ML16004A355

Response to NRC Request for Additional Information Regarding License Amendment Request 15-03, Regulatory Guide 1.183 Conformance Tables, Revision 1

ML16004A355
Person / Time
Site:	Diablo Canyon
Issue date:	12/17/2015
From:	Pacific Gas & Electric Co
To:	Office of Nuclear Reactor Regulation
Shared Package
ML16004A363	List: DCL-15-152, Response to NRC Request for Additional Information Regarding License Amendment Request 15-03 - Updated Final Safety Analysis Report Markup, Revision 1. Part 3 of 3 ML16004A354 ML16004A355 ML16004A356 ML16004A357 ML16004A361
References
DCL-15-152, TAC MF6399, TAC MF6400
Download: ML16004A355 (62)

Text

Enclosure Attachment 3 PG&E Letter DCL-l5-152 License Amendment Request 15-03, Attachment 3 Regulatory Guide 1.183 Conformance Tables, Revision 1 In Table A5-A through A5 -H, the text shown in "RG Position" columns is taken from Regulatory Guide (RG) 1.183; therefore, references to footnotes, tables, and numbered references, may be found in RG 1.183.Only Pressurized Walter 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)

Table A5-A: Conformance with Regulatory Guide 1.183 Main Sections RG TDCP Section jRG Position Analysis Comments_______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. (Aft. 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/ORIG EN-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-i137, equilibrium will not be reached prior to fuel offload. Thus, the maximum________inventory at the end of life should be used._________________________

A5-1 Revision 1 Table A5-A: Conformance with Regulatory Guide 1.183 Main Sections RG DCPP Section RG Position Analysis Comments 3.1 For the DBA LOCA, all fuel assemblies in the core are assumed Conforms For the DBA LOCA, all fuel assemblies in I2 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 Revision 1 Table A5-A: Conformance with Regulatory Guide 1.183 Main Sections RG 1(DCPP 1 Section] RG Position _ _ _ ______Analysis

]Comments 3.2¶1 The core inventory release fractions, by radionuclide groups, for the gap release and early in-vessel damage phases for IDBA LOCAs are listed in Table 1 for BWRs and Table 2 for PWRs.These fractions are applied to the equilibrium core inventory described in Regulatory Position 3.1. (Footnote 10 applies to entire RG Section 3.2.)Conforms The release fractions from Regulatory Position 3.2, Table 2 are used. Footnote 10 criterion is met in that peak fuel rod burnup is limited to 62,000 MWD/MTU.The equilibrium core average isotopic inventory that meets regulatory Position 3.1 was used for LOCA.(Att. 4 Section 7.2.3.2.6, Table 7.2-1)RG 1.183, Table 2 PWR Core Inventory Fraction Released Into Containment 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 0 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 containincl mixed oxide (MOX) fuel.A5-3 Revision 1 Table A5-A: Conformance with Regulatory Guide 1.183 Main Sections RG 1DCPP Section RG Position Analysis jComments 3.2 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 3"1 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 1 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 provided that the maximum linear heat generation rate does not exceed 6.3 kw/ft peak rod average power for burnups exceeding 54 GWDIMTU. 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 with respect to thE potential of having fuel rods that exceed the RG 1.183, Revision 0 linear heat generation rate criteria, and since DCPP falls within, and intends to operate within, the maximum allowable power operating envelop for PWRs shown in Figure 1 of Draft Guide (DG)-1 199, the fuel gap activity fractions used for the DCPP Non-LOCA events that experience fuel damag (with the exception of the CREA) are based on Table 3 of DG-1 199.In accordance with Table 3, Note 11 and Appendix H of RG 1.183, the CREA gap fractions are assumed to be 10% for iodines and noble gases..(Att. 4, Section 2.1 and Section 4.3)A5-4 Revision I Table A5-A: Conformance with Regulatory Guide 1.183 Main Sections RG DCPP Section RG Position Analysis Comments 3.3 Table 4 tabulates the onset and duration of each sequential Conforms The core inventory release timing for gap¶J1 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 gap and 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 Revision 1 Table A5-A: Conformance with Regulatory Guide 1.183 Main Sections RG DCPP Section RG Position Analysis Comments 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 Lantanies a, Z, N, E, Nb Pm Pr Sin YCmadditional nuclides that were deemed to be Lataie amr d u b PPSYm dose significant (i.e., Br-82, B3r-84, Rb-88, Cerium Ce, Pu, Np Rb-89, Te-1 33, Te-133m, Te-1 34, 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% particulate 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 OBAs 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 Revision 1 Table A5-A: Conformance with Regulatory Guide 1.183 Main Sections RG DCPP Section RG Position Analysis 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 including 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 Revision 1 Table A5-A: Conformance with Regulatory Guide 1.183 Main Sections RG DCPP Section RG Position Analysis Comments 4.1.2 The exposure-to-CEDE factors for inhalation of radioactive Conforms The CEDE is calculated using the material should be derived from the data provided in ICRP inhalation dose conversion factors Publication 30, "Limits for Intakes of Radionuclides by Workers" provided in Table 2.1 of Federal Guidance (Ref. 19). Table 2.1 of Federal Guidance Report 11, "Limiting Report 11, "Limiting Values of Values of Radionuclide Intake and Air Concentration and Dose Radionuclide Intake and Air Concentration Conversion Factors for Inhalation, Submersion, and Ingestion" and Dose Conversion Factors for (Ref. 20), provides tables of conversion factors acceptable to the Inhalation, Submersion, and Ingestion".

