ML062830417

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Enclosure 4, NAI Report No. NAI-1149-027, Revision 1, AST Licensing Technical Report for Palisades.
ML062830417
Person / Time
Site: Palisades Entergy icon.png
Issue date: 09/15/2006
From: Guidotti T, Harrell J, Thomasson S
Numerical Applications
To:
Office of Nuclear Reactor Regulation
References
NAI-1149-027, Rev 1
Download: ML062830417 (129)


Text

ENCLOSURE 4 Numerical Applications, Inc.

NAI Report No. NAI-1 149-027 AST Licensing Technical Report for Palisades Revision 1 Dated: September 15, 2006

Page 1of 84 NMI Report Release Report Number~ NAT-i 149-027 Revision Number: I

Title:

AST Licensing Technical Report for Palisades

==

Description:==

This report documents the results of the analyses and evaluations performed by Numerical Applications, Inc. in support of the Palisades licensing project to implement alternative radiological source terms. Design basis accidents and radiological consequences are evaluated using the AST methodology to support control room habitability. The analyses and evaluations performed by NAT are based on the guidance of Regulatory Guide 1.183.

Revision 1of this report is issued to incorporate the latest revisions to the accident analysis calculations and to improve the clarity of the report based on feedback received from the NRC during a pre-submittal meeting.

R.L A 'te ,I - 16'- O

'uor(James R. Harrell) Date ktA' &.- f' C)/157/O Reviewer (Steven Thomasson) Date NAT Management (Tim Quidotti) Date

Numerical Applications, Inc. Page 2 of 84 AST Licensing Technical Report for Palisades Report Number: NA1-1 1149-027 Table of Contents 1.0 Radiological Consequences Utilizing the Alternative Source Term Methodology........................ 4 1.1. Introduction .............................................................................................. 4 1.2. Evaluation Overview and Objective................................................................... 4 1.3. Proposed Changes to the Palisades Licensing Basis................................................. 4 1.4. Compliance with Regulatory Guidelines ............................................................. 5 1.5. Computer Codes ......................................................................................... 5 1.6. Radiological Evaluation Miethodlology ................................................................ 6 1.6.1. Analysis Input Assumptions....................................................................... 6 1.6.2. Acceptance Criteria ................................................................................ 7 1.6.3. Control Room HVAC System Description ....................................................... 7 1.6.3.1. Control Room Dose Calculation Model .................................................... 8 1.6.4. Direct Shine Dose................................................................................. 10 1.7. Radiation Source Terms............................................................................... 11 1.7.1. Fission Product Inventory ........................................................................ 11 1.7.2. Primary Coolant Source Term ................................................................... 12 1.7.3. Secondary Side Coolant Source Term........................................................... 12 1.7.4. LOCA Source Term............................................................................... 12 1.7.5. Fuel Handling Accident Source Term........................................................... 13 1.7.6. Spent Fuel Cask Drop Source Terms ............................................................ 13 1.8. Atmospheric Dispersion (X/Q) Factors .............................................................. 14 1.8.1. Onsite XIQDetermination........................................................................ 14 1.8.2. Offsite X/Q Determination........................................................................ 15 1.8.3. Meteorological Data .............................................................................. 16 2.0 Radiological Consequences -Event Analyses.............................................................. 16 2.1. Loss of Coolant Accident (LOCA) ................................................................... 16 2.2. Fuel Handling Accident (FIIA)....................................................................... 24 2.3. Main Steamline Break (MISLB) ...................................................................... 27 2.4. Steam Generator Tube Rupture (SGTR) ........................................................... 30 2.5. Small Line Break Outside of Containment (SLBOC) ............................................. 34 2.6. Control Rod Ejection (CRE).......................................................................... 36 2.7. Spent Fuel Cask Drop ................................................................................. 39 2.8. Environmental Qualification (EQ)................................................................... 41 3.0 Summary of Results .......................................................................................... 41 4.0 Conclusion..................................................................................................... 41 5.0 References..................................................................................................... 41

Figures and Tables FIGURE 1.8.1-1 ONSITE RELEASE-RECEPTOR LOCATION SKETCHI................................................. 44 TABLE 1.6.3-1 CONTROL Room VENTILATION SYSTEM PARAMETERS ........................................... 45 TABLE 1.6.3-2 LOCA DIRECT SHINE DOSE .......................................................................... 46 TABLE 1.7.2-1 PRIMARY COOLANT SOURCE TERM ................................................................. 47 TABLE 1.7.3-1 SECONDARY SIDE SOURCE TERM .................................................................... 49 TABLE 1.7.4-1 LOCA SOURCE TERM ................................................................................. 50 TABLE 1.7.5-1 FUEL HANDLING ACCIDENT SOURCE TERM ........................................................ 52 TABLE 1.7.6-1 SPENT FUEL CASK DROP SOURCE TERMS*.......................................................... 53 TABLE 1.8.1-1 RELEASE-RECEPTOR COMBINATION PARAMETERS FOR ANALYSIS EVENTS.................... 56 TABLE 1.8.1-2 ONSITE ATMOSPHERIC DISPERSION (X7Q) FACTORS FOR ANALYSIS EVENTS.................. 58 TABLE 1.8.1-3 RELEASE-RECEPTOR POINT PAIRS ASSUMED FOR ANALYSIS EVENTS.......................... 59 TABLE 1.8.2-1 OFFSITE ATMOSPHERIC DISPERSION (AYOQ) FACTORS FOR ANALYSIS EVENTS................. 60 TABLE 2. 1-1 Loss OF COOLANT ACCIDENT (LOCA) - INPUTS AND ASSUMPTIONS ............................ 61 TABLE 2.1-2 LOCARELEASE PHASES................................................................................. 64 TAB3LE 2.1-3 TIME DEPENDENT SIRWT PH.................................................. *64 TABLE 2.1-4 TIME DEPENDENT SIRWT TOTAL IODINE CONCENTRATION ...................................... 65 TABLE 2.1-5 TIME DEPENDENT SIRWT LIQUID TEMPERATURE................................................... 66 TABLE 2.1-6 TIME DEPENDENT SIRWT ELEMENTAL IODINE FRACTION..........................................67 TABLE 2.1-7 TIME DEPENDENT SIRWT PARTITION COEFFICIENT................................................. 68 TABLE 2.1-8 ADJUSTED RELEASE RATE FROM SIRWT......................................................... 69 TABLE 2.1-9 LOCA DOSE CONSEQUENCES ........................................................................... 69 TABLE 2.2-1 FUEL HANDLING ACCIDENT (FHA) - INPUTS AND ASSUMPTIONS ................................. 70 TABLE 2.2-2 FUEL HANDLING ACCIDENT DOSE CONSEQUENCES ................................................. 70 TABLE 2.3-1 MAIN STEAM LINE BREAK (MSLB) - INPUTS AND ASSUMPTIONS................................. 71 TABLE 2.3-2 INTACT SG STEAM RELEASE RATE .................................................................... 72 TABLE 2.3-3 MSLB DOSE CONSEQUENCES ........................................................................... 72 TABLE 2.4-1 STEAM GENERATOR TUBE RUPTURE (SGTR) - INPUTS AND ASSUMPTIONS ..................... 73 TABLE 2.4-2 SGTR INTEGRATED MASS RELEASES (1 ...................................................................... 74 TABLE 2.4-3 SGTR FLASHING FRACTION FOR FLOW FROM BROKEN TUBE ..................................... 74 TABLE 2.4-4 40 .ICiIGM D.E. 1-131 ACTIVITIES ..................................................................... 75 TABLE 2.4-5 IODINE EQUILIBRIUM APPEARANCE ASSUMPTIONS .................................................. 76 TABLE 2.4-6 CONCURRENT (335 X) IODINE SPIKE APPEARANCE RATE........................................... 76 TABLE 2.4-7 AFFECTED STEAM GENERATOR WATER LEVEL AND DECONTAMINATION FACTORS FOR FLASHED FLOW ...................................................................................................... 77 TABLE 2.4-8 SGTR DOSE CONSEQUENCES........................................................................... 77 TABLE 2.5-I SMALL LINE BREAK OUTSIDE OF CONTAINMENT - INPUTS AND ASSUMPTIONS................. 78 TABLE 2.5-2 CONCURRENT (500 X) IODINE SPIKE APPEARANCE RATE........................................... 79 TABLE 2.5-3 SMALL LINE BREAK OUTSIDE OF CONTAINMENT DOSE CONSEQUENCES ........................ 79 TABLE 2.6-1 CONTROL ROD EJECTION - INPUTS AND ASSUMPTIONS............................................. 80 TABLE 2.6-2 CONTROL ROD EJECTION STEAM RELEASE ........................................................... 81 TABLE 2.6-3 CONTROL ROD EJECTION DOSE CONSEQUENCES .................................................... 81 TABLE 2.7-1 SPENT FUEL CASK DROP- INPUTS AND ASSUMPTIONS.............................................. 82 TABLE 2.7-2 SPENT FUEL CASK DROP DOSE CONSEQUENCES ..................................................... 83 TABLE 3- PALISADES

SUMMARY

OF ALTERNATIVE SOURCE TERM ANALYSIS RESULTS......................84

Numerical Applications, Inc. Page 4 of 84 AST Licensing Technical Report for Palisades Report Number: NI-I 1149-027 1.0 Radiological Consequences Utilizing the Alternative Source Term Methodology 1M. Introduction The current Palisades licensing basis for radiological consequences analyses of accidents discussed in Chapter 14 of the Final Safety Analysis Report (FSAR) is based on methodologies and assumptions that are primarily derived from Technical Information Document (TID)-14844 and other early guidance.

Regulatory Guide (RG) 1.183 provides guidance on application of Alternative Source Terms (AST) in revising the accident source terms used in design basis radiological consequences analyses, as allowved by IOCFR5O.67. Because of advances made in understanding the timing, magnitude, and chemical form of fission product releases from severe nuclear power plant accidents, IO0CFR5O.67 was issued to allow holders of operating licenses to voluntarily revise the traditional accident source term used in the design basis accident (DBA) radiological consequence analyses with alternative source terms (ASTs).

1.2. Evaluation Overviews and Objective As documented in NEI 99-03 and Generic Letter 2003-01, several nuclear plants performed testing on control room unfiltered air inleakage that demonstrated leakage rates in excess of amounts assumed in the current accident analyses. While the Palisades tracer gas test demonstrated inleakage less than that assumed in the FSAR analyses, the AST methodology, established in RG 1.183 as supplemented by Regulatory Issue Summary 2006-04, is being used to calculate the offsite and control room radiological consequences for Palisades to support the control room habitability program by establishing a conforming set of radiological analyses.

The following limiting UFSAR Chapter 14 accidents are analyzed:

  • Loss-of-Coolant Accident (LOCA)
  • Small Line Break Outside Containment (SLBOC)
  • Fuel Handling Accident (Fl-A)
  • Spent Fuel Cask Drop Note that Sections 14.7.1.3 and 14.7.1.4 of the Palisades UFSAR state that the radiological consequences of the RCP Seized Rotor event are not analyzed since all applicable acceptance criteria are met. Therefore, the Locked Rotor Accident is not included in this report. Each accident listed above, along with the specific input and assumptions, are described in Section 2.0 of this report. These analyses provide for a bounding allowvable control room unfiltered air inleakage of 10 cfm. The use of 10 cfm as a design basis value will be established to be above the unfiltered inleakage value determined through modification, testing and analysis consistent wvith the resolution of issues identified in NEI 99-03 and Generic Letter 2003-01.

1.3. Proposed Changes to the Palisades Licensing Basis Nuclear Management Company, LLC (NMC) proposes to revise the Palisades licensing basis to implement the AST, described in RG 1.183, through reanalysis of the radiological consequences of the UFSAR Chapter 14 accidents listed in Section 1.2 above. As part of the full implementation of this AST, the following changes are assumed in the analysis:

Numerical Applications, Inc. Page 5 of 84 AST Licensing Technical Report for Palisades Report Number: NAI-1 1149-027 The total effective dose equivalent (TEDE) acceptance criterion of IOCFR50.67(b)(2) replaces the previous whole body and thyroid dose guidelines of I0CFRI 00.1 1.

  • New onsite (Control Room) and offisite atmospheric dispersion factors are developed, as required.
  • Dose conversion factors for inhalation and submersion are from Federal Guidance Reports (FGR)

Nos. II and 12 respectively.

Accordingly, the following changes to the Palisades Technical Specifications (TS) are proposed:

  • The definition of Dose Equivalent 1-131 in Section 1.1 is revised to reference Table 2.1 of Federal Guidance Report No. 11I(FGR 11), "Limiting Values of Radionuclide Intake and Air Concentration and Dose Conversion Factors for Inhalation, Submersion, and Ingestion," 1989, as the source of effective dose conversion factors.

1.4. Compliance wvith Regulatory Guidelines The revised Palisades accident analyses addressed in this report follow the guidance provided in RG 1.183.

Assumptions and methods utilized in this analysis for wvhich no specific guidance is provided in RG 1.183, but for which a regulatory precedent has been established, are as flollows:

" Selection of the Small Line Break Outside Containment dose consequences methodology and acceptance criteria are based on Standard Review Plan Smcion 15.6.2 and Regulatory Guide 1.183.

" Selection of the Spent Fuel Cask Drop dose consequences methodology and acceptance criteria are based on thle Fuel Handling Accident from Regulatory Guide 1.183.

" Use of the MicroShield code to develop direct shine doses to the Control Room. MicroShield is a point kernel integration code used for general-purpose gamma shielding analysis. It is qualified for this application and has been used to support licensing submittals that have been accepted by the NRC. Precedent for this use of MicroShield is established in the Duane Arnold Energy Center submittal dated October 19, 2000 and associated NRC Safety Evaluation dated July 31, 2001.

" Use of the QADMOD-GP code to develop direct shine d:ose to the Control Room from the SIRWT. QAD is recommended for determining shielded dose in Standard Review Plan Section 12.3.

1.5. Computer Codes The following computer codes are used in performing the Alternative Source Term analyses:

Computer Code Version Reference Purpose ARCON96 June 1997 5.15 Atmospheric Dispersion Factors MicroShield 5.05 5.16 D~irect Shine Dose Calculations QADMOD-GP November 5.41 Direct Shine Dose Calculations 1999 ORIGEN 2.1 5.17 Core Fission Product Inventory PA VAN 2.0 5.18 Atmospheric Dispersion Factors RADTRAD-NAI 1.1a(QA) 5.19 Radiological Dose Calculations

Numerical Applications, Inc. Page 6 of 84 AST Licensing Technical Report for Palisades Report Number: NAI-l 1149-027 1.5.1. ARCON96 - used to calculate relative concentrations (XIQ factors) in plumes from nuclear power plants at control room intakes in the vicinity of the release point using plant meteorological data.

1.5.2. MicroShield - used to analyze shielding and estimate exposure from gamma radiation.

1.5.3. QADMOD-GP D computer code used to analyze shielding and estimate exposure from gamma radiation.

1.5.4. ORIGEN - used for calculating the buildup, decay, and processing of radioactive materials.

1.5.5. PAVAN - provides relative air concentration (X/Q) values as functions of direction for various time periods at the EAB and LPZ boundaries assuming ground-level releases or elevated releases from freestanding stacks.

1.5.6. RADTRAD-NAI - estimates the radiological doses at offsite locations and in the control room of nuclear powver plants as consequences of postulated accidents. Tile code considers the timing, physical form (i.e., vapor or aerosol) and chemical species of the radioactive material released into the environment.

RADTRAD-NAI began with versions 3.01 and 3.02 of the NRC's RADTRAD computer code, originally developed by Sandia National Laboratory (SNL). The code is initially modified to compile on a UNIX system. Once compiled, an exten.3ive design reviewv/verif ication and validation process began on the code and documentation. The subject of the reviewv also included the source code for the solver, wvhich is made available in a separate distribution from the NRC. RADTRAD-NAI validation is performed with three different types of tests:

" Comparison of selected Acceptance Test Case results with Excel spreadsheet solutions and hand solutions,

" Separate effects tests, and

" Industry examples.

" The industry examples included prior AST submittals by BWRs and PWVRs, as wvell as other plant examples.

In addition to reviewing the code and incorporating error corrections, several software revisions wvere made. One revision involved the consideration of noble gases generated by decay of isotopes on filters that are returned to the dowvnstream compartment. Another revision involved the modification of the dose conversion and nuclide inventory files to account for 107 isotopes to assure that significant dose contributors were addressed. The dose conversion factors used by RADTRAD-NAI are from Federal Guidance Report Nos. II and 12 (FGR Il and FGR 12).

RADTRAD-NAI is developed and is maintained under Numerical Applications' IOCFR50 Appendix B program.

1.6. Radiological Evaluation Methodology 1.6.1. Analysis Input Assumptions Common analysis input assumptions include those for the control room ventilation system and dose calculation model (Section 1.6.3), direct shine dose (Section 1.6.4), radiation source terms (Section 1.7), and atmospheric dispersion factors (Section 1.8). Event-specific assumptions are discussed in the event analyses in Section 2.0.

Numerical Applications, Inc. Page 7 of 84 AST Licensing Technical Report for Palisades Report Number: NAI- 1149-027 1.6.2. Acceptance Criteria Offsite and Control Room doses must meet the guidelines of RG 1.183 and requirements of IOCFR5O.67. The acceptance criteria for specific postulated accidents are provided in Table 6 of RG 1.183. For analyzed events not addressed in RG 1.183, the basis used to establish the acceptance criteria for the radiological consequences is provided in the discussion of the event in Section 2.0. For Palisades, the events not specifically addressed in RG 1.183 are the Small Line Break Outside Containment and the Spent Fuel Cask Drop.

1.6.3. Control Room HVAC System Description The Control Room HVAC System is required to assure control room habitability. The design of the control room envelope and overall description of the Control Room HVAC System are discussed in the Palisades FSAR Section 9.8.2.

The Control Room Ventilation System consists of two air handling units and a ducted air intake and air distribution system. Outside air is drawn into the air handling units through roughing filters and cooled as required. Conditioned air is then directed back to the rooms through a supply air duct system.

Under emergency conditions, the Control Room HYAC System has the capability to go into thle recirculation mode. In the recirculation mode, fresh and recirculated air is processed through high efficiency particulate (HEPA) filters and charcoal filters to maintain the control room environment at acceptable conditions. The recirculation mode is automatically entered on receipt of a containment high radiation signal or a safety injection signal. Recirculation mode can also be entered manually by operator action. Redundant isolation dampers at the normal outside air intake and exhaust paths are automatically closed, so that the control room envelope is isolated except for filtered fresh air makeup.

The system is designed to perform its safety functions and maintain a habitable environment in the control room envelope during isolation.

In the normal mode, the control room envelope is slightly pressurized relative to the surroundings wvith outside air continuously introduced to the control room envelope. In the recirculation (emergency) mode, the control room is pressurized at a higher rate to maintain a positive pressure differential.

Makeup air for pressurization is filtered before entering the control room. The recirculated air flowv is filtered by the same filters as the makeup air. If offsite powver is lost, the unfiltered infiltration into the control room can occur.

The net volume of the control room envelope serviced by the Control Room HVAC System is approximately 76,451 cubic ft.

1.6.3.1. Control Room Dose Calculation Model The Control Room model includes a recirculation filter model along with filtered air intake, unfiltered air inleakage and an exhaust path. System performance, sequence, and timing of operational evolutions associated with the CR ventilation system are discussed below. Control Room ventilation system parameters assumed in the analyses are provided in Table 1.6.3-1. Thle dispersion factors for use in modeling the Control Room during each mode of operation are provided in Tables 1.8.1-2 and 1.8.1-3. Control Room occupancy factors and assumed breathing rates are those prescribed in RG 1.183. Figure 1.8.1 -1I provides a site sketch showing the Palisades plant layout, including the location of onsite potential radiological release points wvith respect to the control room air intakes. The elevations of release points and intakes used in the Control Room AST dose assessments are provided in Table 1.8.1-1.

The control room ventilation system contains a filtration system for removal of radioactive iodine and particulate material that may enter the CR during the course of the event. Calculation of thle dose to operators in the control room requires modeling, of various system configurations and operating evolutions of the control room ventilation system during the course of the accident. While in the normal mode prior to control room normal air intake and exhaust isolation, a single inlet to thle control room wvith an unfiltered flow rate of 660 cfm is modeled (384.2 cfmn if offsite power is lost). When in the emergency/recirculation mode, the control room model N-,ill define twvo concurrent air inlet paths representing the defined CR ventilation system air intake and thle unfiltered inleakage into the CR. In the emergency/recirculation mode, outside air can enter the control room through the filtration/ventilation system from the emergency ventilation intake location. Unfiltered outside air can also enter the CR directly from various sources. Modeling ol'the Control Room conservatively addresses these factors as they apply to the various release locations for each analyzed event. Details of the CR modeling for each event are described in subsequent event analyses sections.

All unfiltered inleakage is assumed to enter the control room envelope via the normal intakes.

Potential locations for control room envelope unfiltered inleakage include normal intake and purge exhaust isolation damper leakage, air handling unit drain leakage, and swvitchgear and cable spreading room emergency exhaust fan duct leakage. The potential for inleakage is due to the potential for portions of the control room envelope to be at a negative differential pressure with respect to outside air in these locations and conditions.

Equipment drains from the air handling units are connected to a common floor drain header which is routed to the normal waste system and then to the turbine building sump. Thle floor drains for the swvitchgear, cable spreading, diesel generator and battery rooms and turbine building are connected to this header. Loop seals in the air handling unit drain lines prevent the differential pressure from permitting inleakage. The loop seals are checked monthly and filled as needed to ensure the seals are operable and that no inleakage can occur.

Switchgear and cable spreading room emergency exhaust fan ducting passes through the technical support center (TSC), wvhich is part of the control room envelope, and penetrates thle TSC roof. This non-safety-related system exhausts air from the cable spreading room and I C & I D switchgear rooms.

When the emergency exhaust fan is operating, there is a potential for the exhaust duct to leak air into to the TSC. Also, wvith the exhaust fan operating the differential pressure across the control room/cable spreading room boundary wvould increase, wvhich would increase the leakage of air out of the control room envelope and potentially increase the amount of outside air required for pressurization.

Operation of the emergency exhaust fan is not permitted when the Control Room HVAC system is in emergency mode so that no unfiltered inleakage from this path can occur.

Therefore, the only credible location for control room envelope unfiltered inleakage is the isolation dampers. The normal intake isolation damper intake ducting is closer to all release sources than the

purge isolation damper exhaust ducting. For all events, the limiting train (i.e., highest) normal intake atmospheric relative concentrations are used to model the control room envelope unfiltered inleakage.

It is noted that two assumptions for the value of control room envelope unfiltered inleakage have been utilized. For the more limiting events where radiological dose margins are lower, a value of 10 cfm unfiltered inleakage is assumed. The more limiting events are the Maximum Hypothetical Accident/

Loss of Coolant Accident, Main Steam Line Break, and Control Rod Ejection. For less limiting events where radiological dose margins are greater, a value of 100 cfmf unfiltered inleakage is assumed. The less limiting events are the Steam Generator Tube Rupture, Small Line Break Outside Containment, Fuel Handling Accident and Spent Fuel Cask Drop.

For control room envelope unfiltered inleakage surveillance lesting (i.e., tracer gas testing), the lowver assumed unfiltered inleakage value of 10 cfm forms the basis for the acceptance criterion. The different inleakage assumptions are utilized to potentially limit the scope of any operability recommendation analyses in the event that control room integrity is ever called into question. For example (hypothetically), if control room envelope unfiltered inleakage is ever determined to be greater than 10 cfm- but less than 100 cfim, events that assume 100 cfm inleakage would remain bounding and demonstrate radiological limits wvithout re-analyses. The operability recommendation analyses could be limited to re-analysis of the events that assume 10 cf~m inleakage, reducing the scope of the operability assessment analyses and allowving for a more timely response to Operations staff.

For all events, delays in switching to the emergency/recirculation mode from the normal mode are conservatively considered wvith respect to the time required for signal processing, relay actuation, time required for the dampers to move and the system to re-align and diesel generator start time.

Numerical Applications, Inc. Page 10 of 84 AST Licensing Technical Report for Palisades Report Number: NAM- 149-027 1.6.4. Direct Shine Dose The total control room dose also requires the calculation of direct shine dose contributions from:

" the radioactive material on the control room filters,

  • the radioactive plume in the environment, and

" the activity in the primary containment atmosphere

" thle activity in the containment purge lines

" The activity in the SIRWT The contribution to the total dose to the operators from direct radiation sources such as the control room filters, the containment atmosphere, and the released radioactive plume wvere calculated for the LBLOCA event. The 30-day direct shine dose to a person in the control room, considering occupancy, is provided in Table 1.6.4-I. Note that shine doses assumed for other events conservatively bound the values presented in Table 1.6.4-1 for the LBLOCA event.

Direct shine dose is determined from five different sources to the control roomn operator after a postulated LOCA event. These sources are the containment, the control room make-uip and recirculating air filters, the external cloud that envelops the control room, the containment purge line, and the SIRWT. All other sources of direct shine dose are considered negligible. The MicroShield 5 code is used to determine direct shine exposure to a dose point located in the control room for all sources except the SIRWT. Each source required a different MicroShield case structure including different geometries, sources, and materials. The external cloud is assumed to have a length of 1000 meters in thle MicroShield cases to approximate an infinite cloud. A series of cases is run wvith each structure to determine an exposure rate from the radiological source at given points in time. These sources wvere taken from RADTRAD-NAI runs that output thle nuclide activity at a given point in time for the event. The RADTRAD-NAI output provides the time dependent results of the radioactivity retained in the control room filter components, as wvell as the activity inventory in thle environment and the containment.

The RADTRAD-NAI sources were then input into the MicroShield case file where they are either used as is, or 'decayed' (once the release has stopped) in MicroShield to yield the source activity at a later point in time. The exposure results from the series of cases for each source term wvere then corrected for occupancy using the occupancy factors specified in RG 1.183. The cumulative exposure and dose are subsequently calculated to yield the total 30-day direct shine dose from each source. The results of the Direct Shine Dose evaluation are presented in Table 1.6.4-1.

The QADMOD-GP code is used to determine direct shine exposure to a dose point located in the control room for the SIRWT source. QAD is recommended for shielded dose in Standard Reviewv Plan Section 12.3.

Numerical Applications, Inc. Page I11of 84 AST Licensing Technical Report for Palisades Report Number: NAM- 149-027 1.7. Radiation Source Terms 1.7.1. Fission Product Inventory The source term data to be used in performing alternative source term (AST) analyses for Palisades are summarized in the following tables:

Table 1.7.2-1 - Primary Coolant Source Term Table 1.7.3-I - Secondary Side Source Term (non-LOCA)

Table 1.7.4-1 - LOCA Source Term Table 1.7.5-1 - Fuel Handling Accident Source Term Table 1.7.6-I - Spent Fuel Cask Drop Source Terms The Palisades reactor core consists of 204 fuel assemblies. Thle full core isotopic inventory is determined in accordance wvith RO 1.183, Regulatory Position 3. 1, using the ORIGEN-2.1 isotope generation and depletion computer code (part of the SCALE-4.3 system of codes) to develop thle isotopics for the specified burnup, enrichment, and burnup rates (power levels). The plant-specific isotopic source terms are developed using a bounding approach.

Sensitivity studies wvere performed to assess the bounding fuel enrichment and bounding burnup values. The assembly source term is based on 102% of original design power (2703 MWth). For rod average burnups in excess of 54,000 MWD/MTU, the heat generation rate is limited to 6.3 kwv/ft in accordance with RG 1.183. For non-LOCA events with fuel failures, a bounding radial peaking factor of 2.04 is then applied to conservatively simulate the effect of power level differences across the core that might affect the localized fuel failures for assemblies containing thle peak fission product inventory.

The core inventory release fractions for the gap release and early in-vessel damage phases for the design basis LOCAs utilized those release fractions provided in RG 1.183, Regulatory Position 3.2, Table 2, "PWR Core Inventory Fraction Released into Containment." For non-LOCA events, the fractions of the core inventory assumed to be in the gap are consistent wvith RG 1.183, Regulatory Position 3.2, Table 3,"Non-LOCA Fraction of Fission Product Inventory in Gap." In some cases, thle gap fractions listed in Table 3 are modified as required by the event-specific source term requirements listed in the Appendices for RG 1.183.

The following assumptions are applied to the source term calculations:

I1. A conservative maximum fuel assembly uranium loading (440 kilograms) is assumed to apply to all 204 fuel assemblies in the core.

2. Radioactive decay of fission products during refueling outages is ignored in the source term calculation.
3. When adjusting the primary coolant isotopic concentrations to achieve Technical Specification limits, the relative concentrations of fission products in the primary coolant system are assumed to remain constant.

Conservatisms used in the calculation of fission product inventories include the followving. Use of ORIGEN 2.1 with revised data libraries for extended fuel burnup. Use of a core thermal power corresponding to original plant design power plus 2% calorimetric uncertainty. Use of bounding maximum assembly and peak rod burnups. Use of bounding core average equilibrium cycle maximum burnup. Use of a bounding range of average assembly enrichments. Use of a bounding maximum assembly uranium loading. Neglect of decay of fission products during refueling outages.

Numerical Applications, Inc. Page 12 of 84 AST Licensing Technical Report for Palisades Report Number: NA1-l 1149-027 1.7.2. Primary Coolant Source Term The primary coolant source term for Palisades is based on operation with small defects in the cladding of fuel rods generating 1 percent of the core rated power at maximum equilibrium for the fuel cycle.

Corrosion products are derived based on ANSI/ANS-1 8.1-1999.

The iodine activities are adjusted to achieve the Technical Specification 3.4.16 limit of 1.0 pCi/gm dose equivalent 1-131 using the proposed Technical Specification definition of Dose Equivalent 1-131 (DE 1-13 1) and dose conversion factors for individual isotopes from FGR 11. The non-iodine species are adjusted to achieve the Technical Specification limit of I00/E-bar for non-iodine activities.

The dose conversion factors for inhalation and submersion are from Federal Guidance Reports Nos. I1I and 12 respectively.

The final adjusted primary coolant source term is presented in Table 1.7.2-1 ,"Primary Coolant Source Term."

Conservatisms used in the calculation of the primary coolant source term include the followving. Use of conservative fission product inventories as noted above. Use of minimal purification flows. Use of appropriately conservative filter efficiencies. Use of corrosion product inventories derived from ANSI/ANS-I 8.1-1999. Normalization of dose equivalent iodine- 131 to the technical specification maximum of 1.0 pCi/gm using dose conversion factors from FGR-l 1. Normalization of non-iodine species to the technical specification maximum of IQOIE-bar.

1.7.3. Secondary Side Coolant Source Term Secondary coolant system activity is limited to a value of:5 0.10 ttCi/gm dose equivalent 1-131 in accordance with TS 3.7.17. Noble gases entering the secondary coolant system are assumed to be immediately released; thus the noble gas activity concentration in the secondary coolant system is assumed to be 0.0 tiCi/gm.

The secondary side source ternm is presented in Table 1.7.3-1, "Secondary Side Source Term."

Conservatisms used in the calculation of the secondary side source term include the followving. Use of conservative fission product inventories and primary coolant source term as noted above. Scaling to the technical specification maximum of 0.1 pCi/gm using dose conversion factors from FGR-1 1.

1.7.4. LOCA Source Term Per Section 3.1 of Reg. Guide 1.183, the inventory of f ission products in the Palisades reactor core and available for release to the containment is based on the maximum original design powver operation of the core (2703 MW~h wvhich includes 2% uncertainty) and the current licensed values for fuel enrichment, and fuel burnup. The period of irradiation is selected to be of sufficient duration to allowv the activity of dose-significant radionucl ides to reach equilibrium or to reach maximum values. In addition, for the DBA LOCA, all fuel assemblies in the core are assumed to be affected and the core average inventory is used.

During a LOCA, all of the fuel assemblies are assumed to fail; therefore, the source term is based on an "average" assembly with a core average burnup of 39,300 MWD/MTU and an average assembly power* of 13.25 M\Vlh. The fuel enrichment ranges from a minimum of 3.0 w/o to a maximum value of 5.0 wv/o. Itisconservatively assumed that a maximum assembly uranium mass of 440,000 gm applies to allof the fuel assemblies.

  • Average assembly power = (2703 MW,h)(1 / 204 assemblies) = 13.25 M\V 1h / assembly The ORIGEN runs used cross section libraries that correspond to PWR extended burnup fuel. Decay time between cycles is conservatively ignored. For each nuclide, the bounding activity for the allowable range of enrichments is determined.

The LOCA source term is presented in Table l.7.4-l,"LOCA Source Term."

Conservatisms used in the calculation of the LOCA source term include the followving. Use of conservative fission product inventories as noted above. Selection of the maximum activities from the range of enrichments considered.

1.7.5. Fuel Handling Accident Source Term The fuel handling accident for Palisades assumes the failure of one assembly; therefore, the fuel handling accident source term is based on a single "bounding" fuel assembly.

