RC-09-0004, License Amendment Request - LAR 04-02911 License Amendment and Related Technical Specification Changes to Implement Full-scope Alternative Source Term in Accordance with 10 CFR 50.67

License Amendment Request - LAR 04-02911 License Amendment and Related Technical Specification Changes to Implement Full-scope Alternative Source Term in Accordance with 10 CFR 50.67

ML090720887
Person / Time
Site:	Summer
Issue date:	02/17/2009
From:	Archie J South Carolina Electric & Gas Co, South Carolina Public Service Authority
To:	Martin R Document Control Desk, Office of Nuclear Reactor Regulation
References
LAR 04-02911, RC-09-0004
Download: ML090720887 (201)

Text

Jeffrey B. Archie Vice President,Nuclear Operations 803.345.4214 February 17, 2009 RC-09-0004 A SCANA COMPANY Document Control Desk U. S. Nuclear Regulatory Commission Washington, DC 20555 ATTN: R. E. Martin

Subject:

VIRGIL C. SUMMER NUCLEAR STATION (VCSNS)

DOCKET NO. 50/395 OPERATING LICENSE NO. NPF-12 LICENSE AMENDMENT REQUEST - LAR 04-02911 LICENSE AMENDMENT AND RELATED TECHNICAL SPECIFICATION CHANGES TO IMPLEMENT FULL-SCOPE ALTERNATIVE SOURCE TERM IN ACCORDANCE WITH 10 CFR 50.67 South Carolina Electric & Gas Company (SCE&G), acting for itself and as agent for South Carolina Public Service Authority, hereby requests an amendment to the VCSNS licensing basis and Technical Specifications (TS) that supports a full implementation application of an Alternative Source Term (AST) methodology with the following exceptions. The exceptions are that the current TID-14844 accident source term will remain the licensing basis for equipment qualification, NUREG-0737 evaluations other than Control Room Habitability Envelope (CRHE) doses, and FSAR accidents not included in Regulatory Guide (RG) 1.183.

10 CFR 50.67, "Accident Source Term," provides a mechanism for-currently licensed nuclear power reactors to replace the traditional source term used in design basis accident analyses with an alternative source term. Under this provision, licensees who seek to revise the accident source term in design basis radiological consequence analyses must apply for a license amendment under 10 CFR 50.90.

Full implementation AST analyses were performed by SCE&G in accordance with the guidance in Regulatory Guide 1.183, and Section 15.0.1 of the Standard Review Plan (SRP). SCE&G performed AST analyses for the six PWR design basis accidents identified in Regulatory Guide 1.183 that could

, potentially result in significant Control Room and offsite doses. These include the loss of coolant

,:.accident, the main steam line break accident, the refueling accident, the steam generator tube rupture, reactor coolant pump locked rotor and the control rod ejection accident. The analyses demonstrate that using AST methodologies, post-accident Control Room and offsite doses remain within regulatory acceptance limits.

SCE&G proposes implementation of this proposed change through a change to the VCSNS licensing basis, including the TS and associated Bases. Upon approval, conforming changes will, be made to the VCSNS Final Safety Analysis Report (FSAR) and subsequently submitted to the NRC staff in accordance with 10 CFR 50.71 (e) as part of the regular FSAR update process.

SCE&G I Virgil C.Summer Nuclear Station

• P.0. Box 88 . Jenkinsville, South Carolina 29065 1T (803) 345.5209 *www.scana.com

Document Control Desk LAR 04-02911 RC-09-0004 Page 2 of 4 Proposed changes in the licensing basis for VCSNS resulting from application of the AST include the following:

• New Control Room atmospheric dispersion factors (XIQs) based on site specific meteorological data collected between 2002 and 2006.

• Revise the CRHE unfiltered inleakage from 10 scfm to 210 scfm.

• New AST analyses performed in accordance with the guidance in Regulatory Guide 1.183 for the six design basis accidents: loss of coolant accident, the main steam line break accident, the refueling accident, steam generator tube rupture, reactor coolant pump locked rotor and the rod ejection accident.

" Revise the TS to address Improved Standard Technical Specifications Change Traveler (TSTF-51, Revision 2) which permits removal of the Technical Specification requirements for ESF features to be OPERABLE after sufficient radioactive decay has occurred to ensure off-site doses remain below the SRP limits.

• Other TS revisions to reflect the update of the accident source term and associated design basis accidents utilizing the guidance provided in Regulatory Guide 1.183 and the associated Control Room and offsite dose requirements of 10 CFR 50.67.

Table 5-1 of Attachment 5 provides a description of each proposed TS and TS Bases change.

The use of an AST results in changes in the design basis accident radiological consequences; however, the AST methodology has no direct impact on the probability or initiation of the evaluated design basis accidents. Application of AST methodology and the other changes requested by this application for a license amendment do not impact the core damage frequency or the large early release frequency. Therefore, this request for a revision to the VCSNS licensing basis is not being submitted as a "risk-informed approach" using the guidelines in Regulatory Guide 1.174.

Several domestic pressurized water reactors (Byron, Calvert Cliffs, Catawba, Millstone, and Seabrook) have previously provided justification for the use of AST methodology utilizing a similar approach.

These applications of AST methodology have been approved by the NRC.

Attachment 1 to this letter contains the overall description and summary of the proposed changes.

Attachment 2 provides the detailed AST Safety Assessment Report supporting the proposed AST license basis change. Attachments 3 and 4 are compliance tables addressing SCE&G's method of conforming to the regulatory guidance of Regulatory Guides 1.183 and 1.194 respectively. Attachment 5 contains the safety assessment for the proposed TS and Bases changes and their justification.

Attachment 6 provides a mark-up of the current TS. The associated marked up TS Bases changes are provided in Attachment 7 for information only and will be implemented in accordance with the VCSNS Technical Specification Bases Control Program. Attachment 8 contains a list of procedure changes to be completed before AST implementation to support analysis assumptions. Attachment 9 includes the No Significant Hazards Consideration Determination and Environmental Consideration for the proposed changes. Attachment 10 contains non-proprietary calculations that support the Safety Assessment and the updated meteorological data used to calculate the new Control Room X/Qs. Attachment 11 provides the proposed TS changes in final typed format. Attachment 12 delineates that there are no commitments contained in this submittal.

Document Control Desk LAR 04-02911 RC-09-0004 Page 3 of 4 SCE&G has concluded that the proposed changes do not involve a significant hazards consideration.

SCE&G has also determined that the proposed changes satisfy the criteria for a categorical exclusion in accordance with 10 CFR 51.22(c)(9) and does not require an environmental review. Therefore, pursuant to 10 CFR 51.22(b), no environmental impact statement or environmental assessment needs to be prepared for these changes.

Implementation of the AST is desired by mid-2010. To support this schedule, SCE&G requests approval of the proposed License Amendment by April, 2010, with the amendment conditioned to be effective within 90 days.

In accordance with 10 CFR 50.91, a copy of this application, with attachments, is being provided to the designated South Carolina Official.

The proposed change has been reviewed and approved by both the Plant Safety Review Committee and the Nuclear Safety Review Committee.

If you should have any questions regarding this submittal, please contact Mr. Bruce L. Thompson at (803) 931-5042.

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

/&,76?7 Executed on Jeffrey B. r(chie JHW/JBAwm Attachments: (12)

Attachment 1 - Description for the Alternate Source Term License Amendment Attachment 2 - AST Safety Assessment Report Attachment 3 - Regulatory Guide 1.183 Compliance Table Attachment 4 - Regulatory Guide 1.194 Compliance Table Attachment 5 - Safety Assessment for the Proposed Technical Specification and Bases Changes Attachment 6 - Proposed Technical Specification Changes (Mark-ups)

Attachment 7 - For Information - Proposed Technical Specification Bases Changes (Mark-ups)

Attachment 8 - Procedure Changes to be Completed Before AST Implementation Attachment 9 - No Significant Hazards Consideration Determination and Environmental Consideration for the Proposed Changes Attachment 10 - CD of Non-Proprietary Versions of Supporting Calculations (Six Design Basis Accident Calculations and One Control Room Meteorological Calculation) and Meteorological Data Used to Determine New Control Room X/Qs Attachment 11 - Proposed Technical Specification Changes (Re-Typed)

Attachment 12 -Regulatory Commitments

Document Control Desk LAR 04-02911 RC-09-0004 Page 4 of 4 c: K. B. Marsh S. A. Byrne N. S. Carns J. H. Hamilton R. J. White K. J. Browne L. A. Reyes R. E. Martin (with Attachments)

NRC Resident Inspector P. Ledbetter K. M. Sutton T. P. O'Kelley CR (LAR 04-02911)

File (813.20) (with Attachments)

PRSF (RC-09-0004) (with Attachments)

Attachment 1 Description for the Alternative Source Term License Amendment

Attachment 1 Page 1 of 5 DESCRIPTION AND

SUMMARY

OF PROPOSED CHANGES

1.0 INTRODUCTION

South Carolina Gas and Electric Company (SCE&G) hereby proposes to amend the licensing basis of the Virgil C. Summer Nuclear Station (VCSNS) through the full implementation application of an Alternative Source Term (AST) methodology. The exceptions are that the current TID-14844 accident source term will remain the licensing basis for equipment qualification, NUREG-0737 evaluations other than Control Room Habitability Envelope (CRHE) doses, and Final Safety Analysis Report (FSAR) accidents not included in Regulatory Guide 1.183. Applicable, proposed Technical Specification (TS) and TS Bases changes, which are justified by the AST analyses, are included in this application for a license amendment.

This full implementation of AST analyses will modify the VCSNS licensing bases by adopting the AST methodology which replaces the current accident source term with an alternative source term as prescribed in 10 CFR 50.67 and establishes the 10 CFR 50.67 total effective dose equivalent (TEDE) dose limits as a new acceptance criterion. The AST is characterized by the composition and magnitude of the radioactive material, the chemical and physical form of the radionuclides, and the timing of the releases of these radionuclides. The current TID- 14844 accident source term will remain the licensing basis for equipment qualification, NUREG-0737 evaluations other than CRHE doses and radiological consequences for FSAR accidents not included in Regulatory Guide 1.183.

The use of an AST results in changes in the design basis accident (DBA) radiological consequences; however, the AST methodology has no direct impact on the probability or initiation of the evaluated design basis accidents. Application of AST methodology and the other changes requested by this application for a license amendment do not increase the core damage frequency or the large early release frequency. Therefore, this request for a revision to the VCSNS licensing basis is not being submitted as a "risk-informed approach" using the guidelines in Regulatory Guide 1.174.

Several domestic pressurized water reactors (Byron, Calvert Cliffs, Catawba, Millstone, and Seabrook) have previously provided justification for the use of AST methodology utilizing a similar approach. These applications of AST methodology have been approved by NRC.

Regulatory Guide 1.183 recommends that changes to the FSAR that reflect the revised analyses or the actual calculation documentation be submitted to the NRC staff.

Upon issuance of a license amendment, conforming FSAR changes will be completed as required by VCSNS procedures and submitted to the NRC staff in accordance with the regular FSAR update process as required by 10 CFR 50.71(e). In lieu of providing the NRC staff with proposed FSAR changes at this time, the supporting DBA calculations are being provided in Attachment 10.

Attachment 1 Page 2 of 5 The license amendment would revise the following VCSNS licensing bases:

• New Control Room atmospheric dispersion factors (X/Qs) based on site specific meteorological data collected between 2002 and 2006, and Regulatory Guides 1.194 revised methodology.

• Revise the CRHE unfiltered inleakage from 10 scfm to 210 scfm.

• New AST analyses performed in accordance with the guidance in Regulatory Guide 1.183 for the six design basis accidents: loss of coolant accident, the main steam line break accident, the refueling accident, steam generator tube rupture, reactor coolant pump locked rotor and the control rod ejection accident.

0 Revise the TS to address Improved Standard Technical Specifications Change Traveler (TSTF-5 1, Revision 2) which permits removal of the Technical Specification requirements for ESF features to be OPERABLE after sufficient radioactive decay has occurred to ensure off-site doses remain below the Standard Review Plan (SRP) limits.

• Other TS revisions reflect the update of the accident source term and associated design basis accidents utilizing the guidance provided in USNRC Regulatory Guide 1.183 and the associated control room and offsite dose requirements of 10 CFR 50.67.

Implementation of the AST is desired by mid-2010. To support this schedule, SCE&G requests approval of the proposed License Amendment by April, 2010, with the amendment conditioned to be effective within 90 days.

2.0 REGULATORY BACKGROUND The current VCSNS licensing basis for design basis accident (DBA) analysis source terms is U.S. Atomic Energy Commission Technical Information Document TID-14844, "Calculation of Distance Factors for Power and Test Reactor Sites," dated March 23, 1962. This is consistent with 10 CFR Part 100, Section 11 (10 CFR 100.11),

"Determination of Exclusion Area, Low Population Zone, and Population Center Distance," for reactor siting, which contains offsite dose limits in terms of whole body and thyroid dose and further makes reference to TID-14844.

In December 1999, the Nuclear Regulatory Commission (NRC) issued 10 CFR 50.67, "Accident Source Term," which provides a mechanism for licensed power reactors to replace the traditional accident source term used in their DBA analyses with an AST.

Regulatory guidance for the implementation of these ASTs is provided in Regulatory Guide (RG) 1.183, "Alternative Radiological Source Terms for Evaluating Design Basis Accidents at Nuclear Power Reactors." 10 CFR 50.67 requires a licensee seeking to use an AST to apply for a license amendment and requires that the application contain an evaluation of the consequences of affected DBAs.

Attachment 1 Page 3 of 5 As part of the implementation of the AST, the total effective dose equivalent (TEDE) acceptance criterion of 10 CFR 50.67(b)(2) replaces the previous whole body and thyroid dose guidelines of 10 CFR 100.11 and 10 CFR Part 50, Appendix A, General Design Criterion (GDC) 19, "Control Room," for a loss-of-coolant accident (LOCA), main steam line break (MSLB) accident, fuel handling accident (FHA), steam generator tube rupture (SGTR), reactor coolant pump (RCP) locked rotor accident (LRA) and the control rod ejection accident (CREA).

The accident source term is intended to be representative of a major accident involving significant core damage and is typically postulated to occur in conjunction with a large break LOCA. As a result of significant core damage, fission products are available for release into the containment environment. The proposed AST is an accident source term that is different from the accident source term used in the original design and licensing of VCSNS. 10 CFR 50.67, as implemented in accordance with RG 1.183, identifies an AST that is acceptable to the NRC staff for use at operating reactors.

The following regulatory requirements and guidance are also considered within this proposed license amendment:

" GDC 19, "Control Room," of Appendix A to 10 CFR Part 50, insofar as it requires that adequate radiation protection be provided to permit access and occupancy of the Control Room under accident conditions without personnel receiving radiation exposures in excess of allowablevalues.

" NUREG-0800, SRP 15.0.1, "Radiological Consequence Analyses Using Alternative Source Terms, Revision 0," provides guidance for the safety review of the radiological consequences of DBAs associated with implementing an AST. SRP 15.0.1 supports the guidance outlined in RG 1.183.

" NRC Generic letter 2003-01, "Control Room Habitability," requests addressees to submit information that demonstrates that the Control Room at each of their respective facilities complies with the current licensing and design bases and applicable regulatory requirements, and that suitable design, maintenance and testing control measures are in place for maintaining this compliance.