NRC staff. The factors in the column headed "effective" yield The factors in the column headed doses corresponding to the CEDE. "effective" yield doses corresponding to the CEDE and are derived based on ICRP-30.____________________________________________(Att.

4 Section 6.1)4.1.3 For the first 8 hours9.259259e-5 days <br />0.00222 hours <br />1.322751e-5 weeks <br />3.044e-6 months <br />, the breathing rate of persons offsite should Conforms The assumed offsite breathing rates are be assumed to be 3.5 x 1 0-4 cubic meters per second. From 8 to those specified in Section 4.1.3 of RG 24 hours2.777778e-4 days <br />0.00667 hours <br />3.968254e-5 weeks <br />9.132e-6 months <br /> following the accident, the breathing rate should be 1 .183.assumed to be 1.8 x 1 0-4 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- Conforms The submersion EDE is calculated using infinite cloud assumptions with appropriate credit for attenuation the air submersion dose coefficients by body tissue. The DDE is nominally equivalent to the effective provided in Table 111.1 of Federal Guidance dose equivalent (EDE) from external exposure if the whole body Report 12, "External Exposure to is irradiated uniformly.

Since this is a reasonable assumption for Radionuclides in Air, Water, and Soil." submergence exposure situations, EDE may be used in lieu of DDE in determining the contribution of external dose to the TEDE. (Att. 4 Section 6.1)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._____A5-8 Revision 1 Table A5-A: Conformance with Regulatory Guide 1.183 Main Sections RG DCPP Section RG Position Analysis 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 Revision 1 Table A5-A: Conformance with Regulatory Guide 1.183 Main Sections RG DCPP Section RG Position Analysis Comments 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 o 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, oRadiation shine from radioactive material in the reactor (Att. 4, Section 7.2)containment, o 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 Revision 1 Table A5-A: Conformance with Regulato ry Guide 1.183 Main Sections RG DCPP Section RG Position Analysis Comments 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.Revision 1 Table A5-A: Conformance with Regulatory Guide 1.183 Main Sections RG DCPP Section jRG Position _________

Analysis Comments 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 ESE 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 LIRA 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 D3803-1 989.A5-12 Revision 1 Table A5-A: Conformance with Reaulatorv Guide 1.183 Main Sections RG DCPP Section RG Position Analysis Comments 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 hours2.777778e-4 days <br />0.00667 hours <br />3.968254e-5 weeks <br />9.132e-6 months <br /> 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 8 For the duration of the event, the breathing rate of this individual should be assumed to be 3.5 x l0-4cubic meters (Att. 4 Section 6.1)per second.ARCON96 was used for determining CR 1 6 The occupancy is modeled in the X/Q values determined in X/Q values. Occupancy assumptions Reference 22 (Mu rphy-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, DDE~nite, 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)DDfnt DE°°V 0'3 3 8 Equation 1 Dfnte- 1173 A5-13 Revision 1 Table A5-A: Conformance with Reaulatorv Guide 1.183 Main Sections RG DCPP Section RG Position Analysis 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 A5-24 Revision 1 Table A5-A: Conformance with Regulatory Guide 1.183 Main Sections RG DCPP Section RG Position Analysis Comments 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 Revision 1 Table A5-A: Conformance with Reaulatorv Guide 1.183 Main Sections RG --- ---------DCPP Section RG Position Analysis Comments 5.1.2 Credit may be taken for accident mitigation features that are Conforms Credit is taken for the ESF equipment classified as safety-related, are required to be operable by discussed in RO Position 4.2.4. Credit is technical specifications, are powered by emergency power also taken for the PG&E Design Class I sources, and are either automatically actuated or, in limited ABVS and filters, which are also controlled cases, have actuation requirements explicitly addressed in by TS 3.7.12 and TS 5.5.11. Credit is emergency operating procedures.

The single active component taken for operating Containment Spray failure that results in the most limiting radiological consequences during recirculation (Att. 4, Section 2.1, should be assumed. Assumptions regarding the occurrence and item 14). A Time Critical Operator Action timing of a loss of offsite power should be selected with the will be implemented to ensure that the objective of maximizing the postulated radiological consequences.