Per Section 3.1 of Reg. Guide 1.183, the source term methodology for the Fuel Handling Accident is similar to that used for developing the LOCA source term, except that for DRA events that do not involve the entire core, the fission product inventory of each of the damaged fuel rods is determined by dividing the total core inventory by the number of fuel rods in the core. To account for differences in power level across the core, a radial peaking factor of 2.04 is applied in determining the inventory of the damaged rods.

The LOCA source term is based on the activity of 204 fuel assemblies and the radial peaking factor is 2.04. Thus, based on the methodology specified in Reg. Guide 1.183, the fuel handling accident source term is derived by applying a factor of 2.04/204 to the LOCA source term and decaying for 48 hours5.555556e-4 days <br />0.0133 hours <br />7.936508e-5 weeks <br />1.8264e-5 months <br />. To ensure that the "bounding" assembly is identified, the activity of a peak burnup assembly (58,900 MWD/MTU), at 3.0 w/o, 4.0 w/o, and 5.0 w/o, is determined and compared to the source term derived from the LOCA data. For each nuclide, the bounding activity for the allowable range of enrichments and discharge exposure is determined.

The FHA source term is presented inTable 1.7.5-1, "Fuel Handling Accident Source Term."

Conservatisms used in the calculation of the fuel handling accident source term include the following.

Use of conservative fission product inventories as noted above. Selection of the maximum activities from the range of enrichments considered. Selection of maximum activity from the maximum core burnup and maximum assembly burnup cases. Use of the technical specification radial peaking factor limit.

1.7.6. Spent Fuel Cask Drop Source Terms Sections 14.11.3.1.1 of the FSAR describes three cask drop cases:

Case I "Acask drop onto 30 day decayed fuel with the Fuel Handling Building (FHB) Charcoal Filter operating wvith a conservative amount of unfiltered leakage. All "Isolable Unfiltered Leak Paths" are assumed to be isolated prior to event initiation. For scenario 1, charcoal filter bypass of 10%

is assumed to exist for the entire duration of the release."

Case 2 "A cask drop onto 30 day decayed fuel with the Fuel Handling Building (FHB) Charcoal Filter operating with a conservative amount of unfiltered leakage. This scenario wvill determine the

Numerical Applications, Inc. Page 14 of 84 AST Licensing Technical Report for Palisades Report Number: NAI-1 149-027 maximum amount of Non-Isolatable Unfiltered Leakage allowable in order to just meet the offsite dose limits. This scenario also assumes isolation of isolatable leak paths prior to event isolation.

For scenario 2, charcoal filter bypass of 17.5% is assumed to exist for the entire duration of the release."~

Case 3 "Acask drop onto 90 day decayed fuel without FHB Charcoal Filter operating. For scenario 3, charcoal filter bypass of 100% is assumed for the entire duration of the release since the charcoal filters are not operating."

The cask drop source term is determined in a manner similar to that for the FlHA. ORIGEN is used to determine the activities at 30 and 90 days after discharge. Fo'r each nuclide, the bounding activity for the allowable range of enrichments and discharge exposure is determined.

The Spent Fuel Cask Drop source terms are presented in Table 1.7.6-1. Note that these source terms are for a single assembly.

Conservatisms used in the calculation of the spent fuel cask dIrop source terms include the followving.

Use of conservative fission product inventories as noted above. Selection of the maximum activities from the range of enrichments considered. Selection of maximum activity from the maximum core burnup and maximum assembly burnup cases. Use of the technical specification radial peaking factor limit.

1.8. Atmospheric Dispersion (X'Q) Factors 1.8.1. Onsite XIQ Determination NewvX/Q factors for onsite release-receptor combinations are developed using the ARCON96 computer code ("Atmospheric Relative Concentrations in Building Wakes," NUREG/CR-633 1, Rev. I, May 1997, RSTCC Computer Code Collection No. CCC-664). All of the default values in the ARCON96 code were unchanged from the code default values with the exception of the use of 0.2 for the Surface Roughness Length per Table A-2 of RG 1.194, and the use of 4.3 for the Averaging Sector Width Constant. The minimum wvind speed was left at 0.5 m/s per the guidance instruction.

A number of various release-receptor combinations wvere considered for the control room X/Qs. These different cases wvere considered to determine the limiting release-receptor combinations for the various events. A ground level release wvas chosen for each scenario since none of the release points are 2.5 times taller than the closest solid structure as called out in Section 3.2.2 of Reference 5.21 for stack releases. The top of the containment structures is at an elevation of 782 ft. The highest release point is from the top of the plant stack, wvhich is not 2.5 times higher than the nearby containment structure.

The vertical velocity, stack flow, and stack radius terms wvere all set equal to zero since each case is a ground level release. The vent release option wvas not selected for any of the scenarios.

The actual release height wvas used in the cases. No credit was taken for effective release height due to plume rise; therefore, for the releases from the stacks, the release elevations were set equal to the stack top elevation. The elevation difference term was set equal to zero for each case since all elevation points are taken wvith respect to the same datum.

The only cases in this analysis that take credit for the building wake effect are the scenarios where the release is from the containment building and SIRWT. Some of the other scenarios have buildings betwveen the release and receptor points, but for these cases the building wvake wvas not credited for the sake of conservatism. Not crediting wvakes was accomplished by setting the building area term equal to 0.0 1 M2 as stated in Table A-2 of RG 1.194. The building area used is a conservatively determined containment cross sectional area, while the height is taken as the distance between the top of the

  • cylinder portion of the containment structure and the highest auxiliary building roof elevation. This building area is equal to 1,405 in 2.

Figure 1.8.1-1 provides a sketch of the general layout of Palisades that has been annotated to highlight the release and receptor point locations described above. All releases are taken as ground releases per guidance provided in RG 1.145, "Atmospheric Dispersion Models for Potential Accident Consequence Assessment at Nuclear Power Plants," Rev. 1, February 1983.

Table 1.8.1-1, "Release-Receptor Combination Parameters for Analysis Events," provides information related to the relative elevations of the release-receptor combinations, the straight-line horizontal distance between the release point and the receptor location, and the direction (azimuth) from the receptor location to the release point. Angles are calculated based on trigonometric layout of release and receptor points in relation to the North-South and East-West axes. Direction values are corrected for "plant North" offset from "true North" by 22.7450.

Table 1.8.1-2, "Onsite Atmospheric Dispersion Factors (X/Q) for Analysis Events," provides the Control Room X/Q factors for the release-receptor combinations described above.

Conservatisms used in the calculation of the onsite atmospheric relative concentrations include the followving. Use of ARCON96 to calculate dispersion factors. Use of only ground level releases - no elevated or vent releases. Diffuse area releases not assumed for any release pathway. No credit for plume rise taken. Only containment building and SIRWT releases credit building wvake. Use of a conservative building area when building wvake is credited.

1.8.2. Offsite X/Q Determination For offsite receptor locations, the new atmospheric dispersion (X/Q) factors are developed using thle PAVAN computer code ("PA VAN: An Atmospheric Dispersion Program for Evaluating Design Bases Accident Releases of Radioactive Material from Nuclear Power Stations," NUREG/CR-2858, November 1982, RSICC Computer Code Collection No. CCC-445). The offsite maximum X10 factors for the EAB and LPZ are presented in Table 1.8.2-I1, "Offsite Atmospheric Dispersion Factors (A'/Q) for Analysis Events." In accordance wvith Regulatory Position 4 from NUREG/CR-2858, thle maximum value from all downwvind sectors for each time period are compared with the 5% overall site A'/Q values for those boundaries, and the larger of the values are used in evaluations.

The FAB distance used in each of the 16 downwind directions from the site wvas set at 677 m. These distance and direction combinations were chosen to be conservative, not taking credit for the larger distances to the FAB in the various primary directions. The LPZ boundary distance wvas set to 4820 m.

All of the releases are considered ground level releases because the highest possible release height is less than 2.5 times higher than the adjacent containment building; as described above. As such, the release height is set equal to 10.0 meters as required by Table 3.1 of NUREG/CR-2858. The building 2

area used for the offsite building wake term is 2, 011 Mn , which is calculated to be conservatively small in that the height used in the area calculation is from the highest roof elevation of a nearby building to the elevation of the bottom of the containment dome. Release Point elevations are provided in Table 1.8. 1- 1, "Release-Receptor Combination Parameter for Analysis Events."

Conservatisms used in the calculation of the offsite atmospheric relative concentrations include the following. Use of PAVAN to calculate dispersion factors. Selection of maximum value for all downwind sectors for each time period and the 5% overall site dispersion factor values. Use of the minimum distance to site boundary in each downwvind sector. Use of only ground level releases - no elevated releases. Use of a conservative building area for building wvake credit. No exclusion of downwvind sectors for sectors than extend over Lake Michigan.

Numerical Applications, Inc. Page 16 of 84 AST Licensing Technical Report for Palisades Report Number: NAI-1 1149-027 1.8.3. Meteorological Data Meteorological data over a five-year period (1999 through 20103) is used in the development of the new X/Q factors used in the analysis, wvhich meets the guidance set forth in RG 1.194. The Palisades, Meteorological Monitoring Program, complies with RG 1.23; "Onsite Meteorological Programs,"

1972. The Meteorological Monitoring Program is described in Section 2.5 of the Palisades UFSAR.

ARCON96 analyzes the meteorological data file used and lists the total number of hours of data processed and the number of hlours of missing data in the case output. A meteorological data recovery rate may be determined from this information. Since all of the Palisades cases use the same meteorological data file, all of the cases in this analysis have the same data recovery rate. Each of the output listings in ARCON96 files present the number of hours of data processed as 43,824 and the number of missing data hours as 179. This yields a meteorological data recovery rate of 99.6%. No regulatory guidance is provided in Reg. Guide 1.194 and NUREGICR-633 I on the valid meteorological data recovery rate required for use in determining onsite X/Q values. However, Regulatory Position C.5 of RG 1.23 requires a 90% data recovery threshold for measuring and capturing meteorological data. Clearly, the 99.6% valid meteorological data rate for the cases in this analysis exceeds the 90% data recovery limit set forth by RG 1.23. With a data recovery rate of 99.6%

and a total of five years wvorth of data, the contents of the meleorological data file are representative of the long term meteorological trends at the Palisades site.

The meteorological data were also provided in annual joint frequency distribution format for 1999 through 2003. The joint frequency distribution file requires the annual meteorological data to be sorted into several classifications. This is accomplished by using three classifications that include wind direction, wvind speed, and atmospheric stability class. The format for the file conforms to the format provided in Table I of RG 1.23. These data wvere provided for the five years in terms of the percentage of hours of that time period that fell into each classification category. The data for each category (i.e.

wind speed, wind direction, and stability class unique combination) wvere converted from percent to number of hours. These hours are then input into the PA VAN4 code. Other information regarding the joint frequency distribution format for the PA VAN meteorological data may be found in RG 1.23.

2.0 Radiological Consequences - Event Analyses 2.1. Loss of Coolant Accident (LOCA) 2.1.1. Background This event is assumed to be caused by an abrupt failure of the main primary coolant pipe and the ECCS fails to prevent the core from experiencing significant degradation (i.e., melting). This sequence cannot occur unless there are multiple failures, and thus goes beyond the typical design basis accident that considers a single active failure. Activity is released to the containment and from there, released to the environment by means of containment leakage and leakage from the EGGS. This event is described in the Section 14.22.3 of the Palisades FSAR.

Numerical Applications, Inc. Page 17 of 84 AST Licensing Technical Report for Palisades Report Number: NAI- 1149-027 2.1.2. Compliance with RG 1.183 Regulatory Positions The LOCA dose consequence analysis is consistent wvith the guidance provided in RG 1.183, Appendix A, "Assumptions for Evaluating the Radiological Consequences of a L\VR Loss-of-Coolant Accident," as discussed below:

I.Regulatory Position 1 - The total core inventory of the radionuclide groups utilized for determining the source term for this event is based on RG 1.183, Regulatory Position 3. 1, at 102% of core thermal power and is provided in Table 1.7.4-I1. The core inventory release fractions for the gap release and early in-vessel damage phases of the LOCA are consistent with Regulatory Position 3.2 and Table 2 of RG 1.183.

2. Regulatory Position 2 - The sump pH is controlled at a value greater than 7.0 based on the addition of tni-sodium phosphate (TSP) baskets or an alternate buffer. Therefore, the chemical form of thle radioiodine released to the containment is assumed to be 950/ cesiuim iodide (Csl), 4.85% elemental iodine, and 0.15% organic iodide. With the exception of elemental and organic iodine and noble gases, fission products are assumed to be in particulate form.
3. Regulatory Position 3.1 - The activity released from the fuel is assumed to mix instantaneously and homogeneously throughout the free air volume of the containment. The release into the containment is assumed to terminate at the end of the early in-vessel phase.
4. Regulatory Position 3.2 - Reduction of the airborne radioactivity in the containment by natural deposition is credited. The natural deposition removal coeffi.-ient for elemental iodine wvas determined to be 2.3 hir 1 .

A natural deposition removal coefficient of 0.1 hV' is assumed (based on the Industry Degraded Core Rulemaking Program Technical Report 11.3, "Fission Product Transport in Degraded Core Accidents,"

Atomic Industrial Forum, December 1983) for all aerosols. Industry Degraded Core Rulemaking (IDCOR) Program Technical Report 11.3 ("Fission Product Cleanup System," Revision 2, December 1988) documents results from the Containment Systems Experiments. The Containment Systems Experiments examined containment atmosphere cleanup through natural transport processes. A large fraction of aerosols wvere deposited on the floor rather than on the wvalls indicating that sedimentation wvas a dominant removal process during thle tests. IDCOR Program Technical Report 11.3 documents the following in the first sentence of the first full paragraph of page 3: "Settling of aerosols due to gravity is the dominant natural mechanism for fission product retention." The Containment Systems Experiments determined that there was significant sedimentation removal even wvith a relatively low aerosol concentration. The following paragraph is quoted for IDCOR Program Technical Report 11.3, Section 4.2.2.2.4 Sedimentation, pages 25 and 26:

"Most of the available experimental data are from the CSE experiments (see Section B.I.l.4). The highest concentrations of particulates used in the CSE tests were in the range of 10,3 to I0V grams/in 3 wvhereas the initial concentration of aerosols in the RPV can bie expected to be of the order of many hundreds of grams/in 3 and wvere this total quantity of aerosol actually to reach containment, the concentration wvould still be tens of grams/in 3. Consequently, particle sizes and removal rates to be expected inside the primary system would be higher than those predicted by the CSE experiments.

However, even at the low CSE concentrations, very significant removals wvere noticed."

Starting at a conservatively lowv Cesium air concentration of 10 [1g/in 3 on Figure 4-2 of IDCOR Program Technical Report 11.3, the sedimentation removal coefficient was about 0.3/hr (based on data around the 8 hour9.259259e-5 days <br />0.00222 hours <br />1.322751e-5 weeks <br />3.044e-6 months <br /> time period and a 0.92 fraction of Cesium that settles to the floor based on Table 4-7.

Note the initial [time 0] concentration in the experiment wvas only 10,3 grams/in 3 which wvas wvell below the expected "tens of grams/in 3" in the containment from the previous quotation). The Palisades AST analysis conservatively assumes an aerosol natural deposition rate of only 0. 1/hr consistent wvith several prior approved AST submittals. As an example, the NRC found the 0. 1 per hour aerosol

removal rate to be reasonable for Kewaunee based on a study published in NUREG/CR-6 189, 'A Simplified Model of Aerosol Removal by Natural Processes in Reactor containment,' and is therefore, acceptable."

Table 34 of NUREG/CR-61 89 presents decontamination coefficients for design basis accident aerosol deposition. These decontamination coefficients are presented as a function of thermal powver, time range and release phase. Table 36 of NLREG/CR-6189 presents correlations to model these decontamination coefficients as a function of thermal power, time range and release phase.

NUREG/CR-6604 (RADTRAD: A Simplified Model for RADionuclide Transport and Removal And Dose Estimation) Table 2.2.2.1-1 presents correlations for determining the same natural deposition aerosol decontamination coefficients as a function of power, time and release phase (same as Table 36 of NUREG/CR-61 89, but sums the gap and early in-vessel release phases). Thermal power is the only parameter that is varied in this table. The Palisades analyzed thermal power (2703 MWt) is greater than the Kewaunee Nuclear Powver Plant analyzed thermal power (1851.3 MWt) from the above-referenced SER. The values of the decontamination coefficients in Table 2.2.2.1-1 of NUREG-6604 increase with thermal power for these two values, so use of thle aerosol natural deposition rate of 0.1/hr is more conservative for Palisades than for Kewaunee.

No removal of organic iodine by natural deposition is assumed.

5. Regulatory Position 3.3- Per the current licensing basis, there is at least 90% spray coverage of the containment (Reference 5.30); therefore, the containment is treated as a single wvell mixed volume.

Tile method used in the Palisades AST LOCA analysis for determining thle time period required to reach an elemental iodine DF of 200 was based on a containment atmosphere peak iodine concentration equal to 40 percent of the core iodine inventory.

As discussed in the SRP Section 6.5.2, the iodine decontamination factor (DF1) is a function of the effective iodine partition coefficient between the sump and containment atmosphere. Thus, the loss of iodine due to other mechanisms (containment leakage, surface deposition, etc.), wvould not be included in the determination of the time required to reach a DF of 200. In addition, since the iodine in the containment atmosphere and sump are decaying at the same rate, decay should not be included in determining the time to reach a DF of 200. Additional RADTRAD-NAI cases wvere performed for determining the time to reach a decontamination factor of 200.

The first RADTRAD-NAI case wvas used to determine the peak containment atmosphere elemental iodine concentration and amount of aerosol in the containment atmosphere. This case included:

  • No elemental iodine surface deposition
  • No aerosol surface deposition
  • No decay
  • No containment leakage The second RADTRAD-NAI case determined the time required to reach a DF of 200 based on the peak elemental iodine concentration from the first RADTRAI)-NAI case. The second RADTRAD-NAI case included:
  • No surface deposition
  • No decay
  • No containment leakage The second RADTRAD-NAI case showved that a DF of 200 for elemental iodine was reached at a time greater than 2.515 hours0.00596 days <br />0.143 hours <br />8.515212e-4 weeks <br />1.959575e-4 months <br />.

Numerical Applications, Inc. Page 19 of 84 AST Licensing Technical Report for Palisades Report Number: NAI-I 1149-027 A third RADTRAD-NAI case determined the time required to reach a DF of 50 for aerosol based on the peak aerosol mass from the first RADTRAD-NAI case. The third RADTRAD-NAI case included:

  • Aerosol surface deposition credited
  • No decay
  • No containment leakage The third RADTRAD-NAI case showed that a DF of 50 was reached at a time greater than 3.385 hours0.00446 days <br />0.107 hours <br />6.365741e-4 weeks <br />1.464925e-4 months <br />.
6. Regulatory Position 3.4 - Palisades does not have post-accid.-nt in-containment air filtration systems.

Palisades does have containment air coolers, which are not credited for filtration or mixing.

7. Regulatory Position 3.5 - This position relates to suppression pool scrubbing in BWRs, wvhich is not applicable to Palisades.
8. Regulatory Position 3.6 - This position relates to activity retention in ice condensers, which is not applicable to Palisades.
9. Regulatory Position 3.7 - A containment leak rate of 0. 10% per day of the containment air is assumed for the first 24 hours2.777778e-4 days <br />0.00667 hours <br />3.968254e-5 weeks <br />9.132e-6 months <br />. After 24 hours2.777778e-4 days <br />0.00667 hours <br />3.968254e-5 weeks <br />9.132e-6 months <br />, the containment leak rate is reduced to 0.05% per day of the containment air.
10. Regulatory Position 3.8 - The purge system is not considered to be in operation at thle beginning of the event. In addition, containment purge is not used after the beginning of the event for hydrogen control.
11. Regulatory Positions 4.1 through 4.6 apply to facilities wvith dlual containment systems. As such, these positions are not applicable to Palisades.
12. Regulatory Position 5.1 - Engineered Safety Feature (ESF) systems that recirculate wvater outside the primary containment (EGGS systems) are assumed to leak during their intended operation. With the exception of noble gases, all fission products released from the fuel to the containment are assumed to instantaneously and homogeneously mix in the containment sump water at the time of release from the core.
13. Regulatory Position 5.2 - Leakage from the EGGS system to the ESF rooms is taken as two times the Tech. Spec. allowable value of 0.2 gpm. The leakage is assumed to start at the earliest time the recirculation flow occurs in these systems and continue for the 30-day duration. flackleakage to the SIRWT is also considered separately as two times 2.2 gpm until 2 hours2.314815e-5 days <br />5.555556e-4 hours <br />3.306878e-6 weeks <br />7.61e-7 months <br /> when operator action is assumed to cross-tie the LPSI suction headers and eliminate backleaka~ge through the SIRWT discharge lines. After 2 hours2.314815e-5 days <br />5.555556e-4 hours <br />3.306878e-6 weeks <br />7.61e-7 months <br />, the SIWRT backleakage is reduced to two times 0.025 and continues for the remainder of the 30-day duration. SIRWT backleakage is verified through the In-Service Testing program.
14. Regulatory Position 5.3 - With the exception of iodine, all radioactive materials in the recirculating liquid are assumed to be retained in the liquid phase.
15. Regulatory Position 5.4 - A flashing fraction of 3% wvas determined based on the temperature of the containment sump liquid at the time recirculation begins. For EGGS leakage back to the SIRWT, the analysis demonstrates that the temperature of the leaked fluid wvill cool below 212*F prior to release to the SIRWT tank.

Numerical Applications, Inc. Page 20 of 84 AST Licensing Technical Report for Palisades Report Number: NAI-l 1149-027

16. Regulatory Position 5.5 - The iodine available for release at the time recirculation begins is based on the expected sump pH history and temperature (see the Release Inputs in the Methodology section below). For the ECCS leakage to the auxiliary building, 10% of the total iodine in the leaked EGGS fluid is assumed to be available for release and is assumed to become airborne and leak directly to the environment from the initiation of recirculation through 30 days. For the EGGS leakage back to the SIRWT, the sump and SIRWT pH history and temperature are used to evaluate the amount of iodine that enters the SIRWT air space.
17. Regulatory Position 5.6 - For EGGS leakage into the auxiliary building, the form of thle released iodine is 97% elemental and 3% organic. For EGGS- leakage into the SIRWT, the temperature and pH history of the sump and SIRWT are considered in determining the radioiodine available for release and the chemical form. Credit is taken for hold-uip and dilution of activity in the SIRWT as allowed by Regulatory Position 5.6. Per thle current design basis, a 50% reduction of the EGGS activity is taken for the release to the auxiliary building. No credit for holdup or dilution of EGGS leakage into the auxiliary building is taken.
18. Regulatory Position 6 - This position relates to MSSV leakage in BWRs, wvhich is not applicable to Palisades.
19. Regulatory Position 7 - Containment purge is not considered as a means of combustible gas or pressure control in this analysis. In addition, routine containment purge is not active for this event.

2.1.3. Methodology For this event, the Control Room ventilation system cycles through twvo modes of operation (the operational modes are summarized in Table 1.6.3-I). Inputs and assumptions fall into three main categories: Radionuclide Release Inputs, Radionuclide Transport Inputs, and Radionuclide Removal Inputs.

For the purposes of the LOGA analyses, a major LOGA is defined as a rupture of the PGS piping, including the double-ended rupture of the largest piping in the PGS, or of any line connected to that system up to the first closed valve. Should a major break occur, depressurization of the PGS results in a pressure decrease in the pressurizer. A reactor trip signal occurs when the pressurizer low-pressure trip setpoint is reached.

A safety injection system signal is actuated when the appropriate setpoint (high containment pressure) is reached. The following measures wvill limit the consequences of the accident in two wvays:

1. Reactor trip and borated water injection complement void formation in causing rapid reduction of power to a residual level corresponding to fission product decay heat, and
2. Injection of borated water provides heat transfer from the core and prevents excessive cladding temperatures.

Release Inputs The core inventory of the radionuclide groups utilized for this event is based on RG 1.183, Regulatory Position 3. 1, at 102% of core thermal power and is provided as Table 1.7.4- 1. The source term represents end of cycle conditions assuming enveloping initial fuel enrichment and an average core burnup of 39,300 MWD/MTU.

The leakage rate for the containment is 0. 10% of the containment air weight per day. Per RG 1.183, Regulatory Position 3.7, the primary containment leakage rate is reduced by 50% at 24 hours2.777778e-4 days <br />0.00667 hours <br />3.968254e-5 weeks <br />9.132e-6 months <br /> into the LOGA to 0.05% /day based on the post-LOGA primary containment pressure history.

Numerical Applications, Inc. Page 21 of 84 AST Licensing Technical Report for Palisades Report Number: NAI-I 1149-027 The EGGS leakage to the auxiliary building is 0.4 gpm based upon two times the current Tech. Spec.

allowable value of 0.2 gpm. The temperature of the leaka 'ge is based on the sump temperature at and after the time recirculation begins. The leakage is assumed to start at 19 minutes (minimum time to recirculation) into the event and continue throughout the 30-day period. The maximum ECCS leakage flashing fraction is less than 10% and the minimum sump pH followving the start of recirculation is 7.0; therefore, 10% of the total iodine in the leaked EGGS fluid is assumed to be released. The form of the released iodine is 97% elemental and 3% organic. Per the current design basis, a 50% reduction of the leaked EGGS activity is credited. Dilution and holdup of the EGGS leakage are not credited.

The ECGS backleakage to the SIRWT is initially assumed to be 4.4 gpm based upon doubling the previously mentioned value of 2.2 gpm. This leakage is assumed to start at 19 minutes into the event when recirculation starts and continue until 2 hours2.314815e-5 days <br />5.555556e-4 hours <br />3.306878e-6 weeks <br />7.61e-7 months <br /> when operator action is assumed to cross-tie the LPSI suction headers and eliminate backleakage through the SIRWVT discharge lines. After 2 hours2.314815e-5 days <br />5.555556e-4 hours <br />3.306878e-6 weeks <br />7.61e-7 months <br />, the backleakage is reduced to 2 x 0.025 gpm from thle recirculation line only, and continue throughout the 30-day period.

Note that based on the leakage rate and the size of the piping, the leakage wvould not reach the STRWT for an extended period of time after recirculation begins. This time period is conservatively not credited for determining when the leakage reaches thle SIRWT (i.e., the leakage is assumed to reach the SIRWT instantaneously allowing no time for radioactive decay); howvever, this time period is credited for determining the temperature of the leakage reaching the SIRWT. Based on sump pH history and pH control (pH is greater than 7 at the time recirculation begins), the iodine in the sump solution is assumed to all be in nonvolatile iodide or iodate form during the time of interest for this analysis. Precedent for this assumption was previously established in the revised Shearon Harris Alternative Source Term submittal dated August 17, 2001 (specifically page 2.22-11) and the associated Shearon Harris Issuance of Amendment (IA) and Safety Evaluation (SE) for Amendment No. 107 to NPF-63 issued October 12, 2001 (specifically page 31 of the Safety Evaluation Report). -

When introduced into the acidic solution of the SIRWT inventory, there is a potential for the particulate iodine to convert into the elemental form. The initial SIRWT pH is 4.5. Based upon the backleakage of sump water (pH of 7.0), the SIRWT pH slowly increases, reaching a maximum pH of 4.7 at 30 days (see Table 2.1-3). Using the time-dependent SIRWT pH, the amount of iodine converted to the elemental form in the SIRWT was determined based upon the data and equations provided in NUREG-5950. Thle SIRWT total iodine concentration (including stable iodine) as a function of time wvas determined. This iodine concentration ranged from a minimum value of 0 at the beginning of the event to a maximum value of 1.93E-O5gm-atomfliter at 30 days (see Table 2.1-4). Based on the time-dependent SIRWT pH level and based on the amount of water leaked into the SIRWT from the sumnp, the elemental iodine fraction in the SIRWT was determined to range from O at the beginning of the event to a maximum of 0.149 at 30 days (see Table 2.1-6).

The elemental iodine in the liquid leaked into the SIRWT is assumed to become volatile and partition between the liquid and vapor space in the SIRWT based upon the temperature dependent partition coefficient for elemental iodine as presented in NUREG-5950. A GOTHIC model was used to determine the time-dependent SIRWT temperature (see Table 2.1-5). The time-dependent partition coefficient is provided in Table 2.1-7. The SIRWT is a vented tank; therefore, there will be no pressure transient in the air region that would affect the partition coefficient. The particulate portion of the leakage is assumed to be retained in the liquid phase of the SIRWT. Since no boiling occurs in the SIRWT, the release of the activity from the vapor space within the SIRWT is calculated based upon the displacement of air by the incoming leakage. The adjusted iodine release rate is determined as followvs:

A djusted Re leaseRate (Baseline Iodine Flow) x (Iodine Volatile Fraction)

(PartitionCoefficienit) where:

Numerical Applications, Inc. Page 22 of 84 AST Licensing Technical Report for Palisades Report Number: NAM- 149-027 Baseline Iodine Flow = Volumetric flowv from the sump (displacement due to backleakage)

Partition Coefficient (12) =Table 2.1-7 Iodine Volatile Fraction =Table 2.1-6, Elemental Iodine fraction available for release from the leaked wvater The time dependent iodine release rate presented in Table 2.1-8 in then applied to the entire iodine inventory (particulate, elemental and organic) in the containment sump. The iodine released via the SIRWT air vent to the environment wvas effectively set to 100% elemental (the control room filters have the same efficiency for all forms of iodine).

Thle release point for each of the above sources is presented in Table 1.8.1-3.

Transport Inputs During the LOCA event, the activity collected in containment is assumed to be released to the environment via a ground level release from the containment. The activity from the ECCS leakage into the Auxiliary Building is modeled as release from the Auxiliary Building via the plant stack with no filtration. The activity from the SIRWT is modeled as an unfiltered ground level release from the SIRWT.

For this event, the Control Room ventilation system cycles through twvo modes of operation:

" A loss of offsite power is assumed at the beginning of the event; therefore, the initial airflo~v to the control room is assumed to consist of 384.2 cfim of unfiltered air (air infiltration due to loss of control room ventilation).

" Afler the start of the event, and after the Diesel Generators restore power, the Control Room normal air intake is isolated due to a high containment pressure (or radiation) signal. After isolation of the Control Room normal air intake, the air flow distribution consists of 1413.6 cfm of filtered makeup flow through the emergency intake, 10 cfm of unfiltered inleakage, and 1413.6 cfmn of filtered recirculation flow.

" The Control Room ventilation filter efficiencies that are applied to the filtered makeup and recirculation flowvs are 99% for particulIates/aerosols, elemental iodine, and organic iodine.

LOCA Removal Inputs Reduction of the airborne radioactivity in the containment by natural deposition is credited. The natural deposition removal coefficient for elemental iodine is 2.3 fir". A natural deposition removal coefficient of 0.1 hr'l is assumed for all aerosols (applied after containment sprays are turned off at 10 hours1.157407e-4 days <br />0.00278 hours <br />1.653439e-5 weeks <br />3.805e-6 months <br />). No removal of organic iodine by natural deposition is assumed.

Containment spray provides at least 90% coverage therefore, the Palisades containment building atmosphere is considered to be a single, well-mixed volume.

The maximum decontamination factor (DF) for the elemental iodine spray removal coefficient is 200 based on the maximum airborne elemental iodine concentration in the containment. The maximum airborne elemental iodine concentration is based on the release of 40 percent of the core iodine inventory. Based upon the elemental iodine removal rate of 4.8 hr-', the decontamination factor for elemental iodine reaches 200 at just over 2.5 15 hours1.736111e-4 days <br />0.00417 hours <br />2.480159e-5 weeks <br />5.7075e-6 months <br />.

The spray aerosol removal rate is reduced by a factor of 10 when a DF of 50 is reached. Based upon the aerosol removal rate of 2.3 hr-', the time for containment spray to produce an aerosol decontamination factor of 50 with respect to the containment atmosphere is just over 3.385 hours0.00446 days <br />0.107 hours <br />6.365741e-4 weeks <br />1.464925e-4 months <br />.

2.1.4. Radiological Consequences The atmospheric dispersion factors (X/Qs) used for this event for the Control Room dose are based on the postulated release locations and the operational mode of the control room ventilation system. These X/Qs are summarized in Table 1.8.1-2 and Table 1.8.1-3.

For the EAB dose analysis, the X/Q factor corresponding to the 0-2 hour time period was used for the entire duration of the event. The LPZ dose is determined using the X/Q factors for the appropriate time intervals.

These X10 factors are provided in Table 1.8.2-1.

The radiological consequences of the design basis LOCA are analyzed using the RADTRAD-NAI code and the inputs/assumptions previously discussed. In addition, the MicroShield code, Version 5.05, Grove Engineering, is used to develop direct shine doses to the Control Room. MicroShield is a point kernel integration code used for general-purpose gamma shielding analysis. It is qualified for this application and has been used to support licensing submittals that have been accepted by the NRC (for example, see Duane Arnold Energy Center submittal dated October 19, 2000 and assoc.iated NRC Safety Evaluation dated July 31, 2001). The QADMOD-GP code was used for determining the control room direct shine dose from the SIRWT. QADMOD-GP is a 3-D shielding analysis code that is suggested for use in the Standard Review Plan.