" USNRC RG 1.194, "Atmospheric Relative Concentrations for Control Room Radiological Habitability Assessments at Nuclear Power Plants," provides guidance on determining atmospheric relative concentration (X/Q) values in support of design basis Control Room radiological habitability assessments at nuclear power plants. This document describes methods acceptable to the NRC staff for determining X/Q values that will be used in Control Room radiological habitability assessments performed in support of applications for licenses and license amendment requests. Many of the regulatory positions presented in this guide represent substantial changes from procedures previously used to determine atmospheric relative concentrations for assessing the potential Control Room radiological consequences for a range of

Attachment 1 Page 4 of 5 postulated accidental releases of radioactive material to the atmosphere.

These revised procedures are largely based on the NRC sponsored computer code, ARCON96.

3.0 SAFETY ASSESSMENT SCE&G has performed analyses to support a full implementation of the AST as defined in RG 1.183 with the exception that the current TID-14844 accident source term will remain the licensing basis for equipment qualification, NUREG-0737 evaluations other than CRHE doses and radiological consequences for FSAR accidents not included in Regulatory Guide 1.183. A detailed description of the AST analyses, including a safety assessment, is provided in Attachment 2. Copies of AST accident dose calculations and the calculation that determines the atmospheric dispersion coefficients for the Control Room required to support the licensing bases changes for AST are included in Attachment 10.

The detailed Safety Assessment Report associated with AST analyses is provided in Attachment 2. This report is supplemented with RG 1.183 and 1.194 Compliance Tables presented in Attachments 3 and 4.

The basis/safety assessment associated with the proposed changes to the VCSNS Technical Specifications and Bases is included in Table 5-1 of Attachment 5 and supported by the Safety Assessment in Attachment 2 and the calculations in Attachment 10.

A list of procedure changes to be completed before AST implementation is included in Attachment 8.

A No Significant Hazards Consideration Determination and Environmental Consideration for the proposed changes are included in Attachment 9.

4.0 CONCLUSION

In conclusion, based on the considerations discussed above and detailed in the remainder of this submittal, (1) there is reasonable assurance that the health and safety of the public will not be endangered by operation in the proposed manner; (2) such activities will be conducted in compliance with the Commission's regulations; and (3) the issuance of the requested license amendment will not be inimical to the common defense and security or to the health and safety of the public.

A comparison of the calculated AST doses and the RG 1.183 dose limits for the Control Room operator, the exclusion area boundary and the low population zone for the DBA LOCA, MSLB, FHA, SGTR, LRA, and CREA, are provided in the table below. Note that the calculated AST doses provide a conservative estimate of the dose consequences for the postulated event. The conservative models and methodologies that were used to calculate the dose consequences are presented in Attachment 2.

Attachment 1 Page 5 of 5 Summary of Design Basis Accident Radiological Consequences for AST Calculated Dose Versus RG 1.183 Dose Criteria AST Calculated Dose RG 1.183 Dose Criteria (Rem TEDE) (Rem TEDE)

Loss of Coolant CR Operator Dose 1.01 5 Accident EAB Dose 4.87 25 (LOCA) LPZ Dose 0.54 25 Case 1: Concurrent Iodine Spike CR Operator Dose 0.37 5.0 EAB Dose 0.78 2.5 Main Steam LPZ Dose 0.16 2.5 Line Break (MSLB)

Case 2: Pre-existing Iodine Spike CR Operator Dose 1.15 5.0 EAB Dose 1.96 25 LPZ Dose 0.13 25 Case 1: FHA Inside Containment CR Operator Dose 0.76 5.0 EAB Dose 4.29 6.3 Fuel Handling LPZ Dose 1.06 6.3 (FHA) Case 2: FHA Inside Fuel Handling Building CR Operator Dose 0.41 5.0 EAB Dose 4.29 6.3 LPZ Dose 1.06 6.3 Case 1: Concurrent Iodine Spike CR Operator Dose 0.37 5.0 EAB Dose 0.71 2.5 Te Rator LPZ Dose 0.047 2.5 Tube Rupture Case 2: Pre-existing Iodine Spike (SGTR) CR Operator Dose 1.18 '5.0 EAB Dose 2.08 25 LPZ Dose 0.12 25 Locked Rotor CR Operator Dose 2.43 5.0 Accident EAB Dose 2.20 2.5 (LRA) LPZ Dose 0.45 2.5 Case 1: Containment Release Path CR Operator Dose 1.71 5.0 EAB Dose 4.31 6.3 Control Rod LPZ Dose 0.95 6.3 Ejection Accident (CREA) Case 2: Steam Generator Release Path CR Operator Dose 2.38 5.0 EAB Dose 2.52 6.3 LPZ Dose 0.48 6.3

Attachment 2 AST Safety Assessment Report

Attachment 2 Page 1 of 54 TABLE OF CONTENTS Page TABLE OF CONTENTS ................................................................................................. 1 ACRONYM S AND ABBREVIATIONS ..................................................................... 4

1.0 DESCRIPTION

........................................................................................................... 6

2.0 PROPOSED CHANGE

S ................................................................................................ 6

3.0 BACKGROUND

........................................................................................................ 7

4.0 TECHNICAL ANALYSIS

............................................... ...................................... 7 4.1 Atmospheric Dispersion Factors - Offsite and CRHE ........................................ 7 4.1.1 Atmospheric Dispersion Factors - Offsite ............................................. 7 4.1.2 Atmospheric Dispersion Factors - CRHE ............................................... 8 4.1.2.1 Infiltration ................................................................................. 8 4.1.2.2 Calculations ............................................................................. 9 4.2 Accident Source Terms .................................................................................... 11 4.3 Loss of Coolant Accident ................................................................................. 12 4.3.1 Introduction and Background ............................................................... 13 4.3.2 Source Term. ......................................... 14 4.3.3 M itigation ............................................................................................... 14 4.3.4 Radiological Transport M odeling ........................................................ 14 4.3.5 Results - Control Room Operator Dose ............................................... 14 4.3.6 Results - Offsite Doses ........................................................................ 15 4.3.7 Conclusion ............................................................................................. 15 4.3.8 Summary of Calculation Conservatisms ............................................... 15 4.4 Fuel Handling Accident .................................................................................. 19 4.4.1 Introduction and Background .............................. 20 4.4.2 Source Term .......................................................................................... 21 4.4.3 M itigation ...................................... ;..................................................... 22 4.4.4 Radiological Transport M odeling ........................................................ 23 4.4.5 Results - Control Room Operator Dose .............................................. 23 4.4.6 Results - Offsite Doses ....................................................................... 23 4.4.7 Conclusions .............................................. ..................................... 23 4.4.8 Summary of Calculation Conservatisms ............................................... 24 4.5 M ain Steam Line Break ................................................................................... 26 4.5.1 Introduction and Background ............................................................... 27 4.5.2 Source Term .......................................................................................... 27 4.5.3 M itigation ............................................................................................ 28 4.5.4 Radiological Transport M odeling ........................................................ 29 4.5.5 Results - Control Room Operator Dose ............................................... 29 4.5.6 Results - Offsite Doses ....................................................................... 29 4.5.7 Conclusions ......................................................................................... 30 4.5.8 Summary of Calculation Conservatisms ............................................... 30

Attachment 2 Page 2 of 54 TABLE OF CONTENTS Page 4.6 Steam Generator Tube Rupture Accident ....................................................... 32 4.6.1 Introduction and Background ............................................................... 33 4.6.2 Source Term .......................................................................................... 34 4.6.3 M itigation .......................................................................................... . 34 4.6.4 Radiological Transport Modeling ........................................................ 35 4.6.5 Results - Control Room Operator Dose .............................................. 36 4.6.6 Results - Offsite Doses ....................................................................... 36 4.6.7 C onclusions ......................................................................................... 36 4.6.8 Summary of Calculation Conservatisms .............................................. 36 4.7 Reactor Coolant Pump Locked Rotor Accident .............................................. 39 4.7.1 Introduction and Background ............................................................... 40 4.7.2 Source T erm ............................................................................................... 40 4.7.3 M itigation ............................................................................................ 41 4.7.4 Radiological Transport Modeling ........................................................ 42 4.7.5 Results - Control Room Operator Dose .............................................. 42 4.7.6 Results - Offsite Doses ........................................................................... 42 4.7.7 C onclusions ......................................................................................... 43 4.7.8 Summary of Calculation Conservatisms ............................................... 43 4.8 Control Rod Ejection Accident ........................................................................ 45 4.8.1 Introduction and Background ............................................................... 46 4.8.2 Source Term .......................................................................................... 47 4.8.3 M itigation ............................................................................................ 48 4.8.4 Radiological Transport Modeling ........................................................ 49 4.8.5 Results - Control Room Operator Dose .............................................. 49 4.8.6 Results - Offsite Doses ....................................................................... 49 4 .8.7 C onclusions ............................................................................................. 49 4.8.8 Summary of Calculation Conservatisms ............................................... 50 4.9 Equipment Qualification and NUREG-0737 ................................................... 53 5.0 REGULATORY SAFETY ANALYSIS (10 CFR 50.92 Evaluation) .............. 53

6.0 ENVIRONMENTAL CONSIDERATION

S (10 CFR 50.21 Evaluation) ........ 53

7.0 REFERENCES

................................................................................................. 53

Attachment 2 Page 3 of 54 LIST OF TABLES Page Table 4.1-1 Offsite X/Q (sec/m 3) ................................................................................ 8 Table 4.1-2 Release Points to the Environment .............................................................. 9 Table 4.1-3 3 CRHE X/Q (sec/m ) ............................................................................... 10 Table 4.2-1 Core Activity Available for Release - Design Basis ............................ 11 Table 4.3-1 Design Input Comparison - Current Licensing Basis vs. AST D esign - LO C A .................................................................................... 12 Table 4.4-1 Design Input Comparison - Current Licensing Basis vs. AST D esign - FH A ........................................................................................ 19 Table 4.4-2 Design Basis Core Activity @ 72 Hours in Ci/MTU ............................ 21 Table 4.5-1 Design Input Comparison - Current Licensing Basis vs. AST D esign - M SLB ................................................................................... 26 Table 4.6-1 Design Input Comparison - Current Licensing Basis vs. AST D esign - SG TR ..................................................................................... 32 Table 4.7-1 Design Input Comparison - Current Licensing Basis vs. AST D esign - LR A ........................................................................................ 39 Table 4.7-2 Core Activity @ T = 0 Hours ............................................................... 40 Table 4.7-3 Reactor Coolant Fission and Corrosion Product Activity ..................... 41 Table 4.8-1 Design Input Comparison - Current Licensing Basis vs. AST D esign - CREA .................................................................................... 45 Table 4.8-2 Core Activity @ T = 0 Hours ............................................................... 47 Table 4.8-3 Reactor Coolant Fission and Corrosion Product Specific Activity at E quilibrium ........................................................................................ 47 LIST OF FIGURES Page Figure 1 Activity Flow Path Model Developed to Calculate LOCA Doses from Containm ent Leakage ........................................................................... 17 Figure 2 Activity Flow Path Model Developed to Calculate LOCA Doses from ECCS Leakage ........................................ 18 Figure 3 Activity Flow Path Model Developed to Calculate FHA Doses ........... 25 Figure 4 Activity Flow Path Model Developed to Calculate MSLB Doses ........ 31 Figure 5 Activity Flow Path Model Developed to Calculate SGTR Doses ...... 38 Figure 6 Activity Flow Path Model Developed to Calculate RCP LRA Doses ...... 44 Figure 7A RADTRAD Model Developed to. Calculate CREA Doses (Containm ent Release) .......................................................................... 51 Figure 7B RADTRAD Model Developed to Calculate CREA Doses (SG POR V Releases) ............................................................................ 52

Attachment 2 Page 4 of 54 Acronyms and Abbreviations gCi/gm micro-curies per gram 3 X/Q atmospheric dispersion factor in sec/mi acfm actual cubic feet per minute AOP Abnormal Operating Procedure ASHRAE American Society of Heating, Refrigeration and Air Conditioning Engineers AST Alternative Source Term cc cubic centimeter cfm cubic feet per minute CFR Code of Federal Regulations Ci curie CLB current licensing basis CR Control Room CREA control rod ejection accident CRHE Control Room Habitability Envelope CsI cesium iodine d day DBA Design Basis Accident DE Dose Equivalent DF decontamination factor EAB exclusion area boundary EAHS Emergency Air Handling System ECCS emergency core cooling system EDE effective dose equivalent EOP Emergency Operating Procedure ESF engineered safeguard features OF degrees Fahrenheit FHA fuel handling accident FHB Fuel Handling Building FSAR Final Safety Analysis Report ft3 feet ft cubic feet gpm gallons per minute GTP General Test Procedure GWD/MTU gigawatt days/metric tons uranium HEPA high efficiency particulate air hr hour IB Intermediate Building in inch kw kilowatt Ibm pounds-mass LAR License Amendment Request LCO Limiting Condition For Operation LOCA loss-of-coolant accident LPZ low population zone LRA locked rotor accident m meters

Attachment 2 Page 5 of 54 Acronyms and Abbreviations m3 cubic meters min minute MS main steam MSLB main steam line break MTU metric tons uranium MWD/MTU megawatt days/metric tons uranium MWt megawatt thermal NaOH Sodium Hydroxide NRC Nuclear Regulatory Commission PF partition factor PORV power operated relief valves psig pounds per square inch gauge PWR pressurized water reactor RB Reactor Building RCCA rod cluster control assembly RCS reactor coolant system RCP reactor coolant pump Rem roentgen equivalent man RG Regulatory Guide RTP rated thermal power RWST Refueling Water Storage Tank SCE&G South Carolina Gas & Electric scfm standard cubic feet per minute sec second SER Safety Evaluation Report SG steam generator SGTR steam generator tube rupture SI safety injection SRP Standard Review Plan SSV secondary safety valves TEDE total effective dose equivalent TS Technical Specification TSC Technical Support Center USNRC United States Nuclear Regulatory Commission UUI unanticipated unfiltered inleakage VCSNS Virgil C. Summer Nuclear Station

Attachment 2 Page 6 of 54

1.0 DESCRIPTION

In accordance with 10 CFR 50.67, "Accident Source Term," a licensee may voluntarily revise the accident source term used in design basis radiological consequence analyses. Paragraph 50.67(b) requires that applications under this section contain an evaluation of the consequences of applicable design basis accidents (DBAs) previously analyzed in the plant Final Safety Analysis Report (FSAR). Regulatory Guide (RG) 1.183, "Alternative Radiological Source Terms for Evaluating Design Basis Accidents at Nuclear Power Reactors" (Reference 1),

provides guidance to licensees on performing evaluations, and reanalyzes as required to adopt an alternative source term (AST).

VCSNS has performed radiological consequence analyses of the six applicable pressurized water reactor (PWR) DBAs identified in RG 1.183. These DBAs are a loss-of-coolant accident (LOCA), main steam line break (MSLB) accident, fuel handling accident (FHA), steam generator tube rupture (SGTR), reactor coolant pump (RCP) locked rotor accident (LRA) and the control rod ejection accident (CREA). These analyses were performed using the guidance of RG 1.183 and Standard Review Plan (SRP) Section 15.0.1, "Radiological Consequence Analyses Using Alternative Source Terms" (Reference 2). The analyses were prepared, reviewed, and approved in accordance with the VCSNS 10 CFR 50, Appendix B Quality Assurance Program.

Comparison with the guidance contained in RG 1.183 and RG 1.194, "Atmospheric Relative Concentrations for Control Room Radiological Habitability Assessments at Nuclear Power Plants," (Reference 3) is summarized in Attachments 3 and 4 respectively of this license amendment request (LAR).

The supporting analyses consisted of the following steps:

" Determination of the AST based on plant-specific analysis of the fission product inventory.

• Application of the release fractions for the six PVWR DBAs.

• Application of the deposition and removal mechanisms.

" Evaluation of activity transport pathways to the environment.

