Text

Transmittal of Supplement to License Amendment Request 241 Regulatory Guide 1.183 Compliance Matrix
	ML091320437
Person / Time
Site:	Point Beach
Issue date:	05/08/2009
From:	Meyer L; Point Beach
To:	; Document Control Desk, Office of Nuclear Reactor Regulation
References
	NRC 2009-0026
	Download: ML091320437 (57)
	v • d • e

May 8,2009 NRC 2009-0026 10 CFR 50.90 U. S. Nuclear Regulatory Commission ATTN: Document Control Desk Washington, DC 20555 Point Beach Nuclear Plant, Units 1 and 2 Dockets 50-266 and 50-301 Renewed License Nos. DPR-24 and DPR-27 Supplement to License Amendment Request 241 Reaulatow Guide I.I 83 Compliance Matrix

Reference:

(1) FPL Energy Point Beach, LLC, License Amendment Request 241, dated December 8, 2008, Alternative Source Term (ML083450683)

NextEra Energy Point Beach, LLC (NextEra) has enclosed a Regulatory Guide I.I83 Compliance Matrix. This matrix was discussed during a teleconference held with NRC staff on January 13, 2009.

The enclosure supports the review of the NextEra submittal of License Amendment Request 241, "Alternative Source Term."

This supplement does not affect the no significant hazards consideration provided in Reference 1.

This letter contains no new commitments and no revisions to existing commitments.

In accordance with 10 CFR 50.91, a copy of this letter is being provided to the designated Wisconsin Official.

I declare under penalty of perjury that the foregoing is true and correct.

Executed on May 8,2009.

Very truly yours, Enclosure cc: Administrator, Region Ill, USNRC Project Manager, Point Beach Nuclear Plant, USNRC Resident Inspector, Point Beach Nuclear Plant, USNRC PSCW NextEra Energy Point Beach, LLC,6610 Nuclear Road, Two Rivers, WI 54241

ENCLOSURE NEXTERA ENERGY POINT BEACH, LLC POINT BEACH NUCLEAR PLANT, UNITS 1 AND 2 SUPPLEMENT TO LICENSE AMENDMENT REQUEST 241 REGULATORY GUIDE 1.183 COMPLIANCE MATRIX 55 pages follow

Point Beach Nuclear Plant, Units 1 and 2 License Amendment Request 241 Dockets 50-266 and 50-301 Alternative Source Term Regulatory Guide 1.183 Compliance Matrix Regulatory Guidance Comments

3. ACCIDENT SOURCE TERM Fission Product Inventory Complies The inventory of fission products in the reactor core and available The source terms derived from the reactor core fission product for release to the containment should be based on the maximum inventory are listed in Table 5 of LAR 241, Enclosure 3. The full power operation of the core with, as a minimum, current inventory of the nuclides in the reactor core is based on maximum licensed values for fuel enrichment, fuel burnup, and an assumed full-power operation of the core at a power level equal to core power equal to the current licensed rated thermal power times 1811 MWt, which includes a 0.6% power uncertainty. Further, the the ECCS evaluation uncertainty8.The period of irradiation should calculated nuclide inventories are increased by a factor of 1.04 to be of sufficient duration to allow the activity of dose-significant account for potential fuel variations (LAR 241, Enclosure 3, radionuclides to reach equilibrium or to reach maximum valuesg. Section 3.1).

The core inventory should be determined using an appropriate isotope generation and depletion computer code such as ORIGEN The period of irradiation was of sufficient duration to achieve 2 (Ref. 17) or ORIGEN-ARP (Ref. 18). Core inventory factors equilibrium or maximum isotopic concentrations, for a (CiIMWt) provided in TID 14844 and used in some analysis representative operating cycle. The core inventory was determined computer codes were derived for low burnup, low enrichment fuel with the ORIGEN-S code (LAR 241, Enclosure 3, Sections 1.5 and and should not be used with higher burnup and higher enrichment 3.1).

fuels.

For DBA events that do not involve the entire core (Locked Rotor For the DBA LOCA, all fuel assemblies in the core are assumed to Accident (LR), Control Rod Ejection Accident (CRDE), and Fuel be affected and the core average inventory should be used. For Handling Accident (FHA)), a peaking factor of 1.7 is applied to the DBA events that do not involve the entire core, the fission product average assembly inventory (LAR 241, Enclosure 3, Tables 20, 22 inventory of each of the damaged fuel rods is determined by and 23).

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 the Design Basis (DBA) Loss of Coolant Accident (LOCA), all radial peaking factors from the facility's core operating limits report fuel assemblies in the core are assumed to be affected, and the (COLR) or technical specifications should be applied in determining total core inventory is used (LAR 241, Enclosure 3, Table 5).

the inventory of the damaged rods.

No adjustments for less than full power were made. For the FHA, a decay time of 65 hours7.523148e-4 days 0.0181 hours 1.074735e-4 weeks 2.47325e-5 months from the time of shutdown was assumed (LAR 241, Enclosure 3, Table 23).

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Point Beach Nuclear Plant, Units 1 and 2 License Amendment Request 241 Dockets 50-266 and 50-301 Alternative Source Term Regulatory Guide 1.I83 Compliance Matrix No adjustment to the fission product inventory should be made for events postulated to occur during power operations at less than 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.

Footnote 8 - The uncertainty factor used in determining the core inventory should be that value provided in Appendix K to 10 CFR Part 50, typically 1.02.

Footnote 9 - Note that for some radionuclides, such as Cs-137, equilibrium will not be reached prior to fuel offload. Thus, the maximum inventory at the end of life should be used.

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Point Beach Nuclear Plant, Units 1 and 2 License Amendment Request 241 Dockets 50-266 and 50-301 Alternative Source Term Regulatory Guide 1.183 Compliance Matrix 3.2 Release Fractions Complies The core inventory release fractions, by radionuclide groups, for the For the LOCA, release fractions are as given in Table 2 of gap release and early in-vessel damage phases for DBA LOCAs RG 1.I 83 and are applied to the core inventory described in are listed in Table 1 for BWRs and Table 2 for PWRs. These Section 3.1 above.

fractions are applied to the equilibrium core inventory described in Regulatory Position 3.1. For non-LOCA events, see gap inventory discussion provided in LAR 241, Enclosure 3, Section 3.3.

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 PBNP core consists of fuel rods with an initial composition of The release fractions from Table 3 are used in conjunction with the natural or slightly enriched uranium dioxide (U02) as fuel material.

fission product inventory calculated with the maximum core radial PBNP does not use MOX fuel.

peaking factor.

Does Not Comply with Footnote 11 Footnote 10 - The release fractions listed here have been determined to be acceptable for use with currently approved LWR Some fuel was determined to exceed the burnup criteria of fuel with a peak burnup up to 62,000 MWDIMTU. The data in this Footnote 11. Alternate gap fractions were assumed for this fuel section may not be applicable to cores containing mixed oxide (LAR 241, Enclosure 3, Section 3.3),

(MOX) fuel.

Also, see FHA in Appendix B of this document.

Footnote 11- 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 MWDIMTU provided that the maximum linear heat generation rate does not exceed 6.3 kwlft 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.

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Point Beach Nuclear Plant, Units 1 and 2 License Amendment Request 241 Dockets 50-266 and 50-301 Alternative Source Term Regulatory Guide 1.183 Compliance Matrix 3.3 Timing of Release Phases Complies Table 4 tabulates the onset and duration of each sequential release The activity released from the core during each release phase is phase for DBA LOCAs at PWRs and BWRs. The specified onset is modeled in a linear fashion with the RADTRAD code (LAR 241, the time following the initiation of the accident (i.e., time = 0). The Enclosure 3, Section 1.5). Gap release and early-in-vessel release early in-vessel phase immediately follows the gap release phase. are modeled in the RADTRAD release fraction and timing file (rft).

The activity released from the core during each release phase The core fraction specified in the rft file is released linearly during should be modeled as increasing in a linear fashion over the the specified time.

duration of the phase12. For non-LOCA DBAs in which fuel damage is projected, the release from the fuel gap and the fuel The non-LOCA activity release rate is instantaneous, consistent pellet should be assumed to occur instantaneously with the onset with RG 1.183. The non-LOCA events that assume fuel damage of the projected damage. include the CRDE, LR and FHA. Instantaneous release (10~ hours)

~ is shown in the respective rft files for the CRDE and For facilities licensed with leak-before-break methodology, the LR. The FHA release to the containment occurs over 2 hours2.314815e-5 days 5.555556e-4 hours 3.306878e-6 weeks 7.61e-7 months onset of the gap release phase may be assumed to be 10 minutes. (LAR 241, Enclosure 3, Table 23).

A licensee may propose an alternative time for the onset of the gap release phase, based on facility-specific calculations using suitable The RVHD activity release rate is also instantaneous (1 hours).

analysis codes or on an accepted topical report shown to be The RVHD is specific to PBNP; not a RG 1.183 event.

applicable to the specific facility. In the absence of approved alternatives, the gap release phase onsets in Table 4 should be The leak-before-break methodology is not credited in the PBNP used. AST analyses. As such, the onset of the gap release is delayed by 30 seconds, consistent with RG 1.183, Table 4 (LAR 241, Footnote 12 - In lieu of treating the release in a linear ramp Enclosure 3, Table 16).

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.

3.4 Radionuclide Composition complies Table 5 lists the elements in each radionuclide group that should be The elements assumed in each radionuclide group are consistent considered in DBAs. (not identical) with those of Table 5 of RG 1.I 83 (LAR 241, Enclosure 3, Table 5). The specific nuclide selection was performed by Westinghouse. The nuclide grouping is consistent with that used in the RADTRAD users guide (NUREGICR-6604).

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Point Beach Nuclear Plant, Units 1 and 2 License Amendment Request 241 Dockets 50-266 and 50-301 Alternative Source Term Regulatory Guide 1.I 83 Compliance Matrix 3.5 Chemical Form Complies Of the radioiodine released from the reactor coolant system (RCS) The assumed iodine chemical forms are consistent with RG 1.I 83 to the containment in a postulated accident, 95% of the iodine (LAR 241, Enclosure 3, Tables 18 through 23).

released should be assumed to be cesium iodide (Csl), 4.85%

elemental iodine, and 0.15% organic iodide. This includes releases Within this document, refer to the following Appendices:

from the gap and the fuel pellets. With the exception of elemental and organic iodine and noble gases, fission products should be Appendix A, for LOCA assumed to be in particulate form. The same chemical form is Appendix B for FHA assumed in releases from fuel pins in FHAs and from releases from Appendix E for MSLB the fuel pins through the RCS in DBAs other than FHAs or LOCAs. Appendix F for SGTR However, the transport of these iodine species following release Appendix G for LR from the fuel may affect these assumed fractions. The accident- Appendix H for CRDE specific appendices to this regulatory guide provide additional details. For the RVHD, the assumed chemical form of iodine released is 100% elemental. The RVHD is not addressed in RG 1.183 (LAR 241, Enclosure 3, Table 24).

3.6 Fuel Damage in Non-LOCA DBAs Complies The amount of fuel damage caused by non-LOCA design basis For the FHA (1 fuel assembly) and CRDE (10% of core), fuel events should be analyzed to determine, for the case resulting in damage assumptions are based on the current licensing basis the highest radioactivity release, the fraction of the fuel that cases that result in conservative radioactivity releases (LAR 241, reaches or exceeds the initiation temperature of fuel melt and the Enclosure 3, Tables 22 and 23). The CLB values are provided in fraction of fuel elements for which the fuel clad is breached. PBNP FSAR Chapter 14. Applicable FSAR sections are listed in Although the NRC staff has traditionally relied upon the departure LAR 241, Enclosure 3. The locked rotor analysis assumes 30%

from nucleate boiling ratio (DNBR) as a fuel damage criterion, failed fuel (LAR 241, Enclosure 3, Table 20). The locked rotor rods licensees may propose other methods to the NRC staff, such as in DNB methodology is described in LAR 241, Enclosure 3, those based upon enthalpy deposition, for estimating fuel damage Section 6.3.

for the purpose of establishing radioactivity releases.