The post accident doses are the result of three distinct activity releases:

1. Containment leakage.
2. ESF system leakage into the Auxiliary Building.
3. ESF system leakage into the SIRWT.

The dose to the Control Room occupants includes terms for:

I. Contamination of the Control Room atmosphere by intake and infiltration of radioactive material from the containment and ESF leakage.

2. External radioactive plumne shine contribution from the containment and ESF leakage releases.

This term takes credit for Control Room structural shielding.

3. A direct shine dose contribution from the Containment's contained accident activity. This term takes credit for both Containment and Control Room structural shielding.
4. A direct shine dose contribution from the activity collected on the Control Room ventilation filters.
5. A direct shine dose contribution from the activity in the containment purge lines.
6. A direct shine dose contribution from the activity collected in the SIRWT.

As showvn in Table 2.1-9, the sum of the results of all dose contributions for EAB dose, LPZ dose, and Control Room dose are all wvithin the appropriate regulatory acceptance criteria.

Numerical Applications, Inc. Page 24 of 84 AST Licensing Technical Report for Palisades Report Number: NAT-] 149-027 2.2. Fuel Handling Accident (FHA) 2.2.1. Background This event consists of the drop of a single fuel assembly either in the Fuel Handling Building (FHB) or inside of Containment. The FHA is described in Section 14.19 of the FSAR. The FSAR description of the FHA specifies a case that assumes all of the fuel rods in a single fluel assembly are damaged. In addition, a minimum wvater level of 22.5 feet is maintained above the damaged fuiel assembly for both the containment and FHB release locations.

This analysis examined both a FHA inside the containment (wvith the equipment hatch open) and a FHA inside the FHB (wvith varying degrees of release filtration). Three FHA cases in the FHB wvere analyzed:

these cases assumed that 10%, 34%, and 50% of the release passed through the FHB filtration system. Thle source term released from the overlying water pool is the same for both the FlHB and the containment cases. RG 1.183 imposes the same 2-hour criteria for the release of the activity to the environment for either location. The control room wvas assumed to be manually isolated at 20 minutes.

The minimum wvater level above the damaged fuel is 22.5 feet. Since this is less than 23 feet specified in Section 2.0 of Appendix B to Reg. Guide 1.183, the decontamination factor for elemental iodine wvas reduced per the guidance provided in Reference 5.40. Using 22.5 feet of water along wvith the methodology outlined in Reference 5.40, produced an elemental iodine decontamination factor of 252. In combination with an organic iodine decontamination factor of 1.0, the reduced elemental iodine decontamination factor produces an overall iodine decontamination factor of 183.07.

2.2.2. Compliance wvith RG 1.183 Regulatory Positions The FHA dose consequence analysis is consistent wvith the guidance provided in RG 1.183 Appendix B, "Assumptions for Evaluating the Radiological Consequences of a Fuel Handling Accident," as discussed below:

I. Regulatory Position 1.1 - The amount of fuel damage is assumed to be all of the fuel rods in a single fuel assembly.

2. Regulatory Position 1.2 - The fission product release from the breached fuel is based on Regulatory Positions 3.1 and 3.2 of RG 1.183. Section 1.7 provides a discussion of how the FHA source term is developed. A listing of the FHA source term is provided in Table 1.7.5-1. The gap activity available for release is specified by Table 3 of RG 1.183. This activity is assumed to be released from the fuel assembly instantaneously.
3. Regulatory Position 1.3 - The chemical form of radioiodine released from the damaged fuel into the spent fuel pool is assumed to be 95% cesium iodide (CsI), 4.85% elemental iodine, and 0.15% organic iodide. The cesium iodide is assumed to completely dissociate in the spent fuel pool resulting in a final iodine distribution of 99.85% elemental iodine and 0.15% organic iodine.
4. Regulatory Position 2 - A minimum water depth of 22.5 feet is maintained above the damaged fuel assembly. Per Section 2.0 of Appendix B to Reg. Guide 1.183, Reference 5.40 wvas used to calculate the elemental iodine decontamination factor. A factor of 252 wvas calculated based on 22.5 feet of water. A decontamination factor of 1.0 wvas used for organic iodine. These twvo decontamination factors give an overall decontamination factor of 183.07 for iodine with 22.5 feet of water coverage.

Numerical Applications, Inc. Page 25 of 84 AST Licensing Technical Report for Palisades Report Number: NAI-1 1149-027

5. Regulatory Position 3 - All of the noble gas released is assumed to exit the pool without mitigation.

All of the non-iodine particulate nuclides are assumed to be retained by the pool wvater.

6. Regulatory Position 4.1 - The analysis models the release to the environment over a 2-hour period.
7. Regulatory Position 4.2 - No credit is taken for filtration of the release for the FHA in containment.

For the Fl-A in the FHB, three filtration cases were examined. These cases analyzed 10%, 34%, and 50% of the release passing through the FHB filtration system.

8. Regulatory Position 4.3 - No credit is taken for dilution of the release.
9. Regulatory Position 5.1 - Thle containment equipment hatch is assumed to be open at thle time of the fuel handling accident.
10. Regulatory Position 5.2 - No automatic isolation of the containment is assumed for the Fl-A.
11. Regulatory Position 5.3 - The release from the fuel pool is assumed to leak to the environment over a two-hour period.
12. Regulatory Position 5.4 - No ESF filtration of the containment release is credited.
13. Regulatory Position 5.5 - No credit is taken for dilution or mixing in the containment atmosphere.

2.2.3. Methodology The input assumptions used in the dose consequence analysis of thle Fl-A are provided in Table 2.2-1. It is assumed that the fuel handling accident occurs at 48 hours5.555556e-4 days <br />0.0133 hours <br />7.936508e-5 weeks <br />1.8264e-5 months <br /> after shutdown of thle reactor per the current design basis documented in FSAR Section 14.19. 100% of the gap activity specified in Table 3 of RG 1.183 is assumed to be instantaneously released from a single fuel assembly into the pool. A minimum wvater level of 22.5 feet is maintained above the damaged fuel for the duration of thle event. 100% of the noble gas released from the damaged fuel assembly is assumed to escape from the pool. All of the non-iodine particulates released from the damaged fuel assembly are assumed to be retained by the pool. The iodine released from the damaged fuel assembly is assumed to be composed of 99.85% elemental and

0. 15% organic. The activity released from the pool is then assumed to leak to the environment over a two-hour period. No credit for dilution in the containment or Fl-B is taken.

The FHA source term meets the requirements of Regulatory Position I of Appendix B to RG 1.183.

Section 1.7 discusses the development of the FHA source term, wvhich is listed in Table 1.7.5-1.

With Control Room isolation, the Control Room ventilation system cycles through two modes of operation

  • Initially the ventilation system is assumed to be operating in normal mode. The air flow distribution during this mode is 660 cfm of unfiltered fresh air.
  • 20 minutes after the start of the event, the Control Room is manually isolated. After isolation, the air flow distribution consists of 1413.6 cfm of filtered makeup flow from thle outside, 100 cfm of unfiltered inleakage, and 1413.6 cfmn of filtered recirculation flow.
  • The Control Room ventilation filter efficiencies that are applied to the filtered makeup and recirculation flows are 99% for particulate, elemental iodine, and organic iodine.

Numerical Applications, Inc. Page 26 of 84 AST Licensing Technical Report for Palisades Report Number: NAI-l 1149-027 2.2.4. Radiological Consequences The atmospheric dispersion factors (X/Qs) used for this event for the Control Room dose are based on the location of the containment equipment hatch and plant stack and the operational mode of the control room ventilation system. These X/Qs are summarized in Table 1.8.1-2 and Table 1.8.1-3.

When the Control Room Ventilation System is in normal mode, the X/Q corresponds to the normal air intake to the control room. When the ventilation system is isolated, the limiting X/Q corresponds to the Control Room emergency air intake.

The EAB doses are determined using the 0-2 hour X/Q factors for the entire event, and LPZ doses are determined using the X/Q factors for the appropriate time intervals. These X/Q factors are provided in Table 1.8.2-1.

The radiological consequences of the FHA are analyzed using the RADTRAD-NAI code and the inputs/assumptions previously discussed. As showvn in Table 2.2-2 the results for EAB dose, LPZ dose, and Control Room dose are all wvithin the appropriate regulatory acceptance criteria.

Numerical Applications, Inc. Page 27 of 84 AST Licensing Technical Report for Palisades Report Number: NAI-l 149-027 2.3. Main Steamline Break (MSLB) 2.3.1. Background This event consists of a double-ended break of a main steam line outside of containment. The radiological consequences of such an accident bound those of a MSLB inside containment. The affected steam generator (SG) rapidly depressurizes and releases the initial contents of the SG to the environment. Plant cool down is achieved via the remaining unaffected SG. This event is described in FSAR Section 14.14.3.

2.3.2. Compliance with RG 1.183 Regulatory Position~s The MSLB dose consequence analysis is consistent with the guidance provided in RG 1.183, Appendix E, "Assumptions for Evaluating the Radiological Consequences of a PWR Main Steam Line Break Accident,"

as discussed belowv:

I. Regulatory Position I - 2% of the fuel is assumed to experience DNB for thle Palisades MSLB3 event.

2. Regulatory Position 2 - 2% of the fuel is assumed to experience DNB for the Palisades MSLB event.

It wvas determined that the activity released from the damaged fuel wvill exceed that released by the twvo iodine spike cases; therefore, the two iodine spike cases wvere not analyzed.

3. Regulatory Position 3 - The activity released from the fuel is assumed to be released instantaneously and homogeneously through the primary coolant.
4. Regulatory Position 4 - Iodine releases from the steam generators to the environment are assumed to be 97% elemental and 3% organic.
5. Regulatory Position 5.1 - The primary-to-secondary leak rate is 0.3 gpm per SG.
6. Regulatory Position 5.2 - The density used in converting volumetric leak rates to mass leak rates is consistent wvith the basis of surveillance tests used to show compliance wvith the SG leak rate TS (62.4 lb./ft3 ).
7. Regulatory Position 5.3 - Based on the existing licensing basis, the primary-to-secondary leak rate is assumed to continue until the temperature of the leakage is less than 2121F at 12 hours1.388889e-4 days <br />0.00333 hours <br />1.984127e-5 weeks <br />4.566e-6 months <br />. The release of radioactivity from the unaffected SG continues for 8 hours9.259259e-5 days <br />0.00222 hours <br />1.322751e-5 weeks <br />3.044e-6 months <br /> (time to place SDC in operation).
8. Regulatory Position 5.4 - All noble gas radionuclides released from the primary system are assumed to be released to the environment without reduction or mitigation.
9. Regulatory Position 5.5.1 - In the faulted SG, all of the primary-to-secondary leakage is assumed to flash to vapor and be released to the environment with no mitigation. For the unaffected steam generator used for plant cooldown, the primary-to-secondary leakage is assumed to mix with the secondary water without flashing.
12. Regulatory Position 5.5.2 - Any postulated leakage that immediately flashes to vapor is assumed to rise through the bulk wvater of the SG into the steam space and is assumed to be immediately released to the environment wvith no mitigation; i.e., no reduction for scrubbing wvithin the SG bulk water is credited.

Numerical Applications, Inc. Page 28 of 84 AST Licensing Technical Report for Palisades Report Number: NAI-l 1149-027

13. Regulatory Position 5.5.3 - All leakage that does not immediately flash is assumed to mix with the bulk wvater.
14. Regulatory Position 5.5.4 - The radioactivity wvithin the bulk water is assumed to become vapor at a rate that is a function of the steaming rate and the partition coefficient. A partition coefficient of 100 is assumed for the iodine. The retention of particulate radionuclides in the unaffected SG is limited by the moisture carryover from the SG. The same partition coefficient of 100, as used for iodine, is assumed for other particulate radionuclides. This assumption is consistent wvith the SG carryover rate of less than 1%. No reduction in the release is assumed from the faulted SG.
15. Regulatory Position 5.6 - Steam generator tube bundle uncovery is not postulated for thle intact SG.

2.3.3. Other Assumptions I. RG 1.183 does not address secondary coolant activity. This analysis assumes that the equilibrium specific activity on the secondary side of the steam generators is equal to the Tech. Spec. limit of 0.1 pCi/gm Dose Equivalent 1-131.

2. Thle steam mass release rates for the intact SG are provided in Table 2.3-2.
3. This evaluation assumes that the PCS mass remains constant throughout the MSLB event (no change in the PCS mass is assumed as a result of the MSLB or from the safety injection system).
4. The SG secondary side mass in the unaffected SG is assumed to remain constant throughout the event.
5. Releases from the faulted main steam line (and associated SG) are postulated to occur from the main steam line associated with the most limiting atmospheric dispersion factors. Releases from the unaffected SG are postulated to occur from the MSSV or ADV wvith thle most limiting atmospheric dispersion factors.

2.3.4. Methodology Input assumptions used in the dose consequence analysis of the MSLB are provided in Table 2.3-I. The postulated accident assumes a double-ended break of one main steam line outside containment. The radiological consequences of such an accident bound those of a MSLB inside of containment. Upon a MSLB, the affected SG rapidly depressurizes and releases the initial contents of the SG to the environment.

Plant cooldown is achieved via the remaining unaffected SG.

The analysis assumes that the entire fluid inventory from the affected SG is immediately released to the environment. The secondary coolant iodine concentration is assumed to be the maximum value of 0.1 pCi/gm DE 1-131 permitted by Tech. Specs. Primary coolant is also released into the affected steam generator by leakage across the SG tubes. Activity is released to the environment from the affected steam generator, as a result of the postulated primary-to-secondary leakage and the postulated activity levels of the primary and secondary coolants, until the affected steam generator is completely isolated at 12 hours1.388889e-4 days <br />0.00333 hours <br />1.984127e-5 weeks <br />4.566e-6 months <br /> (primary system temperature less than 212'F). Additional activity, due to SG tube leakage, is released via the unaffected SG via steaming from the unaffected SG MSSVs/ADVs for 8 hours9.259259e-5 days <br />0.00222 hours <br />1.322751e-5 weeks <br />3.044e-6 months <br />, which is the time required to initiate SDC. These release assumptions are consistent with the requirements of RG 1.183.

2% of the fuel is assumed to experience DNB. The activity released by the damaged fuel exceeds that released by the twvo iodine spike cases; therefore, the two iodine spike cases wvere not analyzed. The source term for the MSLB wvas determined by adjusting the total core inventory (presented in Table 1.7.4-I) by the following:

  • the fraction of fuel damaged (2%)

Numerical Applications, Inc. Page 29 of 84 AST Licensing Technical Report for Palisades Report Number: NA1-l1149-027

  • the non-LOCA fission product gap fractions from Table 3 of RG 1.183
  • the radial peaking factor of 2.04 For this event, the Control Room ventilation system cycles through twvo modes of operation:

0 Initially the ventilation system is assumed to be operating in normal mode. The air flow distribution during this mode is 660 cfm of unfiltered fresh air.

0 20 minutes after the start of the event, the Control Room is manually isolated. After isolation, the air flowv distribution consists of 1413.6 cfm of filtered makeup flow from the outside, 10 cfM of unfiltered inleakage, and 1413.6 cfm of filtered recirculation flow.

0 The Control Room ventilation filter efficiencies that are applied to the filtered makeup and recirculation flows are 99% for particulate, elemental iodine, and organic iodine.

2.3.5. Radiological Consequences The atmospheric dispersion factors (X/Qs) used for this event for the Control Room dose are based on the postulated release locations and the operational mode of the control room ventilation system. These X/Qs are summarized in Table 1.8.1-2 and Table 1.8.1-3.

Releases from the intact SG are assumed to occur from the MSSV/ADV that produces the most limiting X/Qs. Releases from the faulted SG are assumed to occur from the location on a steam line that produces the most limiting X/Qs.

For the EAB dose analysis, the X/Q factor corresponding to the 0-2 hour time period was used for the entire duration of the event. The LPZ dose is determined using the X/Q factors for the appropriate time intervals.

These X/Q factors are provided in Table 1.8.2-I1.

The radiological consequences of the MSLB Accident are analyzed using the RADTRAD-NAI code and the inputs/assumptions previously discussed. As showvn in Table 2.3-3, the results of both cases for EAB dose, LPZ dose, and Control Room dose are all wvithin the appropriate regulatory acceptance criteria.

The apparent limited margin for the MSLB control room operator dose is mitigated by noting thle conservatisms used in the calculation of the main steam line break results. The assumption of 2% fuel failures is very conservative considering the thermal-hydraulic analysis of thle MSLB demonstrates that no fuel failures would occur. In addition, no credit for the high velocity, buoyant release from the main steam safety valves or atmospheric dump valves is applied to the atmospheric dispersion factors for these releases. This is also a significant conservatism given the close proximity of the MSSV and ADV to the control room normal intakes (source of the control room envelope unfiltered inleakage) and resulting high probability that a significant fraction of the release bypasses the normal intakes.

.2.4. Steam Generator Tube Rupture (SGTR) 2.4.1. Background This event is assumed to be caused by the instantaneous rupture of a Steam Generator tube that relieves to the lower pressure secondary system. No melt or clad breach is postulated for the Palisades SGTR event.

This event is described in FSAR Section 14.15.

2.4.2. Compliance wvith RG 1.183 Regulatory Positions The SGTR dose consequence analysis is consistent wvith the guidance provided in RG 1.183, Appendix F, "Assumptions for Evaluating the Radiological Consequences of a PWR Steam Generator Tube Rupture Accident," as discussed belowv:

1. Regulatory Position I - No fuel damage is postulated to occur for the Palisades SGTR event.
2. Regulatory Position 2 - No fuel damage is postulated to occur for the Palisades SGTR event. Twvo cases of iodine spiking are assumed.
3. Regulatory Position 2.1 - One case assumes a reactor transient prior to the postulated SGTR that raises the primary coolant iodine concentration to the maximum allowed by Tech. Specs, wvhich is a value of 40.0 piCi/gm DE 1-131 for the analyzed conditions. This is the pre-accident spike case.
4. Regulatory Position 2.2 - One case assumes the transient associated with the SGTR causes an iodine spike. The spiking model assumes the primary coolant activity is initially at the Tech. Spec. value of 1.0 pCi/gm DE 1-131. Iodine is assumed to be released from thle fuel into the PCS at arate of335 times the iodine equilibrium release rate for a period of 8 hours9.259259e-5 days <br />0.00222 hours <br />1.322751e-5 weeks <br />3.044e-6 months <br />. This is the accident-induced spike case.
5. Regulatory Position 3 - The activity released from the fuel is assumed to be released instantaneously and homogeneously through the primary coolant.
6. Regulatory Position 4 - Iodine releases from the steam generators to the environment are assumed to be 97% elemental and 3% organic.
7. Regulatory Position 5.1 - The primary-to-secondary leak rate is 0.3 gpm per SG.
8. Regulatory Position 5.2 - The density used in converting volumetric leak rates to mass leak rates is consistent with the basis of surveillance tests used to show compliance wvith the SG leak rate TS (62.4 lbý/113 ).
9. Regulatory Position 5.3 - The release of radioactivity from both the affected and unaffected SGs is assumed to continue until shutdowvn cooling is in operation and steam release from the SGs is terminated (8 hours9.259259e-5 days <br />0.00222 hours <br />1.322751e-5 weeks <br />3.044e-6 months <br /> into the event).
10. Regulatory Position 5.4 - The release of fission products from the secondary system is evaluated with the assumption of a coincident loss of offsite powver (LOOP).
11. Regulatory Position 5.5 - All noble gases released from thle primary system are assumed to be released to the environment without reduction or mitigation.

Numerical Applications, Inc. Page 31 of 84 AST Licensing Technical Report for Palisades Report Number: NA 1-I 1149-027

12. Regulatory Position 5.6 - Regulatory Position 5.6 refers to Appendix E, Regulatory Positions 5.5 and 5.6. The iodine transport model for release from the steam generators is as followvs:

" Appendix E, Regulatory Position 5.5.1 - All steam generators effectively maintain tube coverage.

The primary-to-secondary leakage is assumed to mix wviih the secondary wvater without flashing for all steam generators.

" Appendix E, Regulatory Position 5.5.2 - A portion of the primary-to-secondary ruptured tube flow through the SGTR is assumed to flash to vapor, based on the thermodynamic conditions in the reactor and secondary. The portion that flashes immediately to vapor is assumed to rise through the bulk water of the SG, enter the steam space, and be immediately released to the environment.

Scrubbing of the flashed flow in the affected SG is credited. The methodologies presented in NUREG-0409 are used to determine the amount of scrubbing of the flashed flow.

" Appendix E, Regulatory Position 5.5.3 - All of the SG tube leakage and ruptured tube flowv that does not flash is assumed to mix with the bulk water.

" Appendix E, Regulatory Position 5.5.4 - The radioactivity wvithin the bulk wvater is assumed to become vapor at a rate that is a function of the steaming rate and the partition coefficient. A partition coefficient of 100 is assumed for the iodine. Thle retention of particulate radionuclides in the SGs is limited by the moisture carryover from the SGs.

The same partition coefficient of 100, as used for iodine, is assumed for other particulate radionuclides. This assumption is consistent wvith the SG carryover rate of less than 1%.

" Appendix E, Regulatory Position 5.6 - Steam generator tube bundle uncovery is not postulated for this event for Palisades.

2.4.3. Other Assumptions I1. For the determination of the activity concentrations, the PCS and SG volumes are assumed to remain constant throughout both the pre-accident and the accident-induced iodine spike SGTR events.

2. For the determination of the amount of scrubbing in the affected SG, the time-dependent wvater level in the affected SG is used.
3. Data used to calculate the iodine equilibrium appearance rate are provided in Table 2.4-5, "Iodine Equilibrium Appearance Assumptions."

2.4.4. Methodology Input assumptions used in the dose consequence analysis of the SGTR event are provided in Table 2.4-1.

This event is assumed to be caused by the instantaneous rupture of a steam generator tube releasing primary coolant to the lowver pressure secondary system. In the unlikely event of a concurrent loss of power, the loss of circulating water through the condenser would eventually result in the loss of condenser vacuum.

Valves in the condenser bypass lines would automatically close to protect the condenser thereby causing steam relief directly to the atmosphere from the ADVs or MSS Vs.

A thermal-hydraulic analysis is performed to determine a conservative maximum break flowv, break flashing flowv, and steam release inventory through the faulted SG relief valves. In order to prevent SG overfill in the faulted SG, periodic releases via the ADVs occur from 0.5 to 8.0 hours0 days <br />0 hours <br />0 weeks <br />0 months <br />. At 8 hours9.259259e-5 days <br />0.00222 hours <br />1.322751e-5 weeks <br />3.044e-6 months <br />, all releases from the affected SG cease. Additional activity, based on the proposed primary-to-secondary leakage limits, is released via the unaffected SG via the ADVs until the RI-R system is placed in operation (at 8 hours9.259259e-5 days <br />0.00222 hours <br />1.322751e-5 weeks <br />3.044e-6 months <br />) to continue heat removal from the primary system.

Numerical Applications, Inc. Page 32 of 84 AST Licensing Technical Report for Palisades Report Number: NAI-l 1149-027 Per the Palisades FSAR, Section 14.15, no fuel melt or clad breach is postulated for the SGTR event.

Consistent wvith RG 1.183 Appendix F, Regulatory Position 2, if no or minimal fuel damage is postulated for the limiting event, the activity release is assumed as the maximum allowed by Technical Specifications for two cases of iodine spiking: (1) maximum pre-accident iodine spike, and (2) maximum accident-induced, or concurrent, iodine spike.

For the case of a pre-accident iodine spike, a reactor transient is assumed to have occurred prior to the postulated SGTR event. The primary coolant iodine concentration is increased to the maximum value of 40 p.Ci/gm DE 1-131 permitted by Tech. Specs. The iodine activities for the pre-accident spike case are presented in Table 2.4-4. Primary coolant is released into the ruptured SG by the tube rupture and by the allowable primary-to-secondary leakage. Activity is released to the environment from thle ruptured SG via direct flashing of a fraction of the released primary coolant from the tube rupture and also via steaming from the ruptured SG ADVs until the ruptured steam generator is isolated at 8 hours9.259259e-5 days <br />0.00222 hours <br />1.322751e-5 weeks <br />3.044e-6 months <br />. The unaffected SG is used to cool dowvn the plant during the SGTR event. Primary-to-secondary tube leakage is also postulated into the intact SG. Activity is released via steaming from the unaffected SG ADVs until the decay heat generated in the reactor core can be removed by the SDC system at 8 hours9.259259e-5 days <br />0.00222 hours <br />1.322751e-5 weeks <br />3.044e-6 months <br /> into the event. These release assumptions are consistent wvith the requirements of RG 1.183.

For the case of the accident-induced spike, the postulated STGR event induces an iodine spike. The PCS activity is initially assumed to be 1.0 pCi/gmi DE 1-131 as allowved by Tech. Specs. Iodine is released from the fuel into the PCS at a rate of 335 times the iodine equilibrium release rate for a period of 8 hours9.259259e-5 days <br />0.00222 hours <br />1.322751e-5 weeks <br />3.044e-6 months <br />. WVith iodine activity at equilibrium, the iodine release rate is equal to the rate at which iodine is lost due to decay, purification, and primary system leakage. Parameters used in the determination of the iodine equilibrium release rate are provided in Table 2.4-5. The iodine activities for the accident-induced (concurrent) iodine spike case are presented in Table 2.4-6. All other release assumptions for this case are identical to those for the pre-accident spike case.

For this event, the Control Room ventilation system cycles through two modes of operation:

  • Initially the ventilation system is assumed to be operating in normal mode. The air flow distribution during this mode is 660 cfmn of unfiltered fresh air.
  • 20 minutes'after the start of the event, the Control Room is manually isolated. After isolation, the air flowv distribution consists of 1413.6 cfm of filtered makeup flow from thle outside, 100 cfM of unfiltered inleakage, and 1413.6 cfm of filtered recirculation flow.
  • The Control Room ventilation filter efficiencies that are applied to the filtered makeup and recirculation flows are 99% for particulate, elemental iodine, and organic iodine.

Numerical Applications, Inc. Page 33 of 84 AST Licensing Technical Report for Palisades Report Number: NAl 11149-027 2.4.5. Radiological Consequences The atmospheric dispersion factors (X/Qs) used for this event for the Control Room dose are based on the postulated release locations and the operational mode of the control room ventilation system. These XlQs are summarized in Table 1.8.1-2 and Table 1.8.1-3.

Releases from the intact and faulted SGs are assumed to occur from the MSSV/ADV that produces the most limiting X/Qs wvhen combined with the limiting applicable control room intake.

For the EAB dose analysis, the X/Q factor corresponding to the 0-2 hour time period wvas used for the entire duration of the event. The LPZ dose is determined using tileX/Q factors for the appropriate time intervals.

These X/Q factors are provided in Table 1.8.2-I1.

The radiological consequences of the SGTR Accident are analyzed using the RADTRAD-NAI code and the inputs/assumpt ions previously discussed. Two activity release cases corresponding to the PCS maximum pre-accident iodine spike and the accident-induced iodine spike are analyzed. As shown in Table 2.4-8, the radiological consequences of the Palisades SGTR. event for EAB dose, LPZ dose, and Control Room dose are all wvithin the appropriate regulatory acceptance criteria.

2.5. Small Line Break Outside of Containment (SLIIOC) 2.5.1. Background This event examines the radiological consequences of the rupture of a letdown line in the reactor building.

The activity from the letdown line rupture will be released to the environment by the auxiliary building ventilation system that exhausts via the plant stack. The current FSAR analysis of this event includes a pre-accident iodine spike case, a concurrent iodine spike case, and a case wvith an initial PCS activity of 1.0 ILCi/gm with no iodine spike. Per Section 15.6.2 of the Standard Review Plan, for the small line break outside of containment, only the concurrent iodine spike case is required. In addition, the initial PCS activity case wvill be bounded by the concurrent iodine spike case; therefore, only the concurrent iodine spike case was analyzed. This event is described in Section 14.23 of the FSAR.

2.5.2. Compliance with RG 1.183 Regulatory Positions and other Assumptions No specific guidance for this event is provided by RG 1.183. Therefore, the AST reanalysis for this event wvill followv Section 15.6.2 of the Standard Review Plan wvith appropriate modifications to maintain the intent of RG 1.183.

1. No fuel damage is postulated to occur for the Palisades SLBC)C event.
2. Per SRP 15.6.2 - It is assumed that the transient associated %Nith thle SLBOC causes an iodine spike.

The spiking model assumes the primary coolant activity is initially at the Tech. Spec. value of 1.0 pCi/gm DE 1-13 1. Iodine is assumed to be released from the fuel into the PCS at a rate of 500 times the iodine equilibrium release rate for a period of 8 hours9.259259e-5 days <br />0.00222 hours <br />1.322751e-5 weeks <br />3.044e-6 months <br />.

3. The activity released from the fuel is assumed to be released instantaneously and homogeneously through the primary coolant.
4. Iodine releases from the break to the environment are assumed to be 97% elemental and 3% organic.
5. The temperature and pressure of the break flow (135*F and 35 psia) wvill result in minimal flashing of the break flow. Howvever, consistent wvith Section 5.5 of Appendix A to RG 1.183, a flashing fraction of 10% is conservatively used for determining the iodine release from the break.
6. 100% of thle noble gases in the break flow are assumed to be released.
7. The initial PCS activity is assumed to be at the TS limit Of 1.0 pLCi/gm Dose Equivalent 1-13 1and I00/E-bar gross activity.

2.5.3. Methodology Input assumptions used in the dose consequence analysis of the SLBOC event are provided in Table 2.5-1.

This event is caused by the rupture of a letdowvn line in the auxiliary building. A 160 gpm break flow at 135 0F and 35 psia is postulated. 60 minutes are required to identify and isolate the break. No reactor trip is assumed.

Per Section 15.6.2 of the Standard Reviewv Plan a concurrent iodine spike with a multiplier of 500 on the equilibrium iodine release rate is used to determine the iodine con.-entration in the released fluid. Table 2.5-2 lists the iodine release rates. Table 2.4-5 lists the inputs used] to determine the equilibrium iodine

Numerical Applications, Inc. Page 35 of 84 AST Licensing Technical Report for Palisades Report Number: NAM- 149-027 release rates. The thermodynamic properties of the release flow will result in little flashing; however, per Section 5.5 of Appendix A to RG 1.183 (Assumptions on ESF System Leakage), it is conservatively assumed that 10% of the iodine in the break flow wvill be released to the reactor building airspace. The iodine released to the reactor building airspace is assumed to consist of 97% elemental and 3% organic iodine. Dilution, holdup, and plateout of the release in the reactor building are not credited. No filtration of the release from the reactor building is applied.

For this event, the Control Room ventilation system cycles through two modes of operation:

  • Initially the ventilation system is assumed to be operating in normal mode. The air flow distribution during this mode is 660 cfmn of unfiltered fresh air.
  • 20 minutes after the start of the event, the Control Room is manually isolated. After isolation, the air flow distribution consists of 1413.6 cfmn of filtered makeup flow from the outside, 100 cfmn of unfiltered inleakage, and 1413.6 cfmn of filtered recirculation flow.
  • The Control Room ventilation filter efficiencies that are applied to the filtered makeup and recirculation flows are 99% for particulate, elemental iodine, and organic iodine.

2.5.4. Radiological Consequences The atmospheric dispersion factors (X/Qs) used for this event for the Control Room dose are based on the postulated release locations and the operational mode of the control room ventilation system. These X/Qs are summarized in Table 1.8.1-2 and Table 1.8.1-3.

The release-receptor point locations are chosen to minimize the distance from the release point to the Control Room air intake. When the Control Room Ventilation System is in normal mode, the X/Q corresponds to the normal air intake to the control room. When the ventilation -system is isolated, thle limiting X/Q corr esponds to the Control Room emergency air intake. The X/Q values for thle various combinations of release points and receptor locations are presented in Table 1.8.1-2. Table 1.8.1-3 presents the RelIease- Receptor pairs applicable to the Control Room dose from the SLBOC release point (plant stack) for the different modes of Control Room operation during the event.

For the EAR dose analysis, the X/Q factor corresponding to the 0-2 hour time period was used for the entire duration of the event. The LPZ doses are determined using the XIQ factors for the appropriate time intervals. These A'/Q factors are provided in Table 1.8.2- 1.

The radiological consequences of the SLBOC event are analyzed using the RADTRAD-NAI code and the inputs/assumptions previously discussed. As showvn in Table 2.5-3, the results for EAB dose, LPZ dose, and Control Room dose are all wvithin the appropriate regulatory acceptance criteria.

Numerical Applications, Inc. Page 36 of 84 AST Licensing Technical Report for Palisades Report Number: NAI- 1149-027 2.6. Control Rod Ejection (GRE) 2.6.1. Background This event consists of the ejection of a single control rod assembly. The GRE results in a reactivity insertion that leads to a core power level increase and subsequent reactor trip. Following the reactor trip, plant cooldowvn is affected by steam release to the condenser. Two GRE cases are considered. The first case assumes that 100% of the activity released from the damaged fuel is instantaneously and homogeneously mixed throughout the containment atmosphere. The second case assumes that 100% of the activity released from the damaged fuel is completely dissolved in the primary coolant and is available for release to the secondary system. This event is described in the FSAR, Section 14.16.

