JAFP-19-0077, License Amendment Request for Application of the Alternative Source Term for Calculating Loss-of-Coolant Accident Dose Consequences
| ML19220A043 | |
| Person / Time | |
|---|---|
| Site: | FitzPatrick  |
| Issue date: | 08/08/2019 |
| From: | David Gudger Exelon Generation Co |
| To: | Samson Lee Document Control Desk, Office of Nuclear Reactor Regulation |
| References | |
| L-2019-LLA-0171, JAFP-19-0077 | |
| Download: ML19220A043 (1082) | |
Text
No. DPR-59
NRC Docket No. 50-333
200 Exelon Way
Kennett Square. PA 19348
www.exeloncorp.com
Subject: License Amendment Request for Application of the Alternative Source Term
for Calculating Loss-of-Coolant Accident Dose Consequences
References:
1. Regulatory Guide 1.183, "Alternative Radiological Source Terms for
Evaluating Design Basis Accidents at Nuclear Power Reactors," July 2000
2. Letter from J. M. Whitman (U.S. Nuclear Regulatory Commission) to
Technical Specifications Task Force, "Final Safety Evaluation of Technical
Specifications Task Force Traveler TSTF-551, Revision 3, Revise
Secondary Containment Surveillance Requirements (CAC No. MF5125),"
dated September 21, 2017
The purpose of this letter is for Exelon Generation Company, LLC (EGC) to request Nuclear
Regulatory Commission (NRG) approval for adopting the Alternative Source Term (AST), in
accordance with 1 O CFR 50.67, for use in calculating the Loss-of-Coolant Accident (LOCA)
dose consequences at James A. FitzPatrick Nuclear Power Plant (JAFNPP).
Regulatory guidance for the implementation of the AST methodology is provided in
Regulatory Guide (RG) 1.183 (Reference 1 ). This RG provides guidance to licensees of
operating nuclear plants on acceptable applications of AST. The use of the AST changes
only the regulatory assumptions regarding the analytical treatment of design basis
radiological consequence analyses.
The AST analysis for JAFNPP was performed following the guidance in RG 1.183 and
Standard Review Plan 15.0.1, "Radiological Consequence Analyses Using Alternative
Source Terms." The analysis covers the LOCA. This License Amendment Request (LAA)
for the LOCA is a full implementation scope application of the AST, as provided in RG 1.183,
Section 1.2.1 .
U.S. Nuclear Regulatory Commission
License Amendment Request for Application of the Alternative Source
Term for Calculating Loss-of-Coolant Accident Dose Consequences
August 8, 2019
Page 2
The proposed changes to the current licensing and design basis for JAFNPP include:
• Revisions to several Technical Specifications (TS) and associated Bases to
reflect implementation of AST methodology.
• Deletion of the Main Steam Leakage Collection (MSLC) system TS and
associated Bases.
• Revision to increase the allowable TS leakage for the Main Steam Isolation
Valves (MSIVs).
• Revision of the Standby Liquid Control (SLC) system TS and associated Bases.
• Revision of the Ventilation Filter Testing Program (VFTP) TS.
In addition, the proposed change revises SR 3.6.4.1.1, "Secondary Containment," to address
short-duration conditions during which the secondary containment pressure may not meet
the SR pressure requirement. The proposed change is consistent with Technical
Specifications Task Force Traveler (TSTF) 551, "Revise Secondary Containment
Surveillance Requirements," Revision 3 (Reference 2), which was approved by the NRC
on September 21, 2017. The proposed change adds a Note to SR 3.6.4.1.1 that allows the
Secondary Containment vacuum limit to not be met for a short duration period provided an
analysis demonstrates that one Standby Gas Treatment (SGT) subsystem remains capable
of establishing the required Secondary Containment vacuum.
In conjunction with this LAR, JAFNPP is requesting that the NRC grant an exemption from:
1) the requirements of 10 CFR 50, Appendix J, Option B, Paragraph III.A to allow exclusion
of the Main Steam Isolation Valve (MSIV) leakage from the overall integrated leakage rate
measured when performing a Type A Test; and 2) the requirements of 10 CFR 50, Appendix
J, Option B, Paragraph III.B, to allow exclusion of the MSIV leakage rate of the penetration
valves subject to Type B and C tests.
Attachment 1 provides a description and assessment of the proposed changes, the No
Significant Hazards Consideration evaluation pursuant to 10 CFR 50.90, and the
environmental impact evaluation pursuant to 10 CFR 51.22. Attachment 2 describes the
conformance of this LAR to RG 1.183. Attachment 3 provides the existing TS pages
marked-up to show the proposed TS changes. Attachment 4 provides TS Bases pages
marked up to show the associated TS Bases changes and is provided for information only.
