 0.15% organic iodide. With the be in particulate form. exception of elemental and organic iodine and noble gases, fission products are assumed to be in particulate form.(Att. 4 Section 7.2.3.2.5) Assumptions on Transport in Primary Containment 3.1 The radioactivity released from the fuel should be assumed to Conforms All radioactivity released from the fuel is mix instantaneously and homogeneously throughout the free air assumed to mix instantaneously and volume of the primary containment in PWRs or the drywell in homogeneously throughout the free air BWRs as it is released. This distribution should be adjusted if volume of the primary containment. there are internal compartments that have limited ventilation exchange. The suppression pool free air volume may be (Att. 4 Section 7.2.3.2)included provided there is a mechanism to ensure mixing between the drywell to the wetwell. The release into the containment or drywell should be assumed to terminate at the end of the early in-vessel phase.A5-23 ~~R~~~g~~~d~~torVqNO -Ai~1 ~ pedixA(oso ~In~. ~~~DCPR An~I~si~<Win,~ht Reduction in airborne radioactivity in the containment by natural deposition within the containment may be credited. Acceptable models for removal of iodine and aerosols are described in Chapter 6.5.2, "Containment Spray as a Fission Product Cleanup System," of the Standard Review Plan (SRP), NUREG-0800 (Ref. A-I) and in NUREG/CR-6189, "A Simplified Model of Aerosol Removal by Natural Processes in Reactor Containments" (Ref. A-2). The later model is incorporated into the analysis code RADTRAD (Ref. A-3). The prior practice of deterministically assuming that a 50% plateout of iodine is released from the fuel is no longer acceptable to the NRC staff as it is inconsistent with the characteristics of the revised source terms.Conforms DCPP does not deterministically assume 50% plateout of iodine. The wall deposition removal coefficient for elemental iodine has been calculated using computer program SWNAUA to estimate the time dependent particulate removal coefficients. The guidance of SRP 6.5.2 was used to determine elemental iodine removal coefficients. Credit is taken for gravitational settling of particulates.(Att. 4 Section 3 for code description, Section 7.2.3.2.4 for fission product removal)A5-24 *I T~b~ 5~~?:Cofoman~ ith~jkt~~Gid ~1.1 .pqen01iXjAios, CoIlancidn p n 3.3 Reduction in airborne radioactivity in the containment by Conforms The containment spray system is%1 containment spray systems that have been designed and are currently credited in DCPP licensing maintained in accordance with Chapter 6.5.2 of the SRP (Ref. A- basis for the removal of fission products 1) may be credited. Acceptable models for the removal of iodine from the containment atmosphere. and aerosols are described in Chapter 6.5.2 of the SRP and Therefore, containment spray has been NUREG/CR-5966, "A Simplified Model of Aerosol Removal by reviewed and approved for this use. The Containment Sprays"' (Ref. A-4). This simplified model is AST analysis continues to credit incorporated into the analysis code RADTRAD (Refs. A-1 to A- containment spray system for the 3). removal of iodine and aerosols from the containment atmosphere. In addition,'This document describes statistical formulations with differing credit is now taken for containment spray levels of uncertainty. The removal rate constants selected for during recirculation. Att. 4 Section use in design basis calculations should be those that will 7.2.3.2.4 provides information used for maximize the dose consequences. For BWRs, the simplified determining the fission product removal model should be used only if the release from the core is not coefficients. for processes credited in directed through the suppression pool. Iodine removal in the reducing the radionuclide inventory suppression pool affects the species assumed by the model to available for release from the be present initially, containment. The SWNAUA code was used to estimate the time dependent particulate removal coefficients. Use of the SWNAUA code has been approved in prior AST applications. Att. 4 Table 7.2-2 provides removal coefficient used in the LOCA dose analysis.A5-25 T e,, Ati'tM,,brif' G u i d I ý 1981kA Lpist of C do 1 a41t-Ac.6i d-6 ntl 03 0 3.3 The evaluation of the containment sprays should address areas Conforms The percentage of the total containment T12 within the primary containment that are not covered by the spray free volume that is sprayed is 82.5%.drops. The mixing rate attributed to natural convection between DCPP uses safety-related containment sprayed and unsprayed regions of the containment building, fan cooler units to provide mixing of the provided that adequate flow exists between these regions, is sprayed and unsprayed volumes of the assumed to be two turnovers of the unsprayed regions per hour, containment. The containment mixing unless other rates are justified. The containment building rate between the sprayed and unsprayed atmosphere may be considered a single, well-mixed volume if regions following a LOCA is determined the spray covers at least 90% of the volume and if adequate to be 9.13 turnovers of the unsprayed. mixing of unsprayed compartments can be shown. regions per hour. (Att. 4 Section 7.2.3.2)3.3 The SRP sets forth a maximum decontamination factor (DF) for Conforms Att. 4 Section 7.2.3.2.4 provides¶13 elemental iodine based on the maximum iodine activity in the information used for determining the primary containment atmosphere when the sprays actuate, fission product removal coefficients for divided by the activity of iodine remaining at some time after processes credited in reducing the decontamination. The SRP also states that the particulate iodine radionuclide inventory available for removal rate should be reduced by a factor of 10 when a DF of release from the containment. The 50 is reached. The reduction in the removal rate is not required if removal rate is based on the calculated the removal rate is based on the calculated time-dependent time-dependent airborne aerosol mass.airborne aerosol mass. There is no specified maximum DF for Since the spray removal coefficients are aerosol removal by sprays. The maximum activity to be used in based on calculated time dependent determining the DF is defined as the iodine activity in the airborne aerosol mass, there is no columns labeled "Total" in Tables I and 2 of this guide multiplied restriction on the DF for particulate by 0.05 for elemental iodine and by 0.95 for particulate iodine iodine.(i.e., aerosol treated as particulate in SRP methodology). Att. 4 Table 7.2-2 provides removal coefficient used in the LOCA dose analysis.3.4 Reduction in airborne radioactivity in the containment by in- N/A DCPP does not have post-accident in-containment recirculation filter systems may be credited if these containment air filtration systems.systems meet the guidance of Regulatory Guide 1.52 and Generic Letter 99-02 (Refs. A-5 and A-6). The filter media loading caused by the increased aerosol release associated with the revised source term should be addressed. A5-26 feirm -P<, 1. 60 _11-1 j4jotý ýG_ JQ 110'3'1"'j'rý pp _5-~DCpP P 3.5 Reduction in airborne radioactivity in the containment by N/A Not Applicable for a PWR.suppression pool scrubbing in BWRs should generally not be credited. However, the staff may consider such reduction on an individual case basis. The evaluation should consider the relative timing of the blowdown and the fission product release from the fuel, the force driving the release through the pool, and the potential for any bypass of the suppression pool (Ref. 7).Analyses should consider iodine re-evolution if the suppression pool liquid pH is not maintained greater than 7.3.6 Reduction in airborne radioactivity in the containment by N/A DCPP does not have ice condensers. retention in ice condensers, or other engineering safety features The engineered safety features not addressed above, should be evaluated on an individual case applicable to DCPP are addressed. basis. See Section 6.5.4 of the SRP (Ref. A-I).3.7 The primary containment should be assumed to leak at the peak Conforms Radioactivity is assumed to leak from pressure technical specification leak rate for the first 24 hours, both the sprayed and unsprayed region For PWRs, the leak rate may be reduced after the first 24 hours of the containment to the environment. to 50% of the technical specification leak rate. For BWRs, A containment leak rate, based on DCPP leakage may be reduced after the first 24 hours, if supported by TS 5.5.16, of 0.1% of containment air plant configuration and analyses, to a value not less than 50% of weight per day is assumed for the first 24 the technical specification leak rate. Leakage from hours. After 24 hours, the containment subatmospheric containments is assumed to terminate when the leak rate is reduced by 50% to 0.05% of containment is brought to and maintained at a subatmospheric containment air weight-per day.condition as defined by technical specifications.(Att. 4 Section 7.2.3.2.6 and Table 7.2-1)For BWRs with Mark III containments, the leakage from the drywell into the primary containment should be based on the steaming rate of the heated reactor core, with no credit for core debris relocation. This leakage should be assumed during the two-hour period between the initial blowdown and termination of the fuel radioactivity release (gap and early in-vessel release phases). After two hours, the radioactivity is assumed to be uniformly distributed throughout the drywell and the primary containment. A5-27 _-MTb~ A5_ BCnfranc, 66fth"R'ulAor T~di~18 bedxALs 3.8 If the primary containment is routinely purged during power Conforms TS 3.6.3 allows opening the 12-inch operations, releases via the purge system prior to containment containment vacuum/over pressure relief isolation should be analyzed and the resulting doses summed valves during operating MODES 1, 2, 3, with the postulated doses from other release paths. The purge and 4. Releases of RCS radionuclide release evaluation should assume that 100% of the radionuclide inventory are assumed through this path inventory in the reactor coolant system liquid is released to the until containment is isolated.containment at the initiation of the LOCA. This inventory should Containment isolation occurs prior to the be based on the technical specification reactor coolant system onset of the gap release phase, thus no equilibrium activity. Iodine spikes need not be considered. If the gap releases occur. (Att. 4 Section purge system is not isolated before the onset of the gap release 7.2.3.1)phase, the release fractions associated with the gap release and early in-vessel phases should be considered as applicable. TS 3.6.3 is being revised to require the containment purge system (48-in purge valves) to be sealed closed during MODES 1, 2, 3, and 4. (LAR Enclosure Section 2)Assumptions on Dual Containments

4. For facilities with dual containment systems, the acceptable N/A Regulatory Positions 4.1 through 4.6 assumptions related to the transport, reduction, and release of apply to facilities with dual containment radioactive material in and from the secondary containment or systems. As such, these positions are enclosure buildings are as follows, not applicable to DCPP.Assumptions on ESF System Leakage A5-28 att ESF systems that recirculate sump water outside of the primary containment are assumed to leak during their intended operation.