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Point Beach Nuclear Plant, Units 1 and 2 License Amendment Request 241 Dockets 50-266 and 50-301 Alternative Source Term Regulatory Guide 1.183 Compliance Matrix

4. DOSE CALCULATION METHODOLOGY

4. I Offsite Dose Consequences The following assumptions should be used in determining the TEDE for persons located at or beyond the boundary of the exclusion area (EAB):

4.1 .IThe dose calculations should determine the TEDE. TEDE is Complies the sum of the committed effective dose equivalent (CEDE) from inhalation and the deep dose equivalent (DDE) from external The dose calculations determine the TEDE dose, with all significant exposure. The calculation of these two components of the TEDE progeny included, as the sum of the CEDE and the EDE (LAR 241, should consider all radionuclides, including progeny from the decay Enclosure 3, Section 2.2). TEDE doses were calculated with of parent radionuclides that are significant with regard to dose RADTRAD Version 3.03 which implements the TEDE dose consequences and the released radioactivityI3. methodology (LAR 241, Enclosure 3, Section 1.5).

Footnote 13 - The prior practice of basing inhalation exposure on only radioiodine and not including radioiodine in external exposure calculations is not consistent with the definition of TEDE and the characteristics of the revised source term.

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Point Beach Nuclear Plant, Units 1 and 2 License Amendment Request 241 Dockets 50-266 and 50-301 Alternative Source Term Regulatory Guide 1.183 Compliance Matrix 4.1.2 The exposure-to-CEDE factors for inhalation of radioactive Complies material should be derived from the data provided in ICRP Publication 30, "Limits for Intakes of Radionuclides by Workers" The dose conversion factors (DCF's) used in determining the (Ref. 19). Table 2.1 of Federal Guidance Report 11, "Limiting CEDE dose are from EPA Federal Guidance Report 11. The Values of Radionuclide Intake and Air Concentration and Dose CEDE dose factors are provided in LAR 241, Enclosure 3, Table I.

Conversion Factors for Inhalation, Submersion, and Ingestion" (Ref. 20), provides tables of conversion factors acceptable to the NRC staff. The factors in the column headed "effective1'yield doses corresponding to the CEDE. The exposure-to-CEDE factors for inhalation of radioactive material should be derived from the data provided in ICRP Publication 30, "Limits for Intakes of Radionuclides by Workers" (Ref. 19). Table 2.1 of Federal Guidance Report 11, "Limiting Values of Radionuclide lntake and Air Concentration and Dose Conversion Factors for Inhalation, Submersion, and Ingestion" (Ref. 20), provides tables of conversion factors acceptable to the NRC staff. The factors in the column headed "effective" yield doses corresponding to the CEDE.

4.1.3 For the first 8 hours9.259259e-5 days 0.00222 hours 1.322751e-5 weeks 3.044e-6 months , the breathing rate of persons offsite Complies should be assumed to be 3.5 x cubic meters per second. From 8 to 24 hours2.777778e-4 days 0.00667 hours 3.968254e-5 weeks 9.132e-6 months following the accident, the breathing rate should be The assumed offsite breathing rates are those specified in assumed to be 1.8 x 1o - cubic

~ meters per second. After that and Section 4.1.3 of RG I .183 (LAR 241, Enclosure 3, Table 3).

until the end of the accident, the rate should be assumed to be 2.3 x 1o4 cubic meters per second.

4.1.4 The DDE should be calculated assuming submergence in Complies semi-infinite cloud assumptions with appropriate credit for attenuation by body tissue. EDE may be used in lieu of DDE in EDE is used to determine the submergence dose in a semi-infinite determining the contribution of external dose to the TEDE. cloud. The assumed conversion factors are those of Federal Guidance Report 12 (LAR 241, Enclosure 3, Table 2).

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Point Beach Nuclear Plant, Units 1 and 2 License Amendment Request 241 Dockets 50-266 and 50-301 Alternative Source Term Regulatory Guide 1.183 Compliance Matrix 4.1.5 The TEDE should be determined for the most limiting person Complies at the EAB. The maximum EAB TEDE for any two-hour period following the start of the radioactivity release should be determined The TEDE is determined for the most limiting person for a two-hour and used in determining compliance with the dose criteria in 10 period at the EAB and the maximum two-hour dose is reported.

CFR 50.67'~.The maximum two-hour TEDE should be determined by calculating the postulated dose for a series of small time LAR 241, Enclosure 3, Section 2.3 states that doses at the EAB are increments and performing a "sliding" sum over the increments for for the worst 2-hour period. The EAB X/Q is determined for the successive two-hour periods. The maximum TEDE obtained is worst sector. The combination of the maximum 2-hr dose using the submitted. The time increments should appropriately reflect the worst case X/Q results in the dose to the limiting person.

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).

Footnote 14 -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 Complies the outer boundary of the low population zone (LPZ) and should be used in determining compliance with the dose criteria in The LPZ TEDE is calculated up to the time releases are terminated 10 CFR 50.67. (LAR 241, Enclosure 3, Section 2.3). The XIQ values for the LPZ are based on the worst sector.

4.1.7 No correction should be made for depletion of the effluent Complies plume by deposition on the ground.

No plume depletion due to ground deposition is credited (LAR 241, Enclosure 3, Section 2.2).

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Point Beach Nuclear Plant, Units 1 and 2 License Amendment Request 241 Dockets 50-266 and 50-301 Alternative Source Term Regulatory Guide 1.183 Compliance Matrix 4.2 Control Room Dose Consequences The following guidance should be used in determining the TEDE for persons located in the control room:

4.2.1 The TEDE analysis should consider all sources of radiation Complies.

that will cause exposure to control room personnel. The applicable sources will vary from facility to facility, but typically will include: The LOCA shine dose to CR personnel (due to containment and ECCS leakage) includes doses from containment and the auxiliary e Contamination of the control room atmosphere by the intake building, and doses from contained sources and the external plume or infiltration of the radioactive material contained in the (LAR 241, Enclosure 3, Section 6.1). The LOCA 30-day DDE is radioactive plume released from the facility, 0.28 rem (LAR 241, Enclosure 3, Section 6.1). The non-LOCA CR TEDE doses all have sufficient margin to the 5 rem limit (LAR 241, e Contamination of the control room atmosphere by the intake Enclosure 3, Section 1.2).

or infiltration of airborne radioactive material from areas and structures adjacent to the control room envelope, a Radiation shine from the external radioactive plume released from the facility, Radiation shine from radioactive material in the reactor containment, Radiation shine from radioactive material in systems and components inside or external to the control room envelope, e.g., radioactive material buildup in recirculation filters.

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Point Beach Nuclear Plant, Units 1 and 2 License Amendment Request 241 Dockets 50-266 and 50-301 Alternative Source Term Regulatory Guide 1.I 83 Compliance Matrix 4.2.2 The radioactive material releases and radiation levels used in Does not Comply the control room dose analysis should be determined using the same source term, transport, and release assumptions used for The control room TEDE doses are determined using the same determining the EAB and the LPZ TEDE values, unless these source term, in-plant transport, and release assumptions used for assumptions would result in non-conservative results for the control determining the EAB and the LPZ TEDE values, resulting in room. conservative results for the control room. However, the offsite and CR atmospheric transport assumptions differ. CR XIQ values assume meteorological data from years 2000 through 2005, while the offsite CLB XIQ values use data from years 1991 through 1993 (LAR 241, Enclosure 3, Section 4.4).

PBNP has assessed the CLB X/Q values against XIQ values generated using the PAVAN computer code with the meteorological data collected at PBNP from September 2000 to September 2005 and determined the CLB XIQ's are conservative (LAR 241, Enclosure 3, Section 4.4).

4.2.3 The models used to transport radioactive material into and Complies through the control roomq5,and the shielding models used to determine radiation dose rates from external sources, should be The models used to transport radioactive material into and through structured to provide suitably conservative estimates of the the control room, and the shielding models used to determine exposure to control room personnel. radiation dose rates from external sources, are conservative and consistent with the recommendations of RG 1.I 83.

Footnote 15 - The iodine protection factor (IPF) methodology of Reference 22 may not be adequately conservative for all DBAs and Doses due to airborne radioactivity were calculated with control room arrangements since it models a steady-state control RADTRAD. The S&W PERC2 code was used to generate the room condition. Since many analysis parameters change over the source terms in the containment atmosphere, in the external plume duration of the event, the IPF methodology should only be used passing the CR, and in the CR charcoal and HEPA filters to with caution. The NRC computer codes HABIT (Ref. 23) and determine the shine dose for the CR operator (LAR 241, RADTRAD (Ref. 24) incorporate suitable methodologies. Enclosure 3, Section 1.5).

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Point Beach Nuclear Plant, Units Iand 2 License Amendment Request 241 Dockets 50-266 and 50-301 Alternative Source Term Regulatory Guide 1.I83 Compliance Matrix 4.2.4 Credit for engineered safety features that mitigate airborne Complies radioactive material within the control room may be assumed. Such features may include control room isolation or pressurization, or Credit is taken for CR emergency intake and recirculation filtration intake or recirculation filtration. Refer to Section 6.5.1, "ESF (LAR 241, Enclosure 3, Table 4). The credited filters are qualified Atmospheric Cleanup System," of the SRP (Ref. 3) and Regulatory and acceptable per the PBNP Ventilation Filter Testing Program Guide 1.52, "Design, Testing, and Maintenance Criteria for (PBNP TS 5.5.10).

Postaccident Engineered-Safety-FeatureAtmosphere Cleanup System Air Filtration and Adsorption Units of Light-Water-cooled In addition, the FHA (LAR 241, Enclosure 3, Section 6.6) credits a Nuclear Power Plants" (Ref. 25), for guidance. The control room high CR radiation alarm (RE-101 or RE-235) (LAR 241, design is often optimized for the DBA LOCA and the protection Enclosure 3, Section 5.2) for initiating CREFS within 10 minutes of afforded for other accident sequences may not be as an FHA (LAR 241, Enclosure 3, Section 6.6).

advantageous. In most designs, control room isolation is actuated by engineered safeguards feature (ESF) signals or radiation monitors (RMs). In some cases, the ESF signal is effective only for selected accidents, placing reliance on the RMs for the remaining accidents. Several aspects of RMs can delay the control room isolation, including the delay for activity to build up to concentrations equivalent to the alarm setpoint and the effects of different radionuclide accident isotopic mixes on monitor response.

4.2.5 Credit should generally not be taken for the use of personal Complies protective equipment or prophylactic drugs. Deviations may be considered on a case-by-case basis. No credit is taken for the use of personal protective equipment or prophylactic drugs (LAR 241, Enclosure 3, Section 5.1).

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Point Beach Nuclear Plant, Units 1 and 2 License Amendment Request 241 Dockets 50-266 and 50-301 Alternative Source Term Regulatory Guide 1.183 Compliance Matrix 4.2.6 The dose receptor for these analyses is the hypothetical Complies maximum exposed individual who is present in the control room for 100% of the time during the first 24 hours2.777778e-4 days 0.00667 hours 3.968254e-5 weeks 9.132e-6 months after the event, 60% of The assumed breathing rates and occupancy factors for control the time between 1 and 4 days, and 40% of the time from 4 days to room operator dose are those specified in Section 4.2.6 of 30 daysg6.For the duration of the event, the breathing rate of this RG 1. I 83 (LAR 241, Enclosure 3, Table 4).

individual should be assumed to be 3.5 x 1o4 cubic meters per second.

Footnote 16 - This occupancy is modeled in the XIQ values determined in Reference 22 and should not be credited twice. The ARCON96 Code (Ref. 26) does not incorporate these occupancy assumptions, making it necessary to apply this correction in the dose calculations.