2.6.2. Compliance wvith RG 1.183 Regulatory Positions The GRE dose consequence analysis is consistent with the guidance provided in RG 1.183 Appendix H, "Assumptions for Evaluating the Radiological Consequences of a PWR Rod Ejection Accident," as discussed belowv:

I. Regulatory Position I - The total core inventory of the radionuclide groups utilized for determining the source term for this event is based on RG 1.183, Regulatory Position 3. 1, and is provided in Table 1.7.4-1. The inventory provided in Table 1.7.4-1 is adjusted for the fraction of fuel dama ,ged and a radial peaking factor of 2.04 is applied. The release fractions provided in RG 1.183 Table 3 are adjusted to comply wvith the specific RG 1.183 Appendix H release requirements. For both the containment and secondary release cases, the activity available for release from the fuel gap for fuel that experiences DNB is assumed to be 10% of the noble gas and iodine inventory in the DNB fuel. For the containment release case for fuel that experiences fuel centerline melt (FCM), 100% of the noble gas and 25% of the iodine inventory in the melted fuel is assumed to be released to the containment. For the secondary release case for fuel that experiences FCM, 100% of the noble gas and 50% of thle iodine inventory in the melted fuel is assumed to be released to thle primary coolant.

2. Regulatory Position 2 - Fuel damagVe is assumed for this event.
3. Regulatory Position 3 - For the containment release case, 100%,1 of the activity released from the damaged fuel is assumed to mix instantaneously and homogeneously in the containment atmosphere.

For the secondary release case, 100% of the activity released from the damaged fuel is assumed to mix instantaneously and homogeneously inthe primary coolant and be available for leakage to the secondary side of the SGs.

4. Regulatory Position 4 - The chemical form of radioiodine released from the damaged fuel to the containment is assumed to be 95% cesium iodide (CsI), 4.85% elemental iodine, and 0. 15% organic iodide. Containment sump pH is controlled to 7.0 or higher.
5. Regulatory Position 5 - The chemical form of radioiodine released from the SGs to the environment is assumed to be 97% elemental iodine, and 3% organic iodide.
6. Regulatory Position 6.1 - For the containment leakage case, natural deposition in the containment is credited. Containment spray is not credited.
7. Regulatory Position 6.2 - The containment is assumed to leak at the proposed TS maximum allowvable rate of 0. 10% for thle first 24 hours2.777778e-4 days <br />0.00667 hours <br />3.968254e-5 weeks <br />9.132e-6 months <br /> and 0.05% for the remainder of the event.
8. Regulatory Position 7.1 - The primary-to-secondary leak rate is 0.3 gpm per SG.
9. Regulatory Position 7.2 - The density used in converting volumnetric leak rates to mass leak rates is consistent with the basis of surveillance tests used to showv compliance with the SG leak rate TS (62.4 lb ,/ft').
10. Regulatory Position 7.3 - All of the noble gas released to the secondary side is assumed to be released directly to the environment without reduction or mitigation.

I1I.Regulatory Position 7.4 - Compliance wvith Appendix E Sections 5.5 and 5.6 is discussed below:

" Appendix E, Regulatory Position 5.5.1 - Both steam generators are used for plant cooldown.

Therefore, the primary-to-secondary leakage is assumed to mix wvith the secondary water without flashing.

" Appendix E, Regulatory Position 5.5.2 - None of the SG tube leakage is assumed to flash for this event.

" Appendix E, Regulatory Position 5.5.3 - All of the SG tube leakage is assumed to mix with the bulk water.

" Appendix E, Regulatory Position 5.5.4 - The radioactivity wvithin the bulk wvater is assumed to become vapor at a rate that is a function of the steaming rate and the partition coefficient. A partition coefficient of 100 is assumed for the iodine. The retention of particulate radionuclides in the S~s is limited by the moisture carryover from the SGs.

The same partition coefficient of 100, as used for iodine, is assumed for other particulate radionuclides. This assumption is consistent with the SG carryover rate of less than 1%.

" Appendix E, Regulatory Position 5.6 - Steam generator tube bundle uncovery is not postulated for this event for Palisades.

2.6.3. Other Assumptions I. RG 1.183 does not address secondary coolant activity. This analysis assumed that the equilibrium specific activity on the secondary side of the steam generators is equal to thle TS limit of 0.1 ItCi/gm Dose Equivalent 1-131.

2. The steam mass release rates for the S~s are provided in Table 2.6-2.
3. This evaluation assumed that the PCS mass remains constant throughout the event.
4. The SG secondary side mass in the SGs is assumed to remain constant throughout the event.
5. Steam releases from the S~s are postulated to occur from the MSSV or ADV with the most limiting atmospheric dispersion factors. For the GRE inside of containment release case, releases are assumed to leak out of the containment via the same containment release points discussed for the LOCA in Section 2. 1.

2.6.4. Methodology Input assumptions used in the dose consequence analysis of the GRE are provided in Table 2.6-1. The postulated accident consists of two cases. One case assumes that 100% of the activity released from the damaged fuel is instantaneously and homogeneously mixed throughout the containment atmosphere, and the second case assumes that 100% of the activity released from the damaged fuel is completely dissolved in the primary coolant and is available for release to the secondary system.

Numerical Applications, Inc. Page 38 of 84 AST Licensing Technical Report for Palisades Report Number: NAM- 149-027 For the containment release case, 100% of the activity is released instantaneously to the containment. The releases from the containment correspond to the same leakage points discussed for the LOCA in Section 2.1. Natural deposition of the released activity inside of containment is credited. Removal of activity via containment spray is not credited.

For the secondary release case, primary coolant activity is released into the SGs by leakage across the SG tubes. The activity on the secondary side is then released via the SG MSSVs/ADVs. Additional activity based on the secondary coolant initial iodine concentration is assumed to be equal to the maximum value of

0. 1 pCi/gm DE 1-131 permitted by Tech Specs. Activity is released to the environment from the steam generator as a result of the postulated primary-to-secondary leakage and the postulated activity levels of the primary and secondary coolants, until the release is terminated (at 8 hours9.259259e-5 days <br />0.00222 hours <br />1.322751e-5 weeks <br />3.044e-6 months <br /> for shutdown cooling initiation).

These release assumptions are consistent with the requirements of RG 1.183. The noble gas activity released by the tube leakage is assumed to be released directly to the environment wvithout mitigation.

The CRE is evaluated wvith the assumption that 0.5% of the fuel excperiences FCM and 14.7% of thle fuel experiences DNB. The activity released from the damaged fuel corresponds to the requirements set out in Regulatory Position I of Appendix H to RG 1.183. A radial peaking factor of 2.04 is applied in the development of the source terms.

For this event, the Control Room ventilation system cycles through three modes of operation (thle operational modes are summarized in Table 1.6.3-1):

  • Loss of offsite power is assumed to occur at the beginning of the event. During this time period, the air flow distribution consists of 384.2 cfmn of unfiltered air infiltration.
  • Power to the control room ventilation system is restored at 90 seconds and the normal ventilation flow rate of 660 cfm of unfiltered air is restored.
  • 20 minutes after the start of the event, the Control Room is assumed to be manually isolated.

After isolation, the air flowv distribution consists of 1413.6 cfmn of filtered makeup flowv from thle outside, 10 cfmr of unfiltered inleakage, and 1413.6 cfm of filtered recirculation flow.

  • The Control Room ventilation filter efficiencies that are applied to the filtered makeup and recirculation flows are 99% for particulate, elemental iodine, and organic iodine.

2.6.5. Radiological Consequences The atmospheric dispersion factors (X/Qs) used for this event for the Control Room dose are based on the postulated release locations and the operational mode of the control room ventilation system. These X/Qs are summarized in Table 1.8.1-2 and Table 1.8.1-3.

For the CRE secondary side release case, releases from the SGs are assumed to occur from the MSSV/ADV that produces the most limiting X'Qs. When the Control Room Ventilation System is in normal mode, the X/Q corresponds to the normal air intake to the control room. When the ventilation system is isolated, the limiting X/Q corresponds to the Control Room emergency air intake. For the CRE containment release case, the X/Qs for containment leakage are assumed to be identical to those for the LOCA discussed in Section 2.1.

For the EAB dose analysis, the A'/Q factor corresponding to the 0-2 hour time period was used for the entire duration of the event. The LPZ doses are determined using the XIQ factors for the appropriate time intervals. These A'/Q factors are provided in Table 1.8.2-1.

The radiological consequences of the RCCA Ejection are analyzed using the RADTRAD-NAI code and the inputs/assumptions previously discussed. As showvn in Table 2.6-3, the results of both cases for EAB dose, LPZ dose, and Control Room dose are all wvithin the appropriate regulatory acceptance criteria.

Numerical Applications, Inc. Page 39 of 84 AST Licensing Technical Report for Palisades Report Number: NAM- 149-027 2.7. Spent Fuel Cask Drop 2.7.1. Background The purpose of this analysis is to reanalyze the radiological consequences of the cask drop accident presented in Section 14.11 of the Palisades FSAR. RG 1.183 does not provide any specific guidance for the cask drop event; therefore, the requirements of Appendix B of the RG (fuel handling accident) are followed for the cask drop reanalysis.

Three cask drop cases were analyzed:

  • Case I wvith 90% of the release via the FHB filtration system, 30 days of decay, and the control room initially aligned in emergency recirculation mode.
  • Case 2 with 82.5% of the release via the FHB filtration system, 30 days of decay, and the control room initially aligned in emergency recirculation mode.
  • Case 3 with 0% of the release via the FHB filtration system, 90 days of decay, and no isolation of the control room.

Due to conflicting requirements outlined in Section 2.0 of Appendix B to Reg. Guide 1.183, the cask drop cases wvere analyzed with an elemental iodine decontamination factor of 285, which corresponds to an overall iodine decontamination factor of 200.

2.7.2. Compliance wvith RG 1.183 Regulatory Positions The Cask Drop dose consequence analysis is consistent wvith the guidance provided in RG 1.183, Appendix B, "Assumptions for Evaluating the Radiological Consequences of a Fuel Handling Accident" as discussed below:

I. Regulatory Position 1.1 - The amount of fuel damaged is consistent with the current design basis (73 damaged fuel assemblies).

2. Regulatory Position 1.2 - The fission product release from the breached fuiel is based on Regulatory Positions 3.1 and 3.2 of RG 1.183. Section 1.7.6 provides a discussion ofhowv the Cask Drop source term is developed. A listing of the source terms is provided in Table 1.7.6-1. The gap activity available for release is specified by Table 3 of RG 1.183. This activity is assumed to be released instantaneously.
3. Regulatory Position 1.3 - The chemical form of radioiodine released from the damaged fuel into the spent fuel pool is assumed to be 95% cesium iodide (Csl), 4.85% elemental iodine, and 0.15% organic iodide. The cesium iodide is assumed to completely dissociate in the spent fuel pool resulting in a final iodine distribution of 99.85% elemental iodine and 0.15% organic iodine.
4. Regulatory Position 2 - A minimum water depth of 23 feet is maintained above the damaged fuel assemblies. Therefore, an overall effective decontamination factor of 200 is used for the iodine.
5. Regulatory Position 3 - All of the noble gas released is assumed to exit the pool wvithout mitigation.

All of the non-iodine particulate nuclides are assumed to be retained by the pool water.

6. Regulatory Position 4.1 - The radioactive material released from the fuel pool is assumed to be released from the building to the environment within 2 hours2.314815e-5 days <br />5.555556e-4 hours <br />3.306878e-6 weeks <br />7.61e-7 months <br />.
7. Regulatory Position 4.3 - No dilution is assumed.

Numerical Applications, Inc. Page 40 of 84 AST Licensing Technical Report for Palisades Report Number: NA1-l 1149-027

8. Regulatory Position 5 - The event does not occur in the containment.

2.7.3. Other Assumptions The dose acceptance criteria for the Spent Fuel Cask Drop are assumed to be the same as those for the Fuel Handling Accident.

2.7.4. Methodology The input assumptions used in thle dose consequence analysis of the Cask Drop are provided in Table 2.7-1.

100% of the gap activity specified in Table 3 of RG 1.183 is assumed to be instantaneously released from all of the fuel assemblies into the fuel pool. A minimum water level of 23 feet is maintained above the damaged fuel for the duration of the event. 100% of the noble gas released from the damaged fuel is assumed to escape from the pool. All of the non-iodine particulates released from the damaged fuel are assumed to be retained by the pool. The iodine released from the damaged fuiel is assumed to be composed of 99.85% elemental and 0. 15% organic. The activity released from the pool is assumed to leak to the environment over a two-hour period.

The source term meets the requiremnents of Regulatory Position I of Appendix 13to RG 1.183. Section 1.7.6 discusses the development of the Cask Drop source terms for both cases, wvhich are listed in Table 1.7.6-I.

For Cases I and 2, Control Room ventilation is assumed to already be in the emergency recirculation mode at the start of the event. 100 cfm of unfiltered leakage is assumed throughout the event. For Case 3, Control Room Ventilation remains unisolated. The Control Room ventilation system operational modes are summarized in Table 1.6.3-1). The Control Room ventilation filter efficiencies that are applied to the filtered makeup and recirculation flows are 99% for particulate, elemental iodine, and organic iodine. For the cases listed above wvhich credit filtration by the Fuel Handling Building charcoal filters, efficiencies that are applied to the release are 94% for the elemental and organ~ic iodine. All of the particulates are filtered out by the water in the pools.

2.7.5. Radiological Consequences The atmospheric dispersion factors (X/Qs) used for this event for the Control Room dose are based on the release path from the FHB and the pathway into the control room. These A'/Qs are summarized in Table 1.8.1-2 and Table 1.8.1-3.

For the EAB dose calculation, the 0-2 hour X/Q factors are used for the entire event, and for the LPZ dose calculation, the X/Q factors from Table 1.8.2-I are used.

The radiological consequences of the Spent Fuel Cask Drop are analyzed using the RADTRAD-NAI code and the inputs/ass umpt ions previously discussed. As shown in Table 2.7-2, the results for EAB dose, LPZ dose, and Control Room dose are all within the appropriate regulatory acceptance criteria.

Numerical Applications, Inc. Page 41 of 84 AST Licensing Technical Report for Palisades Report Number: NAI-l 1149-027 2.8. Environmental Qualification (EQ)

The Palisades FSAR, Appendix 7C, discusses equipment EQ due to a radiation environment. RG 1.183, Regulatory Position 6, allows the licensee to use either the AST or TID-14844 assumptions for performing the required EQ analyses until such time as a generic issue related to the effect of increased cesium releases on EQ doses is resolved. The Palisades EQ analyses wvill continue- to be based on TID-14844 assumptions at this time.

3.0 Summary of Results Results of the Palisades radiological consequence analyses using the AST methodology and the corresponding allowvable control room unfiltered inleakage are sumnmarized on Table 3-1.

4.0 Conclusion Full implementation of the Alternative Source Term methodology, as defined in Regulatory Guide 1.183 and Regulatory Issue Summary 2006-04, into the design basis accident analysis has been made to support control room habitability. Analysis of the dose consequences of the Loss-of-Coolant Accident (LOCA),

Fuel Handling Accident (FHA), Main Steam Line Break (MSLB), Steam Generator Tube Rupture (SGTR),

Small Line Break Outside Containment (SLBOC), Control Rod Ejection (CRE) Ejection, and Spent Fuel Cask Drop have been made using the RG 1.183 methodology. The analyses used assumptions consistent wvith proposed changes in the Palisades licensing basis and the calculated doses do not exceed the defined acceptance criteria.

This report supports a maximum allowvable control room unfiltered air inleakage of 10 cfm.

5.0 References 5.1 Palisades FSAR through Revision 25.

5.2 TID-14844, "Calculation of Distance Factors for Power and Test Reactor Sites," March 23, 1962.

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

5.4 Code of Federal Regulations, I0OCFR5O.67, "Accident Source Term," revised 12/03/02.

5.5 NEI 99-03, "Control Room Habitability Guidance," Nuclear Energy Institute, Revision 0 dated June 2001 and Revision I dated March 2003.

5.6 NRC Generic Letter 2003-0 1, "Control Room Habitability," June 12, 2003.

5.7 Federal Guidance Report No. I1I (FGR 11), "Limiting Values of Radionuclide Intake and Air Concentration and Dose Conversion Factors for Inhalation, Submersion, and Ingestion," 1989.

5.8 Federal Guidance Report No. 12 (FGR 12), "External Exposure to Radionuclides in Air, Water, and Soil," 1993.

5.9 Palisades Technical Specifications through Amendment 205.

5.10 Shearon Harris Issuance of Amendment (IA) and Safety Evaluation (SE) for Amendment No. 107 to NPF-63 issued October 12, 200 1.

5.11 Crystal River Issuance of Amendment (IA) and Safety Evaluation (SE) for Amendment No. 199 to DPR-72 issued September 17,2001.

5.12 Fort Calhoun Issuance of Amendment (IA)and Safety Evaluation (SE) for Amendment No. 201 to DPR-40 issued December 5, 200 1.

5.13 Letter from G. Van Middlesworth to USNRC, "Duane Arnold Energy Center, Docket No: 50-33 1, Op. License No: DPR-49, Technical Specification Change Request (TSCR-037): "Alternative Source Term"," October 19, 200 1.

5.14 Duane Arnold Issuance of Amendment (IA) and Safety Evaluation (SE) for Amendment No. 240 to DPR-49 issued July 31, 2001.

5.15 ARCON96 Computer Code ("Atmospheric Relative Concentrations in Building Wakes,"

NUREG/CR-633 1, Rev. I, May 1997, RSICC Computer Code Collection No. CCC-664 and July 1997 errata).

5.16 MicroShield Version 5 "User's Manual "and "Verification & Validation Report, Rev. 5," Grove Engineering, both dated October 1996.

5.17 Oak Ridge National Laboratory, CCC-371, "RSICC Computer Code Collection - ORIGEN 2.1,"

May 1999.

5.18 "PAVAN: An Atmospheric Dispersion Program for Evaluating Design Bases Accident Releases of Radioactive Material from Nuclear Powver Stations," NUREG/CR-2858, November 1982, (RSICC Computer Code Collection No. CCC-445).

5.19 Numerical Applications Inc., NAI-9912-04, Revision 4, "RADTRAD-NAI Version I.lIa(QA)

Documentation," October 2004.

5.20 Numerical Applications Inc., "Dose Methodology Quality Assurance Procedures," Revision 1, June 4, 2001.

5.21 NRC Regulatory Guide 1.194, "Atmospheric Relative Concentrations for Control Room Radiological Habitability Assessments at Nuclear Power Plants," June 2003.

5.22 USNRC, Regulatory Issue Summary 2006-04, Experience With Implementation of Alternate Source Terms, March 2006..

5.23 USNRC, Regulatory Guide 1.145, "Atmospheric Dispersion Models for Potential Accident Consequence Assessments at Nuclear Power Plants," Rev. 1, February 1983.

5.24 USNRC, Regulatory Guide 1.23, "Onsite Meteorological Programs," February 1972.

5.25 Letter from A. C. Bakken, III to USNRC, "Donald C. Cook Nuclear Plant Units I and 2, Partial Response to Second Nuclear Regulatory Commission Request for Additional Information Regarding License Amendment Request for Control Room Habitability," January 15, 2002.

5.26 D. C. Cook Issuance of Amendments (IA)and Safety Evaluation (SE) for Amendment No. 271 to DPR-58 and Amendment No. 252 to DPR-74 issued November 14, 2002.

Numerical Applications, Inc. Pagve 43 of 84 AST Licensing Technical Report for Palisades Report Number: NAI-1 1149-027 5.27 NUREG-0800, USNRC, "Standard Review Plan for the Review of Safety Analysis Reports for Nuclear Powver Plants," September 1981 (or updates of specific sections).

5.28 Industry Degraded Core Rulemaking Program Technical Report 11.3, "Fission Product Transport in Degraded Core Accidents," Atomic Industrial Forum, December 1983.

5.29 Indian Point Nuclear Generating Unit No. 2 Issuance of Amendment (IA) and Safety Evaluation (SE) for Amendment No. 211 to DPR-26 issued July 27, 2000.

5.30 Letter from A. Schwencer (NRC) to D. Bixel (CPCo), "Transmittal of Amendment No. 31 and Safety Evaluation", November 1, 1977.

5.31 Reference deleted.

5.32 Fort Calhoun Issuance of Amendment (IA) and Safety Evaluation (SE) for Amendment No. 198 to DPR-40 issued April 4, 2001.

5.33 USNRC, Regulatory Guide 1.52, "Design, Inspection, and Testing Criteria for Air Filtration and Adsorption Units of Post-Accident Engineered-Safety-Feature Atmosphere Cleanup Systems in Light-Water-Cooled Nuclear Power Plants," Rev. 3, June 200 1.

5.34 NRC Generic Letter 99-02, "Laboratory Testing of Nuclear-Grade Activated Charcoal," June 3, 1999.

5.35 NRC Information Notice 9 1-56, "Potential Radioactive Leakage to Tank Vented to Atmosphere,"

September 19, 1991.

5.36 Letter from J. Scarola to USNRC, "Shearon Harris Nuclear Power Plant, Docket No. 50-400/License No. NPF-63, Revision to Alternate Source Term Methodology Analyses in Support of the Steam Generator Replacement and Powver Uprate License Amendment Applications,:

August 17, 2001.

5.37 NUREG/CR-5950, "Iodine Evolution and pH Control," December 1992.

5.38 Reference deleted.

5.39 NAI Calculation Number NAI-l 10 1-002 Rev. 0, "Qualification of ORIGEN2.l for Florida Powver

& Light AST Applications," August 8, 2002.

5.40 G. Burley, "Evaluation of Fission Product Release and Transport," Staff Technical Paper, 197 1.

5.41 ORNL RSICC Code Package CCC-565, "QADMOD-GP: Point Kernel Gamma-Ray Shielding Code with Geometric Progression Buildup Factors", November 1990.

Numerical Applications, Inc. Page 44 of 84 AST Licensing Technical Report for Palisades Report Number: NAI-I 1149-027 Figure 1.8.1-1 Onsite Release-Receptor Location Sketch (Plant North) <-----~

22*44' 4 N

JI Normal Inta G

'A' and 'B' 11 (Not to scale)

  • - Control Room Intakes / Receptor Point A- Containment Closest Point B- SIRWT Vent C- Plant Stack D- Closest ADV E- Closest SSRV F- Equipment Door G- Turbine Building NE Roof Exhauster H- Turbine Building NW Roof Exhauster

Numerical Applications, Inc. Page 45 of 84 AST Licensing Technical Report for Palisades Report Number: NAI-1 149-027 Table 1.6.3-1 Control Room Ventilation System Parameters Parameter 1Value Control Room Volume ]35,923 ft 3 Normal Operation Filtered Make-up Flowv Rate 0 cfm Filtered Recirculation Flow Rate 0 cfm UnfiterdFlw Rte ad Ileaage Mke-u 660 cfin UnfiterdFl~v Mae-u Rte nd Ileaage 384.2 cfmn during periods wvithout offsite power Emergency Operation Recirculation Mode:

Filtered Make-up Flowv Rate 1413.6 cfmn Filtered Recirculation Flow Rate 1413.6 cfm Unfiltered Make-up Flowv Rate 0 cfm Unfiltered Inleakage (LOCA, MSLB, GRE) 10 cfm Unfiltered Inleakage (SGTR, FHA, Cask 100 cfmn Drop, Small Line Break)

Filter Efficiencies Elemental 99%

Organic 99%

Particulate 99%

Numerical Applications, Inc. Page 46 of 84 AST Licensing Technical Report for Palisades Report Number: NAM- 149-027 Table 1.6.4-1 LOCA Direct Shine Dose Source Direct Shine Dose (rem)

Containment 0.028 CR Filters 0.005 Purge Line 0.000 External Cloud 0.232 SIRWT 1.269 Total 1.534

Numerical Applications, Inc. Page 47 of 84 AST Licensing Technical Report for Palisades Report Number: NAI-1 1149-027 Table 1.7.2-1 Primary Coolant Source Term Nuclide pCi/gm Nuclide pLCigm KR-83M 2.3 4913-01 BA-140 5.211 E-03 KR-85M 8.808E-01 LA-140 2.496E-03 KR-85 3.89013-01 CE- 141 8.074E-04 KR-87 5.505E-01 CE- 143 4.624E-04 KR-88 1.6 15E+00 CE- 144 6.826E-04 XE- 131 M 3.303E-0 1 PR- 143 7.340E-04 XE- 133M 1.24813+00 MN-5-4 I1.762E-03 XE- 133 5.138E+01 FE-55 1.321 E-03 XE- 135 5.432E+00 CO-513 5.138E-03 BR-83 4.844E-02 ZR-97 3.597E-04 BR-84 I1.982E-02 RU-l05 I1.028E-04 RB-86 1.101 E-02 RH-105 4.037E-04 CS-134 3.5 97E+0 I SB3-127 3.1 56E-02 CS-136 3.1 56E+00 SB-I129 I1.762E-02 CS-137 I1.908E+01I XE- 135M 3.523E-01 CS-138 5.138E-01 CS-134M 3.376E-02 SR-89 3.964E-03 RB-88 I1.688E+00 SR-90 3.597E-04 RB-89 4.257E-02 CR-51 3.450E-03 SB-12.4 8.074E-04 FE-59 3.303 E-04 SB- 125 6.753E-03 CO-60 5.872E-04 SB- 126 4.55 1E-04 SR-91 1.10113-03 TE- 131 1.101 E-02 SR-92 4.33 1 E-04 TE-133M 8.808E-03 Y-90 4.698E-04 BA- 141 6.679E-05 Y-91I M 6.533E-04 BA- 139 3.1 56E-04 Y-91I 1.32 1IE-02 LA- 141 I1.762E-04 Y-92 5.285E-04 LA-142 4.991 E-05 Y-93 3.230E-04 ND-147 3.1 56E-04 ZR-95 8.808E-04 NB-97 5.945E-05 NB-95 8.808E-04 NB-95M 6.239E-06 MO-99 3.083E+00 PM- 147 8.808E-05 TC-99M 2.349E+00 PM- 148 I1.248E-04 RU-103 8.808E-04 PM- 149 2.202E-04 RU- 106 3.890E-04 PM-I15 6.092E-05 RH-1 03M 8.808E-04 PM-148M I1.908E-05 TE-1 25M I1.468E-03 PR- 144 6.826E-04 TE-1 27M 5.285E-03 Y-94 I1.028E-05 TE-127 3.37613-02 BR-82 I1.835E-02

Numerical Applications, Inc. Page 48 AST Licensing Technical Report for Palisades of 841 Report Number: NA1-1 1149-027 TE-129M 1.615E-02 1-131 0.8305 TE-129 2.789E-02 1-132 0.1917 TE-131M 2.716E-02 1-133 0.8624 TE-132 3.523E-01 1-134 0.0751 TE-134 1.541 E-02 1-135 0.3673

  • The iodine activities have been adjusted to the Tech. Spec. limit of 1.0 pCi/gm DE 1-13 1. Non-iodine activities have been adjusted to the Tech. Spec. limit of 100/E-bar.

Numerical Applications, Inc. Page 49 of 84 AST Licensing Technical Report for Palisades Report Number: NAM- 149-027 Table 1.7.3-1 Secondary Side Source Term Isotope g.Ci/gm 1-131 0.0830 1-132 0.0192 1-133 0.0862 1-134 0.0075 1-135 0.0367

Numerical Applications, Inc. Page 50 of 84 AST Licensing Technical Report for Palisades Report Number: NAM- 149-027 Table 1.7.4-1 LOCA Source Term

Numerical Applications, Inc. Page 51 of 84 AST Licensing Technical Report for Palisades Report Number: NA1-1 1149-027 Nuclide Curies Nuclide Curies Xe-133 0.1466E+09 Pm-148 0.2144E+08 Xe-135 0.4692E+08 Pm-149 0.45412E+08 Cs-134 0.2037E+08 Pm-IS! 0.1606E+08 Cs-136 0.5873E+07 Pm-148m 0.2999E+07 Cs-137 0. 11 O0E+08 Pr-144 0. 10252E+09 Ba-139 0.1307E+09 Pr-144m 0. 1224E+07 13a-140 0.1260E+09 Sm-153 0.4423E+08 La-140 0.1299E+09 Y-94 0. 1105 E+09 La-141 0. 11932E+09 Y-95 0. 1183 E+09 La- 142 0. 1156E+09 Y-91nm 0.5151 E+08 Ce- 141 0.1212E+09 Br-82 0.5282E+06 Ce-I 43 0. 11152E+09 Br-83 0.91 02E+07 Ce- 144 0. 1020E+09 Br-84 0.1591E+08 Pr-143 0. 11112E+09 Am-242 0.9062E+07 Nd-147 0.4770E+08 Np-238 0.4306E+08 Np-239 0.1830E+10 Pu-243 0.4690E+08 Pu-238 0.3927E+06II

Numerical Applications, Inc. Page 52 AST Licensing Technical Report for Palisades of 841 Report Number: NAI- 149-027 Table 1.7.5-1 Fuel Handling Accident Source Term Bounding Bounding Bounding Nuclide Activities Nuclide Activities Nuclide Activities (Curies) (Curies) (Curies)

Co-58 0.00002+00 1-135 0.8949E4 04 Sb- 126 0.9900E+03 Co-60 0.00002+00 Xe-133 0.1298E407 Te- 131 0.8307E+04 Kr-85 0.1 052E+05 Xe-135 0.820124+05 Te- 133 0.20342-10 Kr-85m 0. 1174E+03 Cs- 134 0.2034E406 Te- 134 0.22 17E-14 Kr-87 0.1647E-05 Cs- 136 0.5284E2405 Te-125m 0.34 17E+04 Kr-88 0.4302E+01 Cs- 137 0. 11002E406 Te-133m 0. 12132E-09 Rb-86 0. 18 19E+04 Ba- 139 0.48612E-04 Ba- 141 0.00002+00 Sr-89 0.7020E+06 Ba- 140 0.1 130E407 Ba- 137m 0. 10412E+06 Sr-90 0.8456E+05 La- 140 0.1235E407 Pd-109 0.2825E+05 Sr-91 0.2679E+05 La-141 0.2730E403 Rh-106 0.5 7712E+06 Sr-92 0.4453E401 La- 142 0.5794E-03 Rhi-I03m 0.1097E+07 Y-90 0. 8623E2+05 Ce-I141 0.1 168E407 Tc- 101 0.00002+00 Y-91I 0.9 107E+06 Ce- 143 0.4 1002+406 Eu- 154 0.1246E+05 Y-92 0.3229E+03 Ce-144 0. 1014E+07 Eu- 155 0.8442E+04 Y-93 0.4 137E+05 Pr- 143 0.1071E+07 Eu-156 0.19352+06 Zr-95 0.12 10E+07 Nd- 147 0.42112E406 La- 143 0.00002+00 Zr-97 0.1 684E+06 Np-239 0.1023E2408 Nb-97 0.16922+06 Nb-95 0.1248E+07 Pu-238 0.44942404 Nb-95m 0.8748E+04 Mo-99 0.8264E+06 Pu-239 0.3578E403 Pm- 147 0.12962+06 Tc-99m 0.7956E+06 Pu-240 0.5406E403 Pm-148 0.1659E+06 Ru- 103 0. 1216E+07 Pu-241 0.1522E406 Pm- 149 0.24812E+06 Ru-1OS 0.5426E+03 Am-241 0. 18972403 Pm- 15 1 0.50 12E+05 Ru-106 0.57712E+06 Cm-242 0.5649E405 Pm-148m 0.2899E405 Rh-lo05 0.3958E+06 Cm-244 0. 1339E405 Pr- 144 0. 1015E2+07 Sb- 127 0.6450E+05 1-130 0.2546E404 Pr- 144m 0.12 17E+05 Sb- 129 0.1 176E+03 Kr-83m 0.3727E400 Sm-153 0.2 171 E+06 Te- 127 0.7344E+05 Xe-138 0.00002400 Y-94 0.00002+00 Te-127m 0.1222E+05 Xe-131m 0.8276E+04 Y-95 0.00002+00 Te- 129 0.2383E+05 Xe- 133m 0.34032+05 Y-91 m 0. 17022+05 Te- 129m 0.3637E+05 Xe- 135m 0.14342+04 Br-82 0.2060E+04 Te- 131 m 0.3690E+05 Cs- 138 0.00002+00 Br-83 0.8833E-01 Te-132 0.6852E+06 Cs- 134m 0.5 122E+00 Br-84 0.00002+00 1-131 0.6424E+06 Rb-88 0.4804E+01 Am-242 0. 11382E+05 1-132 0.7060E+06 Rb-89 0.00002+00 Np-238 0.2238E+06 1-133 0.3019E+06 Sb-124 0.1663E+04 Pu-243 0.56812E+03 1-134 0.208713-09 Sb- 125 0.1566E+05

Numerical Applications, Inc. Page 53 of 84 AST Licensing Technical Report for Palisades Report Number: NAM. 149-027 Table 1.7.6-1 Spent Fuel Cask Drop Source Terms*

Listed source term is for a single assembly.