This release source includes leakage through valve packing glands, pump shaft seals, flanged connections, and other similar components. This release source may also include leakage through valves isolating interfacing systems (Ref. A-7). The radiological consequences from the postulated leakage should be analyzed and combined with consequences postulated for other fission product release paths to determine the total calculated radiological consequences from the LOCA. The following assumptions are acceptable for evaluating the consequences of leakage from ESF components outside the primary containment for BWRs and PWRs.The radiological consequences from the postulated ESF systems leakage are analyzed and combined with consequences postulated for other fission product release paths. ESF systems that recirculate sump fluids outside containment are postulated to leak at twice the sum of the administrative acceptance criteria.(Att. 4 Section 7.2.3.3)5.1 With the exception of noble gases, all the fission products released from the fuel to the containment (as defined in Tables 1 and 2 of this guide) should be assume to instantaneously and homogeneously mix in the primary containment sump water (in PWRs) or suppression pool (in BWRs) at the time of release from the core. In lieu of this deterministic approach, suitably conservative mechanistic models for the transport of airborne activity in containment to the sump water may be used. Note that many of the parameters that make spray and deposition models conservative with regard to containment airborne leakage are nonconservative with reciard to the buildup of sump activity.Conforms I With the exception of noble gases, all fission products released from the fuel to the containment are instantaneously and homogeneously mixed in the sump water at the time of release. Only iodine is released through ESF leakage since the noble gases are not assumed to dissolve in the sump and particulates would remain in the water of ECCS leakage.(Att. 4 Section 7.2.3.3)A5-29 ~omm~ehts ~The leakage should be taken as two times the sum of the simultaneous leakage from all components in the ESF recirculation systems above which the technical specifications, or licensee commitments to item III.D.1.1 of NUREG-0737 (Ref. A-8), would require declaring such systems inoperable. The leakage should be assumed to start at the earliest time the recirculation flow occurs in these systems and end at the latest time the releases from these systems are terminated. Consideration should also be given to design leakage through valves isolating ESF recirculation systems from tanks vented to atmosphere, e.g., emergency core cooling system (ECCS) pump miniflow return to the refueling water storage tank.Leakage from the ESF system can occur via the plant vent and via the penetration area. DCPP procedures, which are controlled by TS 5.5.2, will be updated as part of AST implementation to establish administrative acceptance criteria to ensure leakage is less than 126 cc/min with the following breakdown: (See LAR Enclosure Section 2)-Plant vent area -< 120 cc/min-Penetration area -< 6 cc/min The assumed leakage of 252 cc/min (240 cc/min plus 12 cc/min) is two times the administrative limit of 126 cc/min.The leakage is assumed to start with recirculation. DCPP does not take credit for filters for this ESF leakage.The LOCA dose analysis also includes an RHR pump seal passive failure of 50 gpm for 30 minutes that occurs 24 hours after the LOCA. The pump seal failure release is a filtered release. The LOCA dose analysis also accounts for releases through the RWST vent due to sump back-leakage and ESF leakage that is hard-piped to the MEDT. These releases are not filtered.(Att. 4 Sections 7.2.3.3 -7.2.3.6)A5-30 T a b I ~ A -B : 6 1n f o y m ani W it ' R d t ' ýG q i & A ~ ~ A b n i A ( o s f C o a t A c d n 'de- O'd prv,.DGPP."-'n- tn$10601J, ARR! Gl-ýe`p q15J. tPA", YA M 0 5.3 With the exception of iodine, all radioactive materials in the Conforms With the exception of iodine, all recirculating liquid should be assumed to be retained in the liquid radioactive materials in the recirculating phase. liquid are assumed to be retained in the liquid phase.(Att. 4 S ction 7.2.3.3)5.4 If the temperature of the leakage exceeds 212F, the fraction of Conforms ESF leakage is expected at the initiation total iodine in the liquid that becomes airborne should be assumed of the recirculation mode for safety equal to the fraction of the leakage that flashes to vapor. This injection at 829 seconds. The maximum flash fraction , FF, should be determined using a constant temperature of the recirculation fluid is enthalpy, h, process, based on the maximum time dependent 259.9*F, which has a flash fraction less temperature of the sum water circulating outside the containment: than 10%.FF hf, -hf, (Att. 4 Section 7.2.3.3)hf 9 Where: hf, is the enthalpy of liquid at system design temperature and pressure; hf2 is the enthalpy of liquid at saturation conditions (14.7 psia, 212F); and hf, is the heat of vaporization at 212F.5.5 If the temperature of the leakage is less than 212F or the Conforms The temperature of the ESF leakage and calculated flash fraction is less than 10%, the amount of iodine that RHR pump seal leakage is 259.90F, becomes airborne should be assumed to be 10% of the total iodine which has a flash fraction less than 10%.activity in the leaked fluid, unless a smaller amount can be justified Thus, a flash fraction of 10% is assumed based on the actual sump pH history and area ventilation rates. in the analysis. A calculated flash fraction of less than 10% is addressed for the RWST and MEDT release pathways.ctions 7.2.3.3 -7.2.3.6)A5-31 5.6 The radioiodine that is postulated to be available for release to the Conforms The iodine released from ESF leakage is environment is assumed to be 97% elemental and 3% organic. assumed to be 97% elemental and 3%Reduction in release activity by dilution or holdup within buildings, organic. No credit for holdup or dilution or by ESF ventilation filtration systems, may be credited where of ESF component leakage is taken. A applicable. Filter systems used in these applications should be time dependent iodine partition evaluated against the guidance of Regulatory Guide 1.52 (Ref. A- coefficient is used to determine the 5) and Generic Letter 99-02 (Ref. A-6). iodine released from the RWST liquid and the MEDT liquid.(Att. 4 Sections 7.2.3.3 -7.2.3.6)The leakage from the passive RHR pump seal failure is a filtered release.The filter efficiencies are determined in accordance with guidance provided GL 99-02 and controlled by TS 5.5.11 (VFTP) in accordance with RG 1.52, Revision 2, ANSI N510 1980, and ASTM D3803-1989. The LAR proposed a change to TS 5.5.11, as outlined in Section 2.(Att. 4 Section 7.2.3.4)Assumptions on Main Steam Isolation Valve Leakage in BWRs A5-32 ~~K~i~Thb~e s~A5- B ~o~forri~iandb!~th Ie jatryGu4e `118i AppedxA(oso oIn ciet 6. For BWRs, the main steam isolation valves (MSIVs) have design N/A Regulatory Positions 6.1 through 6.5 leakage that may result in a radioactivity release. The radiological relate to MSIV leakage in BWRs, which consequences from postulated MSIV leakage should be analyzed is not applicable to DCPP.and combined with consequences postulated for other fission product release paths to determine the total calculated radiological consequences from the LOCA. The following assumptions are acceptable for evaluating the consequences of MSIV leakage.Assumption on Containment Purging 7. The radiological consequences from post-LOCA primary N/A DCPP does not require the use of post-containment purging as a combustible gas or pressure control LOCA containment purging as a measure should be analyzed. If the installed containment purging combustible gas or pressure control capabilities are maintained for purposes of severe accident measure. The 48-inch containment management and are not credited in any design basis analysis, purge valves will be sealed closed during radiological consequences need not be evaluated. If the primary MODES 1, 2, 3, and 4, as required by containment purging is required within 30 days of the LOCA, the the proposed revision to TS 3.6.3. See results of this analysis should be combined with consequences LAR Enclosure Section 2.postulated for other fission product release paths to determine the total calculated radiological consequences from the LOCA.Reduction in the amount of radioactive material released via ESF filter systems may be taken into account provided that these systems meet the guidance in Regulatory Guide 1.52 (Ref. A-5)and Generic Letter 99-02 (Ref. A-6).A5-33 Source Term 1. Acceptable assumptions regarding core inventory and the Conforms See Table A5-A above for conformance release of radionuclides from the fuel are provided in Regulatory with Regulatory Guide 1.183 Appendix B Position 3 of this guide. (Fu el Handling Accident) source terms.The radiological source term for the FHA is based on the equilibrium core inventory determined

with the computer code ORIGEN-S as discussed in Att. 4 Section 4.1 and Table 4.1-1. All fuel rods in one fuel assembly are assumed damaged in the FHA. The FHA is postulated to occur 72 hours after shutdown.