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 Revision 1 Table A5-A: Conformance with Regiulatory Guide 1.183 Main Sections RG {FDCPP F Section jRG Position Analysis Comments 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 Revision 1 Table A5-A: Conformance with Reaulatorv Guide 1.183 Main Sections RG DCPP Section RG Position Analysis Comments 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 CER 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 11 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 Revision 1 Tahle A5-A: Cnnfnrmance with Reaulatorv Guide 1.183 Main Sections RG -- -- --- --- -Section RG Position Analysis Comments 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 Revision 1 Table A5-A: Conformance with Regulatory Guide 1.183 Main Sections RG DCPP Section RG Position Analysis Comments 5.3 Atmospheric dispersion values (XIQ) for thle EAB, the LPZ, and Conforms New atmospheric dispersion values (XIQ)¶[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 XIQ (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 Revision 1 Table A5-A: Conformance with Reaulatorv Guide 1.183 Main Sections RG JDCPP Section jRG Position Analysis Comments 5.3 2 References 22 [Murphy -Cam pe] 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 LAB, 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 XIQ values. Meteorological data collected in accordance with the site-specific meteorological measurements program described in the facility FSAR should be used in generating accident XI/Q values. Additional guidance is provided in Regulatory Guide 1.23, "Onsite Meteorological Programs" (Ref.30). All changes in XIQ analysis methodology should be reviewed by the NRC staff.1 9 The ARCON 96 computer code contains processing options that may yield XIQ 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 (XIQ)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. (Aft. 4 Sections 5.1 and 5.2)The offsite X/Q methodology (LAB and LPZ) is based upon the methodology in RG 1.145 for ground level releases using CBI S&W Proprietary code EN-I113. 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)___________________________

I A5-21 Revision I Table A5-A: Conformance with Regulatory Guide 1.183 Main Sections RG DCPP Section RG Position Analysis Comments 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 Qualification," 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. 1 1), 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 1 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 basis to 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-14844 assumptions at this time.either the AST or the TID 14844 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 TID 14844) on EQ doses pending the outcome of the evaluation of the generic issue.A5-22 Revision 1

' ' Table :A5- B: Conformance Re ulator, Guide 1.183, Appendix;A iL~oss of Coolant ,i: Secio R PoitonAnalysis

Comments

-:i ,L-., Source Term Assumptions__________________

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 (Csl), 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 OBA 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 (OsI), 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 Revision 1 T

A5-B: Conformance with Regulatory Guide 1,183, Appendi~x:A (Loss oif CoolanitAccident) ..i 3.2 Reduction in airborne radioactivity in the containment by natural Conforms DCPP does not deterministically assume deposition within the containment may be credited.

Acceptable 50% plateout of iodine. The wall models for removal of iodine and aerosols are described in deposition removal coefficient for Chapter 6.5.2, "Containment Spray as a Fission Product Cleanup elemental iodine has been calculated System," of the Standard Review Plan (SRP), NUREG-0800 using computer program SWNAUA to (Ref. A-I) and in NUREG/CR-6189, "A Simplified Model of estimate the time dependent particulate Aerosol Removal by Natural Processes in Reactor removal coefficients.

The guidance of Containments" (Ref. A-2). The later model is incorporated into SRP 6.5.2 was used to determine the analysis code RADTRAD (Ref. A-3). The prior practice of elemental iodine removal coefficients.

deterministically assuming that a 50% plateout of iodine is Credit is taken for gravitational settling of released from the fuel is no longer acceptable to the NRC staff particulates.

as it is inconsistent with the characteristics of the revised source terms. (Att. 4 Section 3 for code description, Section 7.2.3.2.4 for fission product___________________________________________

________removal)

A5-24 Revision 1

--Table A5-, B: Conformance-with Guidel 1.183, Appendix A (ILoss ,of' Coolant

.-*Section ;RG Position:

AnlyisComet

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"'1 (Ref. A-4). This simplified model is AST analysis continues to credit incorporated into the analysis code RADTRAD (Refs. A-I to A- containment spray system for the 3). removal of iodine and aerosols from the containment atmosphere.

In addition, 1 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.

Aft. 4 Table 7.2-2 provides removal coefficient used_______ _____________________________________________

________in the LOCA dose analysis.A5-25 Revision 1

" -A5- B:-Conformance with, Re ulatory Guide 1l.1l83,.Appendix A (Loss 'f Coolant Accident)

.-.,: ,*Section:

RG:Position

., .... Analysis-Comments --.,,-- --.,"-, 3.3 The evaluation of the containment sprays should address areas Conforms The percentage of the total containment

¶[2 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. (Aft. 4 Section 7.2.3.2)3.3 The SRP sets forth a maximum decontamination factor (OF) for -Conforms Att. 4 Section 7.2.3.2.4 provides¶[3 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 maxim~um activity to be used in based on calculated time dependent determining the OF is defined as the iodine activity in the airborne aerosol mass, there is no columns labeled "Total' in Tables 1 and 2 of this guide multiplied restriction on the OF 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 Revision 1