4.2.7 Control room doses should be calculated using dose Complies conversion factors identified in Regulatory Position 4.1 above for use in offsite dose analyses. The DDE from photons may be Control room doses are calculated using dose conversion factors corrected for the difference between finite cloud geometry in the identified in Regulatory Position 4.1 above. Offsite and CR doses control room and the semi-infinite cloud assumption used in were calculated using RADTRAD (LAR 241, Enclosure 3, calculating the dose conversion factors. Equation 1 may be used to Section 1.5).

correct the semi-infinite cloud dose, DDEinfinity, to a finite cloud dose, DDEfinite, where the control room is modeled as a CR and offsite TEDE doses are calculated with the DCFs provided hemisphere that has a volume, V, in cubic feet, equivalent to that of in LAR 241, Enclosure 3, Tables 1 and 2.

the control room (Ref. 22).

DDEfinite = DDEinfinity

V0.338 / 1173 Equation 1 Page 12 of 55

Point Beach Nuclear Plant, Units 1 and 2 License Amendment Request 241 Dockets 50-266 and 50-301 Alternative Source Term Regulatory Guide 1.I 83 Compliance Matrix 4.3 Other Dose Consequences Complies The guidance provided in Regulatory Positions 4.1 and 4.2 should RG 1.I 83, Item 6, states that NRC staff is assessing the effect of be used, as applicable, in re-assessing the radiological analyses increased cesium releases on EQ doses to determine whether identified in Regulatory Position 1.3.1, such as those in licensee action is warranted. Until such time as this generic issue is NUREG-0737 (Ref. 2). Design envelope source terms provided in resolved, licensees may use either the AST or the TI014844 NUREG-0737 should be updated for consistency with the AST. In assumptions for performing the required EQ analyses. However, general, radiation exposures to plant personnel identified in no plant modifications are required to address the impact of the Regulatory Position 1.3.1 should be expressed in terms of TEDE. difference in source term characteristics (i.e., AST vs. TlD14844)

Integrated radiation exposure of plant equipment should be on EQ doses pending the outcome of the evaluation of the generic determined using the guidance of Appendix I of this guide. issue. See Appendix I of this document.

4.4 Acceptance Criteria Complies The radiological criteria for the EAB, the outer boundary of the LPZ, The appropriate regulatory limits of 10 CFR 50.67 and RG 1. I 83 and for the control room are in 10 CFR 50.67. These criteria are are used. DBA analyses demonstrate that regulatory limits are not stated for evaluating reactor accidents of exceedingly low exceeded. Results are presented in terms of TEDE (LAR 241, probability of occurrence and low risk of public exposure to Enclosure 3, Section 1.2).

radiation, e.g., a large-break LOCA. The control room criterion applies to all accidents. For events with a higher probability of PBNP TS 5.5.8, "Steam Generator (SG) Program," describes occurrence, postulated EAB and LPZ doses should not exceed the alternative repair criteria. Table 6, "Accident Dose Criteria," was criteria tabulated in Table 6 of RG 1.I 83. applied (LAR 241, Enclosure 3, Section 1.2).

The acceptance criteria for the various NUREG-0737 (Ref. 2) items generally reference General Design Criteria 19 (GDC 19) from Appendix A to 10 CFR Part 50 or specify criteria derived from GDC-19. These criteria are generally specified in terms of whole body dose, or its equivalent to any body organ. For facilities 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 I 0 CFR 50.67(b)(Z)(iii).

Footnote 17 - For PWRs with steam generator alternative repair criteria, different dose criteria may apply to steam generator tube rupture and main steam line break analyses.

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Point Beach Nuclear Plant, Units 1 and 2 License Amendment Request 241 Dockets 50-266 and 50-301 Alternative Source Term Regulatory Guide 1.183 Compliance Matrix

5. ANALYSIS ASSUMPTIONS AND METHODOLOGY 5.1 General Considerations 5.1.IAnalysis Quality Complies The evaluations required by 10 CFR 50.67 are re-analyses of the The analyses have been prepared, reviewed and will be maintained design basis safety analyses and evaluations required by in accordance with quality assurance programs that comply with 10 CFR 50.34; they are considered to be a significant input to the 10 CFR Part 50 Appendix B, "Quality Assurance Criteria for evaluations required by 10 CFR 50.92 or 10 CFR 50.59. These Nuclear Power Plants and Fuel Reprocessing Plants." The dose analyses should be prepared, reviewed, and maintained in analyses are consistent with the recommendations of RG 1. I 83, accordance with quality assurance programs that comply with with the exception of the deviations noted in LAR 241, Enclosure 3, Appendix B, "Quality Assurance Criteria for Nuclear Power Plants Section 1.4.

and Fuel Reprocessing Plants," to 10 CFR Part 50.

These design basis analyses were structured to provide a conservative set of assumptions to test the performance of one or more aspects of the facility design. Many physical processes and phenomena are represented by conservative, bounding 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.

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Point Beach Nuclear Plant, Units 1 and 2 License Amendment Request 241 Dockets 50-266 and 50-301 Alternative Source Term Regulatory Guide 1.183 Compliance Matrix 5.1.2 Credit for Engineered Safeguard Features Complies Credit may be taken for accident mitigation features that are Non-safety-related mitigation features are credited in the dose classified as safety-related, are required to be operable by analyses. These systems include CREFS and VNPAB. These technical specifications, are powered by emergency power systems are designed to function following a single active failure sources, and are either automatically actuated or, in limited cases, (LAR 241, Enclosure 3, Section 1.4).

have actuation requirements explicitly addressed in emergency operating procedures. The single active component failure that In general, the LOOP was used .to limit equipment availability.

results in the most limiting radiological consequences should be Where credited, safety related mitigating structures, systems, and assumed. Assumptions regarding the occurrence and timing of a components were assumed to operate consistent with single loss of offsite power should be selected with the objective of failure, design basis, emergency power, and other requirements.

maximizing the postulated radiological consequences.

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Point Beach Nuclear Plant, Units 1 and 2 License Amendment Request 241 Dockets 50-266 and 50-301 Alternative Source Term Regulatory Guide 1.I 83 Compliance Matrix 5.1.3 Assignment of Numeric Input Values Complies The numeric values that are chosen as inputs to the analyses Conservative parameters are assumed when calculating each required by 10 CFR 50.67 should be selected with the objective of contributor in the dose analyses, consistent with the guidance of determining a conservative postulated dose. In some instances, a RG 1.183.

particular parameter may be conservative in one portion of an analysis but be nonconservative in another portion of the same Assumed analysis efficiencies for CREFS HEPA filters and analysis. For example, assuming minimum containment system adsorbers are consistent with the recommendations of RG I.52, spray flow is usually conservative for estimating iodine scrubbing, Rev. 3, Table I(LAR 241, Enclosure 3, Table 4).

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 specificati~ns'~. 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.

Footnote 18-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.

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Point Beach Nuclear Plant, Units 1 and 2 License Amendment Request 241 Dockets 50-266 and 50-301 Alternative Source Term Regulatory Guide 1.I83 Compliance Matrix 5.1.4 Applicability of Prior Licensing Basis Complies The NRC staff considers the implementation of an AST to be a Proposed changes to the PBNP CLB, required to implement the significant change to the design basis of the facility that is AST, are described in LAR 241, Enclosure 3, Section 1.3.

voluntarily initiated by the licensee. In order to issue a license 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.

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Point Beach Nuclear Plant, Units 1 and 2 License Amendment Request 241 Dockets 50-266 and 50-301 Alternative Source Term Regulatory Guide 1.I 83 Compliance Matrix 1 5.2 Accident-Specific Assumptions I Complies The appendices to this regulatory guide provide accident-specific The postulated accident radiological consequence analyses have assumptions that are acceptable to the staff for performing been reanalyzed for AST (LAR 241, Enclosure 3, Section 1.I).

analyses that are required by 10 CFR 50.67. The DBAs addressed in these attachments were selected from accidents that may For LOCA, see Appendix A involve damage to irradiated fuel. This guide does not address For the FHA, see Appendix B DBAs with radiological consequences based on technical For the MSLB, see Appendix E specification reactor or secondary coolant-specific activities only. For the SGTR, see Appendix F The inclusion or exclusion of a particular DBA in this guide should For the LR, see Appendix G not be interpreted as indicating that an analysis of that DBA is For the CRDE, see Appendix H required or not required. Licensees should analyze the DBAs that are affected by the specific proposed applications of an AST. The RVHD is not addressed in RG 1.I 83. It was also analyzed at the uprated conditions.

The NRC staff has determined that the analysis assumptions in the appendices to this guide provide an integrated approach to No changes have been made to analysis assumptions based upon performing the individual analyses and generally expects licensees risk insights.

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.

The NRC is committed to using probabilistic risk analysis (PRA) insights in its regulatory activities and will consider licensee 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 Page 18 of 55

Point Beach Nuclear Plant, Units 1 and 2 License Amendment Request 241 Dockets 50-266 and 50-301 Alternative Source Term Regulatory Guide 1.183 Compliance Matrix accident considerations not adequately addressed by the core damage frequency (CDF) and large early release frequency (LERF) surrogate indicators of overall risk.

5.3 Meteorology Assumptions Complies Atmospheric dispersion values (X/Q) for the EAB, the LPZ, and the The atmospheric dispersion (X/Q) values for the PBNP exclusion control room that were approved by the staff during initial facility area boundary (EAB) and the low population zone (LPZ) are those licensing or in subsequent licensing proceedings may be used in from the current licensing basis. RG 1.194 guidance has been performing the radiological analyses identified by this guide. used for onsite X/Q values (LAR 241, Enclosure 3, Section 4.5).

Methodologies that have been used for determining X/Q values are Fumigation has not been included since no credit is taken for an documented in Regulatory Guides 1.3 and 1.4, Regulatory Guide elevated release. ARCON96 was used for determining onsite X/Q 1.145, "Atmospheric Dispersion Models for Potential Accident values. Meteorological data acquired in accordance with the PBNP Consequence Assessments at Nuclear Power Plants," and the meteorological measurement program for the five-year period from paper, "Nuclear Power Plant Control Room Ventilation System 2000 to 2005 is used to calculate onsite atmospheric dispersion Design for Meeting General Criterion 19" (Refs. 6, 7, 22, and 28). (LAR 241, Enclosure 3, Section 4.5).

References 22 and 28 should be used if the FSAR X/Q values are to be revised or if values are to be determined for new release points or receptor distances. Fumigation should be considered where applicable for the EAB and LPZ. For the EAB, the assumed fumigation period should be timed to be included in the worst 2-hour exposure period. The NRC computer code PAVAN (Ref. 29) implements Regulatory Guide 1.I45 (Ref. 28) and its use is acceptable to the NRC staff. The methodology of the ,NRC computer code ARCON96 (Ref. 26) is generally acceptable to the NRC staff for use in determining control room X/Q values.

Meteorological data collected in accordance with the site-specific meteorological measurements program described in the facility FSAR should be used in generating accident X/Q values. Additional guidance is provided in Regulatory Guide 1.23, "Onsite Meteorological Programs" (Ref. 30). All changes in X/Q analysis methodology should be reviewed by the NRC staff.

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Point Beach Nuclear Plant, Units Iand 2 License Amendment Request 241 Dockets 50-266 and 50-301 Alternative Source Term Regulatory Guide 1.183 Compliance Matrix

6. ASSUMPTIONS FOR EVALUATING THE RADIATION DOSES FOR EQUIPMENT QUALIFICATION The assumptions in Appendix I to this guide are acceptable to the Complies NRC staff for performing radiological assessments associated with equipment qualification. The assumptions in Appendix I will Radiation environmental qualification of equipment analyses are supersede Regulatory Positions 2.c(I) and 2.c(2) and Appendix D not modified by this LAR. PBNP will continue to use the CLB of Revision Iof Regulatory Guide 1.89, "Environmental qualification analyses, which are based on the TID-14844 source Qualification of Certain Electric Equipment Important to Safety for term (LAR 241, Enclosure 3, Section 1.4).