" Analysis of the atmospheric dispersion for the radiological propagation pathways.

• Calculation of the offsite and Control Room personnel Total Effective Dose Equivalent (TEDE).

• Evaluation of other related design and licensing bases pertaining to NUREG-0737 (Reference 4) requirements.

The radiological dose analyses have been performed assuming reactor operation at 2958 MWt (102% of the rated power level of 2900 MWt). This results in a conservative estimate of fission product releases.

2.0 PROPOSED CHANGE

S The licensing and design basis changes included in this LAR are described below. The proposed Technical Specification (TS) and Bases changes are described in Attachment 5 and a mark-up of the affected TS and Bases pages is provided in Attachments 6 and 7 respectively.

Attachment 2 Page 7 of 54

3.0 BACKGROUND

On December 23, 1999, the NRC published 10 CFR 50.67, "Accident Source Term," in the Federal Register. This regulation provides a mechanism for licensed power reactors to replace the current accident source term used in design basis accident (DBA) analyses with an alternative source term. The direction provided in 10 CFR 50.67 is that licensees who seek to revise their current accident source term in design basis radiological consequence analyses must apply for a license amendment under 10 CFR 50.90.

Regulatory Guide (RG) 1.183 and Standard Review Plan Section 15.0.1 were used by SCE&G in preparing the AST analyses. These documents were prepared by the NRC staff to address the use of ASTs at current operating power reactors. The RG establishes the parameters of an acceptable AST and identifies the significant attributes of an AST acceptable to the NRC staff.

In this regard, the RG provides guidance to licensees for operating power reactors on acceptable applications for an AST; the scope, nature, and documentation of associated analyses and evaluations; consideration of impacts on risk; and acceptable radiological analysis assumptions.

The SRP provides guidance to the staff on the review of AST submittals.

Acceptance criteria consistent with that required by 10 CFR 50.67 were used to replace VCSNS current design basis source term acceptance criteria. The AST analyses were performed for the six PWR DBAs identified in RG 1.183 that could potentially result in Control Room and offsite doses. These include the loss-of-coolant accident (LOCA), main steam line break (MSLB) accident, fuel handling accident (FHA), steam generator tube rupture (SGTR), reactor coolant pump (RCP) locked rotor accident (LRA) and the control rod ejection accident (CREA).

4.0 TECHNICAL ANALYSIS

4.1 Atmospheric Dispersion Factors - Offsite and CRHE The x/Q values used in the AST analyses at the EAB and LPZ are the current licensing values described in the VCSNS Updated FSAR section 2.3.4. Use of the x/Q values previously approved by the staff during the initial facility licensing is acceptable for use in the AST analyses as discussed in RG 1.183, Section 5.3.

The x/Q values at the CR intake are calculated using the NRC-sponsored computer codes ARCON96 consistent with the procedures in RG 1.194.

4.1.1 Atmospheric Dispersion Factors - Offsite Table 4.1-1 provides the offsite atmospheric diffusion coefficients (X/Qs) used for the AST.

Attachment 2 Page 8 of 54 Table 4.1-10Offsite :/Os (sec/mn3 )

Time Period EAB LPZ 0 - 2 hours2.314815e-5 days <br />5.555556e-4 hours <br />3.306878e-6 weeks <br />7.61e-7 months <br /> 4.08E-04 0 - 8 hours9.259259e-5 days <br />0.00222 hours <br />1.322751e-5 weeks <br />3.044e-6 months <br /> 2.37E-05 8 - 24 hours2.777778e-4 days <br />0.00667 hours <br />3.968254e-5 weeks <br />9.132e-6 months <br /> 2.44E-06 1 - 4 days 1.11E-06 4 - 30 days 6.28E-07 4.1.2 Atmospheric Dispersion Factors - CRHE A detailed discussion of the design input parameters, assumptions, methodology, analysis, and results supporting the new CRHE X/Qs is provided in Attachment 10 as Calculation DC00040-079, "Atmospheric Dispersion Coefficients for Control Room."

The atmospheric dispersion factors (X/Q) were calculated using' plant specific meteorological data and the ARCON96 (Reference 5) computer code per RG 1.194. The meteorological data files used for the CRHE includes a record of each hourly data containing a location identifier, Julian day (1-366), hour (0 to 23), low-level direction, low-level speed, stability class (I=A to 7=G), upper level direction, and upper level speed. Wind speeds are entered in tenths of a reporting unit with no decimal. Wind directions are from 1 to 360 in degrees. The five yearly files were combined into one file for ease of code execution. This file contains all the data for the VCSNS site from 2002-2006, which satisfies the ARCON96 requirements for having 3 to 5 years of hourly data. The meteorological tower used for collecting the data is located on the plant site. Instrumentation is provided at the 10 and 60 meter level. Attachment 10 of this LAR contains a CD with the updated meteorological data used to calculate the new CRHE atmospheric diffusion coefficients.

4.1.2.1 Infiltration Section 3.3.3, "Infiltration Pathways", of RG 1.194 states that a X/Q should be determined for any infiltration pathway that could result in a significant intake of radioactive contaminated air into the Control Room Habitability Envelope (CRHE). The infiltration pathways actually applicable to a particular facility will be identified via inleakage testing or CRHE inspections and surveillances.

The current licensing basis for the infiltration rate is 10 scfm, attributed to leakage through improperly sealed doors or by personnel ingress and egress, per NUREG-0800, USNRC Standard Review Plan, Section 6.4, "Control Room Habitability Systems," Reference 6.

Attachment 2 Page 9 of 54 A base line ASTM E741 integrated test was performed in March 2005 to measure leakage into the CRHE. As reported in Reference 10 and acknowledged by Reference 7, filtered outside air was found to be within current limits of 1000 scfm per train and the maximum unanticipated unfiltered inleakage (UUI) recorded was 41 scfm. Although the maximum UUI recorded was 41 scfm, a UUI of 200 scfm is utilized in the AST analyses to provide margin for future testing.

This base infiltration rate is augmented by adding to it the estimated contribution from opening and closing of doors associated with such activities in accordance with by the plant emergency plans and procedures. Normally, 10 scfm is used for this additional contribution. A value of 10 scfm is utilized in this evaluation.

4.1.2.2 Calculations The ARCON96 computer code (Reference 5) was used to calculate the Control Room X/Qs. The receptors considered in the calculation are the two Control Room intakes. The 'A' train intake is at elevation 484 ft and the 'B' train is at elevation 507 ft-9 in above grade. Ground level for the VC Summer site is at elevation 436 ft. Therefore, the intake height for Intake 'A' is 48 ft or 14.6 m and Intake 'B' 71.75 ft or 21.9 m. All elevations are measured from the same reference point and the elevation difference is zero.

The release points to the environment evaluated for possible modeling with the ARCON96 computer code are listed in Table 4.1-2.

Table 4.1-2: Release Points to the Environment Direction from Straight-line Release Height Intake to Release Horizontal Release Point Reight - Intak to Distance (m)

(height - 436 ft)m Intake Intake Intake Intake A B A B Main Plant Vent 524 ft-3 in (26.9m) 380 420 89.0 86.9 Purge Exhaust 524 ft-0 in (26.8m) 380 420 87.5 86.3 MS PORV A 504 ft-0 in (20.7m) 640 590 64.0 63.1 MS SSV MS SSV A495 A ft-0 in 45f-in570 640 60.0 60.0 (B,C,D,E) (18.0m) 503 ft-0 in MS SSV A (20 im)570 (A) (20.4m) 640 60.0 60.0 MS PORV B 470 ft-0 in (10.4m) 720 780 80.8 82.0 MS SSV B 495 ft-0 in (18.0m) 710 760 75.6 78.0 MS PORV C 470 ft-0 in (10.4m) 690 730 105.8 105.8 MS SSV C 495 ft-0 in (18.0m) 730 760 104.5 103.6 IB Blowout Panel Releasing 463 ft-0 in (8.2m) 730 780 67.4 69.5 Directly to Environment RB Nearest 0 ft 630 570 61.0 61.0 Point I

Attachment 2 Page 10 of 54 The resulting Control Room X/Qs are given in Table 4.1-3.

Table 4.1-3: CRHE X/Qs (sec/im)

Time Intake A Intake B Time Intake A Intake B RB Nearest Point MS PORV B 0-2 h 1.39E-03 1.30E-03 0-2 h 8.69E-04 8.31E-04 2-8 h 1.17E-03 1.09E-03 2-8 h 7.16E-04 6.89E-04 8-24 h 5.70E-04 5.27E-04 8-24 h 3.44E-04 3.23E-04 1-4 d 4.17E-04 3.90E-04 1-4 d 2.49E-04 2.33E-04 4-30 d 3.OOE-04 2.82E-04 4-30 d 1.84E-04 1.69E-04 Main Plant Vent MS SSV B 0-2 h 7.11E-04 7.43E-04 0-2 h 9.61E-04 9.29E-04 2-8 h 5.05E-04 5.41E-04 2-8 h 7.64E-04 7.08E-04 8-24 h 2.5 1E-04 2.75E-04 8-24 h 3.67E-04 3.39E-04 1-4 d 2.04E-04 2.16E-04 1-4 d 2.65E-04 2.44E-04 4-30 d 1.39E-04 1.49E-04 4-30 d 1.98E-04 1.79E-04 Purge Exhaust MS PORV C 0-2 h 7.23E-04 7.57E-04 0-2 h 5.18E-04 5.10E-04 2-8 h 5.24E-04 5.47E-04 2-8 h 4.37E-04 4.34E-04 8-24 h 2.59E-04 2.78E-04 8-24 h 2.09E-04 2.05E-04 1-4 d 2.13E-05 2.19E-04 1-4 d 1.51E-04 1.49E-05 4-30 d 1.44E-04 1.51E-04 4-30 d 1.11E-04 1.08E-04 IB Blowout Panel MS SSV C 0-2 h 1.22E-03 1.12E-03 0-2 h 5.36E-04 5.44E-04 2-8 h 1.01E-03 9.16E-04 2-8 h 4.10E-04 4.18E-04 8-24 h 4.79E-04 4.32E-04 8-24 h 1.99E-04 2.OOE-04 1-4 d 3.48E-04 3.12E-04 1-4 d 1.44E-04 1.44E-04 4-30 d 2.53E-04 2.24E-04 4-30 d 1.06E-04 1.06E-04 MS SSV A MS PORV A (Relief's B,C,D,E) 0-2 h 1.50E-03 1.51E-03 0-2 h 1.34E-03 1.37E-03 2-8 h 1.15E-03 1.17E-03 2-8 h 1.01E-03 1.03E-03 8-24 h 5.64E-04 5.75E-04 8-24 h 4.97E-04 5.07E-04 1-4 d 4.23E-04 4.18E-04 1-4 d 3.64E-04 3.77E-04 4-30 d 3.03E-04 3.1OE-04 4-30 d 2.69E-04 2.72E-04 MS SSV A (A-relief only) 0-2 h 1.50E-03 1.5 1E-03 2-8 h 1.12E-03 1.15E-03 8-24 h 5.5 1E-04 5.67E-04 1-4 d 4.15E-04 4.13E-04 4-30 d 2.97E-04 3.02E-04

Attachment 2 Page 11 of 54 Based on the calculated Control Room X/Qs, the conservative values used in each of the AST dose analyses are summarized as follows:

LOCA Reactor Building nearest point (Intake A)

MSLB MS SSV A (Reliefs B, C, D, E), Intake B FHA (Inside Containment) RB Nearest Point Intake A FHA (Outside Containment) Main Plant Vent Intake B SGTR MS SSV A (Reliefs B, C, D, E), Intake B RCP LRA MS SSV A (Reliefs B, C, D, E), Intake B CREA (Containment Release) RB Nearest Point Intake A CREA (Steam Generator Release) MS SSV A (Reliefs B, C, D, E), Intake B 4.2 Accident Source Terms The inventory of fission products in the reactor core available for release to the containment is based on a core thermal power of 2958 MWt (102% of RTP). A list of the 60 isotopes used in the AST analysis is given in Table 4.2-1.

The core inventories are taken from the Westinghouse Radiation Analysis Manual, Revision 1 for VCSNS Uprating updated 12/98 (Reference 8). Reference 8 does not provide core inventory values for all of these 60 isotopes. Therefore, the default coreinventories from Table 1.4.3.2-2 of NUREG/CR-6604 (Reference 13), corrected to a core thermal power of 2958MWt, were included in the analysis. In addition, the noble gas and iodine core inventories from the Westinghouse Radiation Analysis Manual and the corrected core inventories from NUREG/CR-6604 were compared and the larger of the two inventories was conservatively used in the subsequent analyses.

Table 4.2-1: Core ctivity Available for Release - Design Basis Isotope Curies Isotope Curies Isotope Curies Co-58 7.55E+05 Ru- 103 1.06E+08 Cs- 136 3.08E+06 Co-60 5.78E+05 Ru- 105 6.92E+07 Cs-137 5.66E+06 Kr-85 8.30E+05 Ru-106 2.42E+07 Ba-139 1.47E+08 Kr-85m 2.72E+07 Rh-105 4.79E+07 Ba- 140 1.46E+08 Kr-87 4.96E+07 Sb-127 6.53E+06 La-140 1.49E+08 Kr-88 6.71E+07 Sb-129 2.31E+07 La-141 1.37E+08 Rb-86 4.43E+04 Te- 127 6.31E+06 La- 142 1.32E+08 Sr-89 8.41E+07 Te-127m 8.35E+05 Ce-141 1.32E+08 Sr-90 4.54E+06 Te- 129 2.17E+07 Ce- 143 1.29E+08 Sr-91 1.08E+08 Te- 129m 5.72E+06 Ce- 144 7.98E+07 Sr-92 1.13E+08 Te-131m 1.1OE+07 Pr-143 1.26E+08 Y-90 4.87E+06 Te-132 1.09E+08 Nd- 147 5.65E+07 Y-91 1.02E+08 1-131 8.20E+07 Np-239 1.51E+09 Y-92 1.13E+08 1-132 1.20E+08 Pu-238 8.58E+04 Y-93 1.28E+08 1-133 1.68E+08 Pu-239 1.94E+04 Zr-95 1.29E+08 1-134 1.80E+08 Pu-240 2.44E+04 Zr-97 1.35E+08 1-135 1.54E+08 Pu-241 4.11E+06 Nb-95 1.22E+08 Xe- 133 1.70E+08 Am-241 2.72E+03 Mo-99 1.43E+08 Xe- 135 3.70E+07 Cm-242 1.04E+06 Tc-99m 1.23E+08 Cs- 134 1.01E+07 Cm-244 6.08E+04

Attachment 2 Page 12 of 54 4.3 Loss of Coolant Accident A detailed discussion of the design input parameters, assumptions, methodology, analysis, and results supporting the LOCA is provided in Attachment 10 as Calculation DC00040097, "Loss of Coolant Accident - AST". Attachment 3 provides a matrix which compares the RG 1.183 regulatory position with the parameters and methodologies utilized to calculate the LOCA CRHE and offsite doses.

Table 4.3-1 provides a comparison of key parameters utilized to determine the existing licensing basis LOCA and the AST LOCA doses.