30 Day 90 Day 1 Nuelide Decay Decay j(Curies) (Curies)

Co-58 0.000013+00 0.0000E3+00 Co-60 0.0000E+00 0.000013+00 Kr-85 0.6563E3+04 0.6493E3+04 Kr-85m 0.000013+00 0.000013+00 Kr-87 0.0000E+00 0.000013+00 Kr-88 0.00001E+00 0.000013+00 Rb-86 0.3205E+03 0.3450E3+02 Sr-89 0.5045E+06 0.2213E3+06 Sr-90 0.5 178E3+05 0.5157E3+05 Sr-9 I 0. 1431 E-I16 0.000013+00 Sr-92 0.000013+00 0.000013+00 Y-90 0.5184E3+05 0.5 159E3+05 Y-91I 0.6805E-+06 0.3344E3+06 Y-92 0.000013+00 0.000013+00 Y-93 0.39641E- 15 0.000013+00 Zr-95 0.8884E3+06 0.4637E3+06 Zr-97 0.1774E3-06 0.000013+00 Nb-95 0.1 146E3+07 0.7766E3+06 Mo-99 0.67 14E3+03 0.18 16E-03 Tc-991n 0.6469E3+03 0.1749E3-03 Ru-I03 0.610213+06 0.2118E3+06 Ru- 105 0.000013+00 0.000013+00 Ru-106 0.2925E3+06 0.2611 E+06 Rh-1OS 0.5545E3+00 0.3056E3-12 Sb- 127 0.3350E3+03 0.6807E3-02 Sb- 129 0.000013+00 0.000013+00 Te-1 27 0.8360E3+04 0.5490E3+04 Te-1 27m 0.8207E3+04 0.5606E3+04 Te-129 0.1138E3+05 0.3301 E+04 Te-I1291n 0.1748E3+05 0.50711E+04 Te-1 31 ii 0.5959E3-02 0.2118E3-16 Te- 132 0.166413+04 0.4757E3-02 1-131 0.5355E3+05 0.3038E3+03 1-132 0.17 15E3+04 0.4902E3-02 1-133 0.5526E3-04 0.7976E3-25 1-134 0.000013+00 0.000013+00

Numerical Applications, Inc. Page 54 of 84 AST Licensing Technical Report for Palisades Report Number: NAI-1 1149-027 30 Day 90 Day Nuclide Decay Decay (Cu ries) (Cu ries) 1-135 0.000013+00 0.000013+00 Xe-133 0.3313E3+05 0.1194 E+02 Xe-135 0.OOOOE+00 0.000013+00 Cs- 134 0.10 16E+06 0.9612E3+05 Cs- 136 0.6310E3+04 0.2638E3+03 Cs-137 0.6483E3+05 0.6459E3+05 Ba- 139 0.OOOOE+00 0.000013+00 13a-l40 0.2450E3+06 0.9480E3+04 La- 140 0.281 9E+06 0. 10911E+05 La- 141 0.000013+00 0.000013+00 La-142 0.000013+00 0.000013+00 Ce- 141 0.6297E+06 0. 1752E3+06 Ce- 143 0.3033E+00 0.2220E3-13 Ce-I 144 0.8358E3+06 0.7222E3+06 Pr- 143 0.2630E-+06 0. 1226E3+05 Nd- 147 0.7 146E3+05 0.1663E3+04 Np-239 0.2042E-+04 0. 1581E3+02 Pu-238 0.21 74E+04 0.2199E3+04 Pu-239 0.2968E3+03 0.2968E3+03 Pu-240 0.2972E+03 0.2972E3+03 Pu-241 0.8941 E+05 0.8870E3+05 Am-241 0.1070E3+03 0. 1260E-+03 Cm-242 0.2428E3+05 0.188213+05 Cm-244 0.2743E+04 0.2726E3+04 1-130 0.5151 E-13 0.000013+00 Kr-83m 0.000013+00 0.000013+00 Xe-138 0.000013+00 0.000013+00 Xe-I 31 m 0.2962E+04 0.120413+03 Xe-133m 0.53 77E+01I 0.3038E3-07 Xe-135m 0.000013+00 0.0000E3+00 Cs- 138 0.000013+00 0.0000E3+00 Cs-134m 0.OOOOE+00 0.000013+00 Rb-88 0.000013+00 0.000013+00 Rb-89 0.000013+00 0.000013+00 Sb-124 0.5599E3+03 0.2806E3+03 Sb- 125 0.8790E3+04 0.8446E3+04 Sb- 126 0.1278E3+03 0.4525E3+01 Te- 131 0. 1341 E-02 0.4765E3-17 Te- 133 0.000013+00 0.000013+00 Te- 134 0.000013+00 0.000013+00

Numerical Applications, Inc. Page 55 AST Licensing Technical Report for Palisades of 841 Report Number: NAl- 1149-027 30 Day 90 Day Nuclide Decay Decay (Cuiries) (Cu ries)

Te-125m 0.1902E+04 0. 1940E+04 Te-1I33m 0.0000E+00 0.0000E+00 13a- 141 0.OOOOE+00 0.00002+00 Ba-137m 0.6132E+05 0.61102E+05 Pd- 109 0. 1499E- 10 0.00002+00 Rh- 106 0.2925E+06 0.26112E+06 Rh-103m 0.5502E+06 0.1908 E+06 Tc-1 01 0.00002+00 0.00002+00 Eu- 154 0.63 12E+04 0.6229E+04 Eu-I 55 0.4225E+04 0.41292+04 Eu-156 0.2550E+05 0. 1649E+04 La-143 0.0000E+00 0.00002+00 Nb-97 0. 1911 E-06 0.00002+00 Nb-95m 0.6589E+04 0.34412E+04 Pm-147 0. 1206E+06 0. 11622E+06 Pm-148 0.4604E+04 0.32212E+03 Pmn-149 0.3193E+02 0.2179E-06 Pm- 151 0.3097E-02 0.1665E- 17 Pin-148m 0. 1557E+05 0.5685E+04 Pr- 144 0.8358 E+06 0.7222E+06 Pr- 144m 0. 1003 E+05 0.8666E+04 Sm- 153 0.5430E+01 0.2827E-08 Y-94 0.00002+00 0.0000E+00 Y-95 0.00002+00 0.00002+00 Y-91 M 0.9094E- 17 0.00002+00 Br-82 0. 19 19E-02 0.10 12E-14

[lr-83 O.OOOOE+00 0.0000E+00 Br-84 0.00002+00 0.00002+00 Am-242 0.1237E+02 0. 1236E+02 Np-238 0. 1195 E+02 0.62112E-01I Pu-243 0.2743E-06 0.2743E-06

Numerical Applications, Inc. Page 56 of 84 AST Licensing Technical Report for Palisades Report Number: NAl-I 1149-027 Table 1.8.1-1 Release-Receptor Combination Parameters for Analysis Events Release Release Receptor Receptor Direction Release Point Receptor Point Height Height Height Height Distance Distance repetht

() (in) (ft) (in) Mt (in) true North Clontinent Normal Control Pontinmet Room Intake "A" 73.8 22.5 73.8 22.5 79.4 24.2 168 Clontinent Normal Control Pontinmet Room Intake "B" 73.8 22.5 73.8 22.5 70.5 21.5 174 Closest Emergency Containment Control Room 48.9 14.9 48.9 14.9 312 95.1 202 Point Intake SIRW Tank Vent NormalIntake Room Control "A" 80.4 24.5 73.8 22.5 3.

348 10.6 157 SIRW Tank Vent Normal Control Room Intake "B" 80.4 24.5 73.8 22.5 25.3 7.7 184 Emergency SIRW Tank Vent Control Room 80.4 24.5 48.9 14.9 266 81.2 214 Intake Normal Control Stack Vent Room Intake "A" 192 58.5 73.8 22.5 73.2 22.3 169 Normal Control 7.

Stack Vent Room Intake "B" 192 58.5 738 22.5 65.0 19.8 181 Emergency Stack Vent Control Room 192 58.5 48.9 14.9 320 97.4 209 Intake Normal Control 6.

ADV Room Intake "A" 57.1 17.4 73.8 22.5 660 20.1 192 Normal Control 6.

ADV Room Intake "B" 57.1 17.4 73.8 22.5 65019.8 207 Emergency ADV Control Room Intake 57.1 17.4 48.9 14.9 329 100.4 214 SSRV Normal Control (Est Bnk Room Intake "A" 57.1 17.4 73.8 22.5 57.7 17.6 187 SSRV Normal Control (West Bank) Room Intake "B" 57.1 17.4 73.8 22.5 55.8 17.0 204

Numerical Applications, Inc. Page 57 of 84 AST Licensing Technical Report for Palisades Report Number: Nl-I 1149-027 Release Release Receptor Receptor wiretho Release Point Receptor Point Height Height Height Height Distance Distance repetht Mt (in (ft) (in Mt (in true North Emergency SSRV Control Room 57.1 17.4 48.9 14.9 318 96.8 213 (East Bank) Intake Containment Normal Intake Equipment Door 66.9 20.4 73.8 22.5 87.9 26.8 152 Containment Normal Intake Equipment Door 'B' 66.9 20.4 73.8 22.5 75.8 23.1 161 Containment Emergency Equipment Door Intake 66.9 20.4 73.8 22.5 313 95.4 204 Turbine Building Normal Intake NE Roof'A899 2. 738 2. 682 28 25 Exhauster899 2. 738 2. 682 08 25 Turbine Building Normal Intake NE Roof B899 2. 738 2. 827 22 26 Exhauster ' 99 2. 38 2. 27 2. 6 Turbine Building Emrec NE Roof Emtaergency4 8. 1. 325 9. 2 Exhauster Inae 8. 274 4. 14991 27 Turbine Building Normal Intake NW Roof 89.9 27.4 73.8 22.5 73.2 22.3 266 Exhauster _____ __________

Turbine Building Normal Intake NW Roof 'B' 89.9 27.4 73.8 22.5 88.6 27.0 271 Exhauster _____ __________

Turbine Building Emergency NW Roof Intake 89.9 27.4 48.9 14.9 324 98.6 229 Exhauster _____ _____ ___________ _____ _____

Numerical Applications, Inc. Page 58 of 84 AST Licensing Technical Report for Palisades Report Number: NA1-1 1149-027 Table 1.8.1-2 Onsite Atmospheric Dispersion (X7Q) Factors for Analysis Events Release - Receptor 0-2 hr 2-8 hr 8-24 hr 1-4 days 4-30 days Receptor Release Point Point XIQ XIQ XIQ XIQ XIQ Pair Containment Normal Intake 14E0 .1-241E0 .3-3 24E0 A IClosest Point 1B. 14E0 1.10241E0I32E03 .9E3 BContainment Emergency 7.26E-04 6.18E-04 2.47E-04 1.77E-04 1.30E-04 B Closest Point Intake et IVr Vent'B CC SIRW Normal Intake 9.57E-02 7.59E-02 2.87E-02 2.19E-02 1.65E-02 DSIRWTrVent Emergency 9.66E-04 7.92E-04 3.13E-04 2.20E-04 1.64E-04 NomIntake E Plant Stack NomB'ntk 6.1 OE-03 4.32E-03 1.73E-03 1.27E-03 9.79E-04 FPlant Stack Emergency 8.32E-04 7.69E-04 2.83E-04 2.15E-04 1.57E-04 NomIntake G Closest ADV Nora'ntk 1.65E-02 1.34E-02 5.40E-03 4.03E-03 2.98E-03 HClosest ADV Emergency 7.36E-04 6.42E-04 2.43E-04 1.75E-04 1.28E-04 NomIntake I Closest SSRV Noral Intake 4.98E-03 3.72E-03

'B' K Closest SSRV Emergency 7.96E-04 6.g1lE-04 2.60E-04 1.90E-04 1.37E-04 Containment Normal Intake 1.25E-02 9.83E-03 3.62E-03 2.86E-03 2.28E-03 LL Equipment Door 'B' Emergency 7.32E-04 6.13E-04 2.45E-04 1.75E-04 1.2gE-04 MMContainment Equipment Door Intake Turbine Building Normal Intake1.1E2 1.3-246E3 28E03 .6-3 N NE Roof'A1.1E0 1.30246E0 2.703 .6E3 Exhauster'A Turbine Building Emrec 0 NE Roof Emtaerec-- 2.58E-04 --

_________ Exhauster Itk Turbine Building Emrec 7.90463E4 P NW Roof Emtaerec 9g-464E0 - 1.75E-04 1.32E-04

______ Exhauster Intake

Numerical Applications, Inc. Page 59 of 84~

AST Licensing Technical Report for Palisades Report Number: NAt-I 1149-027 Table 1.8.1-3 Release-Receptor Point Pairs Assumed for Analysis Events Event LOCA: Prior to CR Isolation Following CR Isolation

- Containment Leakage A B

- ECCS Leakage E F

- SIRWT Backleakage C D FIIA

- Containment Release L M

- FHB Release E F Cask Drop Cases I & 2 Case 3 Filtered Release - Unfiltered E N/A Makeup and Inleakage ____________

Unfiltered Release - Unfiltered L L Makeup and Inleakage Filtered Release - Filtered F N/A Makeup____________

Unfiltered Release - Filtered M N/A Makeup NISLB:

- Break Release N 0 &1P

- MSSV/ADV Release G H J&G K&H SGTR Initial release via SSRVs Initial release via SSRVs switching to AD\'s switching to ADVs RCCA Ejection:

- Containment Leakage A B J&G K &,H

- Secondary Side Release Initial release via SSRVs Initial release via SSRVs switching to AD\1s swvitching to ADVs Small Line Break Outside F E Containment

Numerical Applications, Inc. Page 60 of 84 AST Licensing Technical Report for Palisades Report Number: NA1-1 1149-027 Table 1.8.2-1 Offsitc Atmospheric Dispersion (X/Q) Factors for Analysis Events Time Period EAB X/Q (sec/rn3 ) LPZ X/Q (sec/rn 3) 0-2 hours 5.39E-04 6.66E-5 0-8 hours 3.3 1E-04 3.03E-5 8-24 hours 2.59E-04 2.04E-5 1-4 days 1.53E-04 8.67E-6 4-30 days 7.14E-05 J 2.54E-6 The above table summarizes the maximum X/Q factors for the EAB and LPZ. Note that the 0-2 hour EAB XIQ factor wvas used for the entire event.

Numerical Applications, Inc. Page 61 of 84 AST Licensing Technical Report for Palisades Report Number: NAM- 149-027 Table 2.1-1 Loss of Coolant Accident (LOCA) - Inputs and Assumptions InputAssumtionValue Release Inputs:

Core Powver Level 2703 M\V 1h Core Average Fuel Burnup 39,300 MWDIMTU Fuel Enrichment 3.0 - 5.0 wv/o Initial PCS Equilibrium ActivityTal1.2-(1.0 pCilgm DE 1-131 and I00/E-bar gross activity)Tal1.2-Core Fission Product Inventory Table 1.7.4-1 Containment Leakage Rate 0 to 24 hours2.777778e-4 days <br />0.00667 hours <br />3.968254e-5 weeks <br />9.132e-6 months <br /> 0. 10% (by wveight)fday after 24 hours2.777778e-4 days <br />0.00667 hours <br />3.968254e-5 weeks <br />9.132e-6 months <br /> 0.05% (by weight)/day LOCA release phase timing and duration Table 2.1-2 Core Inventory Release Fractions (gap release and early in- RG 1.183, Sections 3.1, 3.2, and Table 2 vessel damage phases)

ECCS Systems Leakage (from 19 minutes to 30 days)

Sump Volume (minimum) 39,054 ft.'

ECCS Leakage (2 times allowed value) 0.053472 fi 3/Min Flashing Fraction Calculated - 0.03 Used for dose determination - 0. 10 Chemical form of the iodine released from the ECCS 97% elemental, 3% organic leakage Iodine Decontamination Factor 2 (based on current design basis)

No credit taken for dilution or holdup

Numerical Applications, Inc. Page 62 of 84 AST Licensing Technical Report for Palisades Report Number: NAM- 149-027 Input/Assumption Value SIRWT Back-leakace (from 19 minutes to 30 days)

Sump Volume 292,143 gallons (minimum valve for ECCS leakage, maximizes sump iodine concentration) 430,708 gallons (maximum value for SIRWT backleakage to be consist wvith assumption of minimum water level in SIRWT)

ECCS Leakage to SIRWT (2 times allowed value) 4.4 gpm until 2 hours2.314815e-5 days <br />5.555556e-4 hours <br />3.306878e-6 weeks <br />7.61e-7 months <br /> into the event, then 0.05 gpm Flashing Fraction (elemental iodine assumed to be released 0 % based on temperature of fluid reaching SIRWT into tank space based upon partition factor)

Table 2.1-7 SIRWT liquid/vapor elemental iodine partition factor Table 2.1-6 Elemental Iodine fraction in SIRWT Initial SIRWT Liquid Inventory (minimum at time of 4,144 gallons recirculation) ____________________________

Release from SIRWT Vapor Space Table 2.1-8 Removal Inputs:

Containment Aerosol/Particu late Natural Deposition (only 0.1/hour credited in unsprayed regions)

Containment Elemental Iodine Wall Deposition 2.3/hour Containment Spray Coverage >90%

Spray Removal Rates:

Elemental Iodine 4.8/hour Time to reach DF of 200 2.5 15 hours1.736111e-4 days <br />0.00417 hours <br />2.480159e-5 weeks <br />5.7075e-6 months <br /> Aerosol 1.8/hour (reduced to 0. 18 at 3.385 hours0.00446 days <br />0.107 hours <br />6.365741e-4 weeks <br />1.464925e-4 months <br />)

Time to reach DF of 50 3.385 hours0.00446 days <br />0.107 hours <br />6.365741e-4 weeks <br />1.464925e-4 months <br /> Spray Initiation Time 60 seconds (0.0 16667 hours)

Control Room Ventilation System Table 1.6.3-1 Time of automatic control room isolation and switch to emergency mode 90 seconds Control Room Unfiltered Inleakage 10 cf'm Transport Inputs:

Containment Leakage Release Containment closest point ECCS Leakage Plant stack

Numerical Applications, Inc. Page 63 of 84 AST Licensing Technical Report for Palisades Report Number: NAMl 149-027 Input/Assumption Value SIRWT Backleakage SIRWT vent Personnel Dose Conversion Inputs:

Atmospheric Dispersion Factors Table 1.8.2-I Of'fsite Table 1.8.1-2 and Table 1.8.1-3 Onsite ____________________________

Breathing Rates RG 1.183 Sections 4.1.3 and 4.2.6 Control Room Occupancy Factor RG 1.183 Section 4.2.6

Numerical Applications, Inc. Page 64 of 84 AST Licensing Technical Report for Palisades Report Number: NAM. 149-027 Table 2.1-2 LOCA Release Phases 7 Phase Onset Duration Ga ees 30 seconds 0.5 ouirs Early In-Vessel 0.5 hours5.787037e-5 days <br />0.00139 hours <br />8.267196e-6 weeks <br />1.9025e-6 months <br /> 1.3 hours3.472222e-5 days <br />8.333333e-4 hours <br />4.960317e-6 weeks <br />1.1415e-6 months <br />

  • From RG 1.183, Table 4 Table 2.1-3 Time Dependent SIRWT p11 TimeSI NI II (hours) R Tp1 0.00 4.500 0.3167 4.500 0.50 4.505 1.00 4.518 2.00 4.544 2.00 4.544 4.00 4.545 8.00 4.546 16.00 4.548 24.00 4.550 48.00 4.557 72.00 4.563 96.00 4.570 120.00 4.576 144.00 4.583 168.00 4.589 192.00 4.595 240.00 4.607 288.00 4.618 336.00 4.630 384.00 4.641 432.00 4.651 528.00 4.672 624.00 4.692 720.00 4.711

Numerical Applications, Inc. Page 65 of 84 AST Licensing Technical Report for Palisades Report Number: NAI-1 149-027 Table 2.1-4 Time Dependent SIRWT Total Iodine Concentration*

Time SIRNNT Iodine Concentration (hours) (gm-atom/liter) 0.00 0.002+00 0.3167 0.00E+00 0.50 5.77E-07 1.00 2.09E-06 2.00 4.84E-06 2.00 4.84E-06 4.00 4.90E-06 8.00 5.02E-06 16.00 5.25E-06 24.00 5.48E-06 48.00 6.16E-06 72.00 6.82E-06 96.00 7.46E-06 120.00 8.0813-06 144.00 8.68E-06 168.00 9.26E-06 192.00 9.83E-06 240.00 1.092-05 288.00 1.20E-05 336.00 1.29E-05 384.00 1.39E-05 432.00 1.48E-05 528.00 1.642-05 624.00 1.79E-05 L-720.00 1.93E-05

.Includes radioactive and stable iodine isotopes

Numerical Applications, Inc. Page 66 of 84 AST Licensing Technical Report for Palisades Report Number: NAT- 149-027 Table 2.1-5 Time Dependent SIRWT Liquid Temperature 0

Time (hr) Temperature V F) 0.00 100.0 0.3 167 100.0 0.50 100.0 1.00 100.0 2.00 100.0 2.00 100.0 4.00 100.5 8.00 101.3 16.00 102.4 24.00 103.2 48.00 104.7 72.00 105.0 96.00 105.0 120.00 104.9 144.00 104.8 168.00 104.8 192.00 104.7 240.00 104.6 288.00 104.6 336.00 104.5 384.00 104.5 432.00 104.5 528.00 104.4 624.00 104.4 720.00 104.4

Numerical Applications, Inc. Page 67 of 84 AST Licensing Technical Report for Palisades Report Number: NAM-149-027 Table 2.1-6 Time Dependent SJRWT Elemental Iodine Fraction Time (hr) Elemental Iodine Fraction 0.00 0.OOE+00 0.3167 O.00E+00 0.50 1.25E-02 1.00 4.07E-02 2.00 7.95 E-02 2.00 7.95E-02 4.00 8.0213-02 8.00 8.16E-02 16.00 8.42E-02 24.00 8.68E-02 48.00 9.38E-02 72.00 1OOE-0 I 96.00 1.06E-0 I 120.00 1.11IE-01 144.00 1.15E-01 168.00 1.19E-01 192.00 1.23E-01 240.00 1.29E-0 1 288.00 1.34E-01 336.00 1.38E-01 384.00 1.41 E-0 I 432.00 1.44E-01 528.00 1.47E-01 624.00 1.49E-01 720.00 1.49E-0 I

Numerical Applications, Inc. Page 68 of 84 AST Licensing Technical Report for Palisades Report Number: NA1-1 1149-027 Table 2.1-7 Time Dependent SIRWT Partition Coefficient Time(hr)Elemental Iodline Partition Time(hr)Coefficient 0.00 45.65 0.3167 45.65 0.50 45.65 1.00 45.65 2.00 45.65 2.00 45.65 4.00 45.21 8.00 44.53 16.00 43.61 24.00 42.95 48.00 41.74 72.00 41.50 96.00 41.50 120.00 41.58 144.00 41.66 168.00 41.66 192.00 41.74 240.00 41.82 288.00 41.82 336.00 41.89 384.00 41.89 432.00 41.89 528.00 41.97 624.00 41.97 720.00 41.97

Numerical Applications, Inc. Page 69 AST Licensing Technical Report for Palisades of 841 Report Number: NAI-1 149-027 Table 2.1-8 Adjusted Release Rate from SIRWT Time Adjusted Iodine Release Rate (hours) (cfm) 0.3 167 5.7048E-04 2.00 1.1955E-05 8.00 1.2895E-05 24.00 1.4921 E-05 72.00 I1.7737E-05 168.00 1.9907E-05 240.00 2.1376E-05 336.00 2.25012E-05 432.00 2.3366E-05 624.00 2.3737E-05 Table 2.1-9 LOCA Dose Consequences CaeEAB Dose"l) LPZ Dose(2 ) Control Room Dose( 2 )

Case(rein TEDE) (rem TEI)E) (rem TEDE)

LOCA 12.76 3.27 4.01 Acceptance Criteria 25 25 5 (1)Worst 2-hour dose (2) Integrated 30-day dose based on an unfiltered Control Room inleakage rate of 10 cfm

Table 2.2-1 Fuel Handling Accident (FHA) - Inputs and Assumptions Input/Assumption Value Core Power Level Before Shutdown 2703 MWih Core Average Fuel Burnup .39,300 MWD/MTU Discharged Fuel Assembly Burnup 39,300 - 58,900 MWD/MTIJ Fuel Enrichment 3.0 -5.0Ow/o Maximum Radial Peaking Factor 2.04 Number of Fuel Assemblies Damaged I fuel assembly Delay Before Spent Fuel Movement 48 hours5.555556e-4 days <br />0.0133 hours <br />7.936508e-5 weeks <br />1.8264e-5 months <br /> FHA Source Term for a Single Assembly Table 1.7.5-1 Water Level Above Damaged Fuel Assembly 22.5 feet minimum Elemental - 252 Iodine Decontamination Factors Organic -I OverallI - 18 3.07 Noble Gas Decontamination FactorI In oolElemental Idin ormof Chemcal - 99.85%

Chemcal Idin Inormof oolOrganic-0.15%

Atmospheric Dispersion Factors Mf~ite Table 1.8.2-1 Onsite Table 1.8.1-2 Table 1.8.1-3 Control Room Ventilation System Time of Control Room Ventilation System 20 minutes Isolation Control Room Unfiltered Inleakag~e 100 cfm Breathing Rates RG 1.183 Sections 4.1.3 and 4.2.6 Control Room Occupancy Factor RG 1.183 Section 4.2.6 Elemental iodine - 94%

Fl-B Ventilation Filter Efficiencies Organic iodine - 94%

_________________________________ 1 Noble gas - n/a Table 2.2-2 Fuel Handling Accident Dose Consequences One Fuel Assembly Dam aed CaeEAB Dose"l) LPZ Dose( 2 ) Control Room Dose(2 )

Case(rem TEDE) (rem TEDE) (rem TEDE)

Elemental iodine DF=285 __________

FHA in Containment 2.20 0.28 4.04 FHA in FHB With 10% of release via FHB 2.02 0.25 3.68 filtered ventilation FHA in FHB With 34% of release via FHB 1.60 0.20 2.81 filtered ventilation FHA in FHB WVith 50% of release via FHB 13 .722 filtered ventilation 13 .722 Acceptance Criteria 6.3 6.3 5

()Worst 2-hour dose (2 Integrated 30-day dose based on an unfiltered Control Room inleakage rate of 100 cfm

Numerical Applications, Inc. Page 71 of 84 AST Licensing Technical Report for Palisades Report Number: NA1-l 1149-027 Table 2.3-1 Main Steam Line Break (MSLB) - Inputs and Assumptions Input/Assumption Value Core Power Level 2703 MW~h Core Average Burnu p 39,300 MWD/MTU Radial Peaking Factor 2.04 Fuel Damagle 2% DNB 0% Fuel Centerline Melt Steam Generator Tube Leakage Rate 0.3 gpm per SG Time to establish shutdowvn cooling and terminate steam 8 hours9.259259e-5 days <br />0.00222 hours <br />1.322751e-5 weeks <br />3.044e-6 months <br /> release Time for PCS to reach 212'F and terminate SG tube leakage 12 hours1.388889e-4 days <br />0.00333 hours <br />1.984127e-5 weeks <br />4.566e-6 months <br /> PCs Mass 432,977 Ibm Maximum (Hot Zero Power) - 210,759 Ibm, (used for SG Secondary Side Mass faulted SG to maximize release)

Minimum (Hot Full Powver) - 141,065 Ibm (used for intact SG to maximize concentration)

Release from Faulted SG Instantaneous Steam Release from Intact SGs Table 2.3-2 Secondary Coolant Iodine Activity prior to accident 0. 1 pCi/gm DE 1-131 Geeraor ecodar Stea Sid Patiton oeficints Faulted SG - none Steaecodar GeeraorSid Patiton oeficints Intact SGs - 100 Atmospheric Dispersion Factors Offsite Table 1.8.2-I Onsite Table 1.8.1-2 and Table 1.8.1-3 Control Room Ventilation System Table 1.6.3-I Time of manual control room normal intake isolation 20 minutes and switch to emergency mode Control Room Unfiltered Inleakage 10 cfm Breathing Rates Offsite RG 1.183, Section 4.1.3 Control Room RG 1.183, Section 4.2.6 Control Room Occupancy Factors RG 1.183 Section 4.2.6

Numerical Applications, Inc. Page 72 of 84 AST Licensing Technical Report for Palisades Report Number: NAI-1 1149-027 Table 2.3-2 Intact SG Steam Release Rate Table 2.3-3 MSLB Dose Consequences Case EAB Dose (I) LPZ Dose (2) Control Room Dose (2)

Case(rem TEDE) (rem TEDE) j (rem TEDE)

MSLB 2.46 0.77 4.98 Acceptance Criteria 25 (3) 25 (3) 5(4)

(1)Worst 2-hour dose (2) Integrated 30-day dose based on an unfiltered Control Room inleakage rate of 10 cfm

') RG 1.183, Table 6 (4) 1 OCFR50.67

Numerical Applications, Inc. Page 73 of 84 AST Licensing Technical Report for Palisades Report Number: NAI-1 149-027 Table 2.4-1 Steam Generator Tube Rupture (SGTR) - Inputs and Assumptions Input/Assumption Value Core Power Level 2703 MWth Initial PCS Equilibrium ActivityTal1.2-(1.0 piCi/gm DE 1-131 and IOOIE-bar gross activity)Tal1.2-Initial Secondary Side Equilibrium Iodine Activity Table 1.7.3-1 (0. 1 pCi/gm DE 1-13 1)

Maximum pre-accident spike iodine concentration 40 pCi/gm DE 1-131 Maximum equilibrium iodine concentration 1.0 [tCi'gm DE 1-131 Duration of accident-initiated spike 8 hours9.259259e-5 days <br />0.00222 hours <br />1.322751e-5 weeks <br />3.044e-6 months <br /> Steam Generator Tube Leakage Rate 0.3 gpm per SG Time to establish shutdown cooling and terminate steam 8 hours9.259259e-5 days <br />0.00222 hours <br />1.322751e-5 weeks <br />3.044e-6 months <br /> release PCS Mass 529,706 Ibm for pre-accident iodine spike case 459,445 Ibm, for concurrent iodine spike case SG Secondary Side Mass 141,065 Ibm per SG (minimum mass used to maximize concentration)

Integrated Mass Release Table 2.4-2 Secondary Coolant Iodine Activity prior to accident 0. 1 pCi/gm DE 1-131 Faulted SG (flashed tube flow) - Table 2.4-7 Steam Generator Secondary Side Partition Coefficients Faulted SG (non-flashied tube flowv) - 100 Intact SGO- 100 Break Flowv Flash Fraction Table 2.4-3 Atmospheric Dispersion Factors Offsite Table 1.8.2-1 Onsite Table 1.8.1-2 and Table 1.8.1-3 Control Room Ventilation System 20 minutes Time of manual control room normal intake isolation and switch to emergency mode Control Room Unfiltered Inleakage 100 cfm Breathing Rates Offisite RG 1.183, Section 4.1.3 Control Room RG 1.183, Section 4.2.6 Control Room Occupancy Factor RG 1.183 Section 4.2.6

Numerical Applications, Inc. Page 74 of 84 AST Licensing Technical Report for Palisades Report Number: NAM- 149-027 Table 2.4-2 SGTR Integrated Mass Releases Time Break Flow in Steam Release from Ruptured SG Steam Release from (or) Ruptured SG (b)Unaffected SG 0-0.196417 24,011.15 0 0 0.196417 -0.5 37,111.85 44,654 53,574 0.5 - 1.388889 81,281 22,152.3 109,629.6 1.388889-2 40,798 15,229.7 75,370.4 2-3.638889 64,773 75,485.6 145,983.5 3.638889 -8 357,126 200,868.4 388,464.5 8-720 0 0 0 (1)Flowrate assumed to be constant within time period Table 2.4-3 SGTR Flashing Fraction for Flow From Broken Tube Time FlsigFato (seconds) Flsigrato 0 0.110 707.1 0.065 736 0.031 859 0.023 1090 0.006 1800 0.006

Numerical Applications, Inc. Page 75 of 84 AST Licensing Technical Report for Palisades Report Number: NAM- 149-027 Table 2.4-4 40 jtCi/gm D.E. 1-131 Activities IsotopeActivity Isotope (Ci/gm)

Iodine-131 33.2194 lodine-132 7.6660 Iodine-133 34.4971 Iodine-134 3.0025 Iodine-135 14.6932

Table 2.4-5 Iodine Equilibrium Appearance Assumptions Input Assumption Value Maximum Letdown Flow 40 gpm Assumed Letdown Flow

  • 44 gpm at 120'F, 2060 psia Maximum Identified PCS Leakage 10 gpm Maximum Unidentified PCS Leakage I gpm PCS Mass 459,445 Ibul 1-13 1 Decay Constant 5.986968E-5 min'l 1-132 Decay Constant 0.005023 min" 1-133 Decay Constant 0.000555 min-'

1-134 Decay Constant 0.013 178 min" 1-135 Decay Constant 0.00 1748 min"

  • maximum letdown flow plus 10% uncertainty Table 2.4-6 Concurrent (335 x) Iodine Spike Appearance Rate IooeAppearance Rate Time of Depletion Istoe(Cu/min) (hiours)

Iodine-131 58.0966961 >8 Iodine-132 79.8319317 >8 Iodine-133 90.1310904 >8 Iodine-134 74.0318685 >8 lodine-135 68.9790622 >8

Page 77 AST Licensing INumerical Technical Report Applications, Inc. for Palisades of 84 Report Number: NAI-1 149-027 Table 2.4-7 Affected Steam Generator Water Level and Decontamination Factors for Flashed Flow Time NNater Level Above U-Tubes Calculated Decontamination Factor (seconds) (feet) Decontamination Factor Used in Analysis 0 0.0 (assumed)* 1.0 1.0 707.1 0.0 (assumned)* 1.0 1.0 736 0.11 1.002299 1.002299 859 0.55 1.045037 1.045037 1090 1.39 1.452436 1.452436 1800 3.97 1.467378 1.467378 5000 6.79 60.03443 1.467378 7200 9.43 38.01867 1.467378 13100 j12.34 553073.5 58.16008 28800 115.16 58.16008 58.16008

' It is conservatively assumed that no scrubbing occurs until after the reactor trip at 707.1 seconds. Since the U-tubes remain covered throughout the event, it is also conservatively assumed that at the time of trip the wvater level is just above the top of the U-tubes. The time-dependent wvater level after the trip is a function of the allowable primary to secondary leakage, broken tube flow, and MSSV/ADV releases from the affected steam generator. To minimize the wvater level available for scrubbing, the location of the tube break is assumed to be at the top of the U-tubes.