Nuclear Power Plants" (Ref. 1I), for operating reactors that have amended their licensing basis to use an alternative source term.

Except as stated in Appendix I, all other assumptions, methods, and provisions of Revision Iof Regulatory Guide I.89 remain effective.

The NRC staff is assessing the effect of increased cesium releases on EQ doses to determine whether licensee action is warranted.

Until such time as this generic issue is resolved, licensees may use 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. TlD14844) on EQ doses pending the outcome of the evaluation of the generic issue.

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Point Beach Nuclear Plant, Units Iand 2 License Amendment Request 241 Dockets 50-266 and 50-301 Alternative Source Term Regulatory Guide 1.183 Compliance Matrix APPENDIX A ASSUMPTIONS FOR EVALUATING THE RADIOLOGICAL CONSEQUENCES OF A LWR LOSS-OF-COOLANT ACCIDENT

1. Acceptable assumptions regarding core inventory and the Complies release of radionuclides from the fuel are provided in Regulatory Position 3 of this guide. See Section 3, "Accident Source Term."

2. If the sump or suppression pool pH is controlled at values of 7 or Complies greater, the chemical form of radioiodine released to the containment should be assumed to be 95% cesium iodide (Csl), Iodine chemical forms are consistent with the RG, i.e., 95% cesium 4.85 percent elemental iodine, and 0.15 percent organic iodide. iodide (Csl), 4.85 percent elemental iodine, and 0.15 percent Iodine species, including those from iodine re-evolution, for sump organic iodide (LAR 241, Enclosure 3, Table 18).

or suppression pool pH values less than 7 will be evaluated on a case-by-case basis. Evaluations of pH should consider the effect of The sump pH is controlled at a value greater than 7.0 (LAR 241, acids and bases created during the LOCA event, e.g., radiolysis Enclosure 3, Table 18).

products. With the exception of elemental and organic iodine and noble gases, fission products should be assumed to be in particulate form.

3. Acceptable assumptions related to the transport, reduction, and release of radioactive material in and from the primary containment in PWRs or the drywell in BWRs are as follows:

3.1 The radioactivity released from the fuel should be assumed to Complies mix instantaneously and homogeneously throughout the free air volume of the primary containment in PWRs or the drywell in BWRs The activity initially released from the fuel is assumed to mix as it is released. This distribution should be adjusted if there are instantaneously and homogeneously throughout the assumed internal compartments that have limited ventilation exchange. The unsprayed volume of the containment (LAR 241, Enclosure 3, suppression pool free air volume may be included provided there is Section 6.1).

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.

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Point Beach Nuclear Plant, Units 1 and 2 License Amendment Request 241 Dockets 50-266 and 50-301 Alternative Source Term Regulatory Guide 1.I 83 Compliance Matrix 3.2 Reduction in airborne radioactivity in the containment by Complies natural deposition within the containment may be credited.

Acceptable models for removal of iodine and aerosols are For the containment leakage analysis, all activity released from the described in Chapter 6.5.2, "Containment Spray as a Fission fuel is assumed to be in the containment atmosphere until removed Product Cleanup System," of the Standard Review Plan (SRP), by sprays, sedimentation, radioactive decay or leakage from the NUREG-0800 (Ref. A-I) and in NUREGICR-6189, "A Simplified containment. Reduction of the airborne radioactivity in the Model of Aerosol Removal by Natural Processes in Reactor containment by natural deposition is credited. The natural Containments" (Ref. A-2). The latter model is incorporated into the deposition removal coefficient for particulates was determined to be analysis code RADTRAD (Ref. A-3). The prior practice of O.llhr. Instantaneous plate-out is not assumed (LAR 241, deterministically assuming that a 50% plateout of iodine is released Enclosure 3, Section 6.1).

from the fuel is no longer acceptable to the NRC staff as it is inconsistent with the characteristics of the revised source terms.

3.3 Reduction in airborne radioactivity in the containment by Complies containment spray systems that have been designed and are maintained in accordance with Chapter 6.5.2 of the SRP (Ref. A-I) Removal of elemental iodine from the containment atmosphere by may be credited. Acceptable models for the removal of iodine and spray is assumed to be terminated when the airborne inventory aerosols are described in Chapter 6.5.2 of the SRP and drops to 0.5 percent of the total elemental iodine released to the NUREGICR-5966, "A Simplified Model of Aerosol Removal by containment (this is a decontamination factor or DF of 200)

Containment sprays"' (Ref. A-4). This simplified model is (LAR 241, Enclosure 3, Section 6.1).

incorporated into the analysis code RADTRAD (Refs. A-I to A-3). The containment building is modeled as two discrete volumes, sprayed and unsprayed. The volumes are conservatively assumed The evaluation of the containment sprays should address areas to be mixed only by the containment fan coolers (LAR 241, within the primary containment that are not covered by the spray Enclosure 3, Section 6.1).

drops. The mixing rate attributed to natural convection between sprayed and unsprayed regions of the containment building, The particulate removal rate is reduced by a factor of 10 when a DF provided that adequate flow exists between these regions, is of 50 is reached (LAR 241, Enclosure 3, Section 6.1).

assumed to be two turnovers of the unsprayed regions per hour, unless other rates are justified. The containment building atmosphere may be considered a single, well-mixed volume if the spray covers at least 90% of the volume and if adequate mixing of unsprayed compartments can be shown. The SRP sets forth a maximum decontamination factor (DF) for elemental iodine based on the maximum iodine activity in the primary containment Page 22 of 55

Point Beach Nuclear Plant, Units 1 and 2 License Amendment Request 241 Dockets 50-266 and 50-301 Alternative Source Term Regulatory Guide 1.183 Compliance Matrix atmosphere when the sprays actuate, divided by the activity of iodine remaining at some time after decontamination.

The SRP also states that the particulate iodine removal rate should be reduced by a factor of 10 when a DF of 50 is reached. The reduction in the removal rate is not required if the removal rate is based on the calculated time-dependent airborne aerosol mass.

There is no specified maximum DF for aerosol removal by sprays.

The maximum activity to be used in determining the DF is defined as the iodine activity in the columns labeled "Total" in Tables 1 and 2 of this guide multiplied by 0.05 for elemental iodine and by 0.95 for particulate iodine (i.e., aerosol treated as particulate in SRP methodology).

Footnote 1 - This document describes statistical formulations with differing levels of uncertainty. The removal rate constants selected for use in design basis calculations should be those that will maximize the dose consequences. For BWRs, the simplified model should be used only if the release from the core is not directed through the suppression pool, Iodine removal in the suppression pool affects the iodine species assumed by the model to be present initially.

3.4 Reduction in airborne radioactivity in the containment by in- NIA containment recirculation filter systems may be credited if these systems meet the guidance of Regulatory Guide 1.52 and Generic PBNP does not have post-accident in-containment air filtration Letter 99-02 (Refs. A-5 and A-6). The filter media loading caused systems. This position is not applicable to PBNP.

by the increased aerosol release associated with the revised source term should be addressed.

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Point Beach Nuclear Plant, Units 1 and 2 License Amendment Request 241 Dockets 50-266 and 50-301 Alternative Source Term Regulatory Guide 1.183 Compliance Matrix 3.5 Reduction in airborne radioactivity in the containment by NIA suppression pool scrubbing in BWRs should generally not be credited. However, the staff may consider such reduction on an This position relates to suppression pool scrubbing in BWRs. The individual case basis. The evaluation should consider the relative position is not applicable to PBNP.

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 NIA retention in ice condensers, or other engineering safety features not addressed above, should be evaluated on an individual case This position relates to activity retention in ice condensers. This basis. See Section 6.5.4 of the SRP (Ref. A-I). position is not applicable to PBNP.

3.7 The primary containment (i.e., drywell for Mark I and il Complies containment designs) should be assumed to leak at the peak pressure technical specification leak rate for the first 24 hours2.777778e-4 days 0.00667 hours 3.968254e-5 weeks 9.132e-6 months . For PBNP is a PWR. A containment leak rate, based on the proposed PWRs, the leak rate may be reduced after the first 24 hours2.777778e-4 days 0.00667 hours 3.968254e-5 weeks 9.132e-6 months to 50% technical specifications, of 0.2% per day of the containment air is of the technical specification leak rate. For BWRs, leakage may be assumed for the first 24 hours2.777778e-4 days 0.00667 hours 3.968254e-5 weeks 9.132e-6 months . After 24 hours2.777778e-4 days 0.00667 hours 3.968254e-5 weeks 9.132e-6 months , the containment reduced after the first 24 hours2.777778e-4 days 0.00667 hours 3.968254e-5 weeks 9.132e-6 months , if supported by plant configuration leak rate is reduced to 0.1% per day (LAR 241, Enclosure 3, and analyses, to a value not less than 50% of the technical Table 18).

specification leak rate. Leakage from subatmospheric containments is assumed to terminate when the containment is brought to and maintained at a subatmospheric condition as defined by technical specifications.

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.

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Point Beach Nuclear Plant, Units 1 and 2 License Amendment Request 241.

Dockets 50-266 and 50-301 Alternative Source Term Regulatory Guide 1.I 83 Compliance Matrix 3.8 If the primary containment is routinely purged during power NIA operations, releases via the purge system prior to containment isolation should be analyzed and the resulting doses summed with PBNP does not use the containment purge during power operation.

the postulated doses from other release paths. The purge release evaluation should assume that 100% of the radionuclide inventory During normal reactor operation at power, the containment may be in the reactor coolant system liquid is released to the containment vented by use of the containment gaseous and particulate sampling at the initiation of the LOCA. This inventory should be based on the and monitoring penetrations. The system is automatically isolated technical specification reactor coolant system equilibrium activity. in the event of a containment isolation signal.

Iodine spikes need not be considered. If the purge system is not isolated before the onset of the gap release phase, the release fractions associated with the gap release and early in-vessel phases should be considered as applicable.

ASSUMPTIONS ON DUAL CONTAINMENTS

4. For facilities with dual containment systems, the acceptable NIA assumptions related to the transport, reduction, and release of radioactive material in and from the secondary containment or Regulatory Positions 4.1 through 4.6 apply to facilities with dual enclosure buildings are as follows. containment systems. These positions are not applicable to PBNP.

ASSUMPTIONS ON ESF SYSTEM LEAKAGE

5. ESF systems that recirculate sump water outside of the primary Complies containment are assumed to leak during their intended operation.

This release source includes leakage through valve packing The radiological consequences from postulated ESF systems glands, pump shaft seals, flanged connections, and other similar leakage is analyzed and combined with consequences postulated components. This release source may also include leakage through for containment leakage (LAR 241, Enclosure 3, Section 6.1).

valves isolating interfacing systems (Ref. A-7). The radiological consequences from the postulated leakage should be analyzed and combined with consequences postulated for other fission product release paths to determine the total calculated radiological consequences from the LOCA. The following assumptions are acceptable for evaluating the consequences of leakage from ESF components outside the primary containment for BWRs and PWRs.

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Point Beach Nuclear Plant, Units 1 and 2 License Amendment Request 241 Dockets 50-266 and 50-301 Alternative Source Term Regulatory Guide 1.I 83 Compliance Matrix 5.1 With the exception of noble gases, all the fission products Complies released from the fuel to the containment (as defined in Tables 1 and 2 of this guide) should be assumed to instantaneously and Engineered Safety Feature (ESF) systems that recirculate water homogeneously mix in the primary containment sump water (in outside the primary containment (ECCS systems) are assumed to PWRs) or suppression pool (in BWRs) at the time of release from leak during their intended operation. Only iodine is released the core. In lieu of this deterministic approach, suitably through this pathway since the noble gases are not assumed to conservative mechanistic models for the transport of airborne dissolve in the sump and particulates would remain in the water of activity in containment to the sump water may be used. Note that the ECCS leakage (LAR 241, Enclosure 3, Section 6.1).

many of the parameters that make spray and deposition models conservative with regard to containment airborne leakage are nonconservative with regard to the buildup of sump activity.