Table 4.3-1: Design Input Comparison- Current Licensing Basis vs. AST Design - LOCA Parameter CLB Parameter AST Parameter Core Thermal Power Level 2958 MWt 2958 MWt Activity Inventory in Core 18 isotopes (I, Kr and Xe) 60 dose significant isotopes used in RADTRAD Radioisotope Decay Properties Table of Isotopes RADTRAD Table 1.4.3.2-3 Activity Release to Per R.G. 1.4 Per R.G. 1.183 Table 2 Containment (Gap & Early In-Vessel Phases Only)

Release Timing Instantaneous Per R.G. 1.183 Table 4 Radioiodine Chemical Species 91% Elemental 95% Aerosol (CsI) 5% Particulate 4.85% Elemental 4% Organic 0.15% Organic Primary Containment Volume Total free volume = 1,840,000 ft3 Total free volume = 1,840,000 ft3 Primary Containment Cleanup 50% plateout of released iodine Aerosol removal via Natural (Natural Deposition) Deposition only credited when sprays are not operating

( 1 0 th percentile Powers Model)

Containment Sprays ke = 20 hr-1 (maximum DF for the ke = 20 hr' (maximum DF for the elemental iodine spray removal elemental iodine spray removal coefficient is 100) coefficient is 200) kp = 5.68 hr-l(0 - 2176 sec) )p = 5.68 hr-&(0 - 2176 sec)

Xp = 0.568 hr-1 (>2176 sec) 4p = 0.568 hr' (>2176 sec)

No credit is taken for the removal No credit is taken for the removal of organic iodine, of organic iodine.

Containment Recirculation 90 percent for the removal of 90 percent for the removal of Particulate Filter iodine particulates only iodine particulates only Primary Containment Design 0.2%/day for first 24 hours2.777778e-4 days <br />0.00667 hours <br />3.968254e-5 weeks <br />9.132e-6 months <br />; 0.2%/day for first 24 hours2.777778e-4 days <br />0.00667 hours <br />3.968254e-5 weeks <br />9.132e-6 months <br />; Leak Rate 0.1%/day thereafter. 0.1%/day thereafter.

Minimum Sump Volume 55,730 ft3 55,730 ft3 Post LOCA ESF System Leakage Source Iodine only Iodine only Term to Environment ESF Leakage Outside of the RB 12,000 cc/hr 12,000 cc/hr ESF Leakage post-LOCA Time Begins at 1382 sec - ends at 30 Begins at 1460 sec - ends at 30 days days Sump Maximum Temperature < 2450 F < 245 'F ESF Flash Fraction 10% 10%

Iodine Re-evolution None assumed None assumed since pH >7

Attachment 2 Page 13 of 54 Table 4.3-1: Design Input Comparison- Current Licensing Basis vs. AST Design - LOCA Parameter CLB Parameter AST Parameter RB Sump Iodine Species 91% elemental, 5% particulate, & 97% elemental, 3% organic 4% organic Dose Conversion Factors ICRP 30 RADTRAD3 Table 1.4.3.3-2 Control Room Habitability 226,040 ft3 226,040 ft Envelope (CRHE) Volume CRHE Isolation Time 0 0 CRHE Emergency Filtered 1,000 cfm 1,265 cfm Intake Air Flow CRHE Unfiltered Air Inleakage 10 cfm 243 cfm CRHE Filtered Recirculation 19,125 cfm 19,125 cfm Flow Rate CRHE Emergency Filter Bed 2 in. charcoal 2 in. charcoal Depth CRHE Emergency Filter Bed 95% 95%

Removal Efficiency CRHE Operator Breathing 3.5E-04 m 3/sec (0 - 30 days) 3.5E-04 m3/sec (0 - 30 days)

Rates CRHE Operator Occupancy 1.0 0-24 hrs 1.0 0-24 hrs Factors 0.6 1-4 days 0.6 1-4 days 0.4 4-30 days 0.4 4-30 days 4.3.1 Introduction and Assumptions - LOCA The following primary assumptions from previous LOCA analyses continue to apply:

1. Two release pathways to the environment are considered: containment leakage (0.2%/day for first 24 hours2.777778e-4 days <br />0.00667 hours <br />3.968254e-5 weeks <br />9.132e-6 months <br />; 0.1%/day thereafter.) and ECCS recirculation leakage (12,000 cc/hr). Back leakage through the RWST and NaOH Tank is not considered in the LOCA analysis as a potential source of post LOCA leakage based on plant procedures (EOPs) that require closure of the 20" RWST outlet valve (6700) and closure of the 3" NaOH outlet valve (3012) following the transition to cold leg recirculation. This results in 3 valve isolation and a minimum of 2 valve isolation in the long term with a single failure.

2. A single electrical train is assumed to remove the following equipment from service: one containment spray train, one ECCS train and one of the two reactor building cooling units.

3. Containment spray removal coefficients continue to be based on the methodology in NUREG-0800, "Standard Review Plan For The Review Of Safety Analysis Reports For Nuclear Power Plants", Section 6.5.2, "Containment Spray as a Fission Product Cleanup System".

The following DBA LOCA dose contributors to the CRHE are included in this analysis:

1. Contamination of the Control Room atmosphere by the intake or the infiltration of the radioactive material contained in the radioactive plume.

2. Radiation shine from the external radioactive plume released from the facility.

3. Radioactive shine from radioactive materials in containment.

Attachment 2 Page 14 of 54 4.3.2 Source Term The core inventory is provided in Table 4.2-1. The release fractions and timing are taken from RG 1.183, Tables 2 and 4 as shown below.

4.3.3 Mitigation The radiological consequences of the LOCA are actively mitigated by containment spray, natural deposition, reactor building cooling unit particulate filter, radiological decay, limitations on leakage, and control room filtration (i.e., with an assumed charcoal filtration efficiency of 95%).

As described in Section 9.4.1 of the VCSNS updated FSAR, the control room ventilation system is automatically placed in the emergency mode, with filtration of incoming and recirculated air, following receipt of a SI or high radiation signal from the gaseous activity channel of RM-A1.

Since a large break LOCA results in a near instantaneous Reactor Trip and SI actuation [i.e.,

within the first few seconds on either low RCS pressure or high containment pressure (Hi-1)],

the control room is assumed to enter the emergency recirculation mode on event initiation since delay times of this order (i.e., seconds) are inconsequential to the results of the analysis.

4.3.4 Radiological Transport Modeling A simplified radiological release model and individual pathway models developed to calculate LOCA doses utilizing RADTRAD is shown in Figures 1 and 2. Specific model details and the supporting RADTRAD runs are provided in Attachment 10 as Calculation DC00040-097, "Loss of Coolant Accident - AST." Based on the calculated Control Room X/Qs provided in Table 4.1-3, the conservative values used in the LOCA AST dose analyses are taken at the following release point.

0 LOCA Reactor Building nearest point (intake A) 4.3.5 Results - Control Room Operator Dose The following DBA LOCA dose contributors to the CRHE are included in this analysis:

• Contamination of the Control Room atmosphere by the intake or the infiltration of the radioactive material contained in the radioactive plume.

• Radiation shine from the external radioactive plume released from the facility.

• Radioactive shine from radioactive materials in containment.

The contribution from each pathway and the total CR operator dose is shown as follows.

LOCA CR Operator Dose Rem TEDE 0 Containment leakage = 7.04E-01 0 Containment shine = 2.38E-03

• External cloud from Containment leakage = 2.02E-02

• ESF Recirculation leakage = 2.69E-01

• External cloud from ESF Recirculation leakage - 1.42E-02

• Total 1.01E+00

Attachment 2 Page 15 of 54 4.3.6 Results - Offsite Doses The RADTRAD computer code was used to determine the offsite dose. The calculated doses are shown as follows where the EAB dose represents the maximum 2-hour TEDE over the accident period.

LOCA EAB Dose Rem TEDE 0 Containment leakage = 4.52

• ESF Recirculation leakage = 0.35 9 Total = 4.87 LOCA LPZ Dose Rem TEDE

• Containment leakage = 0.43

" ESF Recirculation leakage = 0.11

" Total = 0.54 4.3.7 Conclusion The LOCA Control Room operator dose is below the 5 rem TEDE regulatory limit and the offsite doses are well below the 25 rem TEDE regulatory limit.

4.3.8 Summary of Calculation Conservatisms

" Primary containment cleanup by natural deposition assumes the 1 0 th percentile Power's Aerosol Decontamination Model.

• The analysis is based on the 60 isotopes of NUREG/CR-6604. For those isotopes not included in the VCSNS plant data, the default core inventories from Table 1.4.3.2-2 of NUREG/CR-6604, corrected to a core thermal power of 2958 MWt, are included in this analysis. In addition, the noble gas and iodine core inventories from VCSNS plant data and the corrected core inventories from NUREG/CR-6604 were compared and the larger of the two concentrations is used in this analysis.

" The Control Room unanticipated unfiltered inleakage (UUI) value of 41 scfm, based on tracer tests, was increased to 200 scfm. An uncertainty value of 25 percent is conservatively applied to the Control Room filtered makeup flow rate and the unfiltered makeup flow rate which bypasses the damper.

Attachment 2 Page 16 of 54 The VCSNS technical specifications do not provide a specific limit for operational leakage that is allowed within the recirculation loop. Administrative limits, however, ensure that operational leakage is adequately controlled. A post LOCA recirculation leakage of 12,000 cc/hr (7.063E-03 cfm) is used as input to the dose calculations.

This is twice the operational limit that is used in plant procedures for system leakage assessments (GTP-006, Reference 11). In the event total recirculation loop leakage exceeds 6,000 cc/hr, a Condition Evaluation Report is generated to facilitate a licensing basis impact assessment and an operability determination. No credit is taken for holdup or filtration of this leakage in the Auxiliary Building, i.e., the iodine released by the recirculation loop leakage is assumed to be immediately available for release to the atmosphere.

The containment recirculation HEPA efficiency is assumed to be 90 percent for the removal of iodine particulates only. This is conservative since Regulatory Guide 1.52 permits a 99 percent removal efficiency, based on the filter's characteristics.

• It is assumed that the sprays will operate for the first 4 hours4.62963e-5 days <br />0.00111 hours <br />6.613757e-6 weeks <br />1.522e-6 months <br /> of the accident.

Attachment 2 Page 17 of 54 Figure 1: Activity Flow Path Model Developed to Calculate LOCA Doses from Containment Leakage Emergency Mode Recirculation Filter 95% aerosol, elemental &

vIF organic (0 sec. - 720 hr.)

19,125 cfm @ 0 to 720 hr 5

FP 2 - 1,265 cfm filtered makeup (0 - 720 hr.)

95% aerosol, elemental & organic Volume 3 Volume 2 Control Room Containment Leakage Habitability 0.2%/day (0 - 24 hr.) Environment Envelope 0.1%/day (24 -720 hr.) FP 3 - 243 cfm unfiltered inleakage (0 - 720 hr.)

unfiltered inleakage/damper leakage/ (226,040 ft3) ingress/egress FP 4 - CR Exhaust: makeup plus unfiltered inleakage

?eiemental = 20 hr' (DF 200 stop) 1,508 cfm X.picujae = 5.68 hr' (DF 50 reduce by 10%)

4nauraa de. = 0.1 hr-' (stop when sprays operating)

Attachment 2 Page 18 of 54 Figure 2: Activity Flow Path Model Developed to Calculate LOCA Doses from ESF Leakage Emergency Mode Recirculation Filter 95% aerosol, elemental &

organic 19,125 cfm @ 0 to 720 hr.

FP 5 Volume 1 FP 2 - 1,265 cfm filtered makeup (0 - 720 hr.)

Volume 3 Sump FP 1 Volume 2 Control Room ESF Leal* Habitability 3

Environment 5.573E+04 ft 7.063E-03 cfm @ 0.41 - 720 hr. Envelope Filter 90% eff. to account for FP 3 - 243 cfm unfiltered inleakage (0 - 720 hr.) (226,040 ft3) 10% flash fraction unfiltered inleakage/damper leakage/

97% elemental ingress/egress 3% organic FP 4 - CR exhaust:

1,508 cfm

Attachment 2 Page 19 of 54 4.4 Fuel Handling Accident A fuel handling accident (FHA) during refueling could release a fraction of the fission product inventory in the plant to the environment. Two accident scenarios are considered: (1) a refueling accident occurring inside containment and (2) a refueling accident occurring outside containment. A detailed discussion of the design input parameters, assumptions, methodology, analysis, and results supporting the FHA are provided in Attachment 10 as Calculation DC00040-102, "Fuel Handling Accidents -

AST. Attachment 3 provides a matrix which compares the RG 1.183 regulatory position with the parameters and methodologies utilized to calculate the FHA onsite and offsite TEDE.

Table 4.4-1 provides a comparison of the design inputs utilized to determine the existing licensing basis FHA and the AST FHA doses.

Table 4.4-1: Design Input Comparison - Current Licensing Basis vs. AST Design - FHA Parameter CLB Parameter AST Parameter Core Thermal Power Level 2958 MWt 2958 MWt Earliest Fuel Handling Time 72 hr 72 hr Number of Damaged Fuel Pins 314 314 Core Radial Peaking Factor 1.7 1.7 Fraction of Fission Products in the Iodines except 1-131 0.10 1-131 0.16 Gap Available for Release from Iodine-131 0.12 Other Halogens 0.10 Damaged Rods Kr-85 0.30 Kr-85 0.20 Other nobles gases 0.10 Other Noble Gases 0.10 Alkali Metals 0.24 Core Bumup (MWD/MTU) Up to 70,000 Up to 70,000 Maximum Fuel Rod Pressurization <1200 psig < 1200 psig Release Timing (Gap) Instantaneously released & Instantaneously released &

mixed into pool water mixed into pool water Minimum Pool Water Depth 23 feet 23 feet Iodine Species Released to Pool 99.75%Elemental 99.85%Elemental 0.25% Organic 0.15% Organic Pool DF Noble gases - 1.0 Noble gases - 1.0 Aerosols - infinite Iodine - 100 Iodine - 200 Release Duration 2 hours2.314815e-5 days <br />5.555556e-4 hours <br />3.306878e-6 weeks <br />7.61e-7 months <br /> 2 hours2.314815e-5 days <br />5.555556e-4 hours <br />3.306878e-6 weeks <br />7.61e-7 months <br /> Filter Efficiency FHA Outside 95% for all iodine species No credit for filtration Containment Filter Efficiency FHA Inside No credit for filtration No credit for filtration Containment Containment locks and equipment Closed Open hatch Dose Conversion Factors ICRP 30 RADTRAD Table 1.4.3.3-2

Attachment 2 Page 20 of 54 Table 4.4-1: Design Input Comparison - Current Licensing Basis vs. AST Design - FHA Parameter CLB Parameter AST Parameter Control Room Habitability Envelope Note 1 226,040 ft3 (CRHE) Volume CRHE Isolation Time Note 1 0.5 hours5.787037e-5 days <br />0.00139 hours <br />8.267196e-6 weeks <br />1.9025e-6 months <br /> CRHE Emergency Filtered Intake Note 1 1,265 cfm Air Flow CRHE Unfiltered Air Inleakage Note 1 243 cfm CRHE Filtered Recirculation Flow Note 1 19,125 cfm Rate CRHE Emergency Filter Bed Depth Note 1 2 in. charcoal CRHE Emergency Filter Bed Note 1 95%

Removal Efficiency CRHE Operator Breathing Rates Note 1 3.5E-04 m3/sec (0 - 30 days)

CRI-IE Operator Occupancy Factors 1.0 0-24 hrs 1.0 0-24 hrs 0.6 1-4 days 0.6 1-4 days 0.4 4-30 days 0.4 4-30 days Note 1: The FHA CLB does not include a CR operator dose evaluation.

4.4.1 Introduction and Background Two accident scenarios are considered: (1) a refueling accident occurring inside containment and (2) a refueling accident occurring outside containment. This analysis was performed in accordance with the requirements of USNRC Regulatory Guide 1.183.

The postulated FHA inside containment is the dropping of a spent fuel assembly onto the core during refueling which results in damage to the fuel assemblies. There are numerous administrative controls and physical limitations which are imposed to prevent a fuel handling accident from occurring during refueling operations. Nevertheless, an accident sequence has been postulated with the objective of assessing the potential risk to the public health and safety.