The radionuclides relevant to the dose analysis of the postulated FHA for a single fuel assembly at 72 hours post reactor shutdown are shown in Att.4 Table 7.3-2.The FHA now credits CRVS, which is initiated by radiation monitors. The response time for radiation monitors is dependent on the magnitude of the radiation level and energy spectrum of the airborne cloud at the location of the detector, which are dependent on the fuel decay time. Therefore, a delayed FHA at fuel offload and fuel reload are also evaluated.(Att. 4 Section 7.3)A5-34 --~ -',:~ -The number of fuel rods damaged during the accident should be based on a conservative analysis that considers the most limiting case. The analysis should consider parameters such as the weight of the dropped heavy load or the weight of a dropped fuel assembly (plus any attached handling grapples), the height of the drop, and the compression, torsion, and shear stresses on the irradiated fuel rods. Damage to adjacent fuel assemblies, if applicable (e.g., events over the reactor vessel), should be considered. The DCPP FHA utilizes the current licensing basis assumption that all fuel rods in one assembly are damaged. As documented in the NRC SER for License Amendments 8 and 6 to DCPP Facility Operating License Nos. DPR-80 and DPR-82, respectively, the assumption that all fuel rods in one assembly rupture is conservative because the kinetic energy available for causing damage to a fuel assembly dropped through water is fixed by the drop distance. The kinetic energy associated with the maximum drop height for a fuel handling accident is not considered sufficient to rupture the equivalent number of fuel rods of one assembly in both the dropped assembly and the impacted assembly. (Att. 4 Section 7.3)A5-35 i's" The fission product release from the breached fuel is based on Regulatory Position 3.2 of this guide and the estimate of the number of fuel rods breached. All the gap activity in the damaged rods is assumed to be instantaneously released.Radionuclides that should be considered include xenons, kryptons, halogens, cesiums, and rubidiums. The fission product release from the breached fuel is based on the breach of all fuel rods in one fuel assembly and the following gap fractions. All gap activity is assumed to be instantaneously released.The gap fractions used for the FHA are based on the bounding (higher) values provided per isotope/isotope class in the following: RG 1.25, NUREG/CR 5009, and RG 1.183, Table 3. The gap fractions used have been accepted in prior licensing applications for other utilities.(Att. 4 Sections 4.3, and, 7.3)1.3 The chemical form of radioiodine released from the fuel to the spent fuel pool should be assumed to be 95% cesium iodide (Csl), 4.85% elemental iodine, and 0.15% organic iodide. The CsI released from the fuel is assumed to completely dissociate in the pool water. Because of the low pH of the pool water, the iodine re-evolves as elemental iodine. This is assumed to occur instantaneously. The NRC staff will consider, on a case-by-case basis, justifiable mechanistic treatment of the iodine release from the pool.Conforms The iodine release into the pool from the fuel is assumed to be 95% Csl, 4.85%elemental iodine and 0.15% organic iodine. Due to the acidic nature of the water in the fuel pool (pH less than 7), the Csl is assumed to immediately disassociate and re-evolve as elemental iodine, thus changing the chemical form of iodine to 99.85% elemental and 0.15%organic.(Att. 4 Section 7.3)Water Depth A5-36 -Tw11, A f I3 one mdjR FSi 'H6hdnin~Af! itvry A~nfl i fo -n!2. If the depth of water above the damaged fuel is 23 feet or Conforms The depth of the water above the greater, the decontamination factors for the elemental and damaged fuel is greater than 23 feet. An organic species are 500 and 1, respectively, giving an overall Iodine decontamination factor of 200 is effective decontamination factor of 200 (i.e., 99.5% of the total assumed. The chemical form of the iodine released from the damaged rods is retained by the water). iodines above the pool is 57% elemental This difference in decontamination factors for elemental and 43% organic.(99.85%) and organic iodine (0.15%) species results in the iodine (Att. 4 Section 7.3)above the water being composed of 57% elemental and 43%organic species. If the depth of water is not 23 feet, the decontamination factor will have to be determined on a case-by-case method (Ref. B-i).Noble Gases 3. The retention of noble gases in the water in the fuel pool or Conforms The noble gas DF is assumed as 1 reactor cavity is negligible (i.e., decontamination factor of 1). resulting in negligible retention in water.Particulate radionuclides are assumed to be retained by the All alkali metals in the form of water in the fuel pool or reactor cavity (i.e., infinite particulates are retained in the pool.decontamination factor). (Att. 4 Section 7.3)Fuel Handling Accidents Within The Fuel Building 4.1 The radioactive material that escapes from the fuel pool to the Conforms It has been determined that for the FHA fuel building is assumed to be released to the environment over in the FHB, the actual release rate a 2-hour time period, lambda based on the FHBVS exhaust (i.e., 8.7 hr') is larger than the release rate applicable to "a 2-hr release" (i.e., 3.45 hr 1). Thus the larger exhaust rate lambda associated with FHBVS operation plus the exhaust rate lambda for the 500 cfm outleakage is utilized in the analysis.Att. 4 Section 7.3)A5-37

i.A reduction in the amount of radioactive material released from the fuel pool by engineered safety feature (ESF) filter systems may be taken into account provided these systems meet the guidance of Regulatory Guide 1.52 and Generic Letter 99-02 (Refs. B-2, B-3). Delays in radiation detection, actuation of the ESF filtration system, or diversion of ventilation flow to the ESF filtration system' should be determined and accounted for in the radioactivity release analyses.1 These analyses should consider the time for the radioactivity concentration to reach levels corresponding to the monitor setpoint, instrument line sampling time, detector response time, diversion damper alignment time, and filter system actuation, as applicable.

No ESF filtration is credited for EAB and LPZ doses. The CR dose analysis credits radiation monitors to switch the CRVS from Mode 1 (normal) to Mode 4 (filtered and pressurized). A 22 second delay is conservatively assumed, including instrument loop uncertainties, from the onset of the event to Mode 4 operation. CR filter efficiencies are determined in accordance with the guidance provided in GL 99-02 and are controlled by TS 5.5.11.(Att. 4 Sections 7.1 and 7.3)4.3 The radioactivity release from the fuel pool should be assumed to be drawn into the ESF filtration system without mixing or dilution in the fuel building. If mixing can be demonstrated, credit for mixing and dilution may be considered on a case-by-case basis. This evaluation should consider the magnitude of the building volume and exhaust rate, the potential for bypass to the environment, the location of exhaust plenums relative to the surface of the pool, recirculation ventilation systems, and internal walls and floors that impede stream flow between the surface of the pool and the exhaust plenums.Conforms Noble gases and iodines are released from the pool and mixed in the available air space, but all activity is released to the environment in less than 2-hours.The radionuclides are released to the environment via the Plant Vent with an assumed 500 cfm outleakage to the environment from the FHB.(Att. 4 Section 7.3)Fuell Handling Accidents Within Containment A5-38 ~ib~ A-C ~o~omace&~f~(qua~r ~~uue~ ~Pen'P {e i-iaulnq g 5.1 If the containment is isolated 2 during fuel handling operations, no Conforms Containment isolation is not credited in radiological consequences need to be analyzed. the analysis. Therefore, a radiological consequence analysis is performed. 2 Containment isolation does not imply containment integrity as (Att. 4 Section 7.3)defined-by technical specifications for non-shutdown modes.The term isolation is used here collectively to encompass both containment integrity and containment closure, typically in place during shutdown periods. To be credited in the analysis, the appropriate form of isolation should be addressed in technical specifications. 5.2 If the containment is open during fuel handling operations, but Conforms Automatic containment isolation is not designed to automatically isolate in the event of a fuel handling credited. Therefore, a radiological accident, the release duration should be based on delays in consequence analysis is performed. radiation detection and completion of containment isolation. If it (Att. 4 Section 7.3)can be shown that containment isolation occurs before radioactivity is released to the environment, no radiological consequences need to be analyzed.A5-39 in~j ~4 .r j-~ 4 ~,, ~If the containment is open during fuel handling operations (e.g., personnel air lock or equipment hatch is open)3 , the radioactive material that escapes from the reactor cavity pool to the containment is released to the environment over a 2-hour time period.3 The staff will generally require that technical specifications allowing such operations include administrative controls to close the airlock, hatch, or open penetrations within 30 minutes. Such administrative controls will generally require that a dedicated individual be present, with necessary equipment available, to restore containment closure should a fuel handling accident occur. Radiological analyses should generally not credit this manual isolation. The radioactive material that escapes from the reactor cavity pool to the containment is released to the environment over a 2-hour time period.(Att. 4 Section 7.3)TS 3.9.4 allows the hatches to be open during fuel movements provided that provisions are in place for the hatches to be closed. The TS bases LCO provide guidance on how that capability is available and monitored and the accepted closure time is within 30 minutes. This LAR does not alter TS 3.9.4 or its Bases.