5.2 The leakage should be taken as two times the sum of the Complies simultaneous leakage from all components in the ESF recirculation systems above which the technical specifications, or licensee Leakage from the ECCS system to the ESF rooms is 800 cclmin commitments to item III.D.1.1 of NUREG-0737 (Ref. A-8), would (two times the Program value of 400 cclmin). Recirculation is require declaring such systems inoperable. The leakage should be conservatively initiated at 0 minutes. 500 cclmin is assumed to assumed to start at the earliest time the recirculation flow occurs in flow to the RWST and 300 cclmin to the PAB. The assumption of these systems and end at the latest time the releases from these the ECCS leakage beginning at 0 minutes is not consistent with the systems are terminated. Consideration should also be given to assumption of injection spray termination in the containment design leakage through valves isolating ESF recirculation systems leakage portion of the analysis. However, beginning the ECCS from tanks vented to atmosphere, e.g., emergency core cooling leakage at 0 minutes adds conservatism to the dose system (ECCS) pump miniflow return to the refueling water storage consequences. The leakage continues for the 30-day period tank. following the accident considered in the analysis (LAR 241, Enclosure 3, Section 6.1).

5.3 With the exception of iodine, all radioactive materials in the Complies recirculating liquid should be assumed to be retained in the liquid phase. With the exception of iodine, all radioactive materials in the recirculating liquid (sump water) are assumed to be retained in the liquid phase (LAR 241, Enclosure 3, Section 6.1).

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Point Beach Nuclear Plant, Units 1 and 2 License Amendment Request 241 Dockets 50-266 and 50-301 Alternative Source Term Regulatory Guide 1.183 Compliance Matrix 5.4 If the temperature of the leakage exceeds 212"F, the fraction Complies of total iodine in the liquid that becomes airborne should be assumed equal to the fraction of the leakage that flashes to vapor. The temperature of the leakage was determined to exceed 212"F, This flash fraction, FF, should be determined using a constant and the flashing fraction was calculated based on the temperature enthalpy, h, process, based on the maximum time-dependent of the containment sump liquid at the time recirculation begins.

temperature of the sump water circulating outside the containment:

FF = (hfl - hf2) 1 hfg Where: hfl is the enthalpy of liquid at system design temperature and pressure; hf2 is the enthalpy of liquid at saturation conditions (14.7 psia, 212OF); and hfg is the heat of vaporization at 212OF.

5.5 If the temperature of the leakage is less than 212°F or the Complies calculated flash fraction is less than lo%, the amount of iodine that becomes airborne should be assumed to be 10% of the total iodine For the ECCS leakage to the auxiliary building, 10% of the total activity in the leaked fluid, unless a smaller amount can be justified iodine in the leaked ECCS fluid is assumed to be available for based on the actual sump pH history and area ventilation rates. release and is assumed to become airborne and leak directly to the environment from the initiation of recirculation through 30 days (LAR 241, Enclosure 3, Table 18).

5.6 The radioiodine that is postulated to be available for release to Complies the environment is assumed to be 97% elemental and 3% organic.

Reduction in release activity by dilution or holdup within buildings, For ECCS leakage into the auxiliary building and RWST, the form or by ESF ventilation filtration systems, may be credited where of the released iodine is 97% elemental and 3% organic. No credit applicable. Filter systems used in these applications should be for holdup, filtration or dilution of ECCS leakage into the auxiliary evaluated against the guidance of Regulatory Guide 1.52 (Ref. A-5) building is taken (LAR 241, Enclosure 3, Table 18).

and Generic Letter 99-02 (Ref. A-6).

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Point Beach Nuclear Plant, Units 1 and 2 License Amendment Request 241 Dockets 50-266 and 50-301 Alternative Source Term Regulatory Guide 1.183 Compliance Matrix ASSUMPTIONS ON MAIN STEAM ISOLATION VALVE LEAKAGE IN BWRS

6. For BWRs, the main steam isolation valves (MSIVs) have NIA design leakage that may result in a radioactivity release. The radiological consequences from postulated MSlV leakage should Regulatory Positions 6.1 through 6.5 relate to MSSV leakage in be analyzed and combined with consequences postulated for other BWRs, which are not applicable to PBNP.

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 NIA containment purging as a combustible gas or pressure control measure should be analyzed. If the installed containment purging Containment purge is not considered as a means of combustible capabilities are maintained for purposes of severe accident gas or pressure control for PBNP. Routine containment purge is management and are not credited in any design basis analysis, not assumed for this event.

radiological consequences need not be evaluated. If the primary containment purging is required within 30 days of the LOCA, the results of this analysis should be combined with consequences 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).

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Point Beach Nuclear Plant, Units 1 and 2 License Amendment Request 241 Dockets 50-266 and 50-301 Alternative Source Term Regulatory Guide 1.I 83 Compliance Matrix APPENDIX B ASSUMPTIONS FOR EVALUATING THE RADIOLOGICAL CONSEQUENCES OF A FUEL HANDLING ACCIDENT

1. Source Term Acceptable assumptions regarding core inventory and the release of radionuclides from the fuel are provided in Regulatory Position 3 of this guide.

1.IThe number of fuel rods damaged during the accident should Complies be based on a conservative analysis that considers the most limiting case. This analysis should consider parameters such as the It was assumed that a single fuel assembly is damaged, consistent weight of the dropped heavy load or the weight of a dropped fuel with the CLB (LAR 241, Enclosure 3, Table 23).

assembly (plus any attached handling grapples), the height of the drop, and the compression, torsion, and shear stresses on the irradiated fuel rods. Damage to adjacent fuel assemblies, if applicable (e.g., events over the reactor vessel), should be considered.

1.2 The fission product release from the breached fuel is based on Complies Regulatory Position 3.2 of this guide and the estimate of the number of fuel rods breached. All the gap activity in the damaged Consistent with PBNP CLB, all rods in one damaged fuel assembly rods is assumed to be instantaneously released. Radionuclides that are assumed to be breached.

should be considered include xenons, kryptons, halogens, cesiums, and rubidiums. Does not comply It was assumed that the single damaged fuel assembly exceeds the RG 1.183 Footnote IIlimits. In lieu of the Table 3 Non-LOCA gap fractions, higher gap fractions were applied to the damaged assembly, following a method previously approved by the NRC for Kewaunee Power Station (ML070430020). The gap fractions used are those from Safety Guide 25 with the value for 1-131 increased by 20%, to 0.12, consistent with the recommendation of NUREGICR-5009 (LAR 241, Enclosure 3, Section 3.3).

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Point Beach Nuclear Plant, Units Iand 2 License Amendment Request 241 Dockets 50-266 and 50-301 Alternative Source Term Regulatory Guide 1.183 Compliance Matrix 1.3 The chemical form of radioiodine released from the fuel to the Complies spent fuel pool should be assumed to be 95% cesium iodide (Csl),

4.85 percent elemental iodine, and 0.15 percent organic iodide. The chemical form of radioiodine released from the fuel to the The Csl released from the fuel is assumed to completely dissociate spent fuel pool are assumed to be 95% cesium iodide (Csl),

in the pool water. Because of the low pH of the pool water, the 4.85 percent elemental iodine, and 0.15 percent organic iodide iodine re- evolves as elemental iodine. This is assumed to occur (LAR 241, Enclosure 3, Section 6.6).

instantaneously. The NRC staff will consider, on a case-by-case basis, justifiable mechanistic treatment of the iodine release from the pool.

2.0 Water Depth Complies If the depth of water above the damaged fuel is 23 feet or greater, The minimum depth of water above the reactor vessel flange and the decontamination factors for the elemental and organic species above the top of irradiated assemblies during fuel movement is are 500 and I , respectively, giving an overall effective 23 feet (LAR 241, Enclosure 3, Table 23).

decontamination factor of 200 (i.e., 99.5% of the total iodine released from the damaged rods is retained by the water). This The AST FHA dose analysis assumes an effective iodine DF of difference in decontamination factors for elemental (99.85%) and 200. This is equivalent to an elemental DF of 500 and an organic organic iodine (0.15%) species results in the iodine above the DF of 1 (LAR 241, Enclosure 3, Section 6.6).

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.

3. Noble Gases Complies The retention of noble gases in the water in the fuel pool or reactor Noble gas is not retained in the pool water (LAR 241, Enclosure 3, cavity is negligible (i.e., decontamination factor of 1). Particulate Table 23).

radionuclides are assumed to be retained by the water in the fuel pool or reactor cavity (i.e., infinite decontamination factor). Particulate radionuclides are assumed to be retained by the water in the fuel pool or reactor cavity (LAR 241, Enclosure 3, Section 6.6).

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Point Beach Nuclear Plant, Units 1 and 2 License Amendment Request 241 Dockets 50-266 and 50-301 Alternative Source Term Regulatory Guide 1.183 Compliance Matrix

4. Fuel Handling Accidents Within the Fuel Building NIA FHA in the Spent Fuel Pool was not analyzed. Since the assumptions and parameters used to model the release due to an FHA inside containment are identical to those for a FHA in the spent fuel pool, except for different CR intake atmospheric dispersion factors (X/Qs) for the different release paths, the activity released is the same regardless of the location of the accident. In order to bound the accident, the location with the highest X/Q value is assumed (LAR 241, Enclosure 3, Section 6.6, and Table 23).

I

5. Fuel Handling Accident Within Containment For fuel handling accidents postulated to occur within the containment, the following assumptions are acceptable to the NRC staff.

5.1 If the containment is isolated during fuel handling operations, no radiological consequences need to be analyzed.

The containment is not isolated (LAR 241, Enclosure 3, Section 6.6).

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 The containment is open and not isolated following an FHA radiation detection and completion of containment isolation. If it can (LAR 241, Enclosure 3, Section 6.6).

be shown that containment isolation occurs before radioactivity is released to the environment, no radiological consequences need to be analyzed.

5.3 If the containment is open during fuel handling operations (e.g., Complies personnel air lock or equipment hatch is open), the radioactive material that escapes from the reactor cavity pool to the Activity is released over a 2-hour time period (LAR 241, containment is released to the environment over a 2-hour time Enclosure 3, Table 23).

I period.

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Point Beach Nuclear Plant, Units 1 and 2 License Amendment Request 241 Dockets 50-266 and 50-301 Alternative Source Term Regulatory Guide 1.I 83 Compliance Matrix 5.4 A reduction in the amount of radioactive material released from Complies the containment by ESF filter systems may be taken into account provided that these systems meet the guidance of Regulatory No filtration is assumed (LAR 241, Enclosure 3, Table 23).

Guide 1.52 and Generic Letter 99-02. Delays in radiation detection, actuation of the ESF filtration system, or diversion of ventilation flow to the ESF filtration system should be determined and accounted for in the radioactivity release analyses.

5.5 Credit for dilution or mixing of the activity released from the NIA reactor cavity by natural or forced convection inside the containment may be considered on a case-by-case basis. Such All activity released from the water is released from the credit is generally limited to 50% of the containment free volume. containment to the environment over a 2-hour period. As such, the This evaluation should consider the magnitude of the containment assumed containment volume is not credited with dilution or mixing volume and exhaust rate, the potential for bypass to the (LAR 241, Enclosure 3, Table 23).

environment, the location of exhaust plenums relative to the surface of the reactor cavity, recirculation ventilation systems, and internal walls and floors that impede stream flow between the surface of the reactor cavity and the exhaust plenums.