-...--. Table.: A5- B: Conformance:.

with Regulatory Guide I ..

iof Coolant. Accident).--

-.: -., 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 indiv~idual 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 hours2.777778e-4 days <br />0.00667 hours <br />3.968254e-5 weeks <br />9.132e-6 months <br />. both the sprayed and unsprayed region For PWRs, the leak rate may be reduced after the first 24 hours2.777778e-4 days <br />0.00667 hours <br />3.968254e-5 weeks <br />9.132e-6 months <br /> 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 hours2.777778e-4 days <br />0.00667 hours <br />3.968254e-5 weeks <br />9.132e-6 months <br />, 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 hours2.777778e-4 days <br />0.00667 hours <br />3.968254e-5 weeks <br />9.132e-6 months <br />, 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 Ill 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 Revision 1 S Table. -A5-:B: Conformance .with Reg ulatory Guide 1:.1 83,- Appendix.A (L~oss of Coolant ,Accident)-", -.-.-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 Revision 1

....Table A5- B: iconformance with ,RegulatoryGuide 1l.183;,Appendix A of Coolant

'::i Section :RG Position, .."-....,. -: .:Analysis,'

_,,>-:i :,:, 5. ESF systems that recirculate sump water outside of the primary Conforms The radiological consequences from the containment are assumed to leak during their intended operation.

postulated ESE systems leakage are This release source includes leakage through valve packing analyzed and combined with glands, pump shaft seals, flanged connections, and other similar consequences postulated for other components.

This release source may also include leakage fission product release paths. ESF through valves isolating interfacing systems (Ref. A-7). The systems that recirculate sump fluids radiological consequences from the postulated leakage should be outside containment are postulated to analyzed and combined with consequences postulated for other leak at twice the sum of the fission product release paths to determine the total calculated administrative acceptance criteria.radiological consequences from the LOCA. The following assumptions are acceptable for evaluating the consequences of (Att. 4 Section 7.2.3.3)leakage from ESE components outside the primary containment for BWRs and PWRs.5.1 With the exception of noble gases, all the fission products released Conforms With the exception of noble gases, all from the fuel to the containment (as defined in Tables 1 and 2 of fission products released from the fuel to this guide) should be assume to instantaneously and the containment are instantaneously and homogeneously mix in the primary containment sump water (in homogeneously mixed in the sump water PWRs) or suppression pool (in BWRs) at the time of release from at the time of release. Only iodine is the core. In lieu of this deterministic approach, suitably released through ESF leakage since the conservative mechanistic models for the transport of airborne noble gases are not assumed to dissolve activity in containment to the sump water may be used. Note that in the sump and particulates would many of the parameters that make spray and deposition models remain in the water of ECCS leakage.conservative with regard to containment airborne leakage are nonconservative with regard to the buildup of sump activity.

______(Att.

4 Section 7.2.3.3)A5-29 Revision 1

• Table; A5- B: Conformance with Regulatory Guide,1.1,83,,!Appendix, A (L0SS:of.'Coolant Accident), .-: -Section. RG Pos-ition"::

,- -:.

j ..

.: ::-" 5.2 The leakage shouid 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 llI.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.Exceeds 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/mmn with the following breakdown: (See LAR Enclosure Section 2)-Plant vent area -< 120 cc/mmn-Penetration area -< 6 cc/mmn The assumed leakage of 252 cc/mmn (240 cc/mmn plus 12 cc/mmn) is two times the administrative limit of 126 cc/mmn.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 hours2.777778e-4 days <br />0.00667 hours <br />3.968254e-5 weeks <br />9.132e-6 months <br /> 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 MEOT. These releases are not filtered.(Att. 4 Sections 7.2.3.3 -7.2.3.6)A5-30 Revision 1

..-::. Table: A5- :B: C onformance..with"Regjulatory

!Guide-1.,183,::AppendiX:

A (Loss 'of-C~oolant

..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 Section 7.2.3.3)5.4 If the temperature of the leakage exceeds 21 2F, 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 1 -- hf 2 (All. 4 Section 7.2.3.3)F= hfg Where: hf 1 is the enthalpy of liquid at system design temperature and pressure; hf 2 is the enthalpy of liquid at saturation conditions (14.7 psia, 212F); and hf~q is the heat of vaporization at 212F. _____5.5 If the temperature of the leakage is less than 21 2F 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 259.9 0 F, 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._______(Att.

4 Sections 7.2.3.3 -7.2.3.6)A5-31 Revision 1

• -Table"-A5-B: Conformance-with Regulatory.

G uide, 1;.183, Appendix :A (Loss of Coolant Accident);i-L R G D C. .PP --f .. : ]:: : i ~ ~ ~ .5.6 The radioiodine that is postulated to be available for release to the environment is assumed to be 97% elemental and 3% organic.Reduction in release activity by dilution or holdup within buildings, or by ESF ventilation filtration systems, may be credited where applicable.