Attachments 5 and 6 provide the drawdown analysis and LOCA dose consequence analysis,
respectively, that support the assessment of the proposed changes. Attachment 7 provides
the proposed exemption to 10 CFR 50, Appendix J. Attachment 8 provides responses to the
NRC questions on credit for the SLC system. Attachment 9 provides the Control Room
atmospheric dispersion calculation.
The proposed change has been reviewed by the JAFNPP Plant Operations Review
Committee, in accordance with the requirements of the EGC Quality Assurance Program.
EGC requests approval of the proposed license amendment by August 8, 2020. Once
approved, the amendment shall be implemented within 60 days.
U.S. Nuclear Regulatory Commission
License Amendment Request for Application of the Alternative Source
Term for Calculating Loss-of-Coolant Accident Dose Consequences
August 8, 2019
Page 3
In accordance with 1 O CFR 50.91, "Notice for public comment; State consultation,"
paragraph (b), a copy of this application, with attachments, is being provided to the
designated State Officials.
There are no regulatory commitments contained in this submittal. Should you have any
questions concerning this submittal, please contact Tom Loomis at (610) 765-5510.
I declare under penalty of perjury that the foregoing is true and correct. This statement was
executed on the 9th day of August 2019.
Respectfully,
J_ /,.- J -r L Jr-----
David T. Gudger
Senior Manager - Licensing
Exelon Generation Company, LLC
Attachments:
1. Assessment of the Proposed Change
2. Regulatory Guide 1.183 Conformance Matrix
3. Mark-Up of JAFNPP Technical Specifications Pages
4. Mark-Up of JAFNPP Technical Specifications Bases Pages
5. JAF-CALC-19-00001, Revision O - JAFNPP Secondary Containment
Drawdown Analysis
6. JAF-CALC-19-00005, Revision 0 - JAFNPP Post-LOCA EAB, LPZ, and CR
Dose - AST Analysis
7. Proposed Exemption to 10 CFR 50, Appendix J
8. Response to Standard NRC Questions Concerning Credit for Standby Liquid
Control System
9. JAF-CALC-19-00004, Revision O - Control Room Atmospheric Dispersion for
Turbine Building Release
cc: USNRC Region I, Regional Administrator
USNRC Senior Resident Inspector, JAFNPP
USNRC Senior Project Manager, JAFNPP
A. L. Peterson, NYSERDA
Attachment 1
Assessment of the Proposed Change
Subject: License Amendment Request for Application of the Alternative Source Term for
Calculating Loss-of-Coolant Accident Dose Consequences
SUMMARY DESCRIPTION .............................................................................................. 1
DETAILED DESCRIPTION .............................................................................................. 1
REASON FOR THE PROPOSED CHANGE ............................................................................ 2
DESCRIPTION OF THE PROPOSED CHANGES ..................................................................... 3
TECHNICAL EVALUATION ............................................................................................. 4
INTRODUCTION ............................................................................................................... 4
ATTRIBUTES OF THE JAFNPP AST .................................................................................. 5
Accident Source Term ............................................................................................... 5
Release Fractions ..................................................................................................... 5
TIMING OF RELEASE PHASES........................................................................................... 5
RADIONUCLIDE COMPOSITION ......................................................................................... 5
CHEMICAL FORM ............................................................................................................ 5
KEY AST INPUT PARAMETERS ......................................................................................... 6
ATMOSPHERIC DISPERSION FACTORS (/Q) ..................................................................... 7
OFFSITE DOSE CONSEQUENCES ..................................................................................... 7
CONTROL ROOM DOSE CONSEQUENCE ........................................................................... 8
ENVIRONMENTAL QUALIFICATION (EQ) ............................................................................ 8
LOSS OF COOLANT ACCIDENT ......................................................................................... 9
Recirculation Line Rupture Vs Main Steam Line Rupture ..................................... 10
Source Term ........................................................................................................ 16
Post-LOCA Containment Leakage ....................................................................... 16
Suppression Pool pH ........................................................................................... 17
Reduction in Airborne Activity Inside Containment ............................................... 17
Reactor Building .................................................................................................. 19
Containment Purging ........................................................................................... 20
Post-LOCA ESF Leakage .................................................................................... 20
Post-LOCA MSIV Leakage .................................................................................. 20
Determination of MSIV Leak Rates ...................................................................... 20
Settling Velocity for Aerosol Deposition in the Main Steam Lines ......................... 21
Aerosol Deposition in Horizontal Main Steam Lines ............................................. 23
Elemental Iodine Removal Rate ........................................................................... 24
Control Room Model ............................................................................................ 24
LOCA Analysis Results ........................................................................................ 25
Vital Area Accessibility ......................................................................................... 25
APPLICABILITY OF TSTF-551 SAFETY EVALUATION ........................................................ 26
NRC REGULATORY ISSUE SUMMARY 2006-04 ............................................................... 28
REGULATORY EVALUATION ...................................................................................... 35
APPLICABLE REGULATORY REQUIREMENTS/CRITERIA ..................................................... 35
NO SIGNIFICANT HAZARDS CONSIDERATION ................................................................... 36
ENVIRONMENTAL CONSIDERATION.......................................................................... 37
REFERENCES ............................................................................................................... 38
Attachment 1
Assessment of the Proposed Change
Page 1
SUMMARY DESCRIPTION
In accordance with 10 CFR 50.90, "Application for amendment of license, construction permit,
or early site permit," Exelon Generation Company, LLC (EGC) requests an amendment to the
Technical Specifications (TS) for James A. FitzPatrick Nuclear Power Plant (JAFNPP).