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Point Beach Nuclear Plant, Units 1 and 2 License Amendment Request 241 Dockets 50-266 and 50-301 Alternative Source Term Regulatory Guide I.I83 Compliance Matrix APPENDIX C ASSUMPTIONS FOR EVALUATING THE RADIOLOGICAL CONSEQUENCES OF A BWR ROD DROP ACCIDENT This appendix is not applicable to PBNP.

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Point Beach Nuclear Plant, Units Iand 2 License Amendment Request 241 Dockets 50-266 and 50-301 Alternative Source Term Regulatory Guide 1.183 Compliance Matrix APPENDIX D ASSUMPTIONS FOR EVALUATING THE RADIOLOGICAL CONSEQUENCES OF A BWR MAIN STEAM LINE BREAK ACCIDENT This appendix is not applicable to PBNP.

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Point Beach Nuclear Plant, Units 1 and 2 License Amendment Request 241 Dockets 50-266 and 50-301 Alternative Source Term Regulatory Guide 1.183 Compliance Matrix APPENDIX E ASSUMPTIONS FOR EVALUATING THE RADIOLOGICAL CONSEQUENCES OF A PWR MAIN STEAM LINE BREAK ACCIDENT This appendix provides assumptions acceptable to the NRC staff for evaluating the radiological consequences of a main steam line break accident at PWR light water reactors. These assumptions supplement the guidance provided in the main body of this guide'.

Footnote 1 - Facilities licensed with, or applying for, alternative repair criteria (ARC) should use this section in conjunction with the guidance that is being developed in Draft Regulatory Guide DG-1074, "Steam Generator Tube Integrity," for acceptable assumptions and methodologies for performing radiological analyses.

SOURCE TERMS

1. Assumptions acceptable to the NRC staff regarding core I Complies inventory and the rekase of radionuclides fromthe fuel are provided in Regulatory Position 3 of this regulatory guide. The No fuel damage is postulated for the limiting event. The Source release from the breached fuel is based on Regulatory Position 3.2 Term specifies only initial reactor coolant activity (LAR 241, of this guide and the estimate of the number of fuel rods breached. Enclosure 3, Table 21).

The fuel damage estimate should assume that the highest worth control rod is stuck at its fully withdrawn position.

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Point Beach Nuclear Plant, Units 1 and 2 License Amendment Request 241 Dockets 50-266 and 50-301 Alternative Source Term Regulatory Guide I. 183 Compliance Matrix

2. If no or minimal2fuel damage is postulated for the limiting event, Complies the activity released should be the maximum coolant activity allowed by the technical specifications. Two cases of iodine spiking Two cases of iodine spiking are assumed (LAR 241, Enclosure 3, should be assumed. Section 6.4).

Footnote 2 - The activity assumed in the analysis should be based on the activity associated with the projected fuel damage or the maximum technical specification values, whichever maximizes the radiological consequences. In determining dose equivalent 1-131 (DE1131), only the radioiodine associated with normal operations or iodine spikes should be included. Activity from projected fuel damage should not be included.

2.1 A reactor transient has occurred prior to the postulated main Complies steam line break (MSLB) and has raised the primary coolant iodine concentration to the maximum value (typically 60 yCiIgm DE 1-131) The pre-accident iodine spike case assumed that a reactor permitted by the technical specifications (i.e., a preaccident iodine transient has occurred prior to the MSLB and has raised the RCS spike case). iodine concentration to a conservative value of 60 pCiIgm of DE 1-131 (The TS 3.4.16 limit for a transient is 50 pCiIgm of DE 1-131) (LAR 241, Enclosure 3, Section 6.4 and Table 21).

2.2 The primary system transient associated with the MSLB Complies causes an iodine spike in the primary system. The increase in primary coolant iodine concentration is estimated using a spiking The accident-initiated iodine spike case, assumes that the primary model that assumes that the iodine release rate from the fuel rods system transient associated with the MSLB causes an iodine spike to the primary coolant (expressed in curies per unit time) increases in the RCS which increases the iodine release rate from the fuel to to a value 500 times greater than the release rate corresponding to the RCS to a value 500 times the appearance rate corresponding the iodine concentration at the equilibrium value (typically to a maximum equilibrium RCS concentration of 0.5 pCiIgm of DE 1.0 y Cilgm DE 1-131) specified in technical specifications 1-131 (proposed TS 3.4.16). The spike is allowed to continue until (i.e., concurrent iodine spike case). A concurrent iodine spike need 4 hours4.62963e-5 days 0.00111 hours 6.613757e-6 weeks 1.522e-6 months from the start of the event. After this point in the accident not be considered if fuel damage is postulated. The assumed there is no activity available for release from the gap (LAR 241, iodine spike duration should be 8 hours9.259259e-5 days 0.00222 hours 1.322751e-5 weeks 3.044e-6 months . Shorter spike durations Enclosure 3, Section 3.4, Section 6.4 and Table 21).

may be considered on a case-by-case basis if it can be shown that the activity released by the 8- hour spike exceeds that available for release from the fuel gap of all fuel pins.

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Point Beach Nuclear Plant, Units Iand 2 License Amendment Request 241 Dockets 50-266 and 50-301 Alternative Source Term Regulatory Guide 1.I 83 Compliance Matrix

3. The activity released from the fuel should be assumed to be Complies released instantaneously and homogeneously through the primary coolant. Accident-initiated iodine spike activity is instantaneously and homogeneously.mixedin the primary coolant. This assumption is inherent in the RADTRAD iodine spike model.

4. The chemical form of radioiodine released from the fuel should Complies be assumed to be 95% cesium iodide (Csl), 4.85 percent elemental iodine, and 0.15 percent organic iodide. Iodine releases from the Iodine releases from the steam generators to the environment are steam generators to the environment should be assumed to be assumed to be 97% elemental and 3% organic (LAR 241, 97% elemental and 3% organic. These fractions apply to iodine Enclosure 3, Table 21).

released as a result of fuel damage and to iodine released during normal operations, including iodine spiking.

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Point Beach Nuclear Plant, Units 1 and 2 License Amendment Request 241 Dockets 50-266 and 50-301 Alternative Source Term Regulatory Guide 1.183 Compliance Matrix TRANSPORT^

5. Assumptions acceptable to the NRC staff related to the transport, reduction, and release of radioactive material to the environment are as follows.

Footnote 3 - In 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:

PC = mass of I per unit mass liquid / mass of I per unit mass gas

5. I For facilities that have not implemented alternative repair Complies criteria (see Ref. E-I, DG-1074), the primary-to-secondary leak rate in the steam generators should be assumed to be the leak rate Note that the primary-to-secondary leak rate is specified per SG, limiting condition for operation specified in the technical rather than per plant. Thus, leakage is apportioned according to specifications. For facilities with traditional generator specifications the per SG leakage limit (LAR 241, Enclosure 3, Section 2.2).

(both per generator and total of all generators), the leakage should be apportioned between affected and unaffected steam generators in such a manner that the calculated dose is maximized.

5.2 The density used in converting volumetric leak rates Complies (e.g., gpm) to mass leak rates (e.g.,lbm/hr) should be consistent with the basis of the parameter being converted. The ARC leak rate Appropriate density is used to insure that the accident-induced leak correlations are generally based on the collection of cooled liquid. rate is greater than the operational leak rate (LAR 241, Surveillance tests and facility instrumentation used to show Enclosure 3, Section 2.2).

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/ft3).

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Point Beach Nuclear Plant, Units 1 and 2 License Amendment Request 241 Dockets 50-266 and 50-301 Alternative Source Term Regulatory Guide 1.I 83 Compliance Matrix 5.3 The primary-to-secondary leakage should be assumed to Complies continue until the primary system pressure is less than the secondary system pressure, or until the temperature of the leakage The primary-to-secondary leak rate to the affected SG is assumed is less than 100°C (212°F). The release of radioactivity from to continue until the temperature of the leakage is less than 212"F, unaffected steam generators should be assumed to continue until which is assumed at 60 hours6.944444e-4 days 0.0167 hours 9.920635e-5 weeks 2.283e-5 months . The release of radioactivity from the shutdown cooling is in operation and releases from the steam unaffected SG continues for 30 hours3.472222e-4 days 0.00833 hours 4.960317e-5 weeks 1.1415e-5 months (time to place RHR in generators have been terminated. operation) (LAR 241, Enclosure 3, Table 21).

5.4 All noble gas radionuclides released from the primary system Complies are assumed to be released to the environment without reduction or mitigation. All noble gas activity carried over to the secondary side through SG tube leakage is assumed to be immediately released to the outside atmosphere (LAR 241, Enclosure 3, Section 6.4).

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Point Beach Nuclear Plant, Units 1 and 2 License Amendment Request 241 Dockets 50-266 and 50-301 Alternative Source Term Regulatory Guide 1.183 Compliance Matrix 5.5 The transport model described in this section should be utilized for iodine and particulate releases from the steam generators. This model is shown in Figure E-I and summarized below.

5.5.1 A portion of the primary-to-secondary leakage will flash to Complies vapor, based on the thermodynamic conditions in the reactor and secondary coolant. The affected SG is assumed to boil dry (iodine partition = 1.O) (LAR 241, Enclosure 3, Table 21).

During periods of steam generator dryout, all of the primary-to-secondary leakage is assumed to flash to vapor and be released to Primary-to-secondary leakage to the unaffected SG is assumed to the environment with no mitigation. mix with the secondary water without flashing. The partition factor in the unaffected SG = 0.01 (LAR 241, Enclosure 3, With regard to the unaffected steam generators used for plant Table 21).

cooldown, the primary-to-secondary leakage can be assumed to mix with the secondary water without flashing during periods of total tube submergence.

5.5.2 The leakage that immediately flashes to vapor will rise Complies through the bulk water of the steam generator and enter the steam space. Credit may be taken for scrubbing in the generator, using Flashing leakage is not assumed to the unaffected SG. The the models in NUREG-0409, "Iodine Behavior in a PWR Cooling affected SG is assumed to boil dry. Thus, no credit is taken for System Following a Postulated Steam Generator Tube Rupture scrubbing of primary activity from flashed rupture flow (LAR 241, Accident" (Ref. E-2), during periods of total submergence of the Enclosure 3, Section 6.4). Also, LAR 241, Enclosure 3, Table 21 tubes. does not specify flashing leakage or scrubbing credit.

5.5.3 The leakage that does not immediately flash is assumed to Complies mix with the bulk water.

Leakage to the unaffected SG is assumed to mix with the bulk SG water (LAR 241, Enclosure 3, Section 6.4).

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Point Beach Nuclear Plant, Units 1 and 2 License Amendment Request 241 Dockets 50-266 and 50-301 Alternative Source Term Regulatory Guide 1.183 Compliance Matrix 5.5.4 The radioactivity in the bulk water is assumed to become Complies vapor at a rate that is the function of the steaming rate and the partition coefficient. A partition coefficient for iodine of 100 may be Analysis assumes a partition factor of 0.01. See response to assumed. The retention of particulate radionuclides in the steam Regulatory Position 5.5.1 of this appendix. A particulate retention generators is limited by the moisture carryover from the steam factor of 0.0025 is applied (LAR 241, Enclosure 3, Table 21).

generators.

5.6 Operating experience and analyses have shown that for some Complies steam generator designs, tube uncovery may occur for a short period following any reactor trip (Ref. E-3). The potential impact of Steam generator tube bundle uncovery is not predicted or tube uncovery on the transport model parameters (e.g., flash postulated (unaffected SG discussion for SGTR also applies to fraction, scrubbing credit) needs to be considered. The impact of MSLB) (LAR 241, Enclosure 3, Section 6.2).

emergency operating procedure restoration strategies on steam generator water levels should be evaluated. Water level in the unaffected SG was specifically analyzed for the SGTR.