Filter systems used in these applications should be evaluated against the guidance of Regulatory Guide 1.52 (Ref. A-5) and Generic Letter 99-02 (Ref. A-6).Conforms The iodine released from ESF leakage is assumed to be 97% elemental and 3%organic. No credit for holdup or dilution of ESF component leakage is taken. A time dependent iodine partition coefficient is used to determine the 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)A5-32 Revision I}

TFable. 'A5- B:' Conformance-with Regulatory.

Guide :1,1:83, Appendix (Loss of Coolant Aci~ide nt)!::.-:::h.::::

_______Assumptions on Main Steam Isolation Valve Leakage in BWRs 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 Revision 1 A5-C: Conformance with Regulatory Guide Appe~ndix iB (Fuel =Handling, Acc'ident):: -Sectioni RG Position , -,, ,-, -. *- " -,-* ,, .Analysis, Comments ,i- '.:i::,>.:: , , :...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. (Fuel 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 hours8.333333e-4 days <br />0.02 hours <br />1.190476e-4 weeks <br />2.7396e-5 months <br /> after shutdown.

The radionuclides relevant to the dose analysis of the postulated FHA for a single fuel assembly at 72 hours8.333333e-4 days <br />0.02 hours <br />1.190476e-4 weeks <br />2.7396e-5 months <br /> 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 Revision 1

.. .Table A5-C: "Conformance with: Regulatory.Guide:1

.83, AppendixB: (Fuel Handling ...section: -RG Position -"; ..". , .i. Analysis Co;Gmmen~ts"::i_:f,.::i-'

i.1.1 The number of fuel rods damaged during the accident should be Conforms The DCPP FHA utilizes the current based on a conservative analysis that considers the most limiting licensing basis assumption that all fuel case. The analysis should consider parameters such as the rods in one assembly are damaged. As weight of the dropped heavy load or the weight of a dropped fuel documented in the NRC SER for License assembly (plus any attached handling grapples), the height of Amendments 8 and 6 to DCPP Facilty the drop, and the compression, torsion, and shear stresses on

• Operating License Nos. DPR-80 and the irradiated fuel rods. Damage to adjacent fuel assemblies, if DPR-82, respectively, the assumption applicable (e.g., events over the reactor vessel), should be that all fuel rods in one assembly rupture considered, 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 Revision 1

....::.. .Table-.A5-C:.

Conformance with Regulatory Guide1.%183, B .=(F.uel H~andling ,Accident)

?Setin G ostinAnalysis ::L i: 1.2 The fission product release from the breached fuel is based on Conforms!

The fission product release from the Regulatory Position 3.2 of this guide and the estimate of the Exceeds breached fuel is based on the breach of number of fuel rods breached.

All the gap activity in the all fuel rods in one fuel assembly and the damaged rods is assumed to be instantaneously released.

following gap fractions.

All gap activity is Radionuclides that should be considered include xenons, assumed to be instantaneously released.kryptons, halogens, cesiums, and rubidiums.

Since DCPP falls within, and intends to operate within, the maximum allowable power operating envelop for PWRs shown in Figure 1 of Draft Guide (DG)-1199, the gap fractions used for the FHA are based on the values provided per isotope/isotope class in Table 3 of DG-1199.(Att. 4 Sections 4.3, and, 7.3)1.3 The chemical form of radioiodine released from the fuel to the Conforms The iodine release into the pool from the spent fuel pool should be assumed to be 95% cesium iodide fuel is assumed to be 95% Csl, 4.85%(CsI), 4.85% elemental iodine, and 0.15% organic iodide. The elemental iodine and 0.15% organic CsI released from the fuel is assumed to completely dissociate in iodine. Due to the acidic nature of the the pool water. Because of the low pH of the pool water, the water in the fuel pool (pH less than 7), iodine re-evolves as elemental iodine. This is assumed to occur the Csl is assumed to immediately instantaneously.

The NRC staff will consider, on a case-by-case disassociate and re-evolve as elemental basis, justifiable mechanistic treatment of the iodine release from iodine, thus changing the chemical form the pool. of iodine to 99.85% elemental and 0.15%organic.(Att. 4 Section 7.3)Water Depth A5-36 Revision I

" : A5-C:i Conformancewith Regulatory_

Guide1.1 lt83, iA~ppendix B (;Fuel Handlin~g-Accident)

.-

Section RG Position-Analysis-Comments.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-1) 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 Revision I

"":Table A5-C: Conformance .with Regulatory Guide. 1..183:, Appendx B (Fuel 'Hanidling fo: .Section RG Position :." .°/ Analysis

!.!i 4.2 A reduction in the amount of radioactive material released from Conforms No ESF filtration is credited for EAB and the fuel pool by engineered safety feature (ESF) filter systems LPZ doses. The CR dose analysis may be taken into account provided these systems meet the credits radiation monitors to switch the guidance of Regulatory Guide 1 .52 and Generic Letter 99-02 CRVS from Mode 1 (normal) to Mode 4 (Refs. B-2, B-3). Delays in radiation detection, actuation of the (filtered and pressurized).