+ I-5.4 A reduction in the amount of radioactive material released from the containment by ESF filter systems may be taken into account provided that these systems meet the guidance of Regulatory Guide 1.52 and Generic Letter 99-02 (Refs. B-2 and B-3). Delays in radiation detection, actuation of the ESF filtration system, or diversion of ventilation flow to the ESF filtration system should be determined and accounted for in the radioactivity release analyses.1 1 These analyses should consider the time for the radioactivity concentration to reach levels corresponding to the monitor setpoint, instrument line sampling time, detector response time, diversion damper alignment time, and filter system actuation, as ap)plicable. Conforms No ESF filtration is credited for releases from the containment to the environment. The CR dose analysis credits radiation monitors to switch the CRVS from Mode 1 (normal) to Mode 4 (filtered and pressurized). A 22 second delay is conservatively assumed, including instrument loop uncertainties, from the onset of the event to Mode 4 operation. CR filter efficiencies are determined in accordance with the guidance provided in GL 99-02 and TS 5.5.11.(Att. 4 Sections 7.1 and 7.3)A5-40 T-44bi A5-C o~I1r3 8'iPpr0VjdiiB.~(u IadIig vcdp)5.5 Credit for dilution or mixing of the activity released from the Conforms All airborne activity is released within a 2 reactor cavity by natural or forced convection inside the hour period.containment may be considered on a case-by-case basis. Such (Att. 4 Section 7.3)credit is generally limited to 50% of the containment free volume.This evaluation should consider the magnitude of the containment volume and exhaust rate, the potential for bypass to the environment, the location of exhaust plenums relative to the surface of the reactor cavity, recirculation ventilation systems, and internal walls and floors that impede stream flow between the surface of the reactor cavity and the exhaust plenums.A5-41 bource i erms 1. Assumptions acceptable to the NRC staff regarding core Conforms No fuel melt or clad breach is postulated inventory and the release of radionuclides from the fuel are for the DCPP MSLB event. See Item 2 provided in Regulatory Position 3 of this regulatory guide. The below for source terms.release from the breached fuel is based on Regulatory Position (Att. 4 Section 7.6)3.2 of this guide and the estimate of the number of fuel rods breached. The fuel damage estimate should assume that the highest worth control rod is stuck at its fully withdrawn position.2. If no or minimal 2 fuel damage is postulated for the limiting event, Conforms No fuel damage is postulated for the the activity released should be the maximum coolant activity MSLB. Activity released is based on the allowed by the technical specifications. Two cases of iodine maximum coolant activity allowed by TS.spiking should be assumed. Two cases of iodine spiking are analyzed, pre-accident spike and 2 The activity assumed in the analysis should be based on the accident-initiated spike.activity associated with the projected fuel damage or the maximum technical specification value, whichever maximizes In addition to the activity associated with the radiological consequences. In determining dose equivalent the DEI, the initial primary coolant DEX 1-131 (DE 1-131), only the radioiodine associated with normal is assumed to be 270 pCi/gm (Revised operations or iodine spikes should be included. Activity from TS SR 3.4.16.1 value). The initial projected fuel damage should not be included. secondary coolant iodine activity is assumed to be at the TS limit of 0.1 pCi/gm DEI (TS 3.7.18).(Att. 4 Section 7.6)2.1 A reactor transient has occurred prior to the postulated main Conforms For the pre-accident iodine spike case, it steam line break (MSLB) and has raised the primary coolant is assumed that a reactor transient has iodine concentration to the maximum value (typically 60 pCi/ DE occurred prior to the MSLB and has 1-131) permitted by the technical specifications (i.e., a pre- raised the RCS iodine concentration to a accident iodine spike case). value of 60 pCi/gm of DEI (TS 3.4.16 limit). (Att. 4 Section 7.6)A5-42 i -TMRý,,' -i _ý -1pam 'M Pe K --S -in 0" (ONrlo' n 6ý4ý 0,044w, ý.A, -'A' ndlilikf U 2.2 The primary system transient associated with the MSLB causes Conforms For the accident-initiated iodine spike an iodine spike in the primary system. The increase in primary case, the MSLB causes an iodine spike coolant iodine concentration is estimated using a spiking model in the RCS, which increases the iodine that assumes that the iodine release rate from the fuel rods to release rate from the fuel to the RCS to a the primary coolant (expressed in curies per unit time) increases value 500 times the appearance rate to a value 500 times greater than the release rate corresponding corresponding to a maximum equilibrium to the iodine concentration at the equilibrium value (typically 1.0 RCS concentration of 1.0 pCi/gm of DEL.pCi/gm DE 1-131) specified in technical specifications (i.e., The spike is allowed to continue until 8 concurrent iodine spike case). A concurrent iodine spike need hours from the start of the event. After not be considered if fuel damage is postulated. The assumed this point in the accident there is no iodine spike duration should be 8 hours. Shorter spike durations activity available for release from the may be considered on a case-by-case basis if it can be shown gap.that the activity released by the 8- hour spike exceeds that (Att. 4 Section 7.6)available for release from the fuel gap of all fuel pins.3. The activity released from the fuel should be assumed to be N/A No fuel damage occurs due to a MSLB.released instantaneously and homogeneously through the The activity released to the environment primary coolant. is based on the maximum coolant activity allowed by TS.(Att. 4 Section 7.6)4. The chemical form of radioiodine released from the fuel should Conforms No fuel damage occurs due to a MSLB.be assumed to be 95% cesium iodide (CsI), 4.85 percent The chemical form of iodine released elemental iodine, and 0.15 percent organic iodide. Iodine from the steam generators to the releases from the steam generators to the environment should environment due to the MSLB is be assumed to be 97% elemental and 3% organic. These assumed to be 97% elemental and 3%fractions apply to iodine released as a result of fuel damage and organic.to iodine released during normal operations, including iodine (Att. 4 Section 7.6)spiking.A5-43 3 1n this appendix, ruptured refers to the state of the steam generator in which primary-to-secondary leakage rate has increased to a value greater than technical specifications. Faulted refers to the state of the steam generator in which the secondary side has been depressurized by a MSLB such that protective system response (main steam line isolation, reactor trip, safety injection, etc.) has occurred. Partitioning Coefficient is defined as: mass of 12 per unit mass of liquid PC = 12 per unit mass of gas mass of 12 per unit mass of gas 5.1 For facilities that have not implemented alternative repair criteria Conforms The primary-to-secondary leak rate (see Ref. E-1, DG-1074), the primary-to-secondary leak rate in assumed in the analysis is a total of 0.75 the steam generators should be assumed to be the leak rate gpm for all four steam generators. This limiting condition for operation specified in the technical equates to 1080 gpd from all SGs which specifications. For facilities with traditional generator is greater than the maximum allowable specifications (both per generator and total of all generators), operational leakage of 150 gpd from any the leakage should be apportioned between affected and one SG imposed by TS 3.4.13d.unaffected steam generators in such a manner that the Conservatively, the total 0.75 gpm tube calculated dose is maximized. leakage will be assigned to the faulted SG.(Att. 4 Section 7.6)A5-44 "Orm ýa T rh qipý, g U' $q Gr 11", K rGa, b AP,6 Li nf'5.2 The density used in converting volumetric leak rates (e.g., gpm) Conforms The leakage density is assumed to be to mass leak rates (e.g.,. lbm/hr) should be consistent with the 1.0 gm/cc (62.4 Ibm/ft 3).basis of the parameter being converted. The ARC leak rate correlations are generally based on the collection of cooled (Att. 4 Section 7.6, Table 7.6-1)liquid. Surveillance tests and facility instrumentation used to show compliance with leak rate technical specifications are typically based on cooled liquid. In most cases, the density should be assumed to be 1.0 gm/cc (62.4 Ibm/ft 3).5.3 The primary-to-secondary leakage should be assumed to Conforms The primary to secondary SG tube continue until the primary system pressure is less than the leakage is assumed to occur until the secondary system pressure, or until the temperature of the RCS reaches 212'F, which is leakage is less than 100 C (212 F). The release of radioactivity conservatively estimated to occur 30 from unaffected steam generators should be assumed to hours after the event.continue until shutdown cooling is in operation and releases (Att. 4 Section 7.6)from the steam generators have been terminated. 5.4 All noble gas radionuclides released from the primary system Conforms All noble gases are released freely with are assumed to be released to the environment without no retention or mitigation. reduction or mitigation. (Att. 4 Section 7.6)5.5 The transport model described in the section should be utilized for iodine and particular releases from the steam generators. This model is shown in Figure E-1 and summarized below: A5-45 abeA5-D: Qo~r~nc itW n" ~t# QGid 1A8, M):,di W Q0,~n S mLie Brek , I .--- k- M A--ntIi _% M-M -APT 5.5.1 A portion of the primary-to-secondary leakage will flash to vapor, based on the thermodynamic conditions in the reactor and secondary coolant." During periods of steam generator dryout, all of the primary-to-secondary leakage is assumed to flash to vapor and be released to the environment with no mitigation." With regard to the unaffected steam generators used for plant cooldown, the primary-to-secondary leakage can be assumed to mix with the secondary water without flashing during periods of total tube submergence. Conforms RCS activities are released to the faulted and intact SGs via tube leakage.* Faulted SG: Due to dry-out, the entire inventory of noble gases and iodines in the SG are released to the environment via the steam line break point without mitigation. The maximum allowable primary to secondary SG tube leakage for all SGs is conservatively assumed to occur in the faulted SG. All iodine and noble gas activities in the tube leakage are assumed to be released directly to the environment without hold-up or mitigation.