It ispostulated that a spent fuel assembly is dropped onto the core during refueling resulting in breaching of the fuel rod cladding. As a result of the damage, a portion of the volatile fission gases are released to the water pool covering the core.

Subsequently, a fraction of the water soluble gases are absorbed in the pool with the remainder being transported through the water and into the Reactor Building atmosphere through open personnel and equipment hatches. The escaped gases are assumed to be immediately available for release to the environment and dispersed into the atmosphere.

The fuel handling accident outside containment is postulated as the dropping of a spent fuel assembly into the Spent Fuel Pool which results in damage to the fuel assemblies

Attachment 2 Page 21 of 54 and the release of the volatile gaseous fission products. The conditions and parameters assumed in analyzing the effects and consequences of this accident are identical to those utilized in the FHA inside containment except that the activity released to the environment is released through the Fuel Handling Building Exhaust System. For this evaluation no treatment of the radioactive release by the HEPA and charcoal filters of the Fuel Handling Building Exhaust System is credited. Accordingly, the activity released into the Fuel Handling Building is identical to that presented for the FHA inside containment case.

The NRC approved computer code RADTRAD, endorsed by RG 1.183, is used to calculate the dose to the Control Room operator as well as the doses at the EAB and LPZ.

4.4.2 Source Term Core inventory based on a power level of 2958 MWt and at time equal to 72 hours8.333333e-4 days <br />0.02 hours <br />1.190476e-4 weeks <br />2.7396e-5 months <br /> after shutdown was determined by running the computer code ORIGIN-S/ARP. Results are provided in as follows. The maximum isotopic activity for all fuel exposures is conservatively utilized in the analysis.

Table 4.4-2: Design Basis Core Inventory @ 72 Hours in Ci/MTU Fuel Exposure - MWD/MTU Nuclide 35,000 40,000 45,000 50,000 70,000 maximum 1-131 9.44E+05 9.49E+05 9.53E+05 9.54E+05 9.61E+05 9.61E+05 1-132 9.32E+05 9.38E+05 9.39E+05 9.38E+05- 9.33E+05 9.39E+05 1-133 2.31E+05 2.31E+05 2.30E+05 2.30E+05 2.24E+05 2.31E+05 1-134 0.OOE+00 0.OOE+00 0.OOE+00 0.OOE+00 0.OOE+00 0.OOE+00 1-135 1.18E+03 1.19E+03 1.18E+03 1.18E+03 1.16E+03 1.19E+03 Kr-85 1.06E+04 1.17E+04 1.28E+04 1.38E+04 1.70E+04 1.70E+04 Kr-85m 4.65E+00 4.47E+00 4.25E+00 4.07E+00 3.37E+00 4.65E+00 Kr-87 0.OOE+00 0.OOE+00 0.OOE+00 0.OOE+00 0.OOE+00 0.OOE+00 Kr-88 2.05E-02 1.96E-02 1.86E-02 1.77E-02 1.43E-02 2.05E-02 Xe-131m 1.51E+04 1.56E+04 1.60E+04 1.64E+04 1.78E+04 1.78E+04 Xe-133 1.97E+06 1.90E+06 1.91E+06 1.89E+06 1.92E+06 1.97E+06 Xe-133m 4.41E+04 4.29E+04 4.43E+04 4.28E+04 4.40E+04 4.43E+04 Xe-135 2.57E+04 2.57E+04 2.55E+04 2.54E+04 2.46E+04 2.57E+04 Xe-135m 1.93E+02 1.94E+02 1.93E+02 1.93E+02 1.90E+02 1.94E+02

Attachment 2 Page 22 of 54 Spent fuel source terms are based on reactor core source terms as discussed above, with a conservative factor of 2.0 multiplier to account for the gap fractions of fuel exceeding 54 GWD/MTU burnup with a maximum linear heat generation rate exceeding the 6.3 kW/ft peak rod average power limit to address RG 1.183 footnote 11. This is taken into account in the RADTRAD analysis by multiplying the core inventory gap fractions listed in Table 3 of RG 1.183 by the factor of 2 for input to the RADTRAD computer code.

Group Gap Fraction per Gap Fraction input to RG 1.183 RADTRAD 1-131 0.08 0.16 Kr-85 0.10 0.20 Other Noble Gases 0.05 0.10 Other Halogens 0.05 0.10 Alkali Metals 0.12 0.24 4.4.3 Mitigation The radiological consequences of the FHA are actively mitigated by pool scrubbing of the released iodine as shown in Table 4.4-1 and control room filtration (i.e., with an assumed charcoal filtration efficiency of 95%).

As described in Section 9.4.1 of the VCSNS updated FSAR, the control room ventilation system is automatically placed in the emergency mode, with filtration of incoming and recirculated air, following receipt of a SI or high radiation signal from the gaseous activity channel of RM-A1. Although very sensitive and fast acting, RM-A1 is not credited due to lack of redundancy. Since a safety injection will not occur for the FHA, manual initiation is credited in the AST analysis at 30 minutes. With direct communications maintained between the control room and personnel at the fuel handling stations and local radiations monitors (e.g., RM-G6 on the RB refueling bridge, RM-G8 on the FHB refueling bridge, RM-G17A/B on the RB Manipulator Crane, etc) with indication in the control room, this timing assumption (30 minutes) allows adequate time for the refueling personnel to contact the control room and for the control room operators to assess the significance of the event and the need for protective action to ensure Control Room Habitability.

Attachment 2 Page 23 of 54 4.4.4 Radiological Transport Modeling A simplified radiological release model and individual pathway models developed to calculate FHA doses utilizing RADTRAD is shown in Figure 3. Specific model details and the supporting RADTRAD runs are provided in Attachment 10 as Calculation DC00040-102, "Fuel Handling Accidents - AST." Based on the calculated Control Room x/Qs provided in Table 4.1-3, the conservative values used in the FHA AST dose analyses are taken at the following release points.

" FHA (Inside Containment) RB Nearest Point Intake A

• FHA (Outside Containment) Main Plant Vent Intake B 4.4.5 Results - Control Room Operator Dose The RADTRAD computer code was used to determine the Control Room operator dose for the FHA. The resultant doses are shown as follows.

FHA CR Operator Dose Rem TEDE FHA Inside Containment = 0.76 FHA Outside Containment = 0.41 4.4.6 Results - Offsite Doses The RADTRAD computer code was used to determine the offsite dose. The calculated doses are shown as follows where the EAB dose represents the maximum 2-hour TEDE over the accident period.

FHA EAB Dose Rem TEDE

" FHA Inside Containment = 4.29

" FHA Outside Containment = 4.29 FHA LPZ Dose Rem TEDE

" FHA Inside Containment = 1.06

• FHA Outside Containment = 1.06 4.4.7 Conclusions The FHA Control Room operator dose is below the 5 rem TEDE regulatory limit and each offsite dose is below the 6.3 rem TEDE regulatory limit for the inside Containment and inside the Fuel Handling Building (FHB) release pathways.

Attachment 2 Page 24 of 54 4.4.8 Summary of Calculation Conservatisms

• Mixing and/or dilution of the radioactivity released from the pool in the FHB or Containment is not considered in this calculation.

• Iodine removal by filtration/charcoal for the FHA in the Fuel Handling Building is not credited.

• For the FHA inside and outside Containment, no credit is taken for ESF filter operation, dilution, or mixing in the Reactor Building or the Fuel Handling Building.

• The FHA release inside Containment is assumed to occur as a ground level release from the nearest point to the CR (Reactor Building nearest point).

" Regulatory Guide 1.183, Table 3, non-LOCA fraction of fission products inventory in the gap is conservatively doubled.

• Isotopes considered are restricted to the 60 isotopes addressed in NUREG/CR-6604. For conservative reasons, additional noble gas isotopes are included in this analysis (Xe-131m, Xe-133m, and Xe-135m).

" The Control Room unanticipated unfiltered inleakage (UUI) value of 41 scfm, based on tracer tests, was increased to 200 scfm. An uncertainty value of 25 percent is conservatively applied to the Control Room filtered makeup flow rate and the unfiltered makeup flow rate which bypasses the damper.

Attachment 2 Page 25 of 54 Figure 3: Activity Flow Path Model Developed to Calculate FHA Doses Emergency Mode Recirculation Filter 95% aerosol, elemental, &

organic (0.5 - 720 hours0.00833 days <br />0.2 hours <br />0.00119 weeks <br />2.7396e-4 months <br />) 19,125 cfm @ 0.5 to 720 hours0.00833 days <br />0.2 hours <br />0.00119 weeks <br />2.7396e-4 months <br /> Volume 1 FP 2 - 1,291 cfm makeup (0 - 0.5 hr.)

No Filtration Volume 3 Gap Release FP 1 Volume 2 1,265 cfm filtered makeup (0.5- 720 hr.)

Volume Control Room 1.OE+ 10 cfm 95% aerosol, elemental, & organic Habitability 3 (arbitrary flow rate) Environment 1.0+04 ft Envelope No filtration or holdup* FP 3 - 217 cfm unfiltered inleakage (0 - 0.5 hr.)

243 cfm unfiltered inleakage (0.5 - 720 hr.) (226,040 ft3)

(arbitrary volume) unfiltered inleakage/ damper leakage /

ingress/egress FP 4 - CR exhaust: makeup plus unfiltered inleakage 1,508 cfm (0 - 720 hr.)

Attachment 2 Page 26 of 54 4.5 Main Steam Line Break The MSLB accident is postulated as a break of one of the large steam lines outside the containment leading from a SG. A detailed discussion of the design input parameters, assumptions, methodology, analysis, and results supporting the MSLB accident are provided in Attachment 10 as Calculation DC00040-099, "Main Steam Line Break -

AST". Attachment 3 provides a matrix which compares the RG 1.183 regulatory position with the parameters and methodologies utilized to calculate the MSLB accident CRHE and offsite doses.

Table 4.5-1 provides a comparison of the design inputs utilized to determine the existing licensing basis MSLB and the proposed AST MSLB accident doses.

Table 4.5-1: Design Input Comparison - Current Licensing Basis vs. AST Design - MSLB Parameter CLB Parameter AST Parameter Core Thermal Power Level 2958 MWt 2958 MWt Reactor Coolant Concentrations 1% Failed Fuel (Defects) 1% Failed Fuel (Defects)

Secondary Coolant Secondary coolant specific activity Secondary coolant specific activity Concentrations based on 0.1% VCi/gm DE 1-131 based on 0.1% VCi/gm DE 1-131 secondary activity, secondary activity.

Radioiodine Chemical Species NA 97% Elemental 3% Organic Pre-existing Iodine Spike 60 gtCi/gm DE 1-131 60 jtCi/gm DE 1-131 Concurrent Iodine Spike 500 Times the Iodine Equilibrium 500 Times the Iodine Equilibrium Release Rate with the Reactor Release Rate with the Reactor Coolant Coolant Activity at the 1.0 *tCi/gm Activity at the 1.0 gCi/gm DE 1-131 DE 1-131 assumed for 8 hours9.259259e-5 days <br />0.00222 hours <br />1.322751e-5 weeks <br />3.044e-6 months <br />. assumed for 8 hours9.259259e-5 days <br />0.00222 hours <br />1.322751e-5 weeks <br />3.044e-6 months <br />.

Iodine Partition Factor for Initial 1.0 1.0 Steam Release from Faulted SG Iodine Partition Factor in Intact 100 100 SG Noble Gas Partition Factor in 1.0 1.0 Intact SG Primary to Secondary Leak Rate 1 gpm for 8 hour9.259259e-5 days <br />0.00222 hours <br />1.322751e-5 weeks <br />3.044e-6 months <br /> duration 0.35 gpm for 24 hour2.777778e-4 days <br />0.00667 hours <br />3.968254e-5 weeks <br />9.132e-6 months <br /> duration to Faulted SG Primary to Secondary Leak Rate None 0.65 gpm for 24 hour2.777778e-4 days <br />0.00667 hours <br />3.968254e-5 weeks <br />9.132e-6 months <br /> duration to Intact SGs Total Release Duration 8 hours9.259259e-5 days <br />0.00222 hours <br />1.322751e-5 weeks <br />3.044e-6 months <br /> 24 hours2.777778e-4 days <br />0.00667 hours <br />3.968254e-5 weeks <br />9.132e-6 months <br /> Steam Release from Faulted SG 406,000 Ibm (0 - 30 min) 406,000 Ibm (0 - 30 min)

Steam Release from Intact SG 343,700 Ibm (0 - 2 hr) 343,700 Ibm (0 - 2 hr) 733,900 Ibm (2 - 8 hr) 733,900 Ibm (2 - 8 hr) 1,200,000 Ibm (8 - 24 hr)

Dose Conversion Factors ICRP 30 RADTRAD Table 1.4.3.3-2 Control Room Habitability Note 1 226,040 ft3 Envelope (CRHE) Volume CRHE Isolation Time Note 1 2 hours2.314815e-5 days <br />5.555556e-4 hours <br />3.306878e-6 weeks <br />7.61e-7 months <br /> CRHE Emergency Filtered Note 1 1,265 cfm Intake Air Flow

Attachment 2 Page 27 of 54 Table 4.5-1: Design Innuit Comnarison - Current Licensing Basis vs. AST Design - MSLB Parameter CLB Parameter AST Parameter CRHE Unfiltered Air Inleakage Note 1 243 cfm CRHE Filtered Recirculation Note 1 19,125 cfm Flow Rate CRHE Emergency Filter Bed Note 1 2 in. charcoal Depth CRHE Emergency Filter Bed Note 1 95%

Removal Efficiency CRHE Operator Breathing Rates Note 1 3.5E-04 m3/sec CRHE Operator Occupancy Note 1 1.0, 0-24 hrs Factors 0.6, 1-4 days 0.4, 4-30 days Note 1: CLB for the MSLB did not include a dose evaluation for CRHE.

4.5.1 Introduction and Background The MSLB accident is postulated as a break of one of the large steam lines outside the containment leading from a SG. For the two intact SGs loops, primary-to-secondary coolant leakage transfers activity into the secondary coolant. This makes it available for release into the environment via steaming through the SG PORV or SSVs. For the coolant loop with the broken steam line (i.e., faulted SG), primary-to-secondary coolant leakage is assumed to be released from the RCS directly into the environment without passing through any secondary coolant. Attachment 3 provides a matrix which compares the RG 1.183 regulatory position with the parameters and methodologies utilized to calculate the MSLB accident CRHE and offsite doses.

4.5.2 Source Term No fuel melt or fuel clad breach is postulated for the MSLB event at VCSNS. Consistent with RG 1.183 Appendix E, Section 2.2, if no or minimal fuel damage is postulated for the limiting event, the activity release should be the maximum allowed by technical specification for two cases of iodine spiking (1) maximum pre-existing (pre-accident) iodine spike and (2) maximum concurrent iodine spike.

The reactor coolant DE 1-131 values are based on VCSNS Technical Specification limit of 1.0 RiCi/gm DE 1-131 for the concurrent spike scenario and 60 liCi/gm DE 1-131 for the pre-existing spike scenario. The concurrent spike activity is based on an activity release rate from the fuel to the reactor coolant of 500 times the iodine equilibrium release rate consistent with the limiting condition for operation with the reactor coolant activity at the Technical Specification limit of 1.0 tCi/gm DE 1-131.

Attachment 2 Page 28 of 54 The resulting activity release to the environment is based on the following:

" Activity in secondary coolant released from faulted SG.

" Activity in reactor coolant released due to 0.35 gpm tube leak in faulted SG (0 to 24 hours2.777778e-4 days <br />0.00667 hours <br />3.968254e-5 weeks <br />9.132e-6 months <br />).