A 32 second ESF filtration system, or diversion of ventilation flow to the ESF delay is conservatively assumed, filtration system 1 should be determined and accounted for in the including instrument loop uncertainties, radioactivity release analyses.

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 1 These analyses should consider the time for the radioactivity controlled by TS 5.5.11.concentration to reach levels corresponding to the monitor (Att. 4 Sections 7.1 and 7.3)setpoint, instrument line sampling time, detector response time, diversion damper alignment time, and filter system actuation, as applicable.

4.3 The radioactivity release from the fuel pool should be assumed Conforms Noble gases and iodines are released to be drawn into the ESF filtration system without mixing or from the pool and mixed in the available dilution in the fuel building.

If mixing can be demonstrated, credit air space, but all activity is released to for mixing and dilution may be considered on a case-by-case the environment in less than 2-hours.basis. This evaluation should consider the magnitude of the The radionuclides are released to the building volume and exhaust rate, the potential for bypass to the environment via the Plant Vent with an environment, the location of exhaust plenums relative to the assumed 500 cfm outleakage to the surface of the pool, recirculation ventilation systems, and internal environment from the FHB.walls and floors that impede stream flow between the surface of ~ (Att. 4 Section 7.3)the pool and the exhaust plenums.Fuel Handling Accidents Within Containment A5-38 Revision 1

A5-C:. Conformance with Regulatory Guide .1A;83,.App~eIdixixB';(FueI'.Handlinig Accident)"l, .:. -.::i Section -RG Position ...- '-.*,- , -,..}Analysis 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 designed to automatically isolate in the event of a fuel handling accident, the release duration should be based on delays in radiation detection and completion of containment isolation.

If it can be shown that containment isolation occurs before radioactivity is released to the environment, no radiological consequences need to be analyzed.Conforms Automatic containment isolation is not credited.

Therefore, a radiological consequence analysis is performed.(Att. 4 Section 7.3)A5-39 Revision 1

.Table A5-C: Conformance with Regulatory~Guide 1.183,-AppendixB (Fuel

, Section RG Position "" "... ' -: IAnalysis

Comments

.,i-:. 5.3 If the containment is open during fuel handling operations (e.g., Conforms The radioactive material that escapes personnel air lock or equipment hatch is open)3 , the radioactive from the reactor cavity pool to the material that escapes from the reactor cavity pool to the containment is released to the containment is released to the environment over a 2-hour time environment over a 2-hour time period.period. (Att. 4 Section 7.3)3 The staff will generally require that technical specifications TS 3.9.4 allows the hatches to be open allowing such operations include administrative controls to close during fuel movements provided that the airlock, hatch, or open penetrations within 30 minutes. Such provisions are in place for the hatches to administrative controls will generally require that a dedicated be closed. The TS bases LCO provide individual be present, with necessary equipment available, to guidance on how that capability is restore containment closure should a fuel handling accident available and monitored and the occur. Radiological analyses should generally not credit this accepted closure time is within 30 manual isolation, minutes. This LAR does not alter TS 3.9.4 or its Bases.5.4 A reduction in the amount of radioactive material released from Conforms No ESF filtration is credited for releases the containment by ESF filter systems may be taken into account from the containment to the environment.

provided that these systems meet the guidance of Regulatory The CR dose analysis credits radiation Guide 1.52 and Generic Letter 99-02 (Refs. B-2 and B-3). Delays monitors to switch the CRVS from Mode in radiation detection, actuation of the ESF filtration system, or I (normal) to Mode 4 (filtered and diversion of ventilation flow to the ESF filtration system should be pressurized).

A 32 second delay is determined and accounted for in the radioactivity release conservatively assumed, including analyses.1 instrument loop uncertainties, from the onset of the event to Mode 4 operation.

1 These analyses should consider the time for the radioactivity CR filter efficiencies are determined in concentration to reach levels corresponding to the monitor accordance with the guidance provided setpoint, instrument line sampling time, detector response time, in GL 99-02 and TS 5.5.11.diversion damper alignment time, and filter system actuation, as (Att. 4 Sections 7.1 and 7.3)________ applicable.

A5-40 Revision 1

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 ofthe 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 Revision 1

.T Conformance withRegulatory, 1:.183; Appendix E (PWR'!Main

,:Section:

J RG'

• - --"-, .".i,:"IAnalysis ~-i~:,i.-.:',/,!..i 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 radionucl ides 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 DEl, the initial primary coolant DEX 1-131 (DE 1-131), only the radioiodine associated with normal is assumed to be 270 paCi/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 DEl (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 tpCi/ 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 DEl (TS 3.4.16 limit). (Att. 4 Section 7.6)A5-42 Revision 1

-.. i Table A5-D:: Conformance with Regulatory-Guide 1.183, Appen~dix E (PWR !Main: Steam

-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 IpCi/gm of DEl.pJCi/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 hours9.259259e-5 days <br />0.00222 hours <br />1.322751e-5 weeks <br />3.044e-6 months <br />. 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 (Csl), 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 Revision 1

-. ...Table,. A5-D: Conformance ,with.Re ulatory Guide .1.183, ,Appendix

E (PWR ,Main Steam;Line --Section RRG ii n Po....sition.