The proposed amendment revises the radiological assessment calculational methodology for
the Design Basis Accident (DBA) Loss-of-Coolant Accident (LOCA) at JAFNPP through
application of the Alternative Source Term (AST), in accordance with the provisions of
10 CFR 50.67, "Accident source term." EGC requests the Nuclear Regulatory Commission
(NRC) review and approval of the AST LOCA methodology for application to JAFNPP. This
application represents a full scope implementation of AST, as provided for in Regulatory Guide (RG) 1.183 (Reference 6.1, Section 1.2.1), with exception that Technical Information Document
(TID)-14844, "Calculation of Distance Factors for Power and Test Reactor Sites"
(Reference 6.2), will continue to be used as the radiation dose basis for equipment qualification.
The revised LOCA analysis employs the guidance provided in Regulatory Position 1.3 and
Appendix A of RG 1.183.
The AST LOCA analysis for JAFNPP was performed following the guidance in Standard Review
Plan 15.0.1, "Radiological Consequence Analyses Using Alternative Source Terms" (Reference
6.3), and 10 CFR 50.67. Attachment 2 of this License Amendment Request (LAR) provides an
RG 1.183 conformance matrix. The LOCA AST dose calculation, upon which this LAR is based,
is provided in Attachment 6. This calculation was developed using the NRC approved
RADTRAD Version 3.03 software (Reference 6.4).
Approval of this LAR will replace the current design basis source term assumptions and
radiological criteria for the LOCA radiological consequences. In accordance with the AST
LOCA analysis results, revisions to the JAFNPP Technical Specifications (TS) and TS Bases
are proposed based on the revised safety analysis assumptions for a postulated LOCA.
DETAILED DESCRIPTION
On December 23, 1999, the NRC published regulation 10 CFR 50.67 in the Federal Register.
This regulation provides a mechanism for operating license holders to revise the current
accident source term used in design-basis radiological analyses with an AST. Regulatory
guidance for the implementation of AST is provided in RG 1.183 (Reference 6.1). RG 1.183
provides NRC-accepted guidance for application of AST. The use of AST changes only the
regulatory assumptions regarding the analytical treatment of the DBAs.
The fission product release from the reactor core into the Primary Containment (Drywell) is
referred to as the "source term," and it is characterized by the composition and magnitude of the
radioactive material, the chemical and physical properties of the material, and the timing of the
release from the reactor core as discussed in TID-14844 (Reference 6.2). Since the publication
of TID-14844, significant advances have been made in understanding the composition and
magnitude, chemical form, and timing of fission product releases from severe nuclear power
plant accidents. Many of these insights developed out of the major research efforts started by
the NRC and the nuclear industry after the accident at Three Mile Island. NUREG-1465
(Reference 6.5) was published in 1995 with revised ASTs for use in the licensing of future light
water reactors (LWRs). The NRC, in 10 CFR 50.67, later allowed the use of the ASTs
described in NUREG-1465 at operating plants. This NUREG represents the result of decades
of research on fission product release and transport in LWRs under accident conditions. One of
the major insights summarized in NUREG-1465 involves the timing and duration of fission
product releases.
Attachment 1
Assessment of the Proposed Change
Page 2
The requested license amendment involves a full-scope application of the AST, addressing the
composition and magnitude of the radioactive material, its chemical and physical form, and the
timing of its release as described in RG 1.183.
A limited alternative source term License Amendment was approved in an NRC Safety
Evaluation Report dated September 12, 2002.