• Activity in reactor coolant released due to 0.65 gpm tube leak in intact SGs (0 to 24 hours2.777778e-4 days <br />0.00667 hours <br />3.968254e-5 weeks <br />9.132e-6 months <br />).

" Activity in secondary coolant released from intact SGs.

4.5.3 Mitigation Activity Removal Mechanisms in Containment The design basis MSLB at VCSNS releases activity directly into the RCS, therefore no plateout or other activity deposition is credited.

Decay Credited - MSLB Decay of radioactivity is credited in all compartments, prior to release. This is implemented in RADTRAD using the half-lives in the RADTRAD NIF file. The RADTRAD decay option is used.

Control Room Filtration Credit is taken for control room filtration (i.e., with an assumed charcoal filtration efficiency of 95%). As described in Section 9.4.1 of the VCSNS updated FSAR, the control room ventilation system is automatically placed in the emergency mode, with filtration of incoming and recirculated air, following receipt of a SI or high radiation signal from the gaseous activity channel of RM-A1. Although very sensitive and fast acting, RM-A1 is not credited due to lack of redundancy. For a large MSLB, SI actuation would occur on low RCS pressure, low steam line pressure, or high RB pressure (i.e., for breaks inside containment) within the first minute. However, as break size decreases, the time to SI actuation will increase. Spectrum analyses for breaks outside containment ranging in break size from 0.2 to 4.6 ft2 predict SI actuation within 10 minutes; and, FSAR analyses (Section 15.2.13) for accidental depressurization of the main steam system (i.e., due to a stuck open steam dump or safety valve) also show SI actuation on low RCS pressure within 10 minutes. Based on these analyzed times, the control room is conservatively assumed to enter the emergency recirculation mode, via automatic or manual initiation, within 30 minutes for the AST MSLB analysis.

This timing assumption bounds expected variations due to break size effects and is judged to be adequate for any MSLB accident that might challenge CR habitability.

Attachment 2 Page 29 of 54 Release of Activity to the Environment Per Technical Specification Bases 3/4.4.5, a total primary to secondary leakage for all three steam generators of 1 gpm is used in the evaluation of design basis accidents. Prior to the event this leakage is assumed to be distributed throughout the three steam generators. Recognizing that an extended plant cooldown may be required under natural circulation conditions, the MSLB is conservatively analyzed using 24 hours2.777778e-4 days <br />0.00667 hours <br />3.968254e-5 weeks <br />9.132e-6 months <br /> for the cooldown time. The activity associated with the 1 gpm leak is assumed to be released to the environment via the faulted steam generator at a rate of 0.35 gpm for the 24 hour2.777778e-4 days <br />0.00667 hours <br />3.968254e-5 weeks <br />9.132e-6 months <br /> duration of the event with no credit taken for any reduction or mitigation, i.e. a partition factor of 1.0. This is conservative in that the actual maximum value allowed by Technical Specification 3.4.6.2.c for any one SG is 150 gpd (- 0.104 gpm). The remaining 0.65 gpm is released to the environment via the two intact steam generators for the 24 hour2.777778e-4 days <br />0.00667 hours <br />3.968254e-5 weeks <br />9.132e-6 months <br /> duration of the event crediting a partition factor of 100.

4.5.4 Radiological Transport Modeling A simplified radiological release model and individual pathway models developed to calculate MSLB doses utilizing RADTRAD is shown in Figure 4. Specific model details and the supporting RADTRAD runs are provided in Attachment 10 as Calculation DC00040-099, "Main Steam Line Break - AST." Based on the calculated Control Room X/Qs provided in Table 4.1-3, the conservative values used in the MSLB AST dose analyses are taken at the following release point.

MSLB MS SSV A (Reliefs B, C, D, E), Intake B 4.5.5 Results - Control Room Operator Dose The RADTRAD computer code was used to determine the Control Room operator dose for the MSLB. The resultant doses are shown as follows.

MSLB CR Operator Dose Rem TEDE

• Pre-existing Iodine Spike = 1.15 0 Concurrent Iodine Spike = 0.37 4.5.6 Results - Offsite Doses The RADTRAD computer code was used to determine the offsite dose. The calculated doses are shown as follows where the EAB dose represents the maximum 2-hour TEDE over the accident period.

MSLB EAB Dose Rem TEDE

" Pre-existing Iodine Spike = 1.96

" Concurrent Iodine Spike = 0.78 MSLB LPZ Dose Rem TEDE

• Pre-existing Iodine Spike = 0.13

• Concurrent Iodine Spike = 0.16

Attachment 2 Page 30 of 54 4.5.7 Conclusions The calculated MSLB Control Room operator doses for the pre-existing and concurrent iodine spike cases are below the 5 rem TEDE regulatory limit.

The calculated MSLB offsite (EAB and LPZ) doses for the pre-existing iodine spike case are below the 25 rem TEDE regulatory limit. The calculated MSLB offsite doses for the concurrent iodine spike case are below the 2.5 rem TEDE regulatory limit.

4.5.8 Summary of Calculation Conservatisms The reactor coolant system (RCS) activity conservatively remains constant throughout the pre-existing Iodine Spike MSLB event. For the Concurrent Iodine Spike MSLB event, a similar assumption is made with the exception that the iodine activity increases during the 8 hours9.259259e-5 days <br />0.00222 hours <br />1.322751e-5 weeks <br />3.044e-6 months <br /> of the transient as a result of release from the defective fuel at a rate of 500 times the iodine equilibrium appearance rate.

• Isotopes considered are restricted to the 60 isotopes addressed in NUREG/CR-6604. For conservative reasons, additional isotopes are included in this analysis (Xe-131m, Xe-133m, Xe-135m, Rb-88, Br-83, Br-84, and Cs-138).

The Control Room unanticipated unfiltered inleakage (UUI) value of 41 scfm, based on tracer tests, was increased to 200 scfm. An uncertainty value of 25 percent is conservatively applied to the Control Room filtered makeup flow rate and the unfiltered makeup flow rate which bypasses the damper.

Attachment 2 Page 31 of 54 Figure 4 Activity Flow Path Model Developed to Calculate MSLB Doses Emergency Mode Recirculation Filter 95% aerosol, elemental, &

organic (0.5 - 720 hours0.00833 days <br />0.2 hours <br />0.00119 weeks <br />2.7396e-4 months <br />) 19,125 cfm @ 0 to 720 hours0.00833 days <br />0.2 hours <br />0.00119 weeks <br />2.7396e-4 months <br /> Volume 1 FP 2 - 1,291 cfm filtered makeup (0 - 0.5 hr.)

No Filtration Volume 3 Source FP 1 Volume 2 1,265 cfm filtered makeup (0.5 - 720 hr.)

Volume Control Room 1.OE+8 cfm 95% aerosol, elemental, & organic Environment Habitability 3 (arbitrary flow rate) 1.0+03 ft Envelope No filtration or holdup FP 3 - 217 cfm unfiltered inleakage (0 - 0.5 hr.) (226,040 ft3)

(arbitrary volume) 243 cfm unfiltered inleakage (0.5 - 720 hr.)

unfiltered inleakage/ damper leakage /

ingress/egress FP 4 - CR exhaust: makeup plus unfiltered inleakage 1,508 cfm (0 - 720 hr.)

Attachment 2 Page 32 of 54 4.6 Steam Generator Tube Rupture Accident A detailed discussion of the design input parameters, assumptions, methodology, analysis, and results supporting the SGTR accident are provided in Attachment 10 as Calculation DC00040-098, "Steam Generator Tube Rupture - AST". Attachment 3 provides a matrix which compares the RG 1.183 regulatory position with the parameters and methodologies utilized to calculate the SGTR accident CRHE and offsite doses.

Table 4.6-1 provides a comparison of the design inputs utilized to determine the existing licensing basis SGTR and the proposed AST SGTR accident doses.

Table 4.6-1: Design Input Comparison - Current Licensing Basis vs. AST Design - SGTR Parameter CLB Parameter AST Parameter Core Thermal Power Level 2958 MWt 2958 MWt Reactor Coolant Concentrations 1% Failed Fuel (Defects) 1% Failed Fuel (Defects)

Secondary Coolant Secondary coolant specific activity Secondary coolant specific activity Concentrations based on 0.1% pCi/gm DE 1-131 based on 0.1% VCi/gm DE 1-131 secondary activity, secondary activity.

Radioiodine Chemical Species NA 97% Elemental 3% Organic Pre-existing Iodine Spike 60 gCi/gm DE 1-131 60 gCi/gm DE 1-131 Concurrent Iodine Spike 500 Times the Iodine Equilibrium 335 Times the Iodine Equilibrium Release Rate with the Reactor Release Rate with the Reactor Coolant Coolant Activity at the 1.0 gtCi/gm Activity at the 1.0 gCi/gm DE 1-131 DE 1-131 assumed for 8 hours9.259259e-5 days <br />0.00222 hours <br />1.322751e-5 weeks <br />3.044e-6 months <br /> assumed for 8 hours9.259259e-5 days <br />0.00222 hours <br />1.322751e-5 weeks <br />3.044e-6 months <br /> Iodine Partition Factor for Initial 10 1.0 Steam Release from Faulted SG Iodine Partition Factor in Intact 10 100 SG Noble Gas Partition Factor in 1.0 1.0 Intact SG SG tube leak prior to and during 1.0 gpm for 8 hour9.259259e-5 days <br />0.00222 hours <br />1.322751e-5 weeks <br />3.044e-6 months <br /> duration 1.0 gpm for 24 hour2.777778e-4 days <br />0.00667 hours <br />3.968254e-5 weeks <br />9.132e-6 months <br /> duration the accident assumed all into the two intact SGs Steam Release from Faulted SG 56,800 Ibm (0 - 30 min.) 56,800 Ibm (0 - 30 min)

Steam Release from Intact SG 381,400 lbm (0 - 2 hr) 381,400 lbm (0 - 2 hr) 924,900 Ibm (2 - 8 hr) 924,900 Ibm (2 - 8 hr) 1,200,000 lbm (8 - 24 hr)

Reactor Coolant Release to 92,900 Ibm (0 - 30 min) 19,400 Ibm (0 - 385 sec.)

Faulted SG from Broken Tube 73,500 Ibm (385 sec. - 30 min) or 92,900 lbm (0 - 30 min)

Time at Which Primary to 30 minutes 30 minutes Secondary Leakage from Broken Tube is Assumed to be Terminated Total Release Duration 8 hours9.259259e-5 days <br />0.00222 hours <br />1.322751e-5 weeks <br />3.044e-6 months <br /> 24 hours2.777778e-4 days <br />0.00667 hours <br />3.968254e-5 weeks <br />9.132e-6 months <br /> Total Release Duration 8 hours9.259259e-5 days <br />0.00222 hours <br />1.322751e-5 weeks <br />3.044e-6 months <br /> 24 hours2.777778e-4 days <br />0.00667 hours <br />3.968254e-5 weeks <br />9.132e-6 months <br />

Attachment 2 Page 33 of 54 Table 4.6-1: Design Input Comparison - Current Licensing Basis vs. AST Design - SGTR Parameter CLB Parameter AST Parameter Dose Conversion Factors ICRP 30 RADTRAD Table 1.4.3.3-2 Control Room Habitability Note 1 226,040 ft3 Envelope (CRHE) Volume CRHE Isolation Time Note 1 2 hr CRHE Emergency Filtered Note 1 1,265 cfm Intake Air Flow CRHE Unfiltered Air Inleakage Note 1 243 cfm CRHE Filtered Recirculation Note 1 19,125 cfm Flow Rate CRHE Emergency Filter Bed Note 1 2 in. charcoal Depth CRHE Emergency Filter Bed Note 1 95%

Removal Efficiency CRHE Operator Breathing Rates Note 1 3.5E-04 m 3/sec CRHE Operator Occupancy Note 1 1.0, 0-24 hrs Factors 0.6, 1-4 days 0.4, 4-30 days Note 1: CLB for the SGTR did not include a dose evaluation for CRHE.

4.6.1 Introduction and Background The accident examined is the complete severance of a single steam generator tube. The accident is assumed to take place at power with the reactor coolant contaminated with fission products corresponding to continuous operation with a limited amount of defective fuel rods. The accident leads to an increase in contamination of the secondary system due to leakage of radioactive coolant from the reactor coolant system. In the event of a coincident loss of offsite power, or failure of the condenser steam dump system, discharge of activity to the atmosphere takes place via the steam generator safety and/or power operated relief valves.

Attachment 2 Page 34 of 54 4.6.2 Source Term No fuel melt or fuel clad breach is postulated for the SGTR event at VCSNS. Consistent with RG 1.183 Appendix E, Section 2.2, if no or minimal fuel damage is postulated for the limiting event, the activity release should be the maximum allowed by technical specification for two cases of iodine spiking (1) maximum pre-existing (pre-accident) iodine spike and (2) maximum concurrent iodine spike.

The reactor coolant DE 1-131 values are based on VCSNS Technical Specification limit of 1.0 gCi/gm DE 1-131 for the concurrent spike scenario and 60 g.tCi/gm DE 1-131 for the pre-existing spike scenario. The concurrent spike activity is based on an activity release rate from the fuel to the reactor coolant of 335 times the iodine equilibrium release rate consistent with the limiting condition for operation with the reactor coolant activity at the Technical Specification limit of 1.0 gCi/gm DE 1-13 1.

4.6.3 Mitigation Activity Removal Mechanisms in Containment The design basis SGTR at VCSNS releases activity directly from the RCS to the steam generators, therefore no plateout or other activity deposition in containment is credited.

Decay Credited - SGTR Decay of radioactivity is credited in all compartments, prior to release. This is implemented in RADTRAD using the half-lives in the RADTRAD NIF file. The RADTRAD decay option is used.

Control Room Filtration Credit is taken for control room filtration (i.e., with an assumed charcoal filtration efficiency of 95%). As described in Section 9.4.1 of the VCSNS updated FSAR, the control room ventilation system is automatically placed in the emergency mode, with filtration of incoming and recirculated air, following receipt of a SI or high radiation signal from the gaseous activity channel of RM-A1. Although very sensitive and fast acting, RM-A1 is not credited due to lack of redundancy. The CLB analysis (FSAR Section 15.4.3) assumes the double-ended rupture of a single tube and predicts a reactor trip and SI actuation on low RCS pressure in approximately 6.5 minutes. The FSAR analysis does not examine break size affects as primary to secondary size leakage and consequently doses are maximized for a large double-ended rupture. A timely automatic SI actuation is, however, expected to occur for large to medium size SGTRs; and, the operator is also instructed to initiate an SI (per Reference 12) for any tube leak that is outside of the capacity of the normal charging system. For the AST SGTR analysis, 30 minutes is conservatively assumed for the time after event initiation to initiate the emergency mode of operation. This timing assumption bounds the calculated times from the FSAR analyses and allows for substantial variation to cover break size effects, including the need to manually initiate SI.

Attachment 2 Page 35 of 54 Release of Activity to the Environment Table 4.6-1 provides the mass releases. The SGTR activity release to the environment is based on the following:

Activity in reactor coolant released due to tube rupture in the faulted SG: A portion of the primary to secondary leakage through the SGTR is assumed to flash to vapor, based on the thermodynamic conditions in the reactor and secondary coolant. The leakage that immediately flashes to vapor is assumed to rise through the bulk water of the SG and enter the steam space and is assumed to be immediately released to the environment with no mitigation; i.e., no reduction for scrubbing within the SG bulk water is credited. All leakage that does not immediately flash is assumed to mix with the bulk water. The radioactivity within the bulk water is assumed to become vapor at a rate that is the function of the steaming rate and the partition coefficient. A partition coefficient for

• iodine of 100 is assumed for the leakage from the bulk water.