Ana'- .,:. : l :.ysi-s [Co m mentsAal si , -C m en ....Transport3 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 '2 per unit mass of liquid PC =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-l, DG-1 074), 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 Revision 1

Table A5-D: Conformance with Re ulator Guide:1:.i83,AppendixE(P.WR Main Steam Linre

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 lbm/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 Revision '1

.Table:. AS-D: Conformance with Regulatory Guide 1,.183, Appendix ,E (PWR Main steam 'Line Break).,-.,.

RG DCPP Section RG Position Analysis ~' Comments 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.

Iodines 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 into the intact SGs occurs because all tube leakage is conservatively assumed to occur in the faulted SG.(Att. 4 Section 7.6)A5-46 Revision 1

,"Table A5,D: Conformance with Regulatory Guidel 1.183, A ipendix E (,PWR Main Steam !Line

i section RG-Position

... : /::: AnalysiS iiCommen~ts;:;::;;i:,:[,::;

, 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 SC 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 Iodines 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 Revision 1 STable A5 -E: Conformnance with Regulatory Guide 1.183, Appendix F (PWR Steam Generator Tube 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 is assumed to be 270 pJCi/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 DEl (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 pJCi/gm of DEl (TS 3.4.16 limit).________(Att.

4 Section 7.7)A5-48 Revision 1

-Table. A5 -E:-.Conformance with Reaulatorv Guide 1..183.,ADDendix

F (PWR Steam Generator Tube RUpture) .i Section G o toRG... Position.

Analy,,sis Comments:-, -j Aalss .

.2.2 The primary system transient associated with the SGTR causes Conforms For the accident-initiated iodine spike an iodine spike in the primary system. The increase in primary case, the primary system transient coolant iodine concentration is estimated using a spiking model associated with the SGTR causes an that assumes that the iodine release rate from the fuel rods to iodine spike in the RCS, which increases the primary coolant (expressed in curies per unit time) increases the iodine release rate from the fuel to to a value 335 times greater than the release rate corresponding the RCS to a value 335 times the to the iodine concentration at the equilibrium value (typically

1.0 appearance

rate corresponding to a pCi/gm DE 1-131) specified in technical specifications (i.e., maximum equilibrium RCS concentration concurrent iodine spike case). A concurrent iodine spike need of 1.0 pCi/gm of DEl (TS SR 3.4.16.2).

not be considered if fuel damage is postulated.

The assumed The spike is allowed to continue until 8 iodine spike duration should be 8 hours9.259259e-5 days <br />0.00222 hours <br />1.322751e-5 weeks <br />3.044e-6 months <br />. Shorter spike durations hours from the start of the event.may be considered on a case-by-case basis if it can be shown (Att. 4 Section 7.7)that the activity released by the 8- hour spike exceeds that available for release from the fuel gap of all fuel pins.3. 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)4. 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 Revision 1

...-Table .A5:. E: :Conformance-with.Requlatory.Guide ApPendi~xF Section RG Position ,lA ls.s omet -: ,;:-, " Transport 3 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.

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., lbm/hr) should be consistent with the 1.0 gm/cc (62.4 lbm/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 Ibm/ft 3). ______________________

A5-50 Revision 1

-Table: A5_-E:-Conformance with, Regulatory Guide 1,183, Appendix PWR Steam Generator Tube Rupture);:.:

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 hours2.314815e-5 days <br />5.555556e-4 hours <br />3.306878e-6 weeks <br />7.61e-7 months <br /> after event initiation and continues until shutdown cooling is in operation (10.73 hours8.449074e-4 days <br />0.0203 hours <br />1.207011e-4 weeks <br />2.77765e-5 months <br />).In the intact SGs, release of radioactivity is assumed to continue until shutdown cooling is in operation (10.73 hours8.449074e-4 days <br />0.0203 hours <br />1.207011e-4 weeks <br />2.77765e-5 months <br />).(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. (Art. 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 Revision 1 STable A5- E: Conformance with Regulatory Guide 1.183, AppendixF (PWR SteamLGenerator Tube Rupture) ':RG : .-.T > ., .i /* ..... ' : D :' ..i. :i- i:. ,:: -.Section RG .i -,..* -i .-,Analysis omments;:" -!, ,i°-. .".* "": Regulatory Positions 5.5.1of 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 Revision 1

-Table A5 -E: ;Conformance with Regulatory'Guide-1.183,"Appendix'F

'PWR.Ste'am Generator Tu~be Rupture).--:

SectiOn' RG Position : i:" :. ,i' , Analysis:

Comments.i 11,::"Ii-i ;Regulatory Positions 5.5.4 of Appendix E Conforms In the ruptured SG, the non-flashed portion of the break flow mixes uniformly The radioactivity in the bulk water is assumed to become vapor with the steam generator liquid mass and at a rate that is the function of the steaming rate and the partition is released into the steam space in coefficient.