Intact SGs: With a loss of offsite power (LOOP), the main steam condenser is not available.

lodines in the intact SGs secondary coolant are released to the environment in proportion to the steaming rate and the inverse of the partition coefficient of 100. Noble gases are released without retention. No primary to secondary leakage intothe intact SGs occurs because all tube leakage is conservatively assumed to occur in the faultedSG.(Att. 4 Section 7.6)A5-46 5.5.2 The leakage that immediately flashes to vapor will rise through Conforms No credit is taken for iodine scrubbing in the bulk water of the steam generator and enter the steam the SG bulk water. Any postulated space. Credit may be taken for scrubbing in the generator, using leakage that immediately flashes to the models in NUREG-0409, "Iodine Behavior in a PWR Cooling vapor is assumed to rise through the System Following a Postulated Steam Generator Tube Rupture bulk water of the SG into the steam Accident" (Ref. E-2), during periods of total submergence of the space and is assumed to be immediately tubes. released to the environment.(Att. 4 Section 7.6)5.5.3 The leakage that does not immediately flash is assumed to mix Conforms All leakage is conservatively assumed to with the bulk water. take place in the faulted SG and immediately flash to steam.(Att. 4 Section 7.6)5.5.4 The radioactivity in the bulk water is assumed to become vapor Conforms lodines in the intact SGs secondary at a rate that is the function of the steaming rate and the coolant is assumed to be a TS level for partition coefficient. A partition coefficient for iodine of 100 may secondary system activity. The iodines be assumed. The retention of particulate radionuclides in the are released to the environment in steam generators is limited by the moisture carryover from the proportion to the steaming rate and the steam generators. inverse of the partition coefficient of 100.No fuel damage occurs due to the MSLB, therefore there are no particulate radionuclides available for release.(Att. 4 Section 7.6)5.6 Operating experience and analyses have shown that for some Conforms Steam generator tube bundle uncovery steam generator designs, tube uncover may occur for a short is not predicted or postulated for the period following any reactor trip (Ref. E-3). The potential impact intact SG.of tube uncovery on the transport model parameters (e.g., flash (Att. 4 Section 7.6)fraction, scrubbing credit) needs to be considered. The impact of emergency operating procedure restoration strategies on steam generator water levels should be evaluated. A5-47 n~ii R J~P~tijn g Afayi.. Comets Source Terms 1 Assumptions acceptable to the NRC staff regarding core Conforms No fuel melt or clad breach is postulated inventory and the release of radionuclides from the fuel are for the SGTR event. See Item 2 below provided in Regulatory Position 3 of this regulatory guide. The for source terms.release from the breached fuel is based on Regulatory Position 3.2 of this guide and the estimate of the number of fuel rods breached.2. If no or minimal 2 fuel damage is postulated for the limiting event, Conforms No fuel damage is postulated for the the activity released should be the maximum coolant activity SGTR. Activity released is based on the allowed by the technical specifications. Two cases of iodine maximum coolant activity allowed by TS.spiking should be assumed. Two cases of iodine spiking are analyzed, pre-accident spike and 2 The activity assumed in the analysis should be based on the accident-initiated spike.activity associated with the projected fuel damage or the maximum technical specification values, whichever maximizes In addition to the activity associated with the radiological consequences. In determining dose equivalent I- the DEl, the initial primary coolant DEX 131 (DE 1-131), only the radioiodine associated with normal isassumed to be,270 pCi/gm (Revised operations or iodine spikes should be included. Activity from TS SR 3.4.16.1 value). The initial projected fuel damage should not be included: secondary coolant iodine activity is assumed to be at the TS limit of 0.1 pCi/gm DEI (TS 3.7.18).(Att. 4 Section 7.7)2.1 A reactor transient has occurred prior to the postulated steam Conforms For the pre-accident iodine spike case, it generator tube rupture (SGTR) and has raised the primary is assumed that a reactor transient has coolant iodine concentration to the maximum value (typically 60 occurred prior to the SGTR and has pCi/ DE 1-131) permitted by the technical specifications (i.e., a raised the RCS iodine concentration to a pre-accident iodine.spike case). value of 60 pCi/gm of DEI (TS 3.4.16 limit).(Att. 4 Section 7.7)A5-48 tomnen M*The primary system transient associated with the SGTR causes an iodine spike in the primary system. The increase in primary coolant iodine concentration is estimated using a spiking model that assumes that the iodine release rate from the fuel rods to the primary coolant (expressed in curies per unit time) increases to a value 335 times greater than the release rate corresponding to the iodine concentration at the equilibrium value (typically 1.0 pCi/gm DE 1-131) specified in technical specifications (i.e., concurrent iodine spike case). A concurrent iodine spike need not be considered if fuel damage is postulated. The assumed iodine spike duration should be 8 hours. Shorter spike durations may be considered on a case-by-case basis if it can be shown that the activity released by the 8- hour spike exceeds that available for release from the fuel gap of all fuel pins.For the accident-initiated iodine spike case, the primary system transient associated with the SGTR causes an iodine spike in the RCS, which increases the iodine release rate from the fuel to the RCS to a value 335 times the appearance rate corresponding to a maximum equilibrium RCS concentration of 1.0 pCi/gm of DEI (TS SR 3.4.16.2). The spike is allowed to continue until 8 hours from the start of the event.(Att. 4 Section 7.7)The activity released from the fuel, if any, should be assumed to N/A No fuel damage occurs due to a SGTR.be released instantaneously and homogeneously through the The activity released to the environment primary coolant. is based on the maximum coolant activity allowed by TS.(Att. 4 Section 7.7)Iodine releases from the steam generators to the environment Conforms The iodine releases from the steam should be assumed to be 97% elemental and 3% organic. generators to the environment is assumed to be 97% elemental and 3%organic.(Att. 4 Section 7.7)A5-49 3 1n this appendix, ruptured refers to the state of the steam generator in which primary-to-secondary leakage rate has increased to a value wreater than technical specifications. 5.1 The primary-to-secondary leak rate in the steam generators Conforms The primary-to-secondary leak rate is a should be assumed to be the leak rate limiting condition for total of 0.75 gpm at STP for all four operation specified in the technical specifications. The leakage steam generators. This equates to 1080 should be apportioned between affected and unaffected steam gpm for all 4 SGs which is greater than generators in such a manner that the calculated dose is the maximum allowable operational maximized. leakage of 150 gpd for any one SG imposed by TS 3.4.13d. Conservatively, the total 0.75 gpm tube leakage will be assigned to the 3 intact SGs.(Att. 4 Section 7.7)5.2 The density used in converting volumetric leak rates (e.g., gpm) Conforms The leakage density is assumed to be to mass leak rates (e.g., Ibm/hr) should be consistent with the 1.0 gm/cc (62.4 Ibm/ft 3).basis of surveillance tests used to show compliance with leak rate technical specifications. These tests are typically based on (Att. 4 Section 7.7, Table 7.7-1)cool liquid. Facility instrumentation used to determine leakage is typically located on lines containing cool liquids. In most cases, the density should be assumed to be 1.0 gm/cc (62.4 lbm/ft 3).A5-50 ...... F5~ #. ~of n iIeItr~Gu~ei8,Ad4W Steam ,Ge6nerato 4Tub Ru't~ ~*PP.5.3 The primary-to-secondary leakage should be assumed to Conforms In the ruptured SG, after the reactor trip, continue until the primary system pressure is less than the the radioactivity in the steam is released secondary system pressure, or until the temperature of the to the environment from the MSSVs/1 0%leakage is less than 100 C (212 F). The release of radioactivity ADVs due to the assumption of LOOP. It from unaffected steam generators should be assumed to is assumed that the 10% ADV of the continue until shutdown cooling is in operation and releases from ruptured SG fails open for 30 minutes.the steam generators have been terminated. The fail-open 10% ADV is isolated at 2653 seconds, at which time the ruptured steam loop is isolated. The break flow continues until equilibrium between the primary and secondary side of the ruptured SG is reached (5872 seconds). Manual depressurization of the ruptured SG starts 2 hours after event initiation and continues until shutdown cooling is in operation (10.73 hours).In the intact SGs, release of radioactivity is assumed to continue until shutdown cooling is in operation (10.73 hours).(Att. 4 Section 7.7)5.4 The release of fission products from.the secondary system Conforms A loss of offsite power is assumed to should be evaluated with the assumption of a coincident loss of occur at the time of the reactor trip.offsite power. (Att. 4 Section 7.7)5.5 All noble gas radionuclides released from the primary system are Conforms All noble gases are released freely with assumed to be released to the environment without reduction or no retention or mitigation. mitigation. (Att. 4 Section 7.7)5.6 The transport model described in Regulatory Positions 5.5 and 5.6 of Appendix E should be utilized for iodine and particulates. A5-51 ~ThbI A~ Elc~nj i~nc~i~b ~ Apencii F~WRteamren~rtorbeRpu) Regulatory Positions 5.5.1 of Appendix E Conforms RCS activities are released to the ruptured SG through the tube rupture A portion of the primary-to-secondary leakage will flash to vapor, and to the intact SGs via tube leakage as based on the thermodynamic conditions in the reactor and well as a portion of the break flow carried secondary coolant. over from the ruptured SG via the* During periods of steam generator dryout, all of the primary- condenser before the reactor trip.to-secondary leakage is assumed to flash to vapor and be

Ruptured SG: The noble gases in released to the environment with no mitigation.

the entire break flow and the iodine* With regard to the unaffected steam generators used for in the flashed portion of the break plant cooldown, the primary-to-secondary leakage can be flow are assumed to be released assumed to mix with the secondary water without flashing directly to the environment without during periods of total tube submergence. hold-up or mitigation.

Intact SGs: lodines in the intact SGs secondary coolant including iodine due to tube leakage are released to the environment in proportion to the steaming rate and the inverse of the partition coefficient of 100. Noble gases are released without retention.(Att. 4 Section 7.7)Regulatory Positions 5.5.2 of Appendix E Conforms No credit is taken for iodine scrubbing in the SG bulk water. Any postulated The leakage that immediately flashes to vapor will rise through leakage that immediately flashes to the bulk water of the steam generator and enter the steam vapor is assumed to rise through the space. Credit may be taken for scrubbing in the generator, using bulk water of the SG into the steam the models in NUREG-0409, "Iodine Behavior in a PWR Cooling space and is assumed to be immediately System Following a Postulated Steam Generator Tube Rupture released to the environment.

Accident" (Ref. E-2), during periods of total submergence of the (Att. 4 Section 7.7)tubes.Regulatory Positions 5.5.3 of Appendix E Conforms The non-flashed portion of the break flow mixes uniformly with the steam generator The leakage that does not immediately flash is assumed to mix liquid mass.with the bulk water. .(Att. 4 Section 7.7)A5-52 G -gig n Regulatory Positions 5.5.4 of Appendix E The radioactivity in the bulk water is assumed to become vapor at a rate that is the function of the steaming rate and the partition coefficient. A partition coefficient for iodine of 100 may be assumed. The retention of particulate radionuclides in the steam generators is limited by the moisture carryover from the steam generators. In the ruptured SG, the non-flashed portion of the break flow mixes uniformly with the steam generator liquid mass and is released into the steam space in proportion to the steaming rate and the inverse of the partition coefficient of 100.lodines in the intact SGs secondary coolant are released to the environment in proportion to the steaming rate and the inverse of the partition coefficient of 100. No fuel damage occurs due to the SGTR, therefore there are no particulate radionuclides available for release.(Att. 4 Section 7.7)Regulatory Positions 5.6 of Appendix E Conforms The amount of steam generator tube bundle uncovery is predicted to be Operating experience and analyses have shown that for some insignificant for the intact SGs.steam generator designs, tube uncover may occur for a short (Att. 4 Section 7.7, Table 7.7-1)period following any reactor trip (Ref. E-3). The potential impact of tube uncovery on the transport model parameters (e.g., flash fraction, scrubbing credit) needs to be considered. The impact of emergency operating procedure restoration strategies on steam generator water levels should be evaluated. A5-53