Activity in reactor coolant released due to SG tube leak: Per Technical Specification Bases 3/4.4.5, a total primary to secondary leakage for all three steam generators of 1 gpm is used in the evaluation of design basis accidents. Prior to the event this leakage is assumed to be distributed throughout the three steam generators. Recognizing that an extended plant cooldown may be required under natural circulation conditions, the SGTR is conservatively analyzed using 24 hours2.777778e-4 days <br />0.00667 hours <br />3.968254e-5 weeks <br />9.132e-6 months <br /> for the cooldown time. The activity associated with the 1 gpm leak is conservatively assumed to be released to the environment via the two intact steam generators for the 24 hour2.777778e-4 days <br />0.00667 hours <br />3.968254e-5 weeks <br />9.132e-6 months <br /> duration of the event. A partition coefficient for iodine of 100 is assumed for this leakage. This is conservative in that the actual maximum value allowed by Technical Specification 3.4.6.2.c for any one SG is 150 gpd

(- 0.104 gpm).

Initial activity in secondary coolant released from the faulted and intact SGs: The initial secondary coolant activity is released to the environment with a partition factor of 100.

Noble gases have a partition factor of 1 for all release paths.

4.6.4 Radiological Transport Modeling A simplified radiological release model and individual pathway models developed to calculate SGTR doses utilizing RADTRAD is shown in Figure 5. Specific model details and the supporting RADTRAD runs are provided in Attachment 10 as Calculation DC00040-098, "Steam Generator Tube Rupture - AST." Based on the calculated Control Room X/Qs provided in Table 4.1-3, the conservative values used in the SGTR AST dose analyses are taken at the following release point.

0 SGTR MS SSV A (Relief's B, C, D, E), Intake B

Attachment 2 Page 36 of 54 4.6.5 Results - Control Room Operator Dose The RADTRAD computer code was used to determine the Control Room operator dose for the SGTR. The resultant doses are shown as follows.

SGTR CR Operator Dose Rem TEDE

" Pre-existing Iodine Spike = 1.18

" Concurrent Iodine Spike = 0.37 4.6.6 Results - Offsite Doses The RADTRAD computer code was used to determine the offsite dose. The calculated doses are shown as follows where the EAB dose represents the maximum 2-hour TEDE over the accident period.

SGTR EAB Dose Rem TEDE

• Pre-existing Iodine Spike = 2.08

• Concurrent Iodine Spike = 0.71 SGTR LPZ Dose Rem TEDE

• Pre-existing Iodine Spike = 0.12

" Concurrent Iodine Spike 0.047 4.6.7 Conclusions The calculated SGTR Control Room operator doses for the pre-existing and concurrent iodine spike cases are below the 5 rem TEDE regulatory limit.

The calculated SGTR offsite (EAB and LPZ) doses for the pre-existing iodine spike case are below the 25 rem TEDE regulatory limit. The calculated SGTR offsite doses for the concurrent iodine spike case are below the 2.5 rem TEDE regulatory limit.

4.6.8 Summary of Calculation Conservatisms

• Regulatory Guide 1.183, Table 3, non-LOCA fraction of fission products inventory in the gap are conservatively doubled.

0 Rb-86, Cs-134, Cs-136, and Cs-137 default core inventories from Table 1.4.3.2-2 of NUREG/CR-6604, corrected to a core thermal power of 2958 MWt, are included in this analysis. In addition, VCSNS noble gas and iodine core inventories and the corrected core inventories Table 1.4.3.2-2 of NUREG/CR-6604 were compared and the larger of the two concentrations are used in this analysis.

Attachment 2 Page 37 of 54

" The Control Room unanticipated unfiltered inleakage (UUI) value of 41 scfm, based on tracer tests, was increased to 200 scfm. An uncertainty value of 25 percent is conservatively applied to the Control Room filtered makeup flow rate and the unfiltered makeup flow rate which bypasses the damper.

" The flashed break flow from the ruptured SG tube prior to reactor trip (< 170 seconds) is released to the condenser with an iodine partition factor of 100. The SGTR analysis conservatively does not credit this reduction and assumes the entire flashed break flow is released to the environment with no reduction.

• A total SG tube leak of 1 gpm is assumed for the duration of the accident into the two intact SGs even though the maximum value in any one steam generator is limited to 150 gpd (- 0.104 gpm) in accordance with Technical Specification 3.4.6.2.c.

Attachment 2 Page 38 of 54 Figure 5: RADTRAD Model Developed to Calculate SGTR Accident Doses Emergency Mode Recirculation Filter 95% aerosol, elemental, &

organic

_1 5 (0.5 - 720 hours0.00833 days <br />0.2 hours <br />0.00119 weeks <br />2.7396e-4 months <br />) 19,125 cfm @ 0 to 720 hours0.00833 days <br />0.2 hours <br />0.00119 weeks <br />2.7396e-4 months <br /> Volume 1 FP 2 - 1,291 cfm filtered makeup (0 - 0.5 hr.)

No Filtration Volume 3 Source FP 1 Volume 2 1,265 cfm filtered makeup (0.5 - 720 hr.)

Volume Control Room 1.OE+8 cfm 95% aerosol, elemental, & organic Environment Habitability 3 (arbitrary flow rate) Envelope 1.0+03 ft No filtration or holdup FP 3 - 217 cfm unfiltered inleakage (0 - 0.5 hr.) (226,040 ft3 )

(arbitrary volume) 243 cfm unfiltered inleakage (0.5 - 720 hr.)

unfiltered inleakage/ damper leakage /

ingress/egress FP 4 - CR exhaust: makeup plus unfiltered inleakage 1,508 cfm (0 - 720 hr.)

Attachment 2 Page 39 of 54 4.7 Reactor Coolant Pump Locked Rotor Accident A detailed discussion of the design input parameters, assumptions, methodology, analysis, and results supporting the Locked Rotor accident are provided in Attachment 10 as Calculation DC00040-100, "Reactor Coolant Pump Locked Rotor - AST".

Attachment 3 provides a matrix which compares the RG 1.183 regulatory position with the parameters and methodologies utilized to calculate the Locked Rotor accident CRHE and offsite doses.

Table 4.7-1 provides a comparison of the design inputs utilized to determine the existing licensing basis Locked Rotor and the proposed AST Locked Rotor accident doses.

Table 4.7-1: Design Input Comrnarison - Current Licensing Basis vs. AST Design - LRA Parameter CLB Parameter AST Parameter Core Thermal Power Level 2958 MWt 2958 MWt Percent of Fuel Rods in Core Failed 15% 15%

Fraction of Fission Products in the Gap Iodine 0.10 1-131 0.16 Available for Release Kr-85 0.30 Other Halogens 0.10 Other Noble Gases 0.10 Kr-85 0.20 Other Noble Gases 0.10 Alkali Metals 0.24 Core Radial Peaking Factor 1.0 1.7 Defective Fuel Prior to Event 1% 1%

Steam Released from the 3 SGs 447,900 Ibm (0 - 2 hr) 447,900 Ibm (0 - 2 hr) 868,300 Ibm 02 - 8 hr) 868,300 Ibm (2 - 8 hr) 1,2000,000 Ibm (8 - 24 hr)

Mass of Reactor Coolant System 400,000 Ibm 400,000 Ibm Total Water Mass of the 3 SGs 340,000 Ibm 340,000 Ibm SG Blowdown Flow Rate 12,756 lb/hr 12,756 lb/hr Chemical Form of Radioiodine Released Not used Particulate 95%

from the Damaged Fuel Elemental 4.85%

Organic 0.15%

Chemical Form of Radioiodine Released Not used Elemental 97%

from SGs to Environs Organic 3%

Dose Conversion Factors ICRP 30 RADTRAD Table 1.4.3.3-2 Control Room Habitability Envelope Note 1 226,040 ft3 (CRHE) Total Volume CRHE Isolation Time Note 1 2 hours2.314815e-5 days <br />5.555556e-4 hours <br />3.306878e-6 weeks <br />7.61e-7 months <br /> CRHE Emergency Filtered Intake Air Note 1 1,265 cfm Flow CRHE Unfiltered Air Inleakage Note 1 243 cfm CRHE Filtered Recirculation Flow Rate Note 1 19,125 cfm CRHE Emergency Filter Bed Depth Note 1 2 in. charcoal CRHE Emergency Filter Bed Removal Note 1 95%

Efficiency CRHE Operator Breathing Rates 3.5E-04 m3/sec 3.5E-04 m 3/sec CRHE Operator Occupancy Factors 1.0, 0-24 hrs 1.0, 0-24 hrs 0.6, 1-4 days 0.6, 1-4 days 1 0.4, 4-30 days 0.4, 4-30 days Note 1: The CRHE dose was not calculated in the CLB analysis.

Attachment 2 Page 40 of 54 4.7.1 Introduction and Background The accident postulated is an instantaneous seizure of a reactor coolant pump rotor. Flow through the affected reactor coolant loop is rapidly reduced, leading to an initiation of a reactor trip on a low flow signal. Following initiation of the reactor trip heat stored in the fuel rods continues to be transferred to the coolant causing the coolant to expand. At the same time, heat transfer to the shell side of the steam generators is reduced, first because the reduced flow results in a decreased tube side film coefficient and then because the reactor coolant in the tubes cools down while the shell side temperature increases (turbine steam flow is reduced to zero upon plant trip). The rapid expansion of the coolant in the reactor core, combined with reduced heat transfer in the steam generators causes an insurge into the pressurizer and a pressure increase throughout the RCS. The insurge into the pressurizer compresses the steam volume, actuates the automatic spray system, opens the power operated relief valves, and opens the pressurizer safety valves, in that sequence. The three power operated relief valves are designed for reliable operation and would be expected to function properly during the accident. However, for conservatism, their pressure reducing effect as well as the pressure reducing effect of the spray is not included in the analysis.

4.7.2 Source Term The core inventory of the radionuclide groups required for non-LOCA events, based on RG 1.183, at 102% of the core thermal power are listed in Table 4.7-2.

Table 4.7-2: Core Activity @ T = 0 Hours Activity Activity Isotope (Ci) Isotope (Ci)

Kr-85 8.30E+05 Xe-133 1.70E+08 Kr-85m 2.72E+07 Xe-135 3.70E+07 Kr-87 4.96E+07 Cs-134 1.01E+07 Kr-88 6.71E+07 Cs-136 3.08E+06 Rb-86 4.43E+04 Cs-137 5.66E+06 1-131 8.20E+07 1-132 1.20E+08 1-133 1.68E+08 1-134 1.80E+08 1-135 1.54E+08

Attachment 2 Page 41 of 54 Reactor coolant equilibrium fission and corrosion product specific activity, based on 1%

fuel defects, are summarized in Table 4.7-3 for the applicable isotopes.

Table 4.7-3: Reactor Coolant Fission and Corrosion Product Specific Activity Activity Activity Isotope (PCi/gm) Isotope (PCi/gm)

Kr-85 7.6E+00 Xe-133 2.9E+02 Kr-85m 1.8E+00 Xe-135 8.6E+00 Kr-87 1.1E+00 Cs- 134 4.4E+00 Kr-88 3.2E+00 Cs-136 4.5E+00 Rb-86 3.6E-02 Cs-137 2.1E+00 1-131 3.OE+00 1-132 3.1E+00 1-133 4.6E+00 1-134 6.OE-01 1-135 2.4E+00 Fuel source terms are based on reactor core source terms as discussed above, with a conservative factor of 2.0 multiplier to account for the gap fractions of fuel exceeding 54 GWD/MTU bumup with a maximum linear heat generation rate exceeding the 6.3 kW/ft peak rod average power limit to address RG 1.183 footnote 11. This is taken into account in the RADTRAD analysis by multiplying the core inventory gap fractions listed in Table 3 of RG 1.183 by the factor of 2 for input to the RADTRAD computer code.

Group Gap Fraction Gap Fraction per RG 1.183 Input to RADTRAD 1-131 0.08 0.16 Kr-85 0.10 0.20 Other Noble Gases 0.05 0.10 Other Halogens 0.05 0.10 Alkali Metals 0.12 0.24 4.7.3 Mitigation The primary-to-secondary leakage is assumed to continue until the primary system pressure is less than the secondary system pressure, or until the temperature of the leakage is less than 1000 C (2120 F). The release of radioactivity is assumed to continue until shutdown cooling is in operation and releases from the steam generators have been terminated. In less than twenty four hours after the accident, the Residual Heat Removal System starts operation to cool down the plant. At this point, no steam and activity are released to the environment. All noble gas radionuclides released from the primary system are assumed to be released to the environment without reduction or mitigation.

The radioactivity in the secondary water is assumed to become a vapor at a rate that is the function of the steaming rate and the partition coefficient. A partition coefficient for iodine and the alkali metals of 100 is assumed.

Attachment 2 Page 42 of 54 No credit is taken for the pressure reducing effect of the pressurizer relief valves, pressurizer spray, steam dump or controlled feedwater flow after plant trip. Credit is taken for isolation of the CR via initiation of its emergency mode of operation filtered makeup (i.e., with an assumed charcoal filtration efficiency of 95%) at 0 hours0 days <br />0 hours <br />0 weeks <br />0 months <br /> after the accident.

Credit is taken for control room filtration (i.e., with an assumed charcoal filtration efficiency of 95%). As described in Section 9.4.1 of the VCSNS updated FSAR, the control room ventilation system is automatically placed in the emergency mode, with filtration of incoming and recirculated air, following receipt of a SI or high radiation signal from the gaseous activity channel of RM-Al. Although very sensitive and fast acting, RM-A1 is not credited due to lack of redundancy; and, without an additional failure, a SI actuation is also not anticipated for the LRA. Fuel failure would, however, be indicated via elevation radiation levels and abnormal radiation readings on a number of radiation monitors. Consequently, for the AST LRA analysis, 2 hours2.314815e-5 days <br />5.555556e-4 hours <br />3.306878e-6 weeks <br />7.61e-7 months <br /> from event initiation is conservatively assumed for manual initiation of the emergency mode of operation. This timing assumption provides adequate time for recognition of an ongoing atmospheric relief that requires the use of protective action to ensure Control Room Habitability.

4.7.4 Radiological Transport Modeling A simplified radiological release model and individual pathway models developed to calculate RCP LRA doses utilizing RADTRAD is shown in Figure 6. Specific model details and the supporting RADTRAD runs are provided in Attachment 10 as Calculation DC00040-100, "Reactor Coolant Pump Locked Rotor - AST." Based on the calculated Control Room X/Qs provided in Table 4.1-3, the conservative values used in the MSLB AST dose analyses are taken at the following release point.

LRA MS SSV A (Reliefs B, C, D, E), Intake B 4.7.5 Results - Control Room Operator Dose The RADTRAD computer code was used to determine the Control Room operator dose for the LRA. The resultant dose is shown as follows.

LRA CR Operator Dose 2.43 Rem TEDE 4.7.6 Results - Offsite Doses The RADTRAD computer code was used to determine the offsite dose. The calculated doses are shown as follows where the EAB dose represents the maximum 2-hour TEDE over the accident period.

LRA EAB Dose 2.20 Rem TEDE LRA LPZ Dose 0.45 Rem TEDE

Attachment 2 Page 43 of 54 4.7.7 Conclusions The LRA Control Room operator dose for the 1 gpm primary to secondary release case is below the 5 rem TEDE regulatory limit.

The LRA offsite (EAB and LPZ) doses are below the 2.5 rem TEDE regulatory limit at the EAB and the LPZ.