Table 2.4-8 SGTR Dose Consequences Case EA B Dose (I LPZ Dose (2) Control Room Dose (2)

Case(rem TEDE) (rem TEDE) (rem TEDE)

SGTR pre-accident iodine spike 0.99 0.22 3.79 Acceptance Criteria (pre-accident iodine spike) 25 (3) 25 (3) 5 (4)

SGTR concurrent iodine spike 1.17 0.21 3.48 Acceptance Criteria (concurrent iodine spike) 2.5 (3) 2.5 (3) 5(4)

(1 Worst 2-hour dose (2) Integrated 30-day dose based on an unfiltered Control Room inleakage rate of 100 cfmr 3

( ) RG 1. 183, Table 6 (4)IOCFR5O.67

Numerical Applications, Inc. Page 78 of 84 AST Licensing Technical Report for Palisades Report Number: NAM- 149-027 Table 2.5-1 Small Line Break Outside of Containment - Inputs and Assumptions Input/Assumption Value PCS Equilibrium Activity Table 1.7.2-1 Break Flow Rate 160 gpm Break Temperature I135 0F Break Pressure 35 psia Time required to isolate break 60 minutes Maximum equilibrium iodine concentration 1.0 piCifgm DE 1-131 Iodine appearance rate for concurrent iodine spike (500x) Table 2.5-2 Iodine fraction released from break flow 10%

Reactor building ventilation system filtration None Atmospheric Dispersion Factors Mffite Table 1.S.2-1 Onsite Table 1.S. 1-2 and Table 1.8.1-3 Control Room Ventilation System Time of manual control room normal intake isolation 20 minutes and swvitch to emergency mode Control Room Unfiltered Inleakage 100 cfmr Breathing Rates Mffite RG 1.183 Section 4.1.3 Onsite RG 1.183 Section 4.2.6 Control Room Occupancy Factor RG 1.183 Section 4.2.6

Numerical Applications, Inc. Page 79 of 84 AST Licensing Technical Report for Palisades Report Number: NAI-I1149-027 Table 2.5-2 Concurrent (500 x) Iodine Spike Appearance Rate IsotopeAppearance Rate Isotope(Cu/min) lodine-131 86.7114868 Iodine-l32 119.152137 Iodine-133 134.524016 lodine-134 110.495326 lodine-135 102.953824 Table 2.5-3 Small Line Break Outside of Containiment Dose Consequences CaseEAB Dose (1 LPZ Dose (2) Control Room Dose (2)

Case(rem TEDE) (rem TEDE) (rem TEDE)

Small Line Break 0.41 0.05 0.53 Acceptance Criteria 2.5 2.5 5.0 "I)Worst 2-hour dose (2) Integrated 30-day dose based on an unfiltered Control Room inleakage rate of 100 cfi-n

Numerical Applications, Inc. Page 80 AST Licensing Technical Report for Palisades of 84 Report Number: NAI-1 149-027 Table 2.6-1 Control Rod Ejection - Inputs and Assumptions Input/Assumption Value Core Powver Level 2703 MWth Core Average Fuel Burnup 39,300 MWDIMTU Fuel Enrichment 3.0 -5.0 w/o Maximum Radial Peaking Factor 2.04

%DNB Fuel 14.7%

% Fuel Centerline Melt 0.5%

LOCA Source Term Table 1.7.4-1 Initial PCS Equilibrium ActivityTal1.2-(1.0 pCi/gm DE 1-131 and I00/E3-bar gross activity) TbeI..-

Initial Secondary Side Equilibrium Iodine Activity Table 1.7.3-1 (0. 1 pCi/gm DE 1-] 31) _________ __________

Release From DNB Fuel Section I of Appendix H to RG 1.183 Release From Fuel Centerline Melt Fuel Section 1 of Appendix H to RG 1.183 Steam Generator Secondary Side Partition Coefficient 100 Steam Generator Tube Leakage 0.3 gpm per SG Time to establish shutdowvn cooling 8 hours9.259259e-5 days <br />0.00222 hours <br />1.322751e-5 weeks <br />3.044e-6 months <br /> PCS Mass 432,976.8 Ibm minimumn- 141,065 lbm (per SG)

SG Secondary Side Mass Minimum mass used for SGs to maximize steam release nuclide concentration.

Particulate - 95%

Chemical Form of Iodine Released to Containment Elemental - 4.85%

Organic - 0. 15%

Particulate - 0%

Chemical Form of Iodine Released from SGs Elemental - 97 %

_______________________________________Organic - 3 Atmospheric Dispersion Factors Offsite Table 1.8.2-1 Onsite Table 1.8.1-2 and Table 1.8.1-3 Control Room Ventilation System Time of Control Room Ventilation System Isolation 20 minutes Control Room Unfiltered Inleakage 10 cfm Breathing Rates RG 1.183 Sections 4.1.3 and 4.2.6 Control Room Occupancy Factor RG 1.183 Section 4.2.6 3

Containment Volume I .64E+06Tft Containment Leakage Rate 0 to 24 hours2.777778e-4 days <br />0.00667 hours <br />3.968254e-5 weeks <br />9.132e-6 months <br /> 0. 10% (by wveight)/day after 24 hours2.777778e-4 days <br />0.00667 hours <br />3.968254e-5 weeks <br />9.132e-6 months <br /> 0.05% (by wveight)/day Aerosols-0.l hr" Containment Natural Deposition Coefficients Elemental Iodine - 1.3 hr-'(')

________________________________________Organic Iodine - None

Numerical Applications, Inc. Page 81 of 84 AST Licensing Technical Report for Palisades Report Number: NAM-I 1149-027 (1) Conservatively assumes lower elemental iodine natural deposition coefficient than LOCA Table 2.6-2 Control Rod Ejection Steam Release Time SG Steam Release Time (lb.)

0 - 1100sec 107,158.8 1100 sec - 0.5 hours5.787037e-5 days <br />0.00139 hours <br />8.267196e-6 weeks <br />1.9025e-6 months <br /> 31,336.8 0.5 hr -8 hir 1,007,100

>8 hr 0 Table 2.6-3 Control Rod Ejection Dose Consequences 2

CaseEAB Dose~') LPZ Dose( 2 ) Control Room Dose ()

Case(rem TEDE) (rem TEDE) (rem TEDE)

CEA Ejection - Containment Release 2.70 0.43 1.14 CEA Ejection - Secondary Release 2.61 0.68 1.14 Acceptance Criteria 6.3 6.3 5 (1)Worst 2-hour dose (2) Integrated 30-day dose based on an unfiltered Control Room inleakage rate of 10 cfm

Page 82 AST Licensing INumerical Applications, Inc. for Palisades Technical Report of 841 Report Number: NAI- 1149-027 Table 2.7-1 Spent Fuel Cask Drop- Inputs and Assumptions Input/Assumption Value Core Power Level Before Shutdown 2703 MWth Core Average Fuel Burnup 39,300 MWD/MTU Discharged Fuel Assembly Burnup 39,300 -5 8,900 MWND/MTU Fuel Enrichment 3.0 -5 w/o Number of Fuel Assemblies Damaged 73 Delay Before Cask Drop Cases I & 2 -30days Case 3 -- 90 days Source Terms Table 1.7.6-1 Water Level Above Damaged Fuel Assembly 23.4 feet minimum Elemental - 285 Iodine Decontamination Factors Organic -I Overall - 200 Noble Gas Decontamination FactorI Idin In ormof Chemcal oolElemental - 99.85%

Cheica o IoineInFom oolOrganic -0. 15%

Atmospheric Dispersion Factors Table 1.8.1-2 and Table 1.8.1-3 Control Room Ventilation System Cases I & 2 - 0 seconds Time of Control Room Ventilation System Isolation Case 3 - o Control Room Unfiltered Inleakage10cf Time of Control Room Filtered Makeup Flowv Cases3 - 0seod Breathing Rates RG 1.183 Sections 4.1.3 and 4.2.6 Control Room Occupancy Factor RG 1.183 Section 4.2.6

Numerical Applications, Inc. Page 83 AST Licensing Technical Report for Palisades of 841 Report Number: NAM- 149-027 Table 2.7-2 Spent Fuel Cask Drop D~ose Consequences CaseEAB Dose (1) LPZ Dose (2) Control Room Dose (2)

Case(rem TEDE) (rem TEDE) (rem TEDE)

Cask Drop - Case 1 2.04 0.25 1.37 Cask Drop - Case 2 2.78 0.35 1.99 Cask Drop - Case 3 0.08 0.01 1.67 Acceptance Criteria 6.3 6.3 5 I)Worst 2-hour dose (2) Integrated 30-day dose based on an unfiltered Control Room inleakage rate of 100 cfm

Numerical Applications, Inc. Page 84 of 84 AST Licensing Technical Report for Palisades Report Number: NAI-l 1149-027 Table 3-1 Palisades Summary of Alternative Source Term Analysis Results Unfiltered CR EAB Dose t i) LPZ Dose( 2 ) Control(2 Case Inleakage (rem TEDE) (re TEDE) Room Dose~2 I_______________ (cfm) (rm(rem TEDE)

LOCA 10 12.76 3.27 4.01 MSLB 10 2.46 0.77 4.98 SGTR Pre-accident Iodine Spike 100 0.99 0.22 3.79 Acceptance Criteria _5 25*) 25(' :55 4 SGTR Concurrent Iodine Spike 100 1.17 0.21 3.48 Small Line Break Outside of 100 0.41 0.05 0.53 Containment_____________________________

Acceptance Criteria 2. 3 . 3 ~ _<5(4)

FHA in Containment 100 2.20 0.28 4.04 FHA in FHB 10% Release 100 2.02 0.25 3.68 Filtration_______________

FHA in FHB 34% Release 100 1.60 0.20 2.81 Filtration _______

FHA in FHB 50% Release 100 1.31 0.17 2.22 Filtration Control Rod Ejection - 10 2.70 0.43 1.14 Containment Release Control Rod Ejection - Secondary 10 2.61 0.68 1.14 Release Spent Fuel Cask Drop 30 days Decay 90% of Release via FHB 100 2.04 0.25 1.37 filtration system_______________________

Spent Fuel Cask Drop 30 days Decay 82.5% of Release via FH-B 100 2.78 0.35 1.99 filtration system _______

Spent Fuel Cask Drop 90 days 1000 .116 Decay No Control Room Isolation 1000 .116 Acceptance Criteria <56.3' 13* 6.3(3 5 4 (1)WVorst 2-hour dose (2) Integrated 30-day dose o) RG 1.183, Table 6 (4)IOCFR5O.67

ENCLOSUR3E 5 PALISADES FULL S.COPE AST CD Containing Supporting Calculations and Drawings

ENCLOSUIRE 6 PALISADES FULL SCOPE AST RIS 2006-04 RESOLUTION MATRIX

RIS 2006-04 RESOLUTION MATRIX PALISADES NUCLEAR PLANT RIS Issue Licensee Comments

1. Level of Detail Contained in LARs (1) The AST amendment request should provide justification Provided in the Palisades AST license amendment request.

for each individual proposed change to the technical specifications (TS)

(2) The AST amendment request should identify and justify Provided in Section 2.0 of the AST Licensing Technical Report for each change to the licensing basis accident analyses Palisades.

(3) The AST amendment request should contain enough Sufficient detail is provided in Sections 1.6 - 1.8 and 2.0 of the AST details (e.g., assumptions, computer analyses input and Licensing Technical Report for Palisades and in the included output) to allow the NRC staff to confirm the dose analyses supporting calculations and computer input and output files.

results in independent calculations.

Licensees should identify the most current analyses, The most current analyses and assumptions are identified in Section assumptions, and TS changes in their submittal and 2.0 of the AST Licensing Technical Report for Palisades. The most supplements to the submittal, current TS changes are identified in the Palisades AST license amendment request.

2. Main Steam Isolation Valve (MSIV) Leakage and Fission Not applicable to PWRs.

Product Deposition in Piping

3. Control Room Habitability Use of non-ESE ventilation systems during a DBA should not No credit is taken for use of non-ESF ventilation systems during a be assumed unless the systems have emergency power and DBA. See Section 2.0 of the AST Licensing Technical Report for are part of the ventilation filter testing program in Section 5 of Palisades.

the TS.II Page 1 of 9

RIS 2006-04 RESOLUTION MATRIX PALISADES NUCLEAR PLANT RIS Issue Licensee Comments Generic Letter (GL) 2003-01, "Control Room Habitability" Control room envelope unfiltered inleakage assumptions are requested licensees to confirm the ability of their facility's discussed in Section 1.6.3.1 of the AST Licensing Technical Report control room to meet applicable habitability regulatory for Palisades. Palisades April 2005 tracer gas testing resulted in requirements. The GL placed emphasis on licensees worst-train inleakage of 49 +/-9 cfm. After bubble-tight damper confirming that the most limiting unfiltered inleakage into the installation, post-modification testing is performed to verify analysis control room envelope (CRE) was not greater than the value assumptions.

assumed in the DBA analyses.

Some AST amendment requests proposed operating schemes No changes are proposed for control room and other ventilation for the control room and other ventilation systems which affect systems which affect areas adjacent to the ORE. See the Palisades areas adjacent to the ORE and are different from the manner AST license amendment request.

of operation and performance described in the response to the GL without providing sufficient justification for the proposed changes in the operating scheme.

4. Atmospheric Dispersion Licensees have the option to adopt the generally less Tne Palisades AST license amendment request includes revised conservative (more realistic) updated NRC staff guidance on atmospheric relative concentration values for onsite radiological determining X/O values in support of design basis control consequence analyses developed in conformance with RG 1.194 as room radiological habitability assessments provided in RG described in Section 1.8.1 of the AST Licensing Technical Report for 1.194, "Atmospheric Relative Concentrations for Control Palisades.

Room Radiological Habitability Assessments at Nuclear Power Plants".

Regulatory positions on X/O values for offsite (i.e., exclusion The Palisades AST license amendment request includes revised area boundary and low population zone) accident radiological atmospheric relative concentration values for offsite radiological consequence assessments are provided in RG 1.145, consequence analyses developed in conformance with RG 1.145 as "Atmospheric Dispersion Models for Potential Accident described in Section 1.8.2 of the AST Licensing Technical Report for Consequence Assessments at Nuclear Power Plants". Palisades.

Page 2 of 9

RIS 2006-04 RESOLUTION MATRIX PALISADES NUCLEAR PLANT RIS Issue Licensee Comments The submittal should include a site plan showing true North A site plan indicating true north and locations of release and receptors and indicating locations of all potential accident release points is included in the Palisades AST license amendment request.

pathways and control room intake and unfiltered inleakage See Figure 1.8.1-1 of the AST Licensing Technical Report for pathways (whether assumed or identified during inleakage Palisades and Enclosure 5 of the Palisades license amendment testing). request.

The submittal should include a justification for using control Justification for the atmospheric relative concentrations used for the room intake X/Q values for modeling the unfiltered inleakage, control room envelope unfiltered inleakage is provided in Section if applicable. 1.6.3.1 of the AST Licensing Technical Report for Palisades.

The submittal should include a copy of the meteorological data Meteorological data and program input and output is provided with the inputs and program outputs along with a discussion of atmospheric relative concentration calculations included in the assumptions and potential deviations from staff guidelines. Palisades license amendment request. The electronic data are Meteorological data input files should be checked to ensure verified to be properly converted and formatted.

quality (e.g., compared against historical or other data and against the raw data to ensure that the electronic file has been properly formatted, any unit conversions are correct, and invalid data are properly identified).

When running the control room atmospheric dispersion model No credit is taken for an elevated release.

ARCON96, two or more files of meteorological data representative of each potential release height should be used if X/Q values are being calculated for both ground-level and elevated releases.

Page 3 of 9

RIS 2006-04 RESOLUTION MATRIX

.PALISADES NUCLEAR PLANT RIS Issue Licensee Comments In addition, licensees should be aware that Meteorological data and program input and output is provided with the atmospheric relative concentration calculations included in the Palisades license amendment request. All data formatting requirements have been observed.

(1)two levels of wind speed and direction data should always be provided as input to each data file, (2) fields of "nines" (e.g., 9999) should be used to indicate invalid or missing data, and (3) valid wind direction data should range from 10 to 3600.

Licensees should also provide detailed engineering Buoyancy or mechanical jets of high energy releases are not credited.

information when applying the default plume rise adjustment citled i n RG 1.194 to control room X/O Valucs- to acout for buoyancy or mechanical jets of high energy releases.

This information should demonstrate that the minimum effluent No credit is taken for an elevated release.

velocity during any time of the release over which the adjustment is being applied is greater than the 95 th percentile wind speed at the height of release.

When running the offsite atmospheric dispersion model No credit is taken for an elevated release.

PAVAN, two or more files of meteorological data representative of each potential release height should be used if X/Q values are being calculated for pathways with significantly different release heights (e.g., ground level versus elevated stack).

Page 4 of 9

RIS 2006-04 RESOLUTION MATRIX PALISADES NUCLEAR PLANT RIS Issue Licensee Comments The joint frequency distributions of wind speed, wind direction, Meteorological data and program input and output is provided with the and atmospheric stability data used as input to PAVAN should atmospheric relative concentration calculations included in the have a large number of wind speed categories at the lower Palisades license amendment request. A wind speed resolution of wind speeds in order to produce the best results calm, 0.1, 1.5, 3.0, 5.0, 7.5, 10.0 and 15.0 m/s is used.

5. Modeling of ESF Leakage The radiological consequences from the postulated [ESF] The postulated ESE leakage is analyzed and combined with leakage should be analyzed and combined with consequences postulated for other fission product release paths to consequences postulated for other fission product release determine the total calculated radiological consequences from the paths to determine the total calculated radiological LOCA.

consequences from the [loss-of-coolant accident] LOCA.

Licensees should account for ESF leakage at accident ESE surveillance test leakage is accounted for in the analysis.

conditions in their dose analyses so as not to underestimate the release rate.

In Appendix A to RG 1.183, Regulatory Position 5.5, the NRC For leakage to the ESF room, the RG 1.183 conservative value Of staff provided a conservative value of 10 percent as the 10% for iodine airborne fraction is used for the entire transient. For assumed amount of iodine that may become airborne from leakage to the SIRWT, a detailed model utilizing time-dependent total ESE leakage that is less than 212 OF. iodine concentration, pH, temperature and inleakage rate is justified and used. See Section 2.1 of the AST Licensing Technical Report for IPalisades.

Page 5 of 9

RIS 2006-04 RESOLUTION MATRIX PALISADES NUCLEAR PLANT I

RIS Issue Licensee Comments Figure 3.1 in NUREG/CR-5950 can be used to quantify the The calculation methodology for sump water back leakage to the amount of elemental iodine as a function of the sump water SIRWT pH is based on the approach outlined in NUREG/CR-5950.

pH and the concentration of iodine in the solution. In some Radioactive and non-radioactive forms of iodine are used to determine cases, however, licensees have misapplied this figure. Rather the total concentration of iodine. See Section 2.1 of the AST than using the total concentration of iodine (i.e., stable and Licensing Technical Report for Palisades.

radioactive), licensees based their assessment on only the radioactive iodine in the sump water. By using only the radioactive iodine, licensees have underestimated how much iodine evolves during post-accident conditions.

6. Release Pathways Changes to the plant configuration associated with an LAR No change to plant configuration with respect to open containment (e.g., an "open" containment during refueling) may require a during refueling is being proposed. The Palisades AST license re-analysis of the design basis dose calculations. A request for amendment request includes analysis of the design basis dose TS modifications allowing containment penetrations (i.e., calculations for an open containment during refueling and does not personnel air lock, equipment hatch) to be open during credit containment closure during the event. See Sections 2.2 and 2.7 refueling cannot rely on the current dose analysis if this of the AST Licensing Technical Report for Palisades.

analysis has not already considered these release pathways.

Releases from personnel air locks and equipment hatches exposed to the environment and containment purge releases prior to containment isolation need to be addressed. For the LOCA analysis, no change to the analysis of containment isolation is being proposed. Containment isolation is achieved prior to the onset of the hot rod fuel burst.

Page 6 of 9

RIS 2006-04 RESOLUTION MATRIX PALISADES NUCLEAR PLANT RIS Issue Licensee Comments Licensees are responsible for identifying all release pathways Revised control room, exclusion area boundary, and low population and for considering these pathways in their AST analyses, zone atmospheric relative concentrations (X]Q) for all known leakage consistent with any proposed modification. paths are considered. Containment leakage (closest containment point, equipment hatch), ESF leakage (plant stack, SIRWT vent),

secondary-side leakage (atmospheric dump valves, main steam safety valves, turbine building roof exhausters) are explicitly calculated. See Section 1.8 of the AST Licensing Technical Report for Palisades.

7. Primary to Secondary Leakage Some analysis parameters can be affected by density The density used in converting volumetric leak rates to mass leak changes that occur in the process steam. The NRC staff rates is consistent with the basis of the surveillance tests (62.4 continues to find errors in LAR submittals concerning the lbM/ft3). See Section 2.4 of the AST Licensing Technical Report for modeling of primary to secondary leakage during a postulated Palisades.

accident. This issue is discussed in Information Notice (IN)88-31, "Steam Generator Tube Rupture Analysis Deficiency,"

(Ref. 11) and Item 3.f in RIS 2001 -19. An acceptable methodology for modeling this leakage is provided in Appendix F to RG 1.183, Regulatory Position 5.2.

8. Elemental Iodine Decontamination Factor (DF)

Appendix B to RG 1.183, provides assumptions for evaluating For the fuel handling accident, the depth of water above the damaged the radiological consequences of a fuel handling accident. If fuel is 22.5 feet, less than 23 feet. The method of Burley is used to the water depth above the damaged fuel is 23 feet or greater, determine the appropriate iodine decontamination factor for 22.5 feet Regulatory Position 2 states that "the decontamination factors of overlying water, based on an elemental iodine decontamination for the elemental and organic [iodine] species are 500 and 1, factor of 285 at 23 feet. See Section 2.2 of the AST Licensing respectively, giving an overall effective decontamination factor Technical Report for Palisades.

of 200." However, an overall DF of 200 is achieved when the OF for elemental iodine is 285, not 500.

Page 7 of 9

RIS 2006-04 RESOLUTION MATRIX PALISADES NUCLEAR PLANT RIS Issue Licensee Comments

9. Isotopes Used in Dose Assessments For some accidents (e.g., main steamline break and rod drop), Noble gas and cesium isotopes were included in the dose licensees have excluded noble gas and cesium isotopes from assessment. See Sections 2.3 and 2.6 of the AST Licensing the dose assessment. The inclusion of these isotopes should Technical Report for Palisades.

be addressed in the dose assessments for AST implementation.

10. Definition of Dose Equivalent 1-131 In the conversion to an AST, licensees have proposed a The proposed change to the Palisades TS references FGR 11 for the modification to the TS definition of dose equivalent 1311. dose conversion factors used to calculate dose equivalent iodine-i 31.

Although different references are available for dose The same FGR 11 dose conversion factors are used in the AST conversion factors, the TS definition should be based on the analyses. See Section 1.3 of the AST Licensing Technical Report for same dose conversion factors that are used in the Palisades.

determination of the reactor coolant dose equivalent iodine curie content for the main steamline break and steam generator tube rupture accident analyses.

11. Acceptance Criteria for Off-Gas or Waste Gas System Release As part of full AST implementation, some licensees have No changes are proposed to the waste gas or volume control systems.

included an accident involving a release from their off-gas or See the Palisades license amendment request. Analyses of the waste waste gas system. gas decay tank rupture and volume control tank rupture events are not included with the Palisades license amendment request.

12. Containment Spray Mixing Page 8 of 9

RIS 2006-04 RESOLUTION MATRIX PALISADES NUCLEAR PLANT RIS Issue Licensee Comments Some plants with mechanical means for mixing containment Palisades's containment is modeled as a single, well-mixed volume air have assumed that the containment fans intake air solely since there is at least 90% spray coverage for the Palisades from a sprayed area and discharge it solely to an unsprayed containment. See Section 2.0 of the AST Licensing Technical Report region or vice versa. Without additional analysis, test for Palisades.

measurements or further justification, it should be assumed that the intake of air by containment ventilation systems is supplied proportionally to the sprayed and unsprayed volumes in containment.

Page 9 of 9

ENCLOSURE 7 PALISADES FULL SCOPE AST RG 1.183 COMPLIANCE MATRIX

REGULATORY GUIDE 1.183 COMPLIANCE MATRIX PALISADES NUCLEAR PLANT Notes:

1. Reference to Tables or Sections in this column refers to Enclosure 4, "AST Licensing Technical Report for Palisades" of the license amendment request unless otherwise noted.
2. RG 1.183 does not contain guidance for SLBOC. See Section 2.5 of Enclosure 4
3. RG 1.183 does not contain guidance for SEOD. Applicable FHA guidance is used. See below.

Regulatory Guide 1.183 Main Body RG Section [ Regulatory Position Analysis Comments

3. - ACCIDENT SOURCE TERM_____

3.1 - Fission The inventory of fission products in the reactor core and Conforms The source terms derived from the reactor core Product available for release to the containment should be based on fission product inventory are listed in Tables 1.7.2-Inventory the maximum full power operation of the core with, as a 1, 1.7.3-1, 1.7.4-1, 1.7.5-1, and 1.7.6-1 of minimum, current licensed values for fuel enrichment, fuel Enclosure 4. The assumed enrichment ranges burnup, and an assumed core power equal to the current from 3.0 to 5.0 w/o U235. The assumed bumup licensed rated thermal power times the ECCS evaluation ranges from 39,300 MWD/MTU to 58,900 uncertainty. The uncertainty factor used in determining the MWD/MTU. These values bound currently core inventory should be that value provided in Appendix K to licensed enrichment limit of 5 w0/6U235 and 10 CFR Part 50, typically 1.02. assembly bumup limit of 58.9 GWD/MTU. The assumed core power level is 2703 MWt. This power level is based on the original design power of 2650 MWt plus 2% uncertainty, which bounds the currently licensed power level of 2565.4 MWt plus 0.5925% Appendix K uncertainty.

The period of irradiation should be of sufficient duration to The assumed period of irradiation is based on a 3-allow the activity of dose-significant radionuclides to reach batch equilibrium core with shield bundles in their equilibrium or to reach maximum values. last cycle of operation (i.e., part of the third batch) and all fuel discharged at the current licensed assembly burnup limit of 58.9 GWD/MTU,

___________ _______________________________________________resulting in a core average bumup of 39,300 1 of 32

REGULATORY GUIDE 1.183 COMPLIANCE MATRIX PALISADES NUCLEAR PLANT MWD/MTU. This is sufficient to allow the activity of dose-significant radionuclides to reach equilibrium or to reach maximum values.

The core inventory should be determined using an The core inventory is determined using ORIGEN appropriate isotope generation and depletion computer code v2.1.

such as ORIGEN 2 or ORIGEN-ARP.

For the OBA LOCA, all fuel assemblies in the core are For the DBA LOCA, all fuel assemblies in the core assumed to be affected and the core average inventory are assumed to be affected and the core average should be used. For OBA events that do not involve the entire inventory is be used. For OBA events that do not core, the fission product inventory of each of the damaged involve the entire core, the fission product fuel rods is determined by dividing the total core inventory by inventory of each of the damaged fuel rods is the number of fuel rods in the core. To account for differences determined by dividing the total core inventory by in power level across the core, radial peaking factors from the the number of fuel rods in the core, multiplying by facility's core operating limits report (COLR) or technical the number of damaged rods, and multiplying by specifications should be applied in determining the inventory the radial peaking factor of 2.04 from the of the damaged rods. Palisades COLR to determine the inventory of the damaged rods.

No adjustment to the fission product inventory should be No adjustment to the fission product inventory is made for events postulated to occur during power operations made for events postulated to occur during power at less than full rated power or those postulated to occur at operations at less than full rated power or those the beginning of core life. postulated to occur at the beginning of core life.

For events postulated to occur while the facility is shutdown, Decay times of 48 hours5.555556e-4 days <br />0.0133 hours <br />7.936508e-5 weeks <br />1.8264e-5 months <br /> for the FHA, and 30 days e.g., a fuel handling accident, radioactive decay from the time and 90 days for the SFCD are modeled.

of shutdown may be modeled.

3.2 - Release The core inventory release fractions, by radionuclide groups, Conforms The assumed core inventory release fractions for Fractions for the gap release and early in-vessel damage phases for each radionuclide group for both the gap release DBA LOCAs are listed in Table 2 for PWRs. These fractions and early in-vessel damage phases for OBA are applied to the equilibrium core inventory described in LOCAs are those of Table 2 of RG 1.183. These Regulatory Position 3.1. fractions are applied to the equilibrium core

____________ ________________________________________ _________inventory described in Regulatory Position 3.1.

2 of 32

REGULATORY GUIDE 1.183 COMPLIANCE MATRIX PALISADES NUCLEAR PLANT r r For non-LOCA events, the fractions of the core inventory For non-LOCA events, the assumed fractions of assumed to be in the gap for the various radionuclides are the core inventory in the gap for the various given in Table 3. The release fractions from Table 3 are used radionuclides are those of Table 3 of RG 1.183.

in conjunction with the fission product inventory calculated The release fractions are used in conjunction with with the maximum core radial peaking factor. the fission product inventory calculated with the maximum core radial peaking factor of 2.04. For rod average burnups in excess of 54,000 MWDIMTU, the heat generation rate is limited to 6.3 kw/ft in accordance with footnote 11 of RG 1.183. Note that in some cases, the gap fractions are modified as required by the event-specific source term requirements listed in the Appendices for RG 1.183.

3.3 - Timing of The activity released from the core during each release phase Conforms For the LOCA DBA, the assumed activity release Release should be modeled as increasing in a linear fashion over the from the fuel is modeled as a constant release Phases duration of the phase. over the duration of the release phase.

For non-LOCA OBAs in which fuel damage is projected, the For non-LOCA OBAs, the assumed activity release from the fuel gap and the fuel pellet should be release from the fuel is modeled as an assumed to occur instantaneously with the onset of the instantaneous release from the onset of the projected damage. projected damage.

For facilities licensed with leak-before-break methodology, the The leak-before-break methodology is not onset of the gap release phase may be assumed to be 10 credited in the AST analyses.

minutes. A licensee may propose an alternative time for the onset of the gap release phase, based on facility-specific calculations using suitable analysis codes or on an accepted topical report shown to be applicable to the specific facility. In the absence of approved alternatives, the gap release phase onsets in Table 4 should be used.

3.4 -

Radionuclide 35-C hempo icti ofnh TTable 5 lists the elements in each radionuclide group that should be considered in design basis analyses.

aiidn eesdfo h eco oln ytm Igroup Conforms Cnom The elements assumed in each radionuclide are those of Table 5 of RG 1.183.

h sue hmclfr fidn eesdt 35-Co empoiion l _ _ eeae__ rmtereco oln _ sse ofrs

_ _e Th _ sue _ hmiaomoodn

_ ___ eesdt Form (RCS) to the containment in a postulated accident, 95 percent _______containment is 95% particulate, 4.85% elemental, 3 of 32

REGULATORY GUIDE 1.183 COMPLIANCE MATRIX PALISADES NUCLEAR PLANT of the iodine released should be assumed to be cesium iodide and 0.15% organic. This applies to the LOCA and (Csl), 4.85 percent elemental iodine, and 0.15 percent organic GRE containment releases. With the exception of iodide. This includes releases from the gap and the fuel elemental and organic iodine and noble gases, pellets. With the exception of elemental and organic iodine fission products are assumed to be in particulate and noble gases, fission products should be assumed to be in form.

particulate form.