.T:jA5F F", n:f~iac ihRiabyGi~18, edi~PRL~e

~o cietv Source Terms Assumptions acceptable to the NRC staff regarding core Conforms See Table A5-A above for conformance inventory and the release of radionuclides from the fuel are with Regulatory Guide 1.183, Appendix provided in Regulatory Position 3 of this regulatory guide. The G (PWR Locked Rotor Accident) source release from the breached fuel is based on Regulatory Position terms.3.2 of this guide and the estimate of the number of fuel rods breached.If no fuel damage is postulated for the limiting event, a N/A Since fuel damage is postulated, a radiological analysis is not required as the consequences of the radiological consequence analysis is event are bounded by the consequences projected for the main performed. steam line break outside containment. The activity released from the fuel should be assumed to be Conforms The activity released from the fuel is released instantaneously and homogeneously through the assumed to be released instantaneously primary coolant. and homogeneously through the primary coolant. (Att. 4 Section 7.4)4.I Te chemical form of radioiodine released from the fuel should be 95% cesium iodide (Csl), 4.85 percent elemental iodine, and 0.15 percent organic iodide. Iodine releases from the steam generators to the environment should be assumed to be 97%elemental and 3% organic. These fractions apply to iodine released as a result of fuel damage and to iodine released during normal operation, including iodine spiking.Conforms The chemical form of iodine released from the fuel is assumed to be 95% Csl, 4.85% elemental iodine, and 0.15%organic iodide.The chemical for of iodine released from the SGs to the environment is assumed to be 97% elemental and 3% organic.(Att. 4 Section 7.4)+/- ____________ L ___________________________________________ Release Transport. A5-54 jmblMan "WiF ~~~~rrac~thReglfyGid I43A 040'ii'G PRLbked- Rot AceW-5.1 The primary-to-secondary leak rate in the steam generators Conforms The primary-to-secondary leak rate is a should be assumed to be the leak-rate-limiting condition for total of 0.75 gpm at STP for all four operation specified in the technical specifications. The leakage steam generators. This equates to a should be apportioned between the steam generators in such a total of 1080 gpd, which is greater than manner that the calculated dose is maximized. the maximum allowable operational leakage of 150 gpd for any one SG imposed by TS 3.4.13d.(Att. 4 Section 7.4)5.2 The density used in converting volumetric leak rates (e.g., gpm) Conforms The leakage density is assumed to be to mass leak rates (e.g., Ibm/hr) should be consistent with the 1.0 gm/cc (62.4 Ibm/ft 3).basis of surveillance tests used to show compliance with leak (Att. 4 Section 7.4, Table 7.4-1)rate technical specifications. These tests are typically based on cool liquid. Facility instrumentation used to determine leakage is typically located on lines containing cool liquids. In most cases, the density should be assumed to be 1.0 gm/cc (62.4 lbm/ft 3).5.3 The primary-to-secondary leakage should be assumed to Conforms The primary-to-secondary leakage is continue until the primary system pressure is less than the assumed to continue until shutdown secondary system pressure, or until the temperature of the cooling is in operation and the steam leakage is less than 100 C (212 F). The release of radioactivity release from the SGs is terminated, should be assumed to continue until shutdown cooling is in which occurs at 10.73 hours after the operation and releases from the steam generators have been initiation of the LRA.terminated. (Att. 4 Section 7.4, Table 7.4-1)5.4 The release of fission products from the secondary system Conforms A loss of offsite power is assumed to should be evaluated with the assumption of a coincident loss of occur at the time of the reactor trip.offsite power. (Att. 4 Section 7.4)5.5 All noble gas radionuclides released from the primary system are Conforms All noble gases are released freely with assumed to be released to the environment without reduction or no retention or mitigation. mitigation. (Att. 4 Section 7.4)A5-55 'Loceloo7ciet t2~1.The transport model described in Regulatory Positions 5.5 5.6 of Appendix E should be utilized for iodine particulates. Regulatory Position 5.6 refers to Appendix E, Regulatory Positions 5.5 and 5.6. The iodine transport model for release for the steam generators is as follows: Activity released from the fuel is assumed to be released instantaneously and mixed homogenously through the primary coolant mass and transmitted to the secondary side via primary to secondary SG tube leakage. SG tubes remain covered for the duration of the LRA; therefore, the gap iodines are assumed to have a partition coefficient of 100 in the SG. Because the amount of SG tube uncovery is insignificant, flashing does not occur. The gap noble gasesare released freely to the environment without retention in the SG whereas particulates are carried over in accordance with the design basis SG moisture carryover fraction.(Att. 4 Section 7.4, Table 7.4-1)A5-56 1~I 5 ~:~nf~t p~,~~ REu MRor GfuiZI hif3,Apedx tPW IA -d ,'-4 -Ift Source Terms 1.Assumptions acceptable to the NRC staff regarding core inventory are in Regulatory Position 3 of this guide. For the rod ejection accident, the release from the breached fuel is based on the estimate of the number of fuel rods breached and the assumption that 10% of the core inventory of the noble gases and iodines is in the fuel gap. The release attributed to fuel melting is based on the fraction of the fuel that reaches or exceeds the initiation temperature for fuel melting and the assumption that 100% of the noble gases and 25% of the iodines contained in that fraction are available for release from containment. For the secondary system release pathway, 100%of the noble gases and 50% of the iodines in that fraction are released to the reactor coolant.Conforms The source term for the control rod ejection accident (CREA) is based on RG 1.183, Regulatory Position 3. The radiological source term for the CREA is based on the equilibrium core inv'entory-determined with the computer code ORIGEN-S as discussed in Section 3.2.2.2 and shown in Table 3. The CREA assumes that 10% of the fuel fails. Thus the source terms for the CREA are 10%of the equilibrium core inventory in Table 3. No fuel melt occurs.It is assumed that 10% of the core inventory of the noble gases and iodines are in the fuel gap. Releases are multiplied by a radial peaking factor 1.65.(Att. 4 Section 7.5)2. If no fuel damage is postulated for the-limiting event, a N/A The CREA is assumed to result in a radiological analysis is not required as the consequences of this breach of 10% of the fuel rods in the event are bounded by the consequences projected for the loss- core, thus a radiological consequence of-coolant accident (LOCA), main steam line break, and steam analysis is performed. generator tube rupture.A5-57 ~.Tablle A5G Cnfrm n~ i~tWRqi~ InýrG~d4 A8A~ Ix H'X WR RdEicon ciden)3. Two release cases are to be considered. In the first, 100% of the Conforms Two release cases are analyzed. In the activity released from the fuel should be assumed to be released first, 100% of the activity released from instantaneously and homogeneously through the containment the fuel is assumed to be released atmosphere. In the second, 100% of the activity released from instantaneously and homogeneously the fuel should be assumed to be completely dissolved in the through the containment atmosphere. In primary coolant and available for release to the secondary the second, 100% of the activity released system. from the fuel is assumed to be completely dissolved in the primary coolant and available for release to the secondary system.(Att. 4 Section 7.5)4. The chemical form of radioiodine released to the containment Conforms The chemical form of radioiodine atmosphere should be assumed to be 95% cesium iodide (CsI), released to the containment is assumed 4.85% elemental iodine, and 0.15% organic iodide. If to be 95% CsI, 4.85% elemental iodine, containment sprays do not actuate or are terminated prior to and 0.15% organic iodide. However, no accumulating sump water, or if the containment sump pH is not credit is taken for spray initiation or pH controlled at values of 7 or greater, the iodine species should be control. Therefore, the iodine released evaluated on an individual case basis. Evaluations of pH should from the containment atmosphere to the consider the effect of acids created during the rod ejection environment is assumed to be 97%accident event, e.g., pyrolysis and radiolysis products. With the elemental and 3% organic.exception of elemental and organic iodine and noble gases, (Att. 4.Section 7.5)fission products should be assumed to be in particulate form.5. Iodine releases from the steam generators to the environment Conforms The chemical form of iodine released should be assumed to be 97% elemental and 3% organic. form the steam generators to the environment is 97% elemental and 3%organic. (Att. 4 Section 7.5)Transport From Containment A5-58 Guide U ac7 7A pfdiH (PFIRd ' Iq'noi~ti et 6.1 A reduction in the amount of radioactive material available for Conforms With the exception of decay, no other leakage from the containment that is due to natural deposition, process is credited for fission product containment sprays, recirculating filter systems, dual removal. (Att. 4 Section 7.5)containments, or other engineered safety features may be taken into account. Refer to Appendix A to this guide for guidance on acceptable methods and assumptions for evaluating these mechanisms. 6.2 The containment should be assumed to leak at the leak rate Conforms A containment leak rate, based on DCPP incorporated in the technical specifications at peak accident TS 5.5.16, of 0.1% of containment air pressure for the first 24 hours, and at 50% of this leak rate for the weight per day is assumed for the first 24 remaining duration of the accident. Peak accident pressure is the hours. After 24 hours, the containment maximum pressure defined in the technical specifications for leak rate is reduced by 50% to 0.05% of containment leak testing. Leakage from subatmospheric -containment air weight per day.containments is assumed to be terminated when the containment (Att. 4 Section 7.5, Table 7.5-1)is brought to a subatmospheric condition as defined in technical specifications. Transport from Secondary System 7.1 A leak rate equivalent to the primary-to-secondary leak rate Conforms The primary-to-secondary leak rate is a limiting condition.for operation specified in the technical total of 0.75 gpm at STP for all four specifications should be assumed to exist until shutdown cooling steam generators. This equates to a is in operation and releases from the steam generators have total of 1080 gpd from the SG, which is been terminated. greater than the maximum allowable operational leakage of 150 gpd from any one SG imposed by TS 3.4.13d.Releases from the SG terminate when shutdown cooling is initiated.(Att. 4 Section 7.5)A5-59 Ie 15'- I 7.2 The density used in converting volumetric leak rates (e.g., gpm) Conforms The leakage density is assumed to be to mass leak rates (e.g., Ibm/hr) should be consistent with the 1.0 gm/cc (62.4 Ibm/ft 3).basis of surveillance tests used to show compliance with leak (Att. 4 Section 7.5, Table 7.5-1)rate technical specifications. These tests typically are based on cooled liquid. The facility's instrumentation used to determine leakage typically is located on lines containing cool liquids. In most cases, the density should be assumed to be 1.0 gm/cc (62.4 Ibm/ft 3).7.3 All noble gas radionuclides released to the secondary system. are Conforms All of the noble gas release to the assumed to be released to the environment without reduction or secondary side is assumed to be mitigation. released directly to the environment without reduction or mitigation.(Att. 4 Section 7.5)7.4 The transport model described in assumptions 5.5 and 5.6 of Conforms Regulatory Position 7.4 refers to Appendix E should be utilized for iodine and particulates. Appendix E, Regulatory Positions 5.5 and 5.6. The iodine transport model for release for the steam generators is as follows: Activity released from the fuel is assumed to be released instantaneously and mixed homogenously through the primary coolant mass and transmitted to the secondary side via primary to secondary SG tube leakage. SG tubes remain covered for the duration of the CREA; therefore, the gap iodines are assumed to have a partition coefficient of 100 in the SG. The gap noble gases are released freely to the environment without retention in the SG.(Att. 4 Section 7.5, Table 7.5-1)A5-60 T-4111' ýA5 I- H:;'ýý -ofonac wit Reiý~ ~ G~~c~.8 AýndixI(Eu,"t~aIf~tin1<- 1 -13 This appendix addresses assumptions associated with N/A Regulatory Positions I through 13 apply equipment qualification that are acceptable to the NRC staff for to equipment qualification radiological performing radiological assessments. As stated in Regulatory analyses. The DCPP EQ analysis will Position 6 of this guide, this appendix supersedes Regulatory continue to be based upon TID-14844 Positions 2.c.(1) and 2.c.(2) and Appendix D of Revision 1 of assumptions at this time.Regulatory Guide 1.89, "Environmental Qualification of Certain Electric Equipment Important to Safety for Nuclear Power Plants" (USNRC, June 1984), for operating reactors that have amended their licensing basis to use an alternative source term. Except as stated in this appendix, other assumptions, methods, and provisions of Revision 1 of Regulatory Guide 1.89 remain effective. A5-61 Enclosure Attachment 6 PG&E Letter DCL-15-069 ATTACHMENT 6 Diablo Canyon Power Plant Comparison to NRC Regulatory Issue Summary (RIS) 2006-04 Experience with Implementation of Alternative Source Terms ATTACHMENT 6 Diablo Canyon Power Plant Comparison to NRC Regulatory Issue Summary (RIS) 2006-04 Experience with Implementation of Alternative Source Terms RIS 2006-04 ISSUE ADDRESSED BY 1. Level of Detail Contained in LARs I An AST amendment request should describe the licensee's analyses of the radiological and non-radiological impacts and provide a justification for the proposed modification in sufficient detail to support review by the NRC staff. For example, the AST amendment request should: 1) Provide justification for each individual proposed change to the technical specification (TS), 2) Identify and justify each change to the licensing basis accident analyses, and 3) Contain enough details (e.g., assumptions, computer analyses input and output) to allow the NRC staff to confirm the dose analyses results in independent calculations. Attachment 4 of the LAR summarizes the analyses performed to support implementation of AST. Attachment 4 provides sufficient details to accommodate the stated purposes.Each AST analysis is described in detail with assumptions, inputs and results.Each change to the licensing basis accident analyses is identified and justified in Attachment 4 and the main body of the LAR. Justification for each individual proposed change to the technical specifications is included in the submittal. In addition, Appendix B to Attachment 4 contains analysis input tables comparing the current licensing basis values to new AST values for each of the analyzed design basis events. This information is being provided to aid the NRC staff review of the AST submittal.