EGC has performed radiological consequence analysis of the LOCA to support full-scope
implementation of AST. The implementation consisted of the following tasks:
• Identification of the AST based on plant-specific analysis of core fission product
inventory
• Application of release fractions for the LOCA DBA that could potentially result in Control
Room (CR) and offsite doses
• Analysis of the atmospheric dispersion for the radiological propagation pathways
• Calculation of fission product deposition rates and transport and removal mechanisms
• Calculation of offsite and Control Room personnel Total Effective Dose Equivalent
(TEDE) doses
• Evaluation of Suppression Pool pH to ensure that the iodine deposited into the
Suppression Pool during a DBA LOCA does not re-evolve and become airborne as
elemental iodine.
EGC is requesting the use of AST for several areas of operational relief for systems used in the
event of a DBA, and without crediting the use of certain previously assumed safety
systems/functions.
The proposed changes to the current licensing basis for JAFNPP that are justified by the AST
analysis include:
• Revisions to several Technical Specifications (TS) and associated Bases to reflect
implementation of AST methodology.
• Deletion of the Main Steam Leakage Collection (MSLC) system TS and associated
Bases.
• Revision to increase the allowable TS leakage for the Main Steam Isolation Valves
(MSIVs).
• Revision of the Standby Liquid Control (SLC) system TS and associated Bases.
• Revision of the Ventilation Filter Testing Program (VFTP) TS.
Reason for the Proposed Change
The primary motivation for this amendment request is to incorporate a revised source term
based on the Core Average Exposure (CAVEX). The CAVEX source term allows increased
operational flexibility by bounding a range of core average exposures (GWd/MTU) and fuel
enrichments. A second motivation for this amendment request is the increase in allowable
MSIV leakage. The third reason for this amendment request is to remove the requirement for
post-accident operation of the MSLC system. Refurbishment of an MSIV to meet the current
SR 3.6.1.3.10 leakage rate limit is a labor intensive effort which results in a cumulative worker
radiation dose and expenditure of resources. Increasing the MSIV leakage rate limit would
significantly reduce the amount of rework on the MSIVs. The change would lower personnel
radiation exposure and improve the overall performance integrity of the MSIVs by reducing the
number of maintenance activities associated with restoring the leakage to an overly strict lower
limit. Approval of this proposed change would also be an economic benefit to EGC in terms of
direct costs and a reduction in outage activities.
Attachment 1
Assessment of the Proposed Change
Page 3
Description of the Proposed Changes
The proposed revisions to the JAFNPP TS include:
Table of Contents
The Table of Contents is being revised to delete reference to the Main Steam Leakage
Collection (MSLC) system.
TS 3.1.7, "Standby Liquid Control (SLC) System"
Added MODE 3 to the applicability statement and added the requirement to be in MODE 4
within 12 hours1.388889e-4 days <br />0.00333 hours <br />1.984127e-5 weeks <br />4.566e-6 months <br /> if a required action was not met.
This change is needed to support the use of the SLC system for buffering Suppression Pool pH
as assumed in the LOCA analysis performed in support of this AST LAR (see Attachment 8).
TS 3.3.6.1-1, "Primary Containment Isolation Instrumentation"
Added MODE 3 to the applicable mode column for item d., “SLC System Initiation.”
This change is needed for the reason stated above for TS 3.1.7.
TS 3.6.1.3, "Primary Containment Isolation Valves (PCIVs)"
Revised Surveillance Requirement (SR) 3.6.1.3.10 to increase the combined main steam line
leakage rate from 46 standard cubic feet per hour (scfh) to 200 scfh when tested at greater than
or equal to 25 psig. It is also revised to include a leakage limit of less than or equal to 100 scfh
for a single main steam line when tested at greater than or equal to 25 psig.
The new allowable limit for the combined main steam leakage is a relaxation from the current
requirements. The acceptability of this new limit is demonstrated in the supporting AST
accident analysis. The resulting radiological consequences are within the applicable regulatory
limits.
TS 3.6.1.8, "Main Steam Leakage Collection (MSLC) System"
This TS is deleted in its entirety.
This TS provided operability requirements for the MSLC system. This system is no longer
credited for the mitigation of any DBA in the accident analyses performed in support of this AST
LAR. Therefore, a TS requiring the operability of this system is no longer necessary and this
deletion is consistent with the criteria of 10 CFR 50.36. The criteria given in 10 CFR 50.36.c.2
are addressed below:
Criterion 1. Installed instrumentation that is used to detect, and indicate in the control room, a
significant abnormal degradation of the reactor coolant pressure boundary.
The JAFNPP MSLC system does not provide any detection of abnormal degradation of
the reactor coolant pressure boundary.