The same chemical form is assumed in releases from fuel For the LOCA DBA, the assumed chemical form pins in FHAs and from releases from the fuel pins through the of iodine released from the EGGS is 97%

RCS in OBAs other than FHAs or LOGAs. However, the elemental and 3% organic.

transport of these iodine species following release from the fuel may affect these assumed fractions. The accident- For the FHA, the assumed chemical form of iodine specific appendices to this regulatory guide provide additional in the water pooi is 99.85% elemental and 015%

details. organic. With an elemental iodine OF of 252 and an organic iodine OF of 1,the assumed chemical form in the pool results in a chemical form of 73%

elemental and 27% organic for iodine released from the pool.

For the MSLB, SGTR, SLBOG, GRE (steam generator release only), the assumed chemical for of iodine released is 97% elemental and 3%

organic.

For the SFGD, the assumed chemical form of iodine in the water pool is 99.85% elemental and 015% organic. With an elemental iodine OF of 285 and organic iodine OF of 1,the assumed chemical form in the pool results in a chemical form of 70% elemental and 30% organic for iodine released from the DOOL.

3.6 - Fuel The amount of fuel damage caused by non-LOGA design Gonforms The non-LOGA design bases analyses use DNBR Oamage in basis events should be analyzed to determine, for the case as the fuel damage criterion for clad rupture and Non-LOGA resulting in the highest radioactivity release, the fraction of the LHGR as the fuel damage criterion for fuel melt.

OBAs fuel that reaches or exceeds the initiation temperature of fuel melt and the fraction of fuel elements for which the fuel clad is

_____________breached. Although the NRC staff has traditionally relied upon ______ _____________________

4 of 32

REGULATORY GUIDE 1.183 COMPLIANCE MATRIX PALISADES NUCLEAR PLANT the departure from nucleate boiling ratio (DNBR) as a fuel damage criterion, licensees may propose other methods to the NRC staff, such as those based upon enthalpy deposition, for estimating fuel damage for the purpose of establishing radioactivity releases. ______ _____________________

4. - DOSE CALCULATIONAL METHODOLOGY _________________

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

4.1.1 The dose calculations should determine the TEDE. TEDE is Conforms The dose calculations determine the TEDE dose, the sum of the committed effective dose equivalent (CEDE) with all significant progeny included (107 nuclides from inhalation and the deep dose equivalent (DDE) from considered), as the sum the CEDE and the DDE.

external exposure. The calculation of these two components of the TEDE should consider all radionuclides, including progeny from the decay of parent radionuclides, that are significant with regard to dose consequences and the released radioactivity. ______ ____________________

4.1.2 The exposure-to-CEDE factors for inhalation of radioactive Conforms The assumed exposure-to-CEDE factors are material should be derived from the data provided in ICRP derived from ICRP Publication 30 and Table 2.1 of Publication 30, "Limits for Intakes of Radionuclides by Federal Guidance Report 11.

Workers" (Ref. 19). Table 2.1 of Federal Guidance Report 11, "Limiting Values of Radionuclide Intake and Air Concentration and Dose Conversion Factors for Inhalation, Submersion, and Ingestion" (Ref. 20), provides tables of conversion factors acceptable to the NRC staff.

4.1.3 For the first 8 hours9.259259e-5 days <br />0.00222 hours <br />1.322751e-5 weeks <br />3.044e-6 months <br />, the breathing rate of persons offsite Conforms The assumed offsite breathing rates are those should be assumed to be 3.5 x 10-4 cubic meters per second. specified in Section 4.1.3 of RG 1.183.

From 8 to 24 hours2.777778e-4 days <br />0.00667 hours <br />3.968254e-5 weeks <br />9.132e-6 months <br /> following the accident, the breathing rate should be assumed to be 1.8 x 10*4 cubic meters per second.

After that and until the end of the accident, the rate should be assumed to be 2.3 x 10-4 cubic meters per second. ______

4.1.4 The DDE should be calculated assuming submergence in Conforms The EDE are used to determine the submergence semi-infinite cloud assumptions with appropriate credit for dose in a semi-infinite cloud. The assumed attenuation by body tissue. EDE may be used in lieu of DDE conversion factors are those of Federal Guidance

___________in determining the contribution of external dose to the TEDE. _______Report 12.

5 of 32

REGULATORY GUIDE 1.183 COMPLIANCE MATRIX PALISADES NUCLEAR PLANT 4.1.5 The TEDE should be determined for the most limiting person Conforms The TEDE are determined for the most limiting at the EAB. The maximum EAB TEDE for any two-hour period person for a two-hour period at the EAB and the following the start of the radioactivity release should be maximum two-hour dose is reported.

determined and used in determining compliance with the dose criteria in 10 CFR 50.67. The maximum two-hour TEDE should be determined by calculating the postulated dose for a series of small time increments and performing a "sliding" sum over the increments for successive two-hour periods.

The maximum TEDE obtained issubmitted. The time increments should appropriately reflect the progression of the accident to capture the peak dose interval between the start of the event and the end of radioactivity release.

4.1.6 TEDE should be determined for the most limiting receptor at Conforms The TEDE is determined for the most limiting the outer boundary of the low population zone (LPZ) and person at the LPZ.

should be used in determining compliance with the dose

____________criteria in 10 CFR 50.67. __________________________

4.1.7 No correction should be made for depletion of the effluent Conforms No plume depletion due to ground deposition is

___________plume by deposition on the ground. _______credited.

4.2 - Control The following guidance should be used in determining the Room Dose TEDE for persons located in the control room.

Consequences______________________________

4.2.1 The TEDE analysis should consider all sources of radiation Conforms The radiation dose to personnel within the CRE that will cause exposure to control room personnel. (control room viewing gallery, technical support center, mechanical equipment room) includes inhalation and immersion doses due to releases from containment, the ESF rooms, and the SIRWT, and includes direct shine doses from the containment, containment purge lines, SIRWT, filter accumulation, and the external plume.

4.2.2 The radioactive material releases and radiation levels used in Conforms The control room doses are determined using the the control room dose analysis should be determined using same source term, transport, and release the same source term, transport, and release assumptions assumptions used for determining the EAB and used for determining the EAB and the LPZ TEDE values, the LPZ TEDE values, resulting in conservative unless these assumptions would result in non-conservative results for the control room.

_____________results for the control room.

4.2.3 The models used to transport radioactive material into and Conforms The models used to transport radioactive material

____________through the control room, and the shielding models used to into and through the control room, and the 6 of 32

REGULATORY GUIDE 1.183 COMPLIANCE MATRIX PALISADES NUCLEAR PLANT determine radiation dose rates from external sources, should shielding models used to determine radiation dose be structured to provide suitably conservative estimates of the rates from external sources, are structured to exposure to control room personnel. provide suitably conservative estimates of the

____________ _________________________________________exposure to control room personnel.

4.2.4 Credit for engineered safety features that mitigate airborne Conforms Credit istaken for CR emergency intake and radioactive material within the control room may be assumed. recirculation filtration.

Such features may include control room isolation or pressurization, or intake or recirculation filtration. Refer to For the FHA, no credit is taken for any auxiliary Section 6.5.1, "ESE Atmospheric Cleanup System," of the building filtration. For the SFCD, credit is taken for SRP (Ref. 3) and Regulatory Guide 1.52, "Design, Testing, fuel handling building filtration, with conservative and Maintenance Criteria for Postaccident Engineered- estimates of filtration bypass.

Safety-Feature Atmosphere Cleanup System Air Filtration and Adsorption Units of Light-Water-Cooled Nuclear The credited filters are qualified and acceptable

__________Power Plants" (Ref. 25), for guidance. ______per RG 1.52.

4.2.5 Credit should generally not be taken for the use of personal Conforms No credit istaken for the use of personal

___________prtetie equipment or prophylactic drugs. _______protective equipment or prophylactic drugs.

4.2.6 The dose receptor for these analyses isthe hypothetical Conforms The assumed breathing rates and occupancy maximum exposed individual who ispresent in the control factors for control room operator dose are those room for 100% of the time during the first 24 hours2.777778e-4 days <br />0.00667 hours <br />3.968254e-5 weeks <br />9.132e-6 months <br /> after the specified in Section 4.2.6 of RG 1.183.

event, 60% of the time between 1 and 4 days, and 40% of the time from 4 days to 30 days. For the duration of the event, the breathing rate of this individual should be assumed to be 3.5 x 10-4cubic meters per second.

4.2.7 Control room doses should be calculated using dose Conforms Control room doses are calculated using dose conversion factors identified in Regulatory Position 4.1 above conversion factors identified in Regulatory for use in offsite dose analyses. Position 4.1 above.

The equation given in RG 1.183 for Regulatory Position 4.2.7 is utilized for finite cloud correction when calculating immersion doses due to the

___________airborne activity inside the control room.

4.3 - Other The guidance provided in Regulatory Positions 4.1 and 4.2 Conforms The radiation doses used for the NUREG-0737 Dose should be used, as applicable, inre-assessing the radiological analyses were calculated using source terms Consequences analyses identified in Regulatory Position 1.3.1, such as those derived from TID-14844 methodology. AST in NUREG-0737. Design envelope source terms provided in impact on these doses has been considered (See NUREG-0737 should be updated for consistency with the the license amendment request).

_____________AST. In general, radiation exposures to plant personnel______________________

7 of 32

REGULATORY GUIDE 1.183 COMPLIANCE MATRIX PALISADES NUCLEAR PLANT identified in Regulatory Position 1.3.1 should be expressed in terms of TEDE.

Integrated radiation exposure of plant equipment should be The radiation doses used for the current licensing determined using the guidance of Appendix I of this guide. basis environmental qualification analyses were calculated using source terms derived from TID-14844 methodology. AST impact on these doses has been considered.

____________ ________________________________________ _________(See the license amendment request).

4.4 - The radiological criteria for the EAB, the outer boundary of the Conforms The EAB and LPZ acceptance criteria used are Acceptance LPZ, and for the control room are in 10 CFR 50.67. These those of Table 6 of RG 1.183. The control room Criteria criteria are stated for evaluating reactor accidents of acceptance criterion is 5 rem TEDE.

exceedingly low probability of occurrence and low risk of public exposure to radiation, e.g., a large-break LOCA. The control room criterion applies to all accidents. For events with a higher probability of occurrence, postulated EAB and LPZ doses should not exceed the criteria tabulated in Table 6.

The acceptance criteria for the various NUREG-0737 (Ref. 2) See the license amendment request.

items generally reference General Design Criteria 19 (GDC

19) from Appendix Ato 10 CFR Part 50 or specify criteria derived from GDC-1 9. 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 10 CFR 50.67(b)(2)(iii).
5. - ANALYSIS ASSUMPTIONS AND METHODOLOGY 5.1 - General Considerations 5.1.1 The evaluations required by 10 CFR 50.67 are re-analyses Conforms The analyses have been prepared, reviewed and of the design basis safety analyses and evaluations will be maintained inaccordance with quality required by 10 CFR 50.34; they are considered to be a assurance programs that comply with Appendix B, significant input to the evaluations required by 10 CER "Quality Assurance Criteria for Nuclear Power 50.92 or 10 CFR 50.59. These analyses should be Plants and Fuel Reprocessing Plants," to 10 CFR prepared, reviewed, and maintained in accordance with Part 50.

______________quality assurance programs that comply with Appendix B, ______ _____________________

8 of 32

REGULATORY GUIDE 1.183 COMPLIANCE MATRIX PALISADES NUCLEAR PLANT "Quality Assurance Criteria for Nuclear Power Plants and Fuel Reprocessing Plants," to 10 CFR Part 50.

These design basis analyses were structured to provide a The dose analyses have not been limited to a conservative set of assumptions to test the performance of specific set of accident progression assumptions.

one or more aspects of the facility design. Licensees should exercise caution in proposing deviations based upon data from a specific accident sequence since the OBAs were never intended to represent any specific accident sequence

-- the proposed deviation may not be conservative for other

______________accident sequences. ______ _____________________

5.1.2 Credit may be taken for accident mitigation features that are Conforms Only safety-related mitigation features are classified as safety-related, are required to be operable by credited in the dose analyses.

technical specifications, are powered by emergency power sources, and are either automatically actuated or, in limited cases, have actuation requirements explicitly addressed in emergency operating procedures.

The single active component failure that results in the most Aloss of offsite power isassumed concurrent with limiting radiological consequences should be assumed. the LOCA and subsequent single active failures to Assumptions regarding the occurrence and timing of a loss maximize doses are also considered.

of offsite power should be selected with the objective of maximizing the postulated radiological consequences.

5.1.3 The numeric values that are chosen as inputs to the Conforms Conservative parameters are assumed when analyses required by 10 CFR 50.67 should be selected with calculating each contributor in the dose analyses.

the objective of determining aconservative postulated dose.

In some instances, a particular parameter may be conservative in one portion of an analysis but be non-conservative in another portion of the same analysis.

5.1.4 Inorder to issue a license amendment authorizing the use Conforms The analysis assumptions and methods are of an AST and the TEDE dose criteria, the NRC staff must compatible with the ASTs and the TEDE criteria.

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 inthe facility's design basis

______________analyses. Licensees should ensure that analysis_______

9 of 32

REGULATORY GUIDE 1.183 COMPLIANCE MATRIX PALISADES NUCLEAR PLANT assumptions and methods are compatible with the ASTs and the TEDE criteria. + t 5.2 - Accident- Licensees should analyze the DBAs that are affected by the Conforms The postulated accident radiological consequence Specific specific proposed applications of an AST. analyses have been updated for AST. The DBA Assumptions LOCA, FHA, MSLB, SGTR, SLBOC, CRE, and SFCD analyses have been analyzed. The WGDT and VCTR events will remain TID-based since no changes are being proposed to these systems or the acceptance criteria for the events.

The NRC staff has determined that the analysis Assumptions have been addressed, as noted assumptions in the appendices to this guide provide an below.

integrated approach to performing the individual analyses and generally expects licensees to address each assumption or propose acceptable alternatives.

The NRC will consider licensee proposals for changes in No changes have been made to analysis analysis assumptions based upon risk insights. The staff assumptions based upon risk insights.

will not approve proposals that would reduce the defense in depth deemed necessary to provide adequate protection for p~ublic health and safetv. I - I-5.3 - Meteorology Atmospheric dispersion values (X/Q) for the EAB, the LPZ, Conforms All X/Q values have been recalculated for the AST Assumptions and the control room that were approved by the staff during application.

initial facility licensing or in subsequent licensing proceedings may be used in performing the radiological analyses identified by this guide.

References 22 and 28 of this RIG should be used if the RG 1.145 has been used to calculate offsite X/Q FSAR X/Q values are to be revised or if values are to be values. RG 1.194 has been used for onsite X/Q determined for new release points or receptor distances. values.

Fumigation should be considered where applicable for the Fumigation has not been included since no credit EAB and LPZ. For the EAB, the assumed fumigation period is taken for an elevated release.

should be timed to be included in the worst 2-hour exposure period.

The NRC computer code PAVAN implements Reaulatorv The PAVAN code was used for determininq of 10 of 32

REGULATORY GUIDE 1.183 COMPLIANCE MATRIX PALISADES NUCLEAR PLANT Guide 1.145 and its use is acceptable to the NRC staff. offsite X/Q values.

The methodology of the NRIC computer code ARCON96 is ARCON96 was used for determining onsite X/Q generally acceptable to the NRC staff for use in determining control room X/O values. values for onsite values.

Meteorological data collected in accordance with the site- Meteorological data acquired in accordance with specific meteorological measurements program described the Palisades meteorological measurement in the facility FSAR should be used in generating accident program and FIG 1.23 for the five year period from X10 values. Additional guidance is provided in Regulatory 1999 to 2003 is used to calculate onsite and Guide 1.23. offsite atmospheric dispersion.

The onsite X/Q methodology has changed from All changes in X/Q analysis methodology should be one based upon site-specific wind tunnel data to reviewed by the NRC staff. one based solely upon the methods in FIG 1.194.

I The calculations are submitted for review.

6.- The assumptions in Appendix I to this guide are acceptable Conforms The radiation doses for environmental ASSUMPTIONS to the NRC staff for performing radiological assessments qualification analyses are based on source terms FOR associated with equipment qualification. The assumptions in derived from TID-1 4844 methodology.

EVALUATING Appendix I will supersede Regulatory Positions 2.c(1) and (See the license amendment request.)

THE RADIATION 2.c(2) and Appendix D of Revision 1of Regulatory Guide DOSES FOR 1.89, for operating reactors that have amended their EQUIPMENT licensing basis to use an alternative source term. Except as QUALIFICATION stated in Appendix I, all other assumptions, methods, and provisions of Revision 1 of Regulatory Guide 1.89 remain effective.

The NRC staff is assessing the effect of increased cesium See resolution of Generic Issue 187.

releases on EQ doses to determine whether licensee action iswarranted. Until such time as this generic issue is resolved, licensees may use either the AST or the TID14844 assumptions for performing the required EQ analyses. However, no plant modifications are required to address the impact of the difference in source term characteristics (i.e., AST vs TID14844) on EQ doses pending the outcome of the evaluation of the generic issue. I I 11 of 32

REGULATORY GUIDE 1.183 COMPLIANCE MATRIX PALISADES NUCLEAR PLANT Regulatory Guide 1.183 Appendix A: ASSUMPTIONS FOR EVALUATING THE RADIOLOGICAL CONSEQUENCES OF A LWR LOSS-OF-COOLANT ACCIDENT SOURCE TERM ASSUMPTIONS

1. Acceptable assumptions regarding core inventory and the Conforms The total core inventory of the radionuclide groups release of radionuclides from the fuel are provided in utilized for determining the source term for this Regulatory Position 3 of this guide. event is based on RG 1.183, Regulatory Position 3.1, at 102% of core thermal power and is provided in Table 1.7.4-1. The core inventory release fractions for the gap release and early in-vessel damage phases of the LOCA are consistent with Regulatory Position 3.2 and Table 2 of RG 1.183. See responses to Regulatory

___________Position 3.'

2. Ifthe sump or suppression pool pH is controlled at values of 7 Conforms The sump pH is controlled at a value greater than or greater, the chemical form of radioiodine released to the 7.0 based on the addition of an alternate buffer.

containment should be assumed to be 95% cesium iodide The license amendment request contains a (Csl), 4.85 percent elemental iodine, and 0.15 percent organic commitment to demonstrate that sump pH is iodide. Iodine species, including those from iodine re- maintained greater than 7 post-LOCA at the time evolution, for sump or suppression pool pH values less than 7 of recirculation with the alternate buffer introduced will be evaluated on a case-by-case basis. Evaluations of pH as part of the solution to GSI-191. The chemical should consider the effect of acids and bases created during form of the radioiodine released to the the LOCA event, e.g., radiolysis products. With the exception containment isassumed to be 95% cesium iodide of elemental and organic iodine and noble gases, fission (Csl), 4.85% elemental iodine, and 0.15% organic products should be assumed to be in particulate form. iodide. With the exception of elemental and organic iodine and noble gases, fission products

____________ _______________________________________________I are assumed to be in particulate form.

ASSUMPTIONS ON TRANSPOPRT IN PRIMARY CONTAINMENT_______________________

3. Acceptable assumptions related to the transport, reduction, and release of radioactivemnaterial 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 Conforms The activity released from the fuel isassumed to mix instantaneously and homogeneously throughout the free mix instantaneously and homogeneously air volume of the primary containment in PWRs as it is throughout the free air volume of the containment.

released. This distribution should be adjusted if there are The release into the containment is assumed to internal compartments that have limited ventilation exchange. terminate at the end of the early in-vessel phase.

_______________~~~~~ ~ ~~~~~The distribution does not need to be adjusted for I________________________________

12 of 32

REGULATORY GUIDE 1.183 COMPLIANCE MATRIX PALISADES NUCLEAR PLANT internal compartment effects.

3.2 Reduction in airborne radioactivity in the containment by Conforms Reduction of the airborne radioactivity in the natural deposition within the containment may be credited. containment by natural deposition is credited.

The prior practice of deterministically assuming that a 50% The natural deposition removal coefficient for plate-out of iodine is released from the fuel is no longer elemental iodine was determined to be 2.3/hr.

acceptable to the NRC staff as it is inconsistent with the Instantaneous plate-out is not assumed.

_____________characteristics of the revised source terms._______________________

3.3 Reduction in airborne radioactivity in the containment by Conforms Palisades containment spray systems are containment spray systems that have been designed and are designed and maintained in accordance with maintained in accordance with Chapter 6.5.2 of the SRP may Chapter 6.5.2 of the SRP.

be credited.

The evaluation of the containment sprays should address areas within the primary containment that are not covered by Per the current licensing basis, there is at least the spray drops. The mixing rate attributed to natural 90% spray coverage of the containment; convection between sprayed and unsprayed regions of the therefore, the containment istreated as a single containment building, provided that adequate flow exists well mixed volume.

between these regions, is assumed to be two turnovers of the un-sprayed 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 (OF) for elemental iodine based on the maximum iodine activity in the primary containment atmosphere when the sprays actuate, divided by the activity of iodine remaining at some The method used in the Palisades AST LOCA time after decontamination. The SRP also states that the analysis for determining the time period required particulate iodine removal rate should be reduced by a factor to reach an elemental iodine OF of 200 is based of 10 when a OF of 50 is reached. The reduction in the on a containment atmosphere peak iodine removal rate is not required if the removal rate is based on concentration equal to 40 percent of the core the calculated time-dependent airborne aerosol mass. There iodine inventory. As discussed in the SRP is no specified maximum OF for aerosol removal by sprays. Section 6.5.2, the iodine decontamination factor (OF) is a function of the effective iodine partition coefficient between the sump and containment

____________ ________________________________________ _________atmosphere. Thus, the loss of iodine due to other 13 of 32

,REGULATORY GUIDE 1.183 COMPLIANCE MATRIX PALISADES NUCLEAR PLANT mechanisms (containment leakage, surface deposition, etc.), would not be included in the determination of the time required to reach a DF of 200. Inaddition, since the iodine in the containment atmosphere and sump are decaying at the same rate, decay should not be included in determining the time to reach a OF of 200.

Additional RADTRAD-NAI cases determined the time to reach a decontamination factor of 200 to be 2.515 hours0.00596 days <br />0.143 hours <br />8.515212e-4 weeks <br />1.959575e-4 months <br />. An additional RADTRAD-NAI case determined the time to reach a OF of 50 for aerosol based on the peak aerosol mass to be 3.385 hours0.00446 days <br />0.107 hours <br />6.365741e-4 weeks <br />1.464925e-4 months <br />. Prior to reaching these decontamination factors, the elemental iodine spray removal coefficient was determined to be 4.8/hr and the particulate iodine spray removal coefficient was determined to be 2.3/hr. After reaching these decontamination factors, the elemental iodine spray removal coefficient is assumed to be zero and the particulate iodine spray removal coefficient is assumed to be reduced by a factor of 10.

3.4 Reduction in airborne radioactivity in the containment by in- Conforms Palisades does not have post-accident in-containment recirculation filter systems may be credited if containment air filtration systems. Palisades does these systems meet the guidance of Regulatory Guide 1.52 have containment air coolers, which are not and Generic Letter 99-02 (Refs. A-5 and A-6). The filter media credited for filtration or mixing.

loading caused by the increased aerosol release associated

_____________with the revised source term should be addressed.

3.5 Reduction in airborne radioactivity in the containment by N/A This position relates to suppression pool suppression pool scrubbing in BWRs should generally not be scrubbing in BWRs, which is not applicable to credited. However, the staff may consider such reduction on Palisades.

an individual case basis. The evaluation should consider the relative timing of the blowdown and the fission product release from the fuel, the force driving the release through the pool, and the potential for any bypass of the suppression pool (Ref. 7). Analyses should consider iodine re-evolution if the

_____________suppression pool liquid pH is not maintained greater than 7. ______________________________

14 of 32

REGULATORY GUIDE 1.183 COMPLIANCE MATRIX PALISADES NUCLEAR PLANT 3.6 Reduction in airborne radioactivity in the containment by N/A This position relates to activity retention in ice retention in ice condensers, or other engineering safety condensers, which is not applicable to Palisades.

features not addressed above, should be evaluated on an The engineered safety features are addressed individual case basis. See Section 6.5.4 of the SRP (Ref. A- above.

1). ______________________________ _

3.7 The primary containment should be assumed to leak at the Conforms Acontainment leak rate, based on the technical peak pressure technical specification leak rate for the first 24 specifications, of 0.10% per day of the hours. For PWRs, the leak rate may be reduced after the first containment air is assumed for the first 24 hours2.777778e-4 days <br />0.00667 hours <br />3.968254e-5 weeks <br />9.132e-6 months <br />.

24 hours2.777778e-4 days <br />0.00667 hours <br />3.968254e-5 weeks <br />9.132e-6 months <br /> to 50% of the technical specification leak rate. After 24 hours2.777778e-4 days <br />0.00667 hours <br />3.968254e-5 weeks <br />9.132e-6 months <br />, the containment leak rate is

____________ ________________________________________reduced to 0.05% per day of the containment air._

3.8 Ifthe primary containment is routinely purged during power Conforms The purge system is not considered to be in operations, releases via the purge system prior to operation at the beginning of the event. In containment isolation should be analyzed and the resulting addition, containment purge is not used after the doses summed with the postulated doses from other release beginning of the event for hydrogen control.

paths. The purge release evaluation should assume that Containment isolation is achieved prior to the 100% of the radionuclide inventory in the reactor coolant onset of fuel damage.

system liquid is released to the containment at the initiation of the LOCA. This inventory should be based on the technical specification reactor coolant system equilibrium activity.

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. 1For facilities with dual containment systems, the acceptable N/A Regulatory Positions 4.1 through 4.6 apply to Iassumptions related to the transport, reduction, and release facilities with dual containment systems. As such, of radioactive material in and from the secondary containment these positions are not applicable to Palisades.

________________ or enclosure buildings are as follows. _________ _______________________________

ASSUMPTIONS ON ESF SYSTEM LEAKAGE _________________

5. ESF systems that recirculate sump water outside of the Conforms The radiological consequences from the primary containment are assumed to leak during their postulated ESF systems leakage is analyzed and intended operation. This release source includes leakage combined with consequences postulated for other through valve packing glands, pump shaft seals, flanged fission product release paths.

connections, and other similar components. This release source may also include leakage through valves isolating interfacing systems (Ref. A-7). The radiological

_____________consequences from the postulated leakage should be ______

15 of 32

REGULATORY GUIDE 1.183 COMPLIANCE MATRIX PALISADES NUCLEAR PLANT 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 Drimarv containment for BWRs and PWRs.

5.1 With the exception of noble gases, all the fission products Conforms Engineered Safety Feature (ESF) systems that released from the fuel to the containment should be assumed recirculate water outside the primary containment to instantaneously and homogeneously mix in the primary (EGGS systems) are assumed to leak during their containment sump water at the time of release from the core. intended operation. With the exception of noble In lieu of this deterministic approach, suitably conservative gases, all fission products released from the fuel mechanistic models for the transport of airborne activity in to the containment are assumed to containment to the sump water may be used. instantaneously and homogeneously mix in the containment sump water at the time of release from the core.

5.2 The leakage should be taken as two times the sum of the Conforms Leakage from the EGGS system to the ESE simultaneous leakage from all components in the ESF rooms istaken as two times the Tech. Spec.

recirculation systems above which the technical specifications allowable value of 0.2 gpm. The leakage is or licensee commitments to item 111.0.1.1 of NUREG-0737 assumed to start at the earliest time the would require declaring such systems inoperable. The recirculation flow occurs in these systems and leakage should be assumed to start at the earliest time the continue for the 30-day duration. Backleakage to recirculation flow occurs in these systems and end at the the SIRWT is also considered separately as two latest time the releases from these systems are terminated, times 2.2 gpm until 2 hours2.314815e-5 days <br />5.555556e-4 hours <br />3.306878e-6 weeks <br />7.61e-7 months <br /> when operator action Consideration should also be given to design leakage through is assumed to cross-tie the LPSI suction headers valves isolating ESE recirculation systems from tanks vented and eliminate backleakage through the SIRWT to atmosphere, e.g., emergency core cooling system (EGGS) discharge lines. After 2 hours2.314815e-5 days <br />5.555556e-4 hours <br />3.306878e-6 weeks <br />7.61e-7 months <br />, the SIWRT pump mini-flow return to the refueling water storage tank. backleakage is reduced to two times 0.025 and continues for the remainder of the 30-day duration. SIRWT backleakage is verified through

____________ ________________________________________the In-Service Testing program.

5.3 With the exception of iodine, all radioactive materials in the Conforms With the exception of iodine, all radioactive re-circulating liquid should be assumed to be retained in the materials in the recirculating liquid are assumed to liquid phase. be retained in the liquid phase.

5.4 If the temperature of the leakage exceeds 212 0F, the fraction Conforms Aflashing fraction of 3%was determined based of total iodine in the liquid that becomes airborne should be on the temperature of the containment sump liquid assumed equal to the fraction of the leakage that flashes to at the time recirculation begins. For EGGS

____________vapor. This flash fraction, FF, should be determined using a leakage back to the SIRWT, the analysis 16 of 32

REGULATORY GUIDE 1.183 COMPLIANCE MATRIX PALISADES NUCLEAR PLANT constant enthalpy, h, process, based on the maximum time- demonstrates that the temperature of the leaked dependent temperature of the sump water circulating outside fluid will cool below 212OF prior to release to the the containment[.] SIRWT tank.

5.5 If the temperature of the leakage is less than 212OF or the Conforms The iodine available for release at the time calculated flash fraction is less than 10%, the amount of recirculation begins is based on the expected iodine that becomes airborne should be assumed to be 10% sump pH history and temperature. For the ECCS of the total iodine activity in the leaked fluid, unless a smaller leakage to the auxiliary building, 10% of the total amount can be justified based on the actual sump pH history iodine in the leaked ECCS fluid is assumed to be and area ventilation rates. available for release and isassumed to become airborne and leak directly to the environment from the initiation of recirculation through 30 days. For the ECOS leakage back to the SIRWT, the sump and SIRWT pH history and temperature are used to evaluate the amount of iodine that enters the

__________ _________SIRWT air space.

5.6 The radioiodine that is postulated to be available for release Conforms For ECCS leakage into the auxiliary building, the to the environment isassumed to be 97% elemental and 3% form of the released iodine is 97% elemental and organic. Reduction in release activity by dilution or holdup 3%organic. For EGGS leakage into the SIRWT, within buildings, or by ESF ventilation filtration systems, may the temperature and pH history of the sump and be credited where applicable. SIRWT are considered in determining the radioiodine available for release and the chemical form. Credit is taken for hold-up and dilution of activity in the SIRWT as allowed by Regulatory Position 5.6. Per the current design basis, a 50%

reduction of the ECCS activity is taken for the release to the auxiliary building. No credit for holdup, filtration or dilution of EGGS leakage into I___________________________ I______ the auxiliary building is taken.

ASSUMPTIONS ON MAIN STEAM ISOLATION VALVE LEAKAGE IN BWRS _______________________

6. For BWRs, the main steam isolation valves (MSIVs) have N/A Regulatory Positions 6.1 through 6.5 relate to design leakage that may result in a radioactivity release. The MSSV leakage in BWRs, which is not applicable radiological consequences from postulated MSIV leakage to Palisades.

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

___________I evaluating the consequences of MSIV leakage. ______

17 of 32

REGULATORY GUIDE 1.183 COMPLIANCE MATRIX PALISADES NUCLEAR PLANT ASSUMPTION ON CONTAINMENT PURGING

7. The radiological consequences from post-LOCA primary Conforms Containment purge isnot considered as a means containment purging as a combustible gas or pressure control of combustible gas or pressure control in this measure should be analyzed. Ifthe installed containment analysis. Inaddition, routine containment purge is purging capabilities are maintained for purposes of severe not active for this event. Hydrogen control by accident management and are not credited in any design purge is not part of the licensing basis.

basis analysis, radiological consequences need not be evaluated. Ifthe 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).

Regulatory Guide 1.183 Appendix B: ASSUMPTIONS FOR EVALUATING THE RADIOLOGICAL CONSEQUENCES OF A FUEL HANDLING ACCIDENT 1.- SOURCE Acceptable assumptions regarding core inventory and the I Conforms See response to Regulatory Position 3.

TERM release of radionuclides from the fuel are provided in 1.1 The number of fuel rods damaged during the accident should Conforms The amount of fuel damage is assumed to be all be based on a conservative analysis that considers the most of the fuel rods in a single nominal fuel assembly limiting case. (216).