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Point Beach Nuclear Plant, Units 1 and 2 License Amendment Request 241 Dockets 50-266 and 50-301 Alternative Source Term Regulatory Guide 1.183 Compliance Matrix APPENDIX F ASSUMPTIONS FOR EVALUATING THE RADIOLOGICAL CONSEQUENCES OF A PWR STEAM GENERATOR TUBE RUPTURE ACCIDENT This appendix provides assumptions acceptable to the NRC staff for evaluating the radiological consequences of a steam generator tube rupture accident at PWR light-water reactors. These assumptions supplement the guidance provided in the main body of this guide.1 SOURCE TERM

1. Assumptions acceptable to the NRC staff regarding core Complies inventory and the release of radionuclides from the fuel are in Regulatory Position 3 of this guide. The release from the breached No fuel damage is postulated to occur for the SGTR event. The fuel is based on Regulatory Position 3.2 of this guide and the Source Term specifies only initial reactor coolant activity (LAR 241, estimate of the number of fuel rods breached. Enclosure 3, Table 19).

2. if no or minimal2fuel damage is postulated for the limiting event, Complies the activity released should be the maximum coolant activity allowed by technical specification. Two cases of iodine spiking No fuel damage is postulated to occur for the SGTR event. Two should be assumed. cases of iodine spiking are assumed (LAR 241, Table 19).

Footnote 2 - The activity assumed in the analysis should be based on the activity associated with the projected fuel damage or the maximum technical specification values, whichever maximizes the radiological consequences. In determining dose equivalent 1-131 (DE1131), only the radioiodine associated with normal operations or iodine spikes should be included. Activity from projected fuel damage should not be included.

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Point Beach Nuclear Plant, Units 1 and 2 License Amendment Request 241 Dockets 50-266 and 50-301 Alternative Source Term Regulatory Guide 1.I 83 Compliance Matrix 2.1 A reactor transient has occurred prior to the postulated steam Complies generator tube rupture (SGTR) and has raised the primary coolant iodine concentration to the maximum value (typically 60 p Cilgm Case assumes a reactor transient prior to the postulated SGTR that DE 1-131) permitted by the technical specifications (i.e., a raises the RCS iodine concentration to a conservative value of preaccident iodine spike case). 60 pCi/gm of dose equivalent (DE) 1-131. (The Technical Specification (TS 3.4.1 6) limit for a transient is 50 pCi1gm of dose equivalent (DE) 1-131.) This is the pre-accident spike case (LAR 241, Enclosure 3, Table 19).

2.2 The primary system transient associated with the SGTR Complies causes an iodine spike in the primary system. The increase in primary coolant iodine concentration is estimated using a spiking The accident-initiated iodine spike case, assumes that the primary model that assumes that the iodine release rate from the fuel rods system transient associated with the MSLB causes an iodine spike to the primary coolant (expressed in curies per unit time) increases in the RCS which increases the iodine release rate from the fuel to to a value 335 times greater than the release rate corresponding to the RCS to a value 335 times the appearance rate corresponding the iodine concentration at the equilibrium value (typically to a maximum equilibrium RCS concentration of 0.5 pCilgm of 1.0 p Cilgm DE 1-131) specified in technical specifications DE 1-131 (proposed TS 3.4.16). The spike is allowed to continue (i.e., concurrent iodine spike case). A concurrent iodine spike need until 8 hours9.259259e-5 days 0.00222 hours 1.322751e-5 weeks 3.044e-6 months from the start of the event (LAR 241, Enclosure 3, not be considered if fuel damage is postulated. The assumed Table 19).

iodine spike duration should be 8 hours9.259259e-5 days 0.00222 hours 1.322751e-5 weeks 3.044e-6 months . Shorter spike durations may be considered on a case-by-case basis if it can be shown that the activity released by the 8- hour spike exceeds that available for release from the fuel gap of all fuel pins.

3. The activity released from the fuel, if any, should be assumed to Complies be released instantaneously and homogeneously through the primary coolant. The activity released from the fuel is assumed to be released instantaneously and homogeneously through the primary coolant.

This assumption is inherent in the RADTRAD model.

4. Iodine releases from the steam generators to the environment Complies should be assumed to be 97% elemental and 3% organic.

Iodine releases from the steam generators to the environment are assumed to be 97% elemental and 3% organic (LAR 241, Enclosure 3, Table 19).

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Point Beach Nuclear Plant, Units 1 and 2 License Amendment Request 241 Dockets 50-266 and 50-301 Alternative Source Term Regulatory Guide 1.183 Compliance Matrix TRANSPORT^

Footnote 3 - In 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. Assumptions acceptable to the NRC staff related to the transport, reduction, and release of radioactive material to the environment are as follows:

5.1 The primary-to-secondary leak rate in the steam generators Complies should be assumed to be the leak rate limiting condition for operation specified in the technical specifications. The leakage The primary-to-secondary leak rate is specified per SG, rather than should be apportioned between affected and unaffected steam per plant. Thus, leakage is apportioned according to the per SG generators in such a manner that the calculated dose is maximized. leakage limit (LAR 241, Enclosure 3, Section 2.2).

5.2 The density used in converting volumetric leak rates (e.g., Complies gpm) to mass leak rates (e.g., Ibmlhr) should be consistent with the basis of surveillance tests used to show compliance with leak rate The analysis leak rate is significantly greater than the operational technical specifications. These tests are typically based on cool leak rate (LAR 241, Enclosure 3, Section 2.2).

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 gmlcc (62.4 Ibmlft3).

5.3 The primary-to-secondary leakage should be assumed to continue until the primary system pressure is less than the secondary system pressure, or until the temperature of the leakage Complies The release of radioactivity from the affected SG is assumed to I

is less than 100" C (212" F). The release of radioactivity from the continue for 30 minutes. The release of radioactivity from the unaffected steam generators should be assumed to continue until unaffected SG is assumed to continue until shutdown cooling is in shutdown cooling is in operation and releases from the steam operation and the steam release from the SGs is terminated, generators have been terminated. 30 hours3.472222e-4 days 0.00833 hours 4.960317e-5 weeks 1.1415e-5 months into the event (LAR 241, Enclosure 3, Table 19).

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Point Beach Nuclear Plant, Units 1 and 2 License Amendment Request 241 Dockets 50-266 and 50-301 Alternative Source Term Regulatory Guide 1.183 Compliance Matrix 5.4 The release of fission products from the secondary system Complies should be evaluated with the assumption of a coincident loss of offsite power. The release of fission products from the secondary system is evaluated with the assumption of a coincident loss of offsite power (LOOP) (LAR 241, Enclosure 3, Table 19).

5.5 Ail noble gas radionuclides released from the primary system Complies are assumed to be released to the environment without reduction or mitigation. All noble gas activity carried over to the secondary side through SG tube leakage is assumed to be immediately released to the outside atmosphere (LAR 241, Enclosure 3, Section 6.2).

5.6 The transport model described in Regulatory Positions 5.5 and 5.6 of Appendix E should be utilized for iodine and particulates.

Appendix E, Regulatory Position 5.5.1: Complies A portion of the primary-to-secondary leakage will flash to vapor, Pre-trip and post-trip flashing fractions were calculated (LAR 241, based on the thermodynamic conditions in the reactor and Enclosure 3, Section 6.2).

secondary coolant.

Flashing leakage is not assumed to the unaffected SG (LAR 241, o During periods of steam generator dryout, all of the primary- Enclosure 3, Table 19). Table 19 does not cite flashing leakage to to-secondary leakage is assumed to flash to vapor and be the unaffected SG.

released to the environment with no mitigation.

All analyses with unaffected SGs use the same unaffected SG o With regard to the unaffected steam generators used for transport model.

plant cooldown, the primary-to-secondary leakage can be assumed to mix with the secondary water without flashing during periods of total tube submergence.

Sufficient water level is maintained in both SGs (LAR 241, Enclosure 3, Section 6.2).

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Point Beach Nuclear Plant, Units 1 and 2 License Amendment Request 241 Dockets 50-266 and 50-301 Alternative Source Term Regulatory Guide 1. I 83 Compliance Matrix Appendix El Regulatory Position 5.5.2: Complies The leakage that immediately flashes to vapor will rise through the Flashing rupture flow is assumed to the affected SG. No credit is bulk water of the steam generator and enter the steam space. taken for scrubbing of primary activity from flashed rupture flow Credit may be taken for scrubbing in the generator, using the (LAR 241, Enclosure 3, Section 6.2).

models in NUREG-0409, "Iodine Behavior in a PWR Cooling System Following a Postulated Steam Generator Tube Rupture Accident" (Ref. E-2), during periods of total submergence of the tubes.

Appendix El Regulatory Position 5.5.3: Complies The leakage that does not immediately flash is assumed to mix with The fraction of primary coolant iodine that is not assumed to the bulk water. become airborne immediately, mixes with the secondary water, and is assumed to become airborne at a rate proportional to the steaming rate (LAR 241, Enclosure 3, Section 6.2).

Appendix El Regulatory Position 5.5.4: Complies The radioactivity in the bulk water is assumed to become vapor at a Analysis assumes a partition factor of 0.01. A particulate retention rate that is the function of the steaming rate and the partition factor of 0.0025 is applied (LAR 241, Enclosure 3, Table 19).

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

Appendix El Regulatory Position 5.6: Complies Operating experience and analyses have shown that for some Steam generator tube bundle uncovery is not predicted or steam generator designs, tube uncovery may occur for a short postulated (LAR 241, Enclosure 3, Section 6.2).

period following any reactor trip (Ref. E-3). The potential impact of tube uncovery on the transport model parameters (e.g., flash Water level in the unaffected SG was specifically analyzed for the fraction, scrubbing credit) needs to be considered. The impact of SGTR.

emergency operating procedure restoration strategies on steam generator water levels should be evaluated.

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Point Beach Nuclear Plant, Units 1 and 2 License Amendment Request 241 Dockets 50-266 and 50-301 Alternative Source Term Regulatory Guide 1.I 83 Compliance Matrix APPENDIX G ASSUMPTIONS FOR EVALUATING THE RADlOLOGlCAL CONSEQUENCES OF A PWR LOCKED ROTOR ACCIDENT This appendix provides assumptions acceptable to the NRC staff for evaluating the radiological consequences of a locked rotor accident at PWR light water reactors.1 These assumptions supplement the guidance provided in the main body of this guide.

I SOURCE TERM

1. Assumptions acceptable to the NRC staff regarding core I Complies inventory and the release of radionuclides from the fuel are in Regulatory Position 3 of this regulatory guide. The release from the Analysis assumed 30% of the fuel rods in core are breached breached fuel is based on Regulatory Position 3.2 of this guide and (LAR 241, Enclosure 3, Table 20).

the estimate of the number of fuel rods breached.

2. If no fuel damage is postulated for the limiting event, a Complies radiological analysis is not required as the consequences of this event are bounded by the consequences projected for the main See Regulatory Position 1 above.

steam line break outside containment.

3. The activity released from the fuel should be assumed to be Complies released instantaneously and homogeneously through the primary coolant. The activity released from the fuel is assumed to be released instantaneously and homogeneously through the primary coolant.

Instantaneous, homogeneous mixing is inherent in the RADTRAD model.

4. The chemical form of radioiodine released from the fuel should Complies be assumed to be 95% cesium iodide (Csl), 4.85 percent elemental iodine, and 0.15 percent organic iodide. Iodine releases from the Iodine releases from the steam generators to the environment are steam generators to the environment should be assumed to be assumed to be 97% elemental and 3% organic (LAR 241, 97% elemental and 3% organic. These fractions apply to iodine Enclosure 3, Table 20).

released as a result of fuel damage and to iodine released during normal operations, including iodine spiking.

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Point Beach Nuclear Plant, Units 1 and 2 License Amendment Request 241 Dockets 50-266 and 50-301 Alternative Source Term Regulatory Guide 1.183 Compliance Matrix RELEASE TRANSPORT

5. Assumptions acceptable to the NRC staff related to the transport, reduction, and release of radioactive material to the environment are as follows.