Criterion 2. A process variable, design feature, or operating restriction that is an initial condition
of a design basis accident or transient analysis that either assumes the failure of or presents a
challenge to the integrity of a fission product barrier.
The JAFNPP MSLC system does provide a process variable, design feature, or
operating restriction that is an initial condition of a design basis accident.
Attachment 1
Assessment of the Proposed Change
Page 4
Criterion 3. A structure, system, or component that is part of the primary success path and
which functions or actuates to mitigate a design basis accident or transient that either assumes
the failure of or presents a challenge to the integrity of a fission product barrier.
The current JAFNPP MSLC system would be actuated following a DBA to mitigate the
consequences of the accident by directing any MSIV leakage to the Standby Gas
Treatment System. However, the revised LOCA analysis demonstrates that the offsite
and onsite dose consequences following a design basis accident are acceptable and
meet the requirements of 10 CFR 50.67 without operation of this system. Therefore, this
TS can be deleted.
Criterion 4. A structure, system, or component which operating experience or probabilistic risk
assessment has shown to be significant to public health and safety.
The revised LOCA analysis demonstrates that the offsite and onsite dose consequences
of the design basis accident are acceptable and meet the requirements of 10 CFR 50.67
without operation of the MSLC system. Consequently, this system is not significant to
the public health and safety and this TS can be deleted.
TS 3.6.4.1, "Secondary Containment"
The proposed change revises SR 3.6.4.1.1, "Secondary Containment," to address shortduration
conditions during which the Secondary Containment pressure may not meet the
Surveillance Requirement pressure requirement. The proposed change is consistent with
Technical Specifications Task Force Traveler (TSTF) 551 (TSTF-551), "Revise Secondary
Containment Surveillance Requirements," Revision 3, which was approved by the NRC on
September 21, 2017 (Reference 6.23). The proposed change adds a Note to SR 3.6.4.1.1 that
allows the Secondary Containment vacuum limit to not be met for a short duration period
provided an analysis demonstrates that one Standby Gas Treatment (SGT) subsystem remains
capable of establishing the required Secondary Containment vacuum. The portion of TSTF-551
that modifies SR 3.6.4.1.3 is also incorporated into the JAFNPP TS SR 3.6.4.1.3 and is included
in this License Amendment Request.
TS 5.5.8.c, "Ventilation Filter Testing Program (VTFP)"
The proposed change will incorporate new testing requirements for the Standby Gas Treatment
System (SGTS) and the Control Room Emergency Ventilation Air Supply System (CREVAS)
charcoal adsorbers. This change is necessary to make the testing requirements consistent with
the revised design basis analysis.
Attachment 3 contains a marked-up version of the JAFNPP TS showing the proposed changes.
Attachment 4 provides the marked-up TS Bases pages which are being submitted for
information only.
TECHNICAL EVALUATION
Introduction
The Current Licensing Basis (CLB) LOCA analysis utilizes the guidance of TID-14844,
"Calculation of Distance Factors for Power and Test Reactor Sites" (Reference 6.2), thus,
conversion to the AST methodology for the LOCA accident will require a license amendment. In
accordance with RG 1.183, implementation of the AST methodology for the LOCA accident
constitutes full implementation of the AST methodology and future revisions of other dose
calculations (Control Rod Drop, Main Steam Line Break, etc.) will need to implement the AST
methodology.
Attachment 1
Assessment of the Proposed Change
Page 5
The LOCA analysis evaluates the Exclusion Area Boundary (EAB), Low Population Zone (LPZ),
and Control Room (CR) doses for JAFNPP using the methodology of Regulatory Guide 1.183.
JAFNPP currently implements the AST methodology for the Fuel Handling Accident only.
The JAFNPP AST is based on one major accident (i.e., LOCA), hypothesized for the purposes
of design analysis or consideration of possible accidental events that could result in hazards not
exceeded by those from other accidents considered credible. The AST LOCA analysis
addresses events that involve a substantial meltdown of the core with the subsequent release of
appreciable quantities of fission products, the times and rates of appearance of radioactive
fission products released into containment, the types and quantities of the radioactive species
released, and the chemical forms of iodine released.
Accident Source Term
The inventory of fission products in the reactor core that is available for release to the
containment is based on the maximum full power operation of the core with bounding values for
fuel enrichment and fuel burnup. The core power used in the analyses is 102% of the current
licensed thermal power level (i.e., 102% x 2536 MWt). The period of irradiation is of sufficient
duration to allow the activity of dose-significant radionuclides to reach equilibrium or to reach
maximum values. The core inventory is based on a Core Average Exposure (CAVEX) isotopic
inventory for average core exposure of 12 to 43 GWD/MTU.