2. Main Steam Isolation Valve (MSIV) Leakage and Fission Product This item is applicable to BWRs only.Deposition in Piping DCPP is a PWR; therefore this item is not applicable to DCPP.A6-1 RIS 2006-04 ISSUE I ADDRESSED BY 3. Control Room Habitability When implementing an AST, some licensees have proposed that certain engineered safety features (ESF) ventilation systems not be credited as a mitigation feature in response to an accident.

In some cases, the licensee's revised design basis analysis introduced the assumption that normal (non-ESF)ventilation systems are operating during all or part of an accident scenario. Such an assumption is inappropriate unless the non-ESF system meets certain qualities, attributes, and performance criteria as described in RG 1.183, Regulatory Positions 4.2.4 and 5.1.2. For example, credit for the operation of non-ESF ventilation systems should not be assumed unless they have a source of emergency power. In addition, the operation of ventilation systems establishes certain building or area pressures based upon their flowrates. These pressures affect leakage and infiltration rates which ultimately affect operator dose. Therefore, to credit the use of these systems, licensees should incorporate the systems into the ventilation filter testing program in Section 5 of the TS. In summary, use of non-ESF ventilation systems during a DBA should not be assumed unless the systems have emergency power and are part of the ventilation filter testing program in Section 5 of the TS.Generic Letter (G L) 2003-01, "Control Room Habitability" (Ref. 5) requested licensees to confirm the ability of their facility's control room to meet applicable habitability regulatory requirements. In addition, licensees were requested to confirm that control room habitability systems were designed, constructed, configured, operated and maintained in accordance with the facility's design and licensing bases. The GL placed emphasis on licensees confirming that the most limiting unfiltered inleakage into the control room envelope (CRE) was not greater than the value assumed in the DBA analyses. The tests, measurements, and analyses which were performed for this confirmation were to be described in the response to the GL. Some AST amendment requests proposed operating schemes for the control room and other ventilation systems which affect areas adiacent to the CRE and are different from the manner of operation and Operation of the ABVS is relied upon to ensure release of ESF leakage via the Plant vent. The ABVS filters are credited for releases of a RHR system pump seal passive failure following a LOCA when determining dose to offsite individuals, as well as onsite personnel in the control room and the TSC.The operation of the FHBVS following a FHA in the FHB is relied on to ensure a negative pressure in the FHB, which will result in post-accident environmental releases via the plant vent. FHBVS filters are not credited. The FHA now credits CRVS Mode 4 operation (filtered), which is actuated by 1/2- RE-25/26. See Attachment 4, Section 7.3.for more detail.CRVS Mode 4 operation (filtered) is also credited for LOCA, MSLB, SGTR, and CREA. Attachment 4, Section 7.1 provides the control room design, operation and transport model.All of these systems are safety related and have operability requirements in Technical Specifications. ESF filters are controlled by TS 5.5.11, "Ventilation Filter Testing Program (VFTP)." No chanqes to or)eratinq schemes of the A6-2 RIS 2006-04 ISSUE ADDRESSED BY performance described: in the response to the GL without providing sufficient CRVS, ABVS, or FHBVS are proposed.justification for the proposed changes in the operating scheme. In some cases, licensees proposed new modes of operation that lacked confirmation of the The AST CR dose analyses use a CRE inleakage characteristics. Measurements 1 of these characteristics are minimum unfiltered inleakage value of important to confirm inleakage assumptions used in the analyses for an AST 70 cfm, which includes 10 cfm for amendment, even for those situations in which the air in the control room would ingress/egress based on the guidance appear to be stagnant. provided by NUREG-0800. The 70 cfm provides margin above the actual 1 Use of parametric studies in which inleakage rates are varied is not a maximum recorded unfiltered inleakage preferable alternative to CRE inleakage measurements. (see GL 2003-01) of 37 cfm from the December 2012 Control Room Tracer Gas Test.4. Atmospheric Dispersion New x/Q values have been determined for the EAB, LPZ, Control Room, and Licensees may continue to use atmospheric relative concentration (x/Q) values TSC. Attachment 4, Section 5 provides and methodologies from their existing licensing-basis analyses when details of the calculations performed for appropriate. Licensees also have the option to adopt the generally less these new X/Q values. Included in conservative (more realistic) updated NRC staff guidance on determining X/Q Attachment 4, Appendix A, is a figure values in support of design basis control room radiological habitability showing plant north with locations of assessments provided in RG 1.194, "Atmospheric Relative Concentrations for releases and receptors or intakes.Control Room Radiological Habitability Assessments at Nuclear Power Plants" Justification for assumptions used to (Ref. 6). Regulatory positions on ,/Q values for offsite (i.e., exclusion area determine the control room X/Q values boundary and low population zone) accident radiological consequence are provided in Attachment 4, Section 5.assessments are provided in RG 1.145, "Atmospheric Dispersion Models for Meteorological data, and inputs used to Potential Accident Consequence Assessments at Nuclear Power Plants" (Ref. develop the control room and TSC I/Q 7). values are being provided in Attachment 4, Appendix A. See Attachment 4, Based on submittal reviews, the NRC staff identified the following areas of Section 5 and LAR Section 2.1 for improvement for licensee submittals that propose revision of the design basis discussion on RG 1.145 and RG 1.194 atmospheric dispersion analyses for implementing AST. They should include the compliance. following information: A6-3 RIS 2006-04 ISSUE ADDRESSED BY RIS 006-4 ISUE ADRESED B* A site plan showing true North and indicating locations of all potential accident release pathways and control room intake and unfiltered inleakage pathways (whether assumed or identified during inleakage testing)." Justification for using control room intake X/Q values for modeling the unfiltered inleakage, if applicable." A copy of the meteorological data inputs and program outputs along with a discussion of assumptions and potential deviations from staff guidelines. Meteorological data input files should be checked to ensure quality (e.g., compared against historical or other data and against the raw data to ensure that the electronic file has been properly formatted, any unit conversions are correct, and invalid data are properly identified). When running the control room atmospheric dispersion model ARCON96, two or more files of meteorological data representative of each potential release height should be used if X/Q values are being calculated for both ground-level and elevated releases (see RG 1.23, "Onsite Meteorological Programs," Regulatory Position 2 (Ref. 8) and Table A-2 in Appendix A to RG 1.194,"Atmospheric Relative Concentrations for Control Room Radiological Habitability Assessments at Nuclear Power Plants"). In addition, licensees should be aware that (1) two levels of wind speed and direction data should always be provided as input to each data file, (2) fields of "nines" (e.g., 9999)should be used to indicate invalid or missing data, and (3) valid wind direction data should range from 1 0 to 3600. Licensees should also provide detailed engineering information when applying the default plume rise adjustment cited in RG 1.194 to control room X/Q values to account for buoyancy or mechanical jets of high energy releases.This information should demonstrate that the minimum effluent velocity during any time of the release over which the adjustment is being applied is greater than the 95th percentile wind speed at the height of release.A6-4 RIS 2006-04 ISSUE ADDRESSED BY When running the offsite atmospheric dispersion model PAVAN, two or more files of meteorological data representative of each potential release height should be used if X/Q values are being calculated for pathways with significantly different release heights (e.g., ground level versus elevated stack). The joint frequency distributions of wind speed, wind direction, and atmospheric stability data used as input to PAVAN should have a large number of wind speed categories at the lower wind speeds in order to produce the best results (e.g., Section 4.6 of NUREG/CR-2858, "PAVAN: An Atmospheric Dispersion Program for Evaluating Design Basis Accidental Releases of Radioactive Materials from Nuclear Power Stations" (Ref. 9), suggests wind speed categories of calm, 0.5, 0.75, 1.0, 1.25, 1.5, 2.0, 3.0, 4.0 5.0, 6.0, 8.0 and 10.0 meters per second).5. Modeling of ESF Leakage Leakage from ESF systems used during post-LOCA recirculation is included in ESF systems that recirculate sump water outside the primary containment may the dose analyses and results are leak during their intended operation. This release source includes leak-age combined with other sources to through valve packing glands, pump shaft seals, flanged connections, and other determine the LOCA dose similar components. This release source may also include leakage through consequence. (Attachmen't 4 Section valves isolating interfacing systems (e.g., refueling water storage tank). 7.2.2). These include: Appendix A to RG 1.183, Regulatory Position 5, states that "the radiological e Leakage from ESF systems that consequences from the postulated [ESF] leakage should be analyzed and recirculate sump fluids outside combined with consequences postulated for other fission product release paths containment to determine the total calculated radiological consequences from the [loss-of-

RHR pump seal passive failure coolant accident]

LOCA." The allowable ESF leakage is typically contained in

RWST back-leakage the plant's TS or procedures.

The ESF leakage at accident conditions may differ a MEDT leakage from the ESF leakage at normal operating conditions. Licensees should account for ESF leakage at accident conditions in their dose analyses so as not to Attachment 4, Sections 7.2.2 and 7.2.3, underestimate the release rate. In Appendix A to RG 1.183, Regulatory Position describe each of these LOCA post-5.5, the NRC staff provided a conservative value of 10 percent as the assumed accident releases, including release amount of iodine that may become airborne from ESF leakage that is less than paths, flashing fractions, and release 212 0 F. The NRC staff structured this regulatory position to be deterministic and amounts. ESF leakage limits are conservative. The 10 percent value also compensates for the lack of research controlled by TS 5.5.2, "Primary Coolant A6-5 RIS 2006-04 ISSUE ADDRESSED BY concerning iodine speciation beyond the containment and the uncertainties of Sources Outside containment." DCPP applying laboratory data to the post-accident environment of the plant. fully complies with RG 1.183, Regulatory Position 5.5 states that a smaller flash fraction could be justified. Regulatory Position 5.5, as indicated in Some licensees have referenced NUREG/CR-5950, "Iodine Evolution and pH Attachment

5. (Attachment 4, Section Control" (Ref. 10) to justify a smaller flash fraction.