1.2 The fission product release from the breached fuel is based Conforms The fission product release from the breached fuel on Regulatory Position 3.2 of this guide and the estimate of is based on Regulatory Positions 3.1 and 3.2 of the number of fuel rods breached. All the gap activity in the RG 1.183. Section 1.7 provides a discussion of damaged rods is assumed to be instantaneously released. how the FHA source term is developed. A listing Radionuclides that should be considered include xenons, of the FHA source term is provided in Table 1.7.5-kryptons, halogens, cesiums, and rubidiums. 1. The gap activity available for release is specified by Table 3 of RG 1.183. This activity is assumed to be released from the fuel assembly

_________________________________________________________ ___________instantaneously.

1.3 The chemical form of radioiodine released from the fuel to the Conforms The chemical form of radioiodine released from spent fuel pool should be assumed to be 95% cesium iodide the damaged fuel into the spent fuel pool is

___________ Csl, 485 percent elemental iodine, and 0.15 percent organic _______assumed to be 95% cesium iodide (Csl), 4.85%

18 of 32

REGULATORY GUIDE 1.183 COMPLIANCE MATRIX PALISADES NUCLEAR PLANT iodide. The Osi released from the fuel is assumed to elemental iodine, and 0.15% organic iodide. The completely dissociate in the pool water. Because of the low cesium iodide is assumed to completely pH of the pool water, the iodine re-evolves as elemental dissociate in the spent fuel pool resulting in a final iodine. This is assumed to occur instantaneously, iodine distribution of 99.85% elemental iodine and

____________0.15% organic iodine.

2. - WATER Ifthe depth of water above the damaged fuel is 23 feet or Conforms A minimum water depth of 22.5 feet is maintained DEPTH greater, the decontamination factors for the elemental and above the damaged fuel assembly. Per Section organic species are 500 and 1,respectively, giving an overall 2.0 of Appendix B to Reg. Guide 1.183, the effective decontamination factor of 200 (i.e., 99.5% of the total method of Burley was used to calculate the iodine released from the damaged rods is retained by the elemental iodine decontamination factor. Afactor water). This difference in decontamination factors for of 252 was calculated based on 22.5 feet of water.

elemental(99.85%) and organic iodine (0.15%) species Adecontamination factor of 1.0 was used for results in the iodine above the water being composed of 57% organic iodine. These two decontamination elemental and 43% organic species. factors give an overall decontamination factor of 183.07 for iodine with 22.5 feet of water coverage.

The assumed iodine chemical form after decontamination by the water pool is73% organic and 27% elemental.

3. - NOBLE The retention of noble gases inthe water in the fuel pool or Conforms All of the noble gas released is assumed to exit GASES reactor cavity is negligible (i.e., decontamination factor of 1). the pool without mitigation. All of the non-iodine Particulate radionuclides are assumed to be retained by the particulate nuclides are assumed to be retained water in the fuel pool or reactor cavity (i.e., infinite by the pool water.

decontamination factor). ______

4. - FUEL HANDLING ACCIDENT WITHIN THE FUEL BUILDING 4.1 The radioactive material that escapes from the fuel pool to the Conforms The analysis models the release to the fuel building is assumed to be released to the environment environment over a 2-hour period.

over a 2-hour time period. ______

4.2 A reduction in the amount of radioactive material released Conforms No credit is taken for filtration of the release for from the fuel pool by engineered safety feature (ESE) filter the FHA in containment. For the FHA in the FHB, systems may be taken into account provided these systems three filtration cases were examined. These meet the guidance of Regulatory Guide 1.52 and Generic cases analyzed 10%, 34%, and 50% of the Letter 99-02. release passing through the FHB filtration system.

However, the FHA in containment is the bounding FHA analysis.

4.3 The radioactivity release from the fuel pool should be Conforms No credit istaken for dilution of the release.

assumed to be drawn into the ESF filtration system without mixing or dilution in the fuel building. ______________________________

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REGULATORY GUIDE 1.183 COMPLIANCE MATRIX PALISADES NUCLEAR PLANT

5. - FUEL HANDLING ACCIDENT WITHIN CONTAINMENT_______________________

5.1 If the containment is isolated during fuel handling operations, Conforms The containment equipment hatch is assumed to no radiological consequences need to be analyzed. be open at the time of the fuel handling accident

___________and this configuration has been analyzed.

5.2 If the containment is open during fuel handling operations, but Conforms No automatic isolation of the containment is designed to automatically isolate in the event of a fuel assumed for the FHA.

handling accident, the release duration should be based on delays in radiation detection and completion of containment isolation. If it can be shown that containment isolation occurs before radioactivity is released to the environment, no radiological consequences need to be analyzed.

5.3 If the containment is open during fuel handling operations Conforms The release from the reactor cavity pool is (e.g., personnel air lock or equipment hatch isopen), the assumed to leak to the environment over a two-radioactive material that escapes from the reactor cavity pool hour period.

to the containment is released to the environment over a

_____________2-hour time period.

5.4 Areduction in the amount of radioactive material released Conforms No credit istaken for filtration of the release for from the containment by ESF filter systems may be taken into the FHA in containment. For the FHA in the FHB, account provided that these systems meet the guidance of three filtration cases were examined. These Regulatory Guide 1.52 and Generic Letter 99-02. cases analyzed 10%, 34%, and 50% of the release passing through the FHB filtration system.

However, the FHA in containment is the bounding

___________ _______________________________________________FHA analysis.

5.5 Credit for dilution or mixing of the activity released from the Conforms No credit istaken for dilution or mixing in the reactor cavity by natural or forced convection inside the containment atmosphere.

____________containment may be considered on a case-by-case basis. I____________________________

Regulatory Guide 1.183 Appendix E: ASSUMPTIONS FOR EVALUATING THE RADIOLOGICAL CONSEQUENCES OF A PWR MAIN STEAM LINE BREAK ACCIDENT SOURCE TERMS 1.Assumptions acceptable to the NRC staff regarding core Conforms 2%of the fuel is assumed to experience DNB for inventory and the release of radionuclides from the fuel are the Palisades MSLB event. See response to provided in Regulatory Position 3 of this regulatory guide. The Regulatory Position 3.

release from the breached fuel isbased on Regulatory Position 3.2 of this guide and the estimate of the number of fuel rods breached. The fuel damage estimate should assume

____________I that the highest worth control rod is stuck at its fully withdrawn______________________________

20 of 32

REGULATORY GUIDE 1.183 COMPLIANCE MATRIX PALISADES NUCLEAR PLANT

____ ____ position. _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _

2. If no or minimal fuel damage ispostulated for the limiting Conforms 2% of the fuel is assumed to experience DNB for event, the activity released should be the maximum coolant the Palisades MVSLB event. It was determined activity allowed by the technical specifications. Two cases of that the activity released from the damaged fuel iodine spiking should be assumed. will exceed that released by the two iodine spike cases; therefore, the two iodine spike cases in Regulatory Positions 2.1 and 2.2 were not

_____________explicitly analyzed.

3. The activity released from the fuel should be assumed to be Conforms The activity released from the fuel is assumed to released instantaneously and homogeneously through the be released instantaneously and homogeneously primar coolant. _______through the primary coolant.
4. The chemical form of radioiodine released from the fuel Conforms Regulatory Position 4 - Iodine releases from the should be assumed to be 95% cesium iodide (OsI), 4.85 steam generators to the environment are percent elemental iodine, and 0.15 percent organic iodide. assumed to be 97% elemental and 3%organic.

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

These fractions apply to iodine released as a result of fuel damage and to iodine released during normal operations, Iincluding iodine spiking._______

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 For facilities that have not implemented alternative repair Conforms The primary-to-secondary leak rate is 0.3 gpm per criteria (see Ref. E-1, DG- 1074), the prim ary-to-secondary SG.

leak rate in the steam generators should be assumed to be the leak rate limiting condition for operation specified in the technical specifications. For facilities with traditional generator specifications (both per generator and total of all generators),

the leakage should be apportioned between affected and unaffected steam generators insuch a manner that the calculated dose is maximized. ______

5.2 The density used in converting volumetric leak rates (e.g., Conforms The density used in converting volumetric leak gpm) to mass leak rates (e.g., lbm/hr) should be consistent rates to mass leak rates is consistent with the with the basis of the parameter being converted. The ARC basis of surveillance tests used to show leak rate correlations are generally based on the collection of compliance with the SG leak rate TS (62.4 1cooled liquid. Surveillance tests and facility instrumentation lbmn/ft3).

21 of 32

REGULATORY GUIDE 1.183 COMPLIANCE MATRIX PALISADES NUCLEAR PLANT used to show compliance with leak rate technical specifications are typically based on cooled liquid. In most cases, the density should be assumed to be 1.0 gm/cc (62.4 Ibm/fPt).

5.3 The prim ary-to-secondary leakage should be assumed to Conforms Based on the existing licensing basis, the primary-continue until the primary system pressure is less than the to-secondary leak rate is assumed to continue secondary system pressure, or until the temperature of the until the temperature of the leakage is less than leakage is less than 1OO 0C (21 20F). The release of 212OF at 12 hours1.388889e-4 days <br />0.00333 hours <br />1.984127e-5 weeks <br />4.566e-6 months <br />. The release of radioactivity radioactivity from unaffected steam generators should be from the unaffected SG continues for 8 hours9.259259e-5 days <br />0.00222 hours <br />1.322751e-5 weeks <br />3.044e-6 months <br /> assumed to continue until shutdown cooling is in operation (time to place SDC in operation).

and releases from the steam generators have been

_____________terminated.______________________________

5.4 All noble gas radionuclides released from the primary system Conforms All noble gas radionuclides released from the are assumed to be released to the environment without primary system are assumed to be released to the

___________reduction or mitigation. environment without reduction or mitigation.

5.5 The transport model described in this section should be Conforms The transport model in this section is used for utilized for iodine and particulate releases from the steam iodine and particulate release from the steam generators. This model is shown in Figure E-1 and generators.

summarized below. ______ ____________________

5.5.1 A portion of the prim ary-to-secondary leakage will flash to Conforms Inthe faulted SG, all of the prim ary-to-secon dary vapor, based on the thermodynamic conditions in the reactor leakage is assumed to flash to vapor and be and secondary coolant. During periods of steam generator released to the environment with no mitigation.

dryout, all of the prim ary-to-secondary leakage isassumed to For the unaffected steam generator used for plant flash to vapor and be released to the environment with no cooldown, the prim ary-to-secondary leakage is mitigation. With regard to the unaffected steam generators assumed to mix with the secondary water without used for plant cooldown, the prim ary-to-secondary leakage flashing.

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 Conforms Any postulated leakage that immediately flashes through the bulk water of the steam generator and enter the to vapor is assumed to rise through the bulk water steam space. Credit may be taken for scrubbing in the of the SG into the steam space and isassumed to generator, using the models in NUREG-0409, "Iodine be immediately released to the environment with Behavior ina PWR Cooling System Following a Postulated no mitigation; i.e., no reduction for scrubbing Steam Generator Tube Rupture Accident" (Ref. E-2), during within the SG bulk water is credited.

____________I periods of total submergence of the tubes. ______ _____________________

5.5.3 The leakage that does not immediately flash is assumed to Conforms All leakage that does not immediately flash is

__________I mix with the bulk water. I_____ I_assumed to mix with the bulk water.

22 of 32

REGULATORY GUIDE 1.183 COMPLIANCE MATRIX PALISADES NUCLEAR PLANT 5.5.4 5.5.4 The radioactivity in the bulk water is assumed to Conforms The radioactivity within the bulk water isassumed become vapor at a rate that isthe function of the steaming to become vapor at a rate that is a function of the rate and the partition coefficient. A partition coefficient for steaming rate and the partition coefficient. A iodine of 100 may be assumed. The retention of particulate partition coefficient of 100 is assumed for the radionuclides in the steam generators is limited by the iodine. The retention of particulate radionuclides moisture carryover from the steam generators. in the unaffected SG is limited by the moisture carryover from the SG. The same partition coefficient of 100, as used for iodine, is assumed for other particulate radionuclides. This assumption is consistent with the SG carryover rate of less than 1%. No reduction in the release is assumed from the faulted SG.

5.6 Operating experience and analyses have shown that for some Conforms Steam generator tube bundle uncovery is not steam generator designs, tube uncovery may occur for a predicted or postulated for the intact SG.

short period following any reactor trip (Ref. E-3). The potential impact of tube uncovery on the transport model parameters (e.g., flash fraction, scrubbing credit) needs to be considered.

The impact of emergency operating procedure restoration strategies on steam generator water levels should be

___________ I evaluated.

Regulatory Guide 1.183 Appendix F: ASSUMPTIONS FOR EVALUATING THE RADIOLOGICAL CONSEQUENCES OF A PWR STEAM GENERATOR TUBE RUPTURE ACCIDENT SOURCE TERM ___________________________

1. Assumptions acceptable to the NRC staff regarding core Conforms No fuel damage is postulated to occur for the inventory and the release of radionuclides from the fuel are in Palisades SGTR event. See response for Regulatory Position 3 of this guide. The release from the Regulatory Position 3.

breached fuel is based on Regulatory Position 3.2 of this guide and the estimate of the number of fuel rods breached.

2. If no or minimal fuel damage is postulated for the limiting Conforms No fuel damage is postulated to occur for the event, the activity released should be the maximum coolant Palisades SGTR event. The two cases of iodine activity allowed by technical specification. Two cases of iodine spiking in Regulatory Positions 2.1 and 2.2 are spiking should be assumed. _______assumed.

2.1 A reactor transient has occurred prior to the postulated steam Conforms Case assumes a reactor transient prior to the generator tube rupture (SGTR) and has raised the primary postulated SGTR that raises the primary coolant coolant iodine concentration to the maximum value (typically iodine concentration to the maximum allowed by 160 ipCi/gm DE 1-131) permitted by the technical specifications _______Tech. Specs, which is a value of 40.0 pCi/gmn DE 23 of 32

REGULATORY GUIDE 1.183 COMPLIANCE MATRIX PALISADES NUCLEAR PLANT (i.e., a preaccident iodine spike case). 1-131 for the analyzed conditions. This isthe pre-

_________________________________________accident spike case.

2.2 The primary system transient associated with the SGTR Conforms Case assumes the transient associated with the causes an iodine spike in the primary system. The increase in SGTR causes an iodine spike. The spiking model primary coolant iodine concentration isestimated using a assumes the primary coolant activity is initially at spiking model that assumes that the iodine release rate from the Tech. Spec. value of 1.0 pCi/gmn DE 1-131.

the fuel rods to the primary coolant (expressed in curies per Iodine is assumed to be released from the fuel unit time) increases to a value 335 times greater than the into the PCS at a rate of 335 times the iodine release rate corresponding to the iodine concentration at the equilibrium release rate for a period of 8 hours9.259259e-5 days <br />0.00222 hours <br />1.322751e-5 weeks <br />3.044e-6 months <br />.

equilibrium value (typically 1.0 pCi/gmn DE 1-131) specified in This is the accident-induced spike case.

technical specifications (i.e., concurrent iodine spike case). A concurrent iodine spike need not be considered if fuel damage is postulated. The assumed iodine spike duration should be 8 hours9.259259e-5 days <br />0.00222 hours <br />1.322751e-5 weeks <br />3.044e-6 months <br />. 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 Conforms The activity released from the fuel is assumed to to be released instantaneously and homogeneously through be released instantaneously and homogeneously the primary coolant. _______through the primary coolant.
4. Iodine releases from the steam generators to the environment Conforms Iodine releases from the steam generators to the should be assumed to be 97% elemental and 3%organic. environment are assumed to be 97% elemental

_______________________________________ _________and 3% organic.

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 prim ary-to-secondlary leak rate in the steam generators Conforms The prim ary-to-secondary leak rate is 0.3 gpm per should be assumed to be the leak rate limiting condition for SG.

operation specified in the technical specifications. The leakage should be apportioned between affected and unaffected steam generators insuch a manner that the calculated dose is maximized.

5.2 The density used in converting volumetric leak rates (e.g., Conforms The density used in converting volumetric leak gpm) to mass leak rates (e.g., Ibm/hr) should be consistent rates to mass leak rates is consistent with the with the basis of surveillance tests used to show compliance basis of surveillance tests used to show

____________with leak rate technical specifications. These tests are compliance with the SG leak rate TS (62.4 24 of 32

REGULATORY GUIDE 1.183 COMPLIANCE MATRIX PALISADES NUCLEAR PLANT typically based on cool liquid. Facility instrumentation used to Ibm/ft~).

determine leakage istypically located on lines containing cool liquids. In most cases, the density should be assumed to be 1.0 gm/cc (62.4 Ibm/ft).

5.3 The prim ary-to-secondary leakage should be assumed to Conforms The release of radioactivity from both the affected continue until the primary system pressure is less than the and unaffected SGs is assumed to continue until secondary system pressure, or until the temperature of the shutdown cooling is in operation and steam leakage is less than 10000C (2120 F). The release of release from the SGs is terminated (8hours into radioactivity from the unaffected steam generators should be the event).

assumed to continue until shutdown cooling is in operation and releases from the steam generators have been terminated.______________________

5.4 The release of fission products from the secondary system Conforms The release of fission products from the should be evaluated with the assumption of a coincident loss secondary system is evaluated with the of offsite power. assumption of a coincident loss of offsite power (LOOP).

5.5 All noble gas radionuclides released from the primary system Conforms All noble gases released from the primary system are assumed to be released to the environment without are assumed to be released to the environment

___________reduction or mitigation. without reduction or mitigation.

5.6 The transport model described in Regulatory Positions 5.5 Conforms Regulatory Position 5.6 refers to Appendix E, and 5.6 of Appendix E should be utilized for iodine and Regulatory Positions 5.5 and 5.6. The iodine particulates. transport model for release from the steam generators is as follows:

Appendix E, Regulatory Position 5.5.1 - All steam generators effectively maintain tube coverage.

The prim ary-to-secondary leakage is assumed to mix with the secondary water without flashing for all steam generators.

Appendix E, Regulatory Position 5.5.2 - Aportion of the prim ary-to-secondary ruptured tube flow through the SGTR is assumed to flash to vapor, based on the thermodynamic conditions in the reactor and secondary. The portion that flashes immediately to vapor is assumed to rise through

____________ ________________________________________ _________the bulk water of the SG, enter the steam space, 25 of 32

REGULATORY GUIDE 1.183 COMPLIANCE MATRIX PALISADES NUCLEAR PLANT and be immediately released to the environment.

Scrubbing of the flashed flow in the affected SG is credited. The methodologies presented in NUREG-0409 are used to determine the amount of scrubbing of the flashed flow.

Appendix E, Regulatory Position 5.5.3 - All of the SG tube leakage and ruptured tube flow that does not flash is assumed to mix with the bulk water.

Appendix E, Regulatory Position 5.5.4 - The radioactivity within the bulk water is assumed to become vapor at a rate that is a function of the steaming rate and the partition coefficient. A partition coefficient of 100 is assumed for the iodine. The retention of particulate radionuclides in the SGs is limited by the moisture carryover from the SGs. The same partition coefficient of 100, as used for iodine, is assumed for other particulate radionuclides. This assumption is consistent with the SG carryover rate of less than 1%.

Appendix E, Regulatory Position 5.6 - Steam generator tube bundle uncovery is not postulated

____________ _________________________________________ __________for this event for Palisades.

Regulatory Guide 1.183 Appendix G: ASSUMPTIONS FOR EVALUATING THE RADIOLOGICAL CONSEQUENCES OF A PWR LOCKED ROTOR ACCIDENT SOURCE TERM __________________________

1. Assumptions acceptable to the NRC staff regarding core N/A See Regulatory Position 2 below.

inventory and the release of radionuclides from the fuel are in Regulatory Position 3 of this regulatory guide. The release from the breached fuel is based on Regulatory Position 3.2 of this guide and the estimate of the number of fuel rods

____________breached._______

2. Ifno fuel damage is postulated for the limiting event, a Conforms No fuel damage is postulated for the limiting 26 of 32

REGULATORY GUIDE 1.183 COMPLIANCE MATRIX PALISADES NUCLEAR PLANT radiological analysis is not required as the consequences of locked rotor event for Palisades and no locked this event are bounded by the consequences projected for the rotor radiological analysis has been performed.

main steam line break outside containment. See Section 1.2 of the license amendment

______________request.

3.- 5. This appendix provides assumptions acceptable to the NRC N/A See Regulatory Position 2 above.

staff for evaluating the radiological consequences of a locked rotor accident at PWR light water reactors. These assumptions supplement the guidance provided in the main

____________ I body of this guide. ______ ____________________

Regulatory Guide 1.183 Appendix H: ASSUMPTIONS FOR EVALUATING THE RADIOLOGICAL CONSEQUENCES OF A PWR ROD EJECTION ACCIDENT SOURCE TERM

1. Assumptions acceptable to the NRC staff regarding core Conforms The total core inventory of the radionuclide groups inventory are in Regulatory Position 3 of this guide. For the utilized for determining the source term for this rod ejection accident, the release from the breached fuel is event is based on RG 1.183, Regulatory Position based on the estimate of the number of fuel rods breached 3.1, and is provided in Table 1.7.4-1. The and the assumption that 10% of the core inventory of the inventory provided in Table 1.7.4-1 is adjusted for noble gases and iodines is in the fuel gap. The release the fraction of fuel damaged and a radial peaking attributed to fuel melting is based on the fraction of the fuel factor of 2.04 is applied. The release fractions that reaches or exceeds the initiation temperature for fuel provided in RG 1.183 Table 3 are adjusted to melting and the assumption that 100% of the noble gases and comply with the specific RG 1.183 Appendix H 25% of the iodines contained inthat fraction are available for release requirements. For both the containment release from containment. For the secondary system release and secondary release cases, the activity pathway, 100% of the noble gases and 50% of the iodines in available for release from the fuel gap for fuel that that fraction are released to the reactor coolant. experiences DNB is assumed to be 10% of the noble gas and iodine inventory in the DNB fuel.

For the containment release case for fuel that experiences fuel centerline melt (FCM), 100% of the noble gas and 25% of the iodine inventory in the melted fuel is assumed to be released to the containment. For the secondary release case for fuel that experiences FCM, 100% of the noble gas and 50% of the iodine inventory in the melted fuel is assumed to be released to the primary coolant.

See the resoonse for Reaulatorv Position 3.

2. 1if no fuel damage ispostulated for the limiting event, a 1Conforms IFuel damage of 0.5% DNB and 14.7% fuel melt is-27 of 32

REGULATORY GUIDE 1.183 COMPLIANCE MATRIX PALISADES NUCLEAR PLANT radiological analysis is not required as the consequences of assumed for this event.

this event are bounded by the consequences projected for the loss-of-coolant accident (LOCA), main steam line break, and steam aenerator tube ruDture.

3. Two release cases are to be considered. Inthe first, 100% of Conforms For the containment release case, 100% of the the activity released from the fuel should be assumed to be activity released from the damaged fuel is released instantaneously and homogeneously through the assumed to mix instantaneously and containment atmosphere. Inthe second, 100% of the activity homogeneously in the containment atmosphere.

released from the fuel should be assumed to be completely For the secondary release case, 100% of the dissolved in the primary coolant and available for release to activity released from the damaged fuel is the secondary system. assumed to mix instantaneously and homogeneously in the primary coolant and be available for leakage to the secondary side of the SGs.

4. The chemical form of radioiodine released to the containment Conforms The chemical form of radioiodine released from atmosphere should be assumed to be 95% cesium iodide the damaged fuel to the containment is assumed (OsI), 4.85% elemental iodine, and 0.15% organic iodide. If to be 95% cesium iodide (CsI), 4.85% elemental containment sprays do not actuate or are terminated prior to iodine, and 0.15% organic iodide. Containment accumulating sump water, or if the containment sump pH is sump pH is controlled to 7.0 or higher.

not controlled at values of 7 or greater, the iodine species should be evaluated on an individual case basis. Evaluations of pH should consider the effect of acids created during the rod ejection accident event, e.g., pyrolysis and radiolysis products. With the exception of elemental and organic iodine and noble gases, fission products should be assumed to be in Iparticulate form.

5. Iodine releases from the steam generators to the environment Conforms The chemical form of radioiodine released from should be assumed to be 97% elemental and 3%organic. the SGs to the environment is assumed to be 97%

____________ ________________________________________elemental iodine, and 3%organic iodide.

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 Conforms For the containment leakage case, natural leakage from the containment that isdue to natural deposition in the containment is credited.

deposition, containment sprays, recirculating filter systems, Containment spray is not credited.

dual containments,_or other engineered safety features may I______ I_____________________

28 of 32

REGULATORY GUIDE 1.183 COMPLIANCE MATRIX PALISADES NUCLEAR PLANT be taken into account. Refer to Appendix Ato this guide for guidance on acceptable methods and assumptions for evaluatina these mechanisms.

6.2 The containment should be assumed to leak at the leak rate Conforms The containment is assumed to leak at the incorporated in the technical specifications at peak accident proposed TS maximum allowable rate of 0.10%

pressure for the first 24 hours2.777778e-4 days <br />0.00667 hours <br />3.968254e-5 weeks <br />9.132e-6 months <br />, and at 50% of this leak rate for for the first 24 hours2.777778e-4 days <br />0.00667 hours <br />3.968254e-5 weeks <br />9.132e-6 months <br /> and 0.05% for the remainder the remaining duration of the accident. Peak accident of the event.

pressure is the maximum pressure defined in the technical specifications for containment leak testing. Leakage from subatmospheric containments isassumed to be terminated when the containment is brought to a subatmospheric

_____________ condition as defined in technical specifications._______ _______________________

TRANSPORT FROM SECONDARY SYSTEM

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

7.1 Aleak rate equivalent to the prim ary-to-secondary leak rate Conforms The prim ary-to-secondary leak rate is0.3 gpm per limiting condition for operation specified in the technical SG.

specifications should be assumed to exist until shutdown cooling is in operation and releases from the steam generators have been terminated. ______

7.2 The density used in converting volumetric leak rates (e.g., Conforms The density used in converting volumetric leak gpm) to mass leak rates (e.g., lbm/hr) should be consistent rates to mass leak rates is consistent with the with the basis of surveillance tests used to show compliance basis of surveillance tests used to show with leak rate technical specifications. These tests typically compliance with the SG leak rate TS (62.4 are based on cooled liquid. The facility's instrumentation used Ibm/ft').

to determine leakage typically islocated on lines containing cool liquids. Inmost cases, the density should be assumed to be 1.0 gm/cc (62.4 Ibm/f t3). ______

7.3 All noble gas radionuclides released to the secondary system Conforms All of the noble gas released to the secondary are assumed to be released to the environment without side is assumed to be released directly to the reduction or mitigation. _______environment without reduction or mitigation.

7.4 The transport model described in assumptions 5.5 and 5.6 of Conforms Compliance with Appendix E Sections 5.5 and 5.6 Appendix E should be utilized for iodine and particulates. is discussed below:

Appendix E, Regulatory Position 5.5.1 - Both

_________________________________________ _________steam generators are used for plant cooldown.

29 of 32

REGULATORY GUIDE 1.183 COMPLIANCE MATRIX PALISADES NUCLEAR PLANT Therefore, the prim ary-to-secondary leakage is assumed to mix with the secondary water without flashing.

Appendix E, Regulatory Position 5.5.2 - None of the SG tube leakage is assumed to flash for this event.

Appendix E, Regulatory Position 5.5.3 - All of the SG tube leakage isassumed to mix with the bulk water.

Appendix E, Regulatory Position 5.5.4 - The radioactivity within the bulk water is assumed to become vapor at a rate that is a function of the steaming rate and the partition coefficient. A partition coefficient of 100 isassumed for the iodine. The retention of particulate radionuclides in the SGs is limited by the moisture carryover from the SGs. The same partition coefficient of 100, as used for iodine, is assumed for other particulate radionuclides. This assumption is consistent with the SG carryover rate of less than 1%.

Appendix E, Regulatory Position 5.6 - Steam generator tube bundle uncovery is not postulated

____________ _________________________________________ __________for this event for Palisades.

Regulatory Guide 1.183 Appendix B: ASSUMPTIONS FOR EVALUATING THE RADIOLOGICAL CONSEQUENCES OF A FUEL HANDLING ACCIDENT - Applied to the Spent Fuel Cask Drop Event 1.- SOURCE Acceptable assumptions regarding core inventory and the Conforms Se response for Regulatory Position 3.

TERM release of radionuclides from the fuel are provided in

_____________Regulatory Position 3 of this guide. ______ _____________________

1.1 The number of fuel rods damaged during the accident should Conforms The amount of fuel damaged is consistent with the be based on a conservative analysis that considers the most current design basis (73 damaged fuel

__________limiting case. I_____ I_assemblies).

30 of 32

REGULATORY GUIDE 1.183 COMPLIANCE MATRIX PALISADES NUCLEAR PLANT 1.2 The fission product release from the breached fuel is based Conforms The fission product release from the breached fuel on Regulatory Position 3.2 of this guide and the estimate of is based on Regulatory Positions 3.1 and 3.2 of the number of fuel rods breached. All the gap activity in the RG 1.183. Section 1.7.6 provides a discussion of damaged rods is assumed to be instantaneously released. how the Cask Drop source term is developed. A Radionuclides that should be considered include xenons, listing of the source terms is provided in Table kryptons, halogens, cesiums, and rubidiums. 1.7.6-1. The gap activity available for release is specified by Table 3 of RG 1.183. This activity is

____________ __________assumed to be released instantaneously.

1.3 The chemical form of radioiodine released from the fuel to the Conforms The chemical form of radioiodine released from spent fuel pool should be assumed to be 95% cesium iodide the damaged fuel into the spent fuel pool is (Csl), 4.85 percent elemental iodine, and 0.15 percent organic assumed to be 95% cesium iodide (Csl), 4.85%

iodide. The CsI released from the fuel isassumed to elemental iodine, and 0.15% organic iodide. The completely dissociate in the pool water. Because of the low cesium iodide is assumed to completely pH of the pool water, the iodine re-evolves as elemental dissociate in the spent fuel pooi resulting in a final iodine. This is assumed to occur instantaneously, iodine distribution of 99.85% elemental iodine and

__________0.15% organic iodine.

2. - WATER If the depth of water above the damaged fuel is 23 feet or Conforms Aminimum water depth of 23 feet is maintained DEPTH greater, the decontamination factors for the elemental and above the damaged fuel assemblies. Therefore, organic species are 500 and 1,respectively, giving an overall an overall effective decontamination factor of 200 effective decontamination factor of 200 (i.e., 99.5% of the total is used for the iodine.

iodine released from the damaged rods is retained by the water). This difference in decontamination factors for elemental(99.85%) and organic iodine (0.15%) species results in the iodine above the water being composed of 57%

elemental and 43% organic species.

3. - NOBLE The retention of noble gases inthe water in the fuel pool or Conforms All of the noble gas released is assumed to exit GASES reactor cavity isnegligible (i.e., decontamination factor of 1). the pool without mitigation. All of the non-iodine Particulate radionuclides are assumed to be retained by the particulate nuclides are assumed to be retained water in the fuel pool or reactor cavity (i.e., infinite by the pool water.

decontamination factor)._______

4. - FUEL HANDLING ACCIDENT WITHIN THE FUEL BUILDING _____

4.1 The radioactive material that escapes from the fuel pool to the Conforms The radioactive material released from the fuel fuel building is assumed to be released to the environment pool isassumed to be released from the building

__________over a 2-hour time period. _______-to the environment within 2 hours2.314815e-5 days <br />5.555556e-4 hours <br />3.306878e-6 weeks <br />7.61e-7 months <br />.

4.2 A reduction in the amount of radioactive material released Conforms No credit istaken for filtration of the release for from the fuel pool by engineered safety feature (ESE) filter the 90-day decay case. For the 30-day decay

____________systems may be taken into account provided these systems _______cases, two filtration scenarios were examined.

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REGULATORY GUIDE 1.183 COMPLIANCE MATRIX PALISADES NUCLEAR PLANT meet the guidance of Regulatory Guide 1.52 and Generic These cases analyzed 82.5% and 90% of the

__________Letter 99-02. _______release passing through the FHB filtration system.

4.3 The radioactivity release from the fuel pool should be Conforms No dilution is assumed.

assumed to be drawn into the ESF filtration system without

_____________mixing or dilution in the fuel building._______________________

5. - FUEL HANDLING ACCIDENT WITHIN CONTAINMENT N/A Regulatory Positions 5.1 through 5.5 relate to events within containment. The Spent Fuel Cask Drop event does not occur inside containment so

____________________________________________________ _________that these positions are not applicable.

Regulatory Guide 1.183 Appendix I: ASSUMPTIONS FOR EVALUATING RADIATION DOSES FOR EQUIPMENT QUALIFICATION 1.-13. This appendix addresses assumptions associated with N/A Regulatory Positions 1through 13 apply to equipment qualification that are acceptable to the NRC staff equipment qualification radiological analyses.

for performing radiological assessments. As stated in Equipment qualification radiological analyses are Regulatory Position 6 of this guide, this appendix supersedes not being submitted. See the license amendment Regulatory Positions 2.c.(1) and 2.c.(2) and Appendix D of request.

Revision 1 of Regulatory Guide 1.89, "Environmental Qualification of Certain Electric Equipment Important to Safety for Nuclear Power Plants" (USNRC, June 1984), for operating reactors that have amended their licensing basis to use an alternative source term. Except as stated in this appendix, other assumptions, methods, and provisions of Revision 1 of

_____________Regulatory Guide 1.89 remain effective.

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