Release Fractions
The core inventory release fractions, by radionuclide groups, for the gap release and early invessel
damage phases for the DBA LOCA listed in Table 1 of RG 1.183 (Reference 6.1) for
boiling water reactors (BWRs) are used. These fractions are applied to the equilibrium core
inventory developed for JAFNPP.
Timing of Release Phases
Table 4 of RG 1.183 tabulates the onset and duration of each sequential release phase for DBA
LOCAs. The specified onset is the time following the initiation of the accident (i.e., time = 0).
The early in-vessel phase immediately follows the gap release phase. The activity released
from the core during each release phase is conservatively modeled as increasing in a linear
fashion over the duration of the phase. The JAFNPP AST analysis conservatively assumes
releases during each phase occur at the beginning of the release phase.
Radionuclide Composition
The elements and radionuclide groups listed in Table 5 of RG 1.183 are used in the JAFNPP
AST analysis.
Chemical Form
Of the radioiodine released from the Reactor Coolant System (RCS) to the containment in a
postulated accident, which includes releases from the gap and the fuel pellets, 95% of the
iodine released is assumed to be cesium 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. However, the transport of these iodine species
following release from the fuel may affect these assumed fractions. The accident-specific
descriptions that follow provide additional details.
Attachment 1
Assessment of the Proposed Change
Page 6
Key AST Input Parameters
Key baseline parameters, associated changes in the LOCA analysis parameters, and
associated license change objectives are summarized in Table 3.6-1.
Table 3.6-1
General AST Parameter or Method
Parameter Pre-AST Value AST Value Comments
Core Power Level 2,535.8 MWt
+ 2% margin =
2586.52 MWt
2,536 MWt
+ 2% margin =
2586.7 MWt
Licensed power
level unchanged
(rounded up)
Leakage
1.5 wt%/day 1.5wt%/day Pre-AST Primary
Containment
Leakage Included
MSIV Leakage
MSIV Leak Rate 46 scfh at 25 psig
200 scfh at 25 psig,
or
270 scfh at 45 psig
Total for all four
accident pressure
MSIV Leakage
Pathway
Routed by MSLC
system to SGTS
Release to Turbine
Building with
release to
environment as
unfiltered ground
level release
New /Q values
established
Aerosol Deposition
Model
Not Applicable 20-group probabilistic distribution of
aerosol settling velocity for MSIV leakage
based on AEB 98-03 (Reference 6.6) and
including RIS 2006-04 (Reference 6.7)
guidance
Portion of Main Steam
Lines Credited for
Aerosol Deposition
Not Applicable Credited between the reactor pressure
vessel (RPV) nozzle and turbine stop
Elemental Removal
Model
Not Applicable Time and temperature dependent
removal efficiency based on J. E. Cline
methodology (Reference 6.8)
Attachment 1
Assessment of the Proposed Change
Page 7
Table 3.6-1
General AST Parameter or Method
Reduction in Airborne
Activity Inside
Containment
Not Credited Credit taken for Drywell sprays in the
removal of aerosols and elemental iodine
based on Standard Review Plan (SRP)
6.5.2 guidance
Standby Gas
Treatment Filtration
90% HEPA
90% Charcoal
97% HEPA
97% Charcoal
TS Penetration
requirements
modified accordingly
Control Room Intake
Filtration
90% HEPA
90% Charcoal
97% HEPA
97% Charcoal
TS Penetration
requirements
modified accordingly
Atmospheric Dispersion Factors (/Q)
RG 1.183 regulatory position 5.3 states that "Atmospheric dispersion values (/Q) for the EAB,
the LPZ, and the Control Room that were approved by the staff during initial facility licensing or
in subsequent licensing proceedings may be used in performing the radiological analyses
identified by this guide." In accordance with this guidance, atmospheric dispersion values
(/Qs) for the EAB, the LPZ, and the Control Room that were previously approved by the Staff
are used in the LOCA analysis. These atmospheric dispersion values were based on
meteorological data from 1985-1992 and provided /Qs for stack releases and ground level
releases from the Reactor Building. These analyses were based on the guidance of Regulatory
Guides 1.111 and 1.145.
Because the MSIV ground level Turbine Building release is a new release pathway, new /Qs
were developed for the Control Room and Technical Support Center (TSC) for this release
pathway. This analysis used the meteorological data from 1985-1992 to maintain consistency
with the other atmospheric dispersion calculations of record and used the ARCON96 computer
code. The 1985-1992 meteorological data is the same data used in the limited scope Alternate
Source Term license amendment for the Fuel Handling Accident which was previously
approved by the NRC (Reference 6.19). A benchmark was performed comparing the
meteorological data from 2014-2018 to the 1985-1992 data, which confirmed the 1985-1992
data is still representative of current site conditions and remains adequate for calculating
atmospheric dispersion factors. The atmospheric dispersion factors were developed using the
NRC sponsored computer code ARCON96 and the guidance from Regulatory Guide 1.194.