However, NUREG/CR-5950 7.2.3.2.5) was developed for very specific laboratory conditions and the results have a degree of uncertainty. The mechanism for release of the fluid is also uncertain. Leaked fluid may spray onto surfaces and evaporate, or be sprayed in fine droplets into the air. A value of less than 10 percent can be justified by including considerations for plant-specific variables, including the post-accident environment. (e.g., impurities in the water or the presence of organic substances) and the uncertainties in the application of research situations to plant environments. Figure 3.1 in NUREG/CR-5950 can be used to quantify the amount of elemental iodine as a function of the sump water pH and the concentration of iodine in the solution. In some cases, however, licensees have misapplied this figure. Rather than using the total concentration of iodine (i.e., stable and radioactive), licensees based their assessment on only the radioactive iodine in the sump water. By using only the radioactive iodine, licensees have underestimated how much iodine evolves during post-accident conditions. A6-6 RIS 2006-04 ISSUE ADDRESSED BY 6. Release Pathways No similar plant configuration changes as described by Issue 6 are associated Changes to the plant configuration associated with an LAR (e.g., an "open" with this LAR. Attachment 4, Section 5 containment during refueling) may require a re-analysis of the design basis dose describes the determination of ,/Q calculations. A request for TS modifications allowing containment penetrations values. The description provides the (i.e., personnel air lock, equipment hatch) to be open during refueling cannot potential release pathways for the rely on the current dose analysis if this analysis has not already considered various DBAs. The most limiting these release pathways. RG 1.194, Regulatory Position 3.2.4.2 supports review pathway is used for determining dose of penetration pathways, by stating that "leakage is more likely to occur at a consequences. penetration, [and that the] analysts must consider the potential impact of leakage from building penetrations exposed to the environment." Therefore, releases from personnel air locks and equipment hatches exposed to the environment and containment purge releases prior to containment isolation need to be addressed. Some licensees have identified unique release pathways that had not been previously considered. For example, a recent submittal noted that containment hatches and containment plugs may be removed during refueling. The removal of these barriers creates new release pathways.Licensees are responsible for identifying all release pathways and for considering these pathways in their AST analyses, consistent with any proposed modification.

7. Primary to Secondary Leakage The methodology for modeling primary-to-secondary leakage conforms to the Some analysis parameters can be affected by density changes that occur in the methodology provided in Appendix F to process steam. The NRC staff continues to find errors in LAR submittals RG 1.183, Position 5.2. See concerning the modeling of primary to secondary leakage during a postulated Attachment 5 for the RG 1.183 accident.

This issue is discussed in Information Notice (IN) 88-31, "Steam compliance tables.Generator Tube Rupture Analysis Deficiency," (Ref. 11) and Item 3.f in RIS 2001-19. An acceptable methodology for modeling this leakage is provided in Appendix F to RG 1.183, Regulatory Position 5.2.A6-7 RIS 2006-04 ISSUE ADDRESSED BY 8. Elemental Iodine Decontamination Factor (DF) An iodine decontamination factor of 200 is assumed for the FHA in the FHB and Appendix B to RG 1.183 provides assumptions for evaluating the radiological the FHA in containment. The chemical consequences of a fuel handling accident. If the water depth above the form of iodines above the pool is 57%damaged fuel is 23 feet or greater, Regulatory Position 2 states that "the elemental and 43% organic.decontamination factors for the elemental and organic [iodine] species are 500 Attachment 4, Section 7.3 provides and 1, respectively, giving an overall effective decontamination factor of 200." details of the FHA analyses.However, an overall DF of 200 is achieved when the DF for elemental iodine is 285, not 500.9. Isotopes Used in Dose Assessments The AST analyses consider the radionuclides in RG 1.183, Section 3.4, For some accidents (e.g., main steamline break and rod drop), licensees have Table 5. Noble gases are included in all excluded noble gas and cesium isotopes from the dose assessment. The dose analyses. Non-LOCA gap inclusion of these isotopes should be addressed in the dose assessments for fractions, with the exception of the AST implementation. CREA, include noble gas and cesium isotopes. The gap releases for the CREA include iodines and noble gases, as provided in RG 1.183, Appendix H.Cesium and rubidium are not included in the non-LOCA radiological consequence analyses that do not result in fuel failures since these isotopes would have a negligible impact on the analysis results. Attachment 4, Section 4.0, provides details for determining the radiation source terms.A6-8 RIS 2006-04 ISSUE ADDRESSED BY 10. Definition of Dose Equivalent 1-131 As part of the AST license submittal, a change is proposed to the DCPP In the conversion to an AST, licensees have proposed a modification to the TS Technical Specification definition of DEI.definition of dose equivalent 1-131. Some have modified the definition to base it The change will reflect that the dose upon the thyroid dose conversion factors of International Commission on conversion factors used to determine Radiation Protection (ICRP) Publication 2, "Report of Committee II on. DEI will be the same as those used to Permissible Dose for Internal Radiation" (Ref. 12) or ICRP Publication 30, determine the reactor-coolant dose"Limits for Intakes of Radionuclides by Workers" (Ref. 13). Others have equivalent iodine curie content for the proposed a definition which is a combination of different iodine dose conversion MSLB and SGTR accident analyses.factors, (e.g., RG 1.109, Revision 1, "Calculation of Annual Doses to Man from Routine Releases of Reactor Effluents for the Purpose of Evaluating Compliance with 10 CFR [Part] 50, Appendix I" (Ref. 14), ICRP Publication 2, Federal Guidance Report 11, "Limiting Values of Radionuclide Intake and Air Concentration and Dose Conversion Factors for Inhalation, Submersion, and Ingestion" (Ref. 15). Although different references are available for dose conversion factors, the TS definition should be based on the same dose conversion factors that are used in the determination of the reactor coolant dose equivalent iodine curie content for the main steamline break and steam generator tube rupture accident analyses.11. Acceptance Criteria for Off-Gas or Waste Gas System Release No design change to the waste gas tank or systems is being proposed with this 2 As part of full AST implementation, some licensees have included an accident AST submittal. Therefore, an analysis involving a release from their off-gas or waste gas system. For this accident, of a release from the waste gas system they have proposed acceptance criteria of 500 millirem (mrem) total effective is not included in this AST submittal. dose equivalent (TEDE). The acceptance criteria for this event is that associated with the dose to an individual member of the public as described in 10 CFR Part 20, "Standards for Protection Against Radiation."' 3 When the NRC revised 10 CFR Part 20 to incorporate a TEDE dose, the offsite dose to an individual member of the public was changed from 500 mrem whole body to 100 mrem TEDE. Therefore, any licensee who chooses to implement AST for an off-gas or waste gas system release should base its acceptance criteria on 1.00 A6-9 RIS 2006-04 ISSUE ADDRESSED BY RIS 2006-04 ISSUE ADDRESSED BY mrem TEDE. Licensees may also choose not to implement AST for this accident and continue with their existing analysis and acceptance criteria of 500 mrem whole body.2 An off-gas or waste gas system release does not need to be addressed for a full AST implementation unless a design change is being proposed for the waste gas tank or systems at the same time.3 Branch Technical Position ETSB 11-5, "Standard Review Plan for the Review of Safety Analysis Reports for Nuclear Power Plants" (Ref. 16).12. Containment Spray Mixing Some plants with mechanical means for mixing containment air have assumed that the containment fans intake air solely from a sprayed area and discharge it solely to an unsprayed region or vice versa. Without additional analysis, test measurements or further justification, it should be assumed that the intake of air by containment ventilation systems is supplied proportionally to the sprayed and unsprayed volumes in containment. Attachment 4, Section 7.2.3.2.3, discusses mixing between the sprayed and unsprayed regions of containment. The locations of the intake and exhaust registers of the operating CFCUs, as well as the major openings in the containment structure, were considered when determining the mixing rate between the sprayed and unsprayed regions. A review of the layout and arrangement of the intake and exhaust registers of the operating fan coolers demonstrates that the CFCUs intake air solely from the sprayed region and discharge air solely to the unsprayed region of the containment. =A6-10 Commitment 1: Commitment 2: Commitment 3: Commitment 4: Commitment 5: Enclosure Attachment 7 PG&E Letter DCL-15-069 ATTACHMENT 7 Diablo Canyon Power Plant List of Regulatory Commitments for Alternative Source Term Implementation Install shielding material, equivalent to that provided by the Control Room outer walls, at the external concrete west wall of the Control Room briefing room prior to implementation of Alternate Source Term.Install a high efficiency particulate air filter in the Technical Support Center normal ventilation system.Re-classify a portion of the 40-inch Containment Penetration Area (GE/GW) Ventilation line from PG&E Design Class II to PG&E Design Class I and upgrade the damper actuators, pressure switches, and the damper solenoid valves to PG&E Design Class I prior to implementation of Alternate Source Term.Re-classify a portion of the 2-inch gaseous radwaste system line which connects to the Plant Vent as PG&E Design Class I prior to implementation of Alternate Source Term.Update setpoints for the redundant safety related gamma sensitive area radiation monitors (1-RE 25/26, 2-RE 25/26) prior to implementation of Alternate Source Term.}}

DCL-15-069, Diablo Canyon, Units 1 and 2 - License Amendment Request 15-03 Application of Alternative Source Term - Summary of Dose Analyses and Results. Part 2 of 8

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1.0 INTRODUCTION

SUMMARY

9.0 CONCLUSION

10.0 REFERENCES

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SUMMARY

9.0 CONCLUSION

10.0 REFERENCES

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DCL-15-069, Diablo Canyon, Units 1 and 2 - License Amendment Request 15-03 Application of Alternative Source Term - Summary of Dose Analyses and Results. Part 2 of 8

Text

1.0 INTRODUCTION

SUMMARY

9.0 CONCLUSION

10.0 REFERENCES

1.0 INTRODUCTION

SUMMARY

9.0 CONCLUSION

10.0 REFERENCES

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