5.1 The primary-to-secondary leak rate in the steam generators Complies should be assumed to be the leak-rate-limiting condition for operation specified in the technical specifications. The leakage The primary-to-secondary leak rate is specified per SG, rather than should be apportioned between the steam generators in such a per plant. Thus, leakage is apportioned according to the per SG manner that the calculated dose is maximized. leakage limit (LAR 241, Enclosure 3, Section 2.2).

5.2 The density used in converting volumetric leak rates (e.g., Complies gpm) to mass leak rates (e.g., Ibmlhr) should be consistent with the basis of surveillance tests used to show compliance with leak rate Appropriate density is used to insure that the accident-induced leak technical specifications. These tests are typically based on cool rate is greater than the operational leak rate (LAR 241, liquid. Facility instrumentation used to determine leakage is Enclosure 3, Section 2.2).

typically located on lines containing cool liquids. In most cases, the density should be assumed to be 1.0 gmlcc (62.4 Ibmlft3).

5.3 The primary-to-secondary leakage should be assumed to Complies continue until the primary system pressure is less than the secondary system pressure, or until the temperature of the leakage The primary-to-secondary leakage is assumed to continue until is less than 100" C (212" F). The release of radioactivity should be shutdown cooling is in operation and the steam release from the assumed to continue until shutdown cooling is in operation and SGs is terminated (30 hours3.472222e-4 days 0.00833 hours 4.960317e-5 weeks 1.1415e-5 months into the event) (LAR 241, Enclosure 3, releases from the steam generators have been terminated. Section 6.3).

5.4 The release of fission products from the secondary system Complies should be evaluated with the assumption of a coincident loss of offsite power. The release of fission products from the secondary system is evaluated with the assumption of a coincident loss of offsite power (LOOP) (LAR 241, Enclosure 3, Section 1.4).

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Point Beach Nuclear Plant, Units 1 and 2 License Amendment Request 241 Dockets 50-266 and 50-301 Alternative Source Term Regulatory Guide 1.I 83 Compliance Matrix 5.5 All noble gas radionuclides released from the primary system Complies are assumed to be released to the environment without reduction or mitigation. All noble gas activity carried over to the secondary side through SG tube leakage is assumed to be immediately released to the outside atmosphere (LAR 241, Enclosure 3, Section 6.3).

5.6 The transport model described in assumptions 5.5 and 5.6 of Complies Appendix E should be utilized for iodine and particulates.

The SG activity transport and release model for the LR is consistent with the unaffected SG assumptions of the MSLB. See Appendix F of this document.

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Point Beach Nuclear Plant, Units 1 and 2 License Amendment Request 241 Dockets 50-266 and 50-301 Alternative Source Term Regulatory Guide 1.I 83 Compliance Matrix APPENDIX H I

ASSUMPTIONS FOR EVALUATING THE RADIOLOGICAL CONSEQUENCES OF A PWR ROD EJECTION ACCIDENT This appendix provides assumptions acceptable to the NRC staff for evaluating the radiological consequences of a rod ejection accident at PWR light water reactors'. These assumptions supplement the guidance provided in the main body of this guide.

Footnote 1 - Facilities licensed with, or applying for, alternative repair criteria (ARC) should use this section in conjunction with the guidance that is being developed in Draft Regulatory Guide DG-I 074, "Steam Generator Tube Integrityi1(USNRC, December 1998), for acceptable assumptions and methodologies for performing radiological analyses.

I SOURCE TERM

1. Assumptions acceptable to the NRC staff regarding core Complies inventory are in Regulatory Position 3 of this guide. For the rod ejection accident, the release from the breached fuel is based on The fraction of activity released from melted fuel to containment is the estimate of the number of fuel rods breached and the 50% for iodine. This is conservative with respect to RG 1.183, assumption that 10% of the core inventory of the noble gases and Appendix H, Position 1 which states that for the containment iodines is in the fuel gap. The release attributed to fuel melting is leakage release path model, the available inventory from the based on the fraction of the fuel that reaches or exceeds the melted fuel is 25% for iodine (LAR 241, Enclosure 3, Section 6.5).

initiation temperature for fuel melting and the assumption that 100% of the noble gases and 25% of the iodines contained in that fraction are available for release from containment. For the secondary system release pathway, 100% of the noble gases and 50% of the iodines in that fraction are released to the reactor coolant.

2. If no fuel damage is postulated for the limiting event, a Complies radiological analysis is not required as the consequences of this event are bounded by the consequences projected for the loss-of- Fuel damage is postulated. See Position 1, above.

coolant accident (LOCA), main steam line break, and steam generator tube rupture.

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Point Beach Nuclear Plant, Units 1 and 2 License Amendment Request 241 Dockets 50-266 and 50-301 Alternative Source Term Regulatory Guide 1.I 83 Compliance Matrix

3. Two release cases are to be considered. In the first, 100% of Complies the activity released from the fuel should be assumed to be released instantaneously and homogeneously through the TWOrelease cases are considered (LAR 241, Enclosure 3, containment atmosphere. In the second, 100% of the activity Section 6.5).

released from the fuel should be assumed to be completely dissolved in the primary coolant and available for release to the The assumption of instantaneous and homogeneous mixing of secondary system. activity released from the fuel into the containment atmosphere and primary coolant is inherent in the RADTRAD model.

4. The chemical form of radioiodine released to the containment Complies atmosphere should be assumed to be 95% cesium iodide (Csl),

4.85% elemental iodine, and 0.15% organic iodide. If containment The chemical form of radioiodine released from the damaged fuel sprays do not actuate or are terminated prior to accumulating sump to the containment is assumed to be 95% cesium iodide (Csl),

water, or if the containment sump pH is not controlled at values of 7 4.85% elemental iodine, and 0.15% organic iodide (LAR 241, or greater, the iodine species should be evaluated on an individual Enclosure 3, Table 22).

case basis. Evaluations of pH should consider the effect of acids created during the rod ejection accident event, e.g., pyrolysis and Containment sump pH is controlled to a value of 7 or greater radiolysis products. With the exception of elemental and organic (LAR 241, Enclosure 3, Table 18).

iodine and noble gases, fission products should be assumed to be in particulate form.

5. Iodine releases from the steam generators to the environment Complies should be assumed to be 97% elemental and 3% organic.

The chemical form of radioiodine released from the SGs to the environment is assumed to be 97% elemental iodine, and 3%

organic iodide (LAR 41, Enclosure 3, Table 22).

Page 51 of 55

Point Beach Nuclear Plant, Units 1 and 2 License Amendment Request 241 Dockets 50-266 and 50-301 Alternative Source Term Regulatory Guide 1.183 Compliance Matrix TRANSPORT FROM CONTAINMENT

6. Assumptions acceptable to the NRC staff related to the transport, reduction, and release of radioactive material in and from the containment are as follows.

6.1 A reduction in the amount of radioactive material available for Complies leakage from the containment that is due to natural deposition, containment sprays, recirculating filter systems, dual containments, For the containment leakage case, sedimentation of alkali metal or other engineered safety features may be taken into account. particulates in containment is credited. Containment spray is not Refer to Appendix A to this guide for guidance on acceptable credited (LAR 241, Enclosure 3, Table 22).

methods and assumptions for evaluating these mechanisms.

6.2 The containment should be assumed to leak at the leak rate Complies incorporated in the technical specifications at peak accident pressure for the first 24 hours2.777778e-4 days 0.00667 hours 3.968254e-5 weeks 9.132e-6 months , and at 50% of this leak rate for the The containment is assumed to leak at the proposed TS maximum remaining duration of the accident. Peak accident pressure is the allowable rate of 0.2% for the first 24 hours2.777778e-4 days 0.00667 hours 3.968254e-5 weeks 9.132e-6 months and 0.1% for the maximum pressure defined in the technical specifications for remainder of the event (LAR 241, Enclosure 3, Table 22).

containment leak testing. Leakage from subatmospheric containments is assumed to be terminated when the containment is brought to a subatmospheric condition as defined in technical specifications.

Page 52 of 55

Point Beach Nuclear Plant, Units Iand 2 License Amendment Request 241 Dockets 50-266 and 50-301 Alternative Source Term Regulatory Guide 1.183 Compliance Matrix TRANSPORT FROM THE SECONDARY SYSTEM

7. Assum~tionsacce~tableto the NRC staff related to the transport, reduction, and release of radioactive material in and from the secondary system are as follows.

7.1 A leak rate equivalent to the primary-to-secondary leak rate Complies limiting condition for operation specified in the technical specifications should be assumed to exist until shutdown cooling is The assumed accident-induced leak rate is greater than the in operation and releases from the steam generators have been operational leak rate (LAR 241, Enclosure 3, Section 2.2).

terminated.

Does not comply Primary-to-secondary leakage is assumed to continue for 2000 seconds. This time is based on the time to equalize primary and secondary pressures, for a 2" LOCA (LAR 241, Enclosure 3, Section 6.5).

Steam releases are assumed to continue for 30 hours3.472222e-4 days 0.00833 hours 4.960317e-5 weeks 1.1415e-5 months (LAR 241, Enclosure 3, Table 22).

7.2 The density used in converting volumetric leak rates (e.g., Complies gpm) to mass leak rates (e.g., Ibmlhr) should be consistent with the basis of surveillance tests used to show compliance with leak rate Appropriate density is used to insure that the accident-induced leak technical specifications. These tests typically are based on cooled rate is greater than the operational leak rate (LAR 241, liquid. The facility's instrumentation used to determine leakage Enclosure 3, Section 2.2).

typically is located on lines containing cool liquids. In most cases, the density should be assumed to be 1.0 gmlcc (62.4 Ibmlft3).

gas radionuclides released to the secondary system are assumed to be released to the environment without reduction or mitigation.

Complies All of the noble gas released to the secondary side is assumed to I

be released directly to the environment (LAR 241, Enclosure 3, Section 6.5).

Page 53 of 55

Point Beach Nuclear Plant, Units 1 and 2 License Amendment Request 241 Dockets 50-266 and 50-301 Alternative Source Term Regulatory Guide 1.183 Compliance Matrix 7.4 The transport model described in assumptions 5.5 and 5.6 of Complies Appendix E should be utilized for iodine and particulates.

The SG activity transport and release model for the CRDE is consistent with the unaffected SG assumptions of the MSLB and SGTR. See Appendix F of this document.

Page 54 of 55

Point Beach Nuclear Plant, Units 1 and 2 License Amendment Request 241 Dockets 50-266 and 50-301 Alternative Source Term Regulatory Guide I .183 Compliance Matrix APPENDIX I ASSUMPTIONS FOR EVALUATING RADIATION DOSES FOR EQUIPMENT QUALIFICATION This appendix addresses assumptions associated with equipment NIA qualification that are acceptable to the NRC staff for performing radiological assessments. As stated in Regulatory Position 6 of this Radiation environmental qualification of Equipment analyses are guide, this appendix supersedes Regulatory Positions 2.c.(l) and not modified by this M R . PBNP will continue to use the CLB 2.c.(2) and Appendix D of Revision 1 of Regulatory Guide 1.89, qualification analyses, which are based on the TlD-14844 source "Environmental Qualification of Certain Electric Equipment term (LAR 241, Enclosure 1, Section 5.4).

Important to Safety for Nuclear Power Plants" (USNRC, June 1984), for operating reactors that have amended their licensing The NRC Staff concluded that there was no clear basis for basis to use an alternative source term. Except as stated in this backfitting the requirement to modify the design basis for appendix, other assumptions, methods, and provisions of Revision equipment qualification to adopt the AST.

1 of Regulatory Guide 1.89 remain effective.

Post-accident vital access shielding is addressed in LAR 241, Enclosure I , Section 5.6.

Page 55 of 55

NRC 2009-0026, Transmittal of Supplement to License Amendment Request 241 Regulatory Guide 1.183 Compliance Matrix

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