Details of the atmospheric dispersion factors calculated for the MSIV ground level Turbine
Building release are provided in Attachment 9.
Offsite Dose Consequences
The following assumptions are used in determining the TEDE for the maximum exposed
individual at EAB and LPZ locations.
Attachment 1
Assessment of the Proposed Change
Page 8
• The offsite dose is determined as a TEDE, which is the sum of the Committed Effective
Dose Equivalent (CEDE) from inhalation and the Deep Dose Equivalent (DDE) from
external exposure from all radionuclides that are significant with regard to dose
consequences and the released radioactivity. The RADTRAD computer code performs
this summation to calculate the TEDE.
• The offsite dose analysis uses the CEDE Dose Conversion Factors (DCFs) for inhalation
exposure. Table 2.1 of Federal Guidance Report 11, "Limiting Values of Radionuclide
Intake and Air Concentration and Dose Conversion Factors for Inhalation, Submersion,
and Ingestion" (Reference 6.9) provides tables of conversion factors acceptable to the
NRC. The factors in the column headed "effective" yield doses corresponding to the
CEDE.
• Because RADTRAD calculates DDE using whole body submergence in a semi-infinite
cloud with appropriate credit for attenuation by body tissue, the DDE can be assumed
nominally equivalent to the Effective Dose Equivalent (EDE) from external exposure.
Therefore, the offsite dose analysis uses EDE in lieu of DDE DCFs in determining
external exposure. Table III.1 of Federal Guidance Report 12, "External Exposure to
Radionuclides in Air, Water, and Soil" (Reference 6.10), provides external EDE
conversion factors acceptable to the NRC. The factors in the column headed "effective"
yield doses corresponding to the EDE.
• The maximum EAB TEDE for any two-hour period following the start of the radioactivity
release is determined and used in determining compliance with the dose acceptance
criteria in 10 CFR 50.67.
• TEDE is determined for the most limiting receptor at the outer boundary of the LPZ and
is used in determining compliance with the dose criteria in 10 CFR 50.67.
• No correction is made for depletion of the effluent plume by deposition on the ground.
Control Room Dose Consequence
The following dose contributions were considered in determining the TEDE for maximum
exposed individuals located in the CR:
• Contamination of the Control Room atmosphere by the filtered intake of radioactive
material contained in the radioactive plume released from the facility.
• Contamination of the Control Room atmosphere by the unfiltered infiltration of airborne
radioactive material from areas and structures adjacent to the Control Room envelope.
• Radiation shine from the external radioactive plume released from the facility (i.e.,
external airborne cloud).
• Radiation shine from radioactive material in the Reactor Building.
• Radiation shine from radioactive material in systems and components external to the
Control Room envelope (e.g., radioactive material buildup on ventilation filters).
The radioactivity releases and radiation levels used for the Control Room dose are determined
using the same source term, transport, and release assumptions used for determining the EAB
No credit for potassium iodide pills or respiratory protection is taken.
Environmental Qualification (EQ)
Regulatory Position 6 of RG 1.183 (Reference 6.1) states: "The NRC staff is assessing the
effect of increased cesium releases on EQ doses to determine whether licensee action is
warranted. Until such time as this generic issue is resolved, licensees may use either the AST
or the TID-14844 assumptions for performing the required EQ analyses. However, no plant
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modifications are required to address the impact of the difference in source term characteristics
(i.e., AST vs. TID-14844) on EQ doses." This generic issue has been resolved in a memo dated
April 30, 2001 (ADAMS Accession No. ML011210348) and in Supplement 25 to NUREG-0933,
Generic Issue 187. These documents showed that exposure to containment atmosphere
sources based on traditional source term methodology and AST methodology, produced similar
integrated doses and that the integrated AST doses from exposure to post-LOCA sump fluid did
not exceed those based on TID-14844 assumptions until 145 days after an event at a BWR.
The NRC staff concluded in the memo and NUREG-0933 that there was no clear basis for backfitting
the requirement to modify the design basis for EQ to adopt the AST and there would be
no discernable risk reduction associated with such a requirement. The staff also concluded that
longer term equipment operability issues associated with severe fuel damage accidents (with
which the AST is associated) could also be addressed under accident management or plant
recovery actions as necessary. A 145-day plant recovery period provides time to bring in
significant external resources to supplement installed plant equipment.