A partition coefficient for iodine of 100 may be proportion to the steaming rate and the assumed. The retention of particulate radionuclides in the steam inverse of the partition coefficient of 100.generators is limited by the moisture carryover from the steam generators.

iodines 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 Revision 1

A5- F: -Conformance with Regulatory Guide .G (RWR iLocked Rotor *Accident):"., -RGci~ jG [DCP....P:

.. ... ;., I Analysis [lComments -:*Source Terms 1. 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.2. 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.

3. 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. The chemical form of radioiodine released from the fuel should Conforms The chemical form of iodine released be 95% cesium iodide (Csl), 4.85 percent elemental iodine, and from the fuel is assumed to be 95% Csl, 0.15 percent organic iodide. Iodine releases from the steam 4.85% elemental iodine, and 0.15%generators to the environment should be assumed to be 97% organic iodide.elemental and 3% organic. These fractions apply to iodine released as a result of fuel damage and to iodine released during The chemical for of iodine released from normal operation, including iodine spiking. the SGs to the environment is assumed to be 97% elemental and 3% organic.(Att. 4 Section 7.4)Release Transport A5-54 Revision 1

..':Table -A5 .-F:. Conformance~with.

Regulatory Guide :Appendix

G `:PWR Locked 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., lbm/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 Ibm/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 hours8.449074e-4 days <br />0.0203 hours <br />1.207011e-4 weeks <br />2.77765e-5 months <br /> 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 Revision 1 S A5,- F:

iReg-ulatory Guide

PWR iLocked Rotor

--;:,: Section RG Position : , ; ---: ;Analysis -:,, J' , ,:: 5.6 The transport model described in Regulatory Positions 5.5 and Conforms Regulatory Position 5.6 refers to 5.6 of Appendix E should be utilized for iodine 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 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 gases are 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 Revision 1 Table ,A5--.G!:Conformance~with Reciulatorv Guide 1.183. App1enldix H !(P.W R~od Ejection

________Source Terms 1. Assumptions acceptable to the NRC staff regarding core Conforms The source term for the control rod inventory are in Regulatory Position 3 of this guide. For the rod ejection accident (CREA) is based on ejection accident, the release from the breached fuel is based on RG 1.183, Regulatory Position 3. The the estimate of the number of fuel rods breached and the radiological source term for the CREA is assumption that 10% of the core inventory of the noble gases based on the equilibrium core inventory and iodines is in the fuel gap. The release attributed to fuel determined with the computer code melting is based on the fraction of the fuel that reaches or ORIGEN-S as discussed in Section exceeds the initiation temperature for fuel melting and the 3.2.2.2 and shown in Table 3. The CREA assumption that 100% of the noble gases and 25% of the iodines assumes that 10% of the fuel fails. Thus contained in that fraction are available for release from the source terms for the CREA are 10%containment.

For the secondary system release pathway, 100% of the equilibrium core inventory in Table of the noble gases and 50% of the iodines in that fraction are 3. No fuel melt occurs.released to the reactor coolant.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 Revision 1

---Table A5,-G:, Conformance with Regulatory H Rod

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 radio iodine 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% Csl, 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. (Art. 4 Section 7.5)A5-58 Revision 1

• - Table A5 -G:: Conformance with. Regulatory Guide i1.183, :AppendixH

(PWR Rod Ejection

,:.i4, Section...RG Position v'Analysis.

Comments

-, ," -..., Transport From Containment 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 hours2.777778e-4 days <br />0.00667 hours <br />3.968254e-5 weeks <br />9.132e-6 months <br />, 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 hours2.777778e-4 days <br />0.00667 hours <br />3.968254e-5 weeks <br />9.132e-6 months <br />, 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 Revision 1 Table .A6- G:. Conformance.

With;Rea ulatorv Guide ;1.18.-

!H PWR Rod, Eiection Section :T, -- -T .. IT-f:T -=;: --- RG Position Analysi 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., lbm/hr) should be consistent with the 1.0 gm/cc (62.4 lbm/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 Revision 1 L.-.i :Table: A5 .-H:T Conformance with .Regulatory.

Guide-i:.1l83,!Appendx.l

,(EquipmentQualification)

-.--. 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 DC;PP EQ analysis will Position 6 of this guide, this appendix supersedes Regulatory continue to be based upon TIID-14844 Positions 2.c.(1) and 2.c.(2) and Appendix D of Revision I 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 I .89 remain________effective.________________________

A5-61 Revision 1