Additionally, qualification of safety-related equipment from the radiation environment resulting
from a DBA LOCA will continue to be based on the original TID-14844 based accident treatment
resulting from a DBA. This practice is recognized as acceptable because of the minimal public
health and safety benefit and substantial cost of re-evaluation of radiation environment
characterization with AST based assumptions of core releases and timing. The changes in
plant parameters in the LOCA calculation do not impact conclusions reached or the general
underlying parameters related to Primary Containment sources, Secondary Containment
airborne sources, and engineered safety feature (ESF) piping sources.
For the above reasons, it is not necessary to revise the JAFNPP equipment qualification
program to convert to alternative source term assumptions and the JAFNPP EQ Program will
continue to be based on the TID-14844 assumptions and methodology.
The increased allowable MSIV leakage was evaluated for potential impacts to the radiation
environments utilized in the EQ program and it was determined that this change does not
impact the doses currently evaluated in the EQ program. However, a new EQ evaluation is
required because the MSLC system is no longer credited to prevent leakage into the Turbine
Building post-LOCA. Previously this leakage pathway was not considered because the MSLC
system, as described in UFSAR Section 9.19, directed any MSIV leakage to the SGTS. The
impact of this change to any safety-related equipment in the Turbine Building was evaluated
and it was determined that there are no safety-related components in the Turbine Building that
would need to be added to the EQ program due to elimination of credit of the MSLC system.
This conclusion is based on specific dose analyses of the safety-related cables and components
in the Turbine Building general area or Electrical Switchgear Bays that are required post-LOCA.
This analysis determined that if the cables and components are required to perform a safetyrelated
function in the post-LOCA environment, the post-LOCA Total Integrated Dose (TID) to
these cables and components is less than 1E4 Rads, and therefore, the components are
located in a mild radiation environment post-LOCA.
Loss of Coolant Accident
Appendix A, "General Design Criteria for Nuclear Power Plants," to 10 CFR Part 50 defines
LOCAs as those postulated accidents that result from a loss of coolant inventory at rates that
exceed the capability of the reactor coolant makeup system. Leaks up to a double-ended
rupture of the largest pipe of the RCS are included. The LOCA is a conservative surrogate
accident that is intended to challenge selective aspects of the facility design. Analyses are
performed using a spectrum of break sizes to evaluate fuel and Emergency Core Cooling
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System (ECCS) performance. Regarding radiological consequences, a large-break LOCA is
assumed as the design basis case for evaluating the performance of release mitigation systems
and the containment and for evaluating the proposed siting of a facility.
The JAFNPP LOCA was analyzed using a conservative set of assumptions and as-built design
input parameters compatible for AST and the TEDE dose criteria. The numeric values of the
critical design inputs were conservatively selected to assure an appropriate prudent safety
margin against unpredicted events in the course of an accident and to compensate for large
uncertainties in facility parameters, accident progression, radioactive material transport, and
atmospheric dispersion.
The design inputs used for the design analyses were extracted from JAFNPP licensing basis
documents, Updated Final Safety Analysis Report (UFSAR) sections, existing calculations,
design basis documents, and regulatory guidance documents. Key parameters used in the
LOCA analysis are summarized in Table 3.11-1.
Recirculation Line Rupture Vs Main Steam Line Rupture
Appendix A, "General Design Criteria for Nuclear Power Plants," to 10 CFR Part 50 defines
LOCAs as those postulated accidents that result from a loss of coolant inventory at rates that
exceed the capability of the reactor coolant makeup system. Leaks up to a double-ended
rupture of the largest pipe of the reactor coolant system are included. The LOCA, as with all
DBAs, is a conservative surrogate accident that is intended to challenge selective aspects of the
facility design. With regard to radiological consequences, a large-break LOCA is assumed as
the design basis case for evaluating the performance of release mitigation systems and the
containment response. Therefore, a recirculation line rupture is considered as the initiating
event rather than a main steam line rupture.
Per the JAFNPP FSAR Update, Section 6.5.3.1, the DBA LOCA is defined as the instantaneous
guillotine rupture of the recirculation pipe with displacement of both ends so that blowdown
occurs from both ends. This LOCA leads to a specific combination of dynamic, quasi-static, and
static loads in time. The thermal transient due to other postulated events including the steam
line break inside the Drywell does not impose maximum challenge to Drywell pressure boundary
and fuel integrity. The DBA LOCA results in the maximum core damage and fission product
releases as shown in the RG 1.183, Table 1. Therefore, a recirculation line rupture is
considered to be the limiting event with respect to radiological consequences.