4.7.8 Summary of Calculation Conservatisms 0 Regulatory Guide 1.183, Table 3, non-LOCA fraction of fission products inventory in the gap is conservatively doubled.

• Rb-86, Cs-134, Cs-136, and Cs-137 default core inventories from Table 1.4.3.2-2 of NUREG/CR-6604, corrected to a core thermal power of 2958 MWt, are included in this analysis. In addition, VCSNS noble gas and iodine core inventories and the corrected core inventories Table 1.4.3.2-2 of NUREG/CR-6604 were compared and the larger of the two concentrations are used in this analysis.

9 The Control Room unanticipated unfiltered inleakage (UUI) value of 41 scfm, based on tracer tests, was increased to 200 scfm. An uncertainty value of 25 percent is conservatively applied to the Control Room filtered makeup flow rate and the unfiltered makeup flow rate which bypasses the damper.

Attachment 2 Page 44 of 54 Figure 6: RADTRAD Model Developed to Calculate RCP LRA Doses Emergency Mode Recirculation Filter 95% aerosol, Vt elemental, &

organic (2 - 720 hours0.00833 days <br />0.2 hours <br />0.00119 weeks <br />2.7396e-4 months <br />) 19,125 cfm @ 0 to 720 hours0.00833 days <br />0.2 hours <br />0.00119 weeks <br />2.7396e-4 months <br /> 5 Volume 1 FP 2 - 1,291 cfm filtered makeup (0 - 2 hr.)

95% aerosol, elemental, & organic Volume 3 Source FP 1 Volume 2 1,265 cfm filtered makeup (2 - 720 hr.)

Volume Control Room 1.OE+ 10 cfm 95% aerosol, elemental, & organic Environment Habitability 3 (arbitrary flow rate) Envelope 1.0+04 ft No filtration or holdup FP 3- 217 cfm unfiltered inleakage (0- 2 hr.) (226,040 ft3)

(arbitrary volume) 243 cfm unfiltered inleakage (2 - 720 hr.)

unfiltered inleakage/ damper leakage /

ingress/egress FP 4 - CR exhaust: makeuD plus unfiltered inleakage 1,508 cfm (0 - 720 hr.)

Attachment 2 Page 45 of 54 4.8 Control Rod Ejection Accident A detailed discussion of the design input parameters, assumptions, methodology, analysis, and results supporting the Control Rod Ejection accident are provided in Attachment 10 as Calculation DC00040-101, "Rod Ejection - AST". Attachment 3 provides a matrix which compares the RG 1.183 regulatory position with the parameters and methodologies utilized to calculate the Control Rod Ejection accident CRHE and offsite doses.

Table 4.8-1 provides a comparison of the design inputs utilized to determine the existing licensing basis Control Rod Ejection and the proposed AST Control Rod Ejection accident doses.

Thhle 4 R-1 Thesi~n Inniit Comnarisnn - Current IIcensin~ Basis vs AST Desirni - CRF.A Parameter CLB Parameter AST Parameter Core Thermal Power Level 2958 MWt 2958 MWt Percent of Fuel Rods in Core Failed 10% 10%

Fraction of Fission Products in the Iodine 0.10 Halogens 0.20 Gap Available for Release from Noble Gases 0.10 Noble Gases 0.20 Failed Rods Alkali Metals 0.24 Core Radial Peaking Factor 1.0 1.7

% of Fuel Melted 0.25% 0.25%

Fraction of Fission Products in the Iodine 0.5 Halogens 0.25 Melted Fuel for Containment Noble Gases 1.0 Noble Gases 1.0 Release Pathway Alkali Metals 0.30 Fraction of Fission Products in the Iodine 0.5 Halogens 0.50 Melted Fuel for Steam Generator Noble Gases 1.0 Noble Gases 1.0 Release Pathway Alkali Metals 0.50 Defective Fuel Prior to Event 1% 1%

Steam Released from the 3 SGs 33,000 Ibm (0 - 150 sec) 447,900 Ibm (0 - 2 hr) 868,300 Ibm (2 - 8 hr) 1,200,000 Ibm (8 - 24 hr)

Mass of Reactor Coolant System 400,000 Ibm 400,000 Ibm Total water mass of the 3 SGs 340,000 Ibm 340,000 lbm SG Blowdown Flow Rate 12,756 lb/hr 12,756 lb/hr Containment Natural Deposition of None 10% Power's Model Aerosols Secondary Coolant Partition Factor Iodine 100 Iodine 100 in SG Alkali Metals 100 Containment Free Volume 1,840,000 ft3 1,840,000 ft3 Containment Leak Rate 0.2%/day (0 - 24 hr) 0.2%/day (0 - 24 hr) 0.1 %/day (24 - 720 hr) 0.1%/day (24 - 720 hr)

Chemical Form of Radioiodine Not Used Particulate 95%

Released from the Damaged Fuel Elemental 4.85%

Organic 0.15%

Chemical Form of Radioiodine Not Used Elemental 97%

Released from SGs to Environs Organic 3%

Dose Conversion Factors ICRP 30 RADTRAD Table 1.4.3.3-2 Control Room Habitability Envelope Note 1 226,040 ft3 (CRHE) Total Volume

Attachment 2 Page 46 of 54 Table 4.8-1: Design Input Comparison - Current Licensing Basis vs. AST Design - CREA Parameter CLB Parameter AST Parameter CRHE Isolation Time Note 1 0.5 hours5.787037e-5 days <br />0.00139 hours <br />8.267196e-6 weeks <br />1.9025e-6 months <br /> Containment Release 2 hours2.314815e-5 days <br />5.555556e-4 hours <br />3.306878e-6 weeks <br />7.61e-7 months <br /> SG Release CRHE Emergency Filtered Intake Note 1 1,265 cfm Air Flow CRHE Unfiltered Air Inleakage Note 1 243 cfm CRHE Filtered Control Room Note 1 19,125 cfm Recirculation Flow Rate CRHE Emergency Filter Bed Depth Note 1 2 in. charcoal CRHE Emergency Filter Bed Note 1 95%

Removal Efficiency CRHE Operator Breathing Rates 3.5E-04 m3/sec 3.5E-04 m 3/sec CRHE Operator Occupancy Factors 1.0, 0-24 hrs 1.0, 0-24 hrs 0.6, 1-4 days 0.6, 1-4 days 0.4, 4-30 days 0.4, 4-30 days Note 1: The CRHE dose was not calculated in the CLB analysis.

4.8.1 Introduction and Background This accident is defined as the mechanical failure of a control rod mechanism pressure housing resulting in the ejection of a rod cluster control assembly (RCCA) and drive shaft. The consequence of this mechanical failure is a rapid positive reactivity insertion together with an adverse core power distribution, possibly leading to localized fuel rod damage. Following the postulated rod ejection accident, two separate scenarios for activity release to the environment are considered: (1) a breach of the RPV and containment leakage for 30 days or (2) primary-to secondary leakage and secondary steaming via the relief valves until cold shutdown.

Attachment 2 Page 47 of 54 4.8.2 Source Term The core inventory of the radionuclide groups required for non-LOCA events, based on RG 1.183, at 102% of the core thermal power are listed in Table 4.8-2.

Table 4.8-2: Core Activity @ T = 0 Hours Activity Activity Isotope (Ci) Isotope (Ci)

Kr-85 8.30E+05 Xe-133 1.70E+08 Kr-85m 2.72E+07 Xe-135 3.70E+07 Kr-87 4.96E+07 Cs-134 1.01E+07 Kr-88 6.71E+07 Cs-136 3.08E+06 Rb-86 4.43E+04 Cs-137 5.66E+06 1-131 8.20E+07 1-132 1.20E+08 1-133 1.68E+08 1-134 1.80E+08 1-135 1.54E+08 Reactor coolant equilibrium fission and corrosion product specific activity, based on 1%

fuel defects, are summarized in Table 4.8-3 for the applicable isotopes.

Table 4.8-3: Reactor Coolant Fission and Corrosion Product Specific Activity at Equilibrium Activity Activity Isotope (PCi/gm) Isotope ([pCi/gm)

Kr-85 7.6E+00 Xe-133 2.9E+02 Kr-85m 1.8E+00 Xe-135 8.6E+00 Kr-87 1.1E+00 Cs- 134 4.4E+00 Kr-88 3.2E+00 Cs-136 4.5E+00 Rb-86 3.6E-02 Cs-137 2.1E+00 1-131 3.OE+00 1-132 3.1E+00 1-133 4.6E+00 1-134 6.OE-01 1-135 2.4E+00 Fuel source terms are based on reactor core source terms as discussed above, with a conservative factor of 2.0 multiplier to account for the gap fractions of fuel exceeding 54 GWD/MTU burnup with a maximum linear heat generation rate exceeding the 6.3 kW/ft peak rod average power limit to address RG 1.183 footnote 11. This is taken into account in the RADTRAD analysis by multiplying the core inventory gap fractions listed in Table 3 of RG 1.183 by the factor of 2 for input to the RADTRAD computer code.

Attachment 2 Page 48 of 54 Group Gap Fraction Gap Fraction per RG 1.183Input to RADTRAD Noble Gases 0.10 0.20 Halogens 0.10 0.20 Alkali Metals 0.12 0.24 4.8.3 Mitigation For the Containment release pathway, leakage is assumed at the Technical Specifications leak rate for peak accident 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 50% of this leak rate for the remaining duration of the accident. Credit is taken for natural deposition removal of aerosols utilizing the RADTRAD 10% Power's model. No credit is assumed in the containment for removal of released activity by the containment sprays or the internal containment recirculation HEPA filters.

For the SG release pathway, per Technical Specification Bases 3/4.4.5, a total primary to secondary leakage for all three steam generators of 1 gpm is used in the evaluation of design basis accidents. Prior to the event this leakage is assumed to be distributed throughout the three steam generators. Recognizing that an extended plant cooldown may be required under natural circulation conditions, the CREA is conservatively analyzed using 24 hours2.777778e-4 days <br />0.00667 hours <br />3.968254e-5 weeks <br />9.132e-6 months <br /> for the cooldown time. The activity associated with the 1 gpm leak is assumed to be released to the environment via the 3 intact SGs until the primary system pressure is less than the secondary system pressure (24 hours2.777778e-4 days <br />0.00667 hours <br />3.968254e-5 weeks <br />9.132e-6 months <br />) at which time the release of radioactivity from the secondary system as a result of steam dump through the relief valve to the atmosphere terminates. At this point, no steam and activity are released to the environment. All noble gas radionuclides released from the primary system are assumed to be released to the environment without reduction or mitigation.

The radioactivity in the secondary water is assumed to become a vapor at a rate that is the function of the steaming rate and the partition coefficient. A partition coefficient for iodine and the alkali metals of 100 is assumed.

Credit is taken for control room filtration (i.e., with an assumed charcoal filtration efficiency of 95%). As described in Section 9.4.1 of the VCSNS updated FSAR, the control room ventilation system is automatically placed in the emergency mode, with filtration of incoming and recirculated air, following receipt of a SI or high radiation signal from the gaseous activity channel of RM-A1. Although very sensitive and fast acting, RM-A1 is not credited due to lack of redundancy. For the CREA with releases to the RB, the emergency mode of operation is assumed to occur as a result of automatic SI actuation. For the maximum possible breakisize (2.75 in diameter),

automatic SI actuation on low RCS pressure would be expected within the first few minutes of the event (per FSAR Section 15.4.6.2.10). However, because the timing of the SI signal is break size dependent, a 30-minute time delay is conservatively assumed to account for break size effects. For the SG release analysis, 2 hours2.314815e-5 days <br />5.555556e-4 hours <br />3.306878e-6 weeks <br />7.61e-7 months <br /> from event initiation is conservatively assumed for manual initiation of the emergency mode of operation. This timing assumption provides adequate time for recognition of an ongoing atmospheric relief that requires the use of protective action to ensure Control Room Habitability.

Attachment 2 Page 49 of 54 4.8.4 Radiological Transport Modeling A simplified radiological release model and individual pathway models developed to calculate CREA doses utilizing RADTRAD is shown in Figure 7. Specific model details and the supporting RADTRAD runs are provided in Attachment 10 as Calculation DC00040-101, "Rod Ejection - AST." Based on the calculated Control Room X/Qs provided in Table 4.1-3, the conservative values used in the MSLB AST dose analyses are taken at the following release point.

" CREA Containment Release RB Nearest Point

• CREA SG Release MS SSV A (Reliefs B, C, D, E), Intake B 4.8.5 Results - Control Room Operator Dose The RADTRAD computer code was used to determine the Control Room operator dose for the CREA. The resultant doses are shown as follows.

CREA CR Operator Dose Rem TEDE

• Case 1 CREA Containment Release= 1.71

• Case 2 CREA SG Release = 2.38 4.8.6 Results - Offsite Doses The RADTRAD computer code was used to determine the offsite dose. The calculated doses are shown as follows where the EAB dose represents the maximum 2-hour TEDE over the accident period.

CREA EAB Dose Rem TEDE

" Case 1 CREA Containment Release= 4.31

• Case 2 CREA SG Release = 2.52 CREA LPZ Dose Rem TEDE

• Case 1 CREA Containment Release= 0.95

" Case 2 CREA SG Release = 0.48 4.8.7 Conclusions The CREA Control Room operator dose is below the 5 rem TEDE regulatory limit.

The CREA offsite (EAB and LPZ) doses are below the 6.3 rem TEDE regulatory limit at the EAB and the LPZ.

Attachment 2 Page 50 of 54 4.8.8 Summary of Calculation Conservatisms

• Regulatory Guide 1.183, Table 3, non-LOCA fraction of fission products inventory in the gap are conservatively doubled.

• Rb-86, Cs-134, Cs-136, and Cs-137 default core inventories from Table 1.4.3.2-2 of NUREG/CR-6604, corrected to a core thermal power of 2958 MWt, are included in this analysis. In addition, VCSNS noble gas and iodine core inventories and the corrected core inventories Table 1.4.3.2-2 of NUREG/CR-6604 were compared and the larger of the two concentrations are used in this analysis.

" The Control Room unanticipated unfiltered inleakage (UUI) value of 41 scfm, based on tracer tests, was increased to 200 scfm. An uncertainty value of 25 percent is conservatively applied to the Control Room filtered makeup flow rate and the unfiltered makeup flow rate which bypasses the damper.

Attachment 2 Page 51 of 54 Figure 7A: RADTRAD Model Developed to Calculate CREA Doses (Containment Release)

Emergency Mode Recirculation Filter 95% aerosol, IiF elemental, &

organic (0.5 - 720 hours0.00833 days <br />0.2 hours <br />0.00119 weeks <br />2.7396e-4 months <br />) 19,125 cfm @ 0 to 720 hours0.00833 days <br />0.2 hours <br />0.00119 weeks <br />2.7396e-4 months <br /> 5 FP 2 - 1,291 cfm filtered makeup (0 -0.5 hr.)

Volume 1 1,265 cfm filtered makeup (0.5 - 720 hr.) Volume 3 Volume 2 95% aerosol, elemental, & organic @ 0.5 hours5.787037e-5 days <br />0.00139 hours <br />8.267196e-6 weeks <br />1.9025e-6 months <br /> Control Room Containment FP 1 Habitability Volume 0.2%/day 0 - 24 hrs. Environment Envelope 0.1%/day 24 - 720 hrs. 3 1.84E+06 ft3 FP 3 - 217 cfm unfiltered inleakage (0 - 0.5 hr.) (226,040 ft )

AL 243 cfm unfiltered inleakage (0.5 - 720 hr.)

unfiltered inleakage/ damper leakage /

ingress/egress FP 4 - CR exhaust: makeup plus unfiltered inleakage 1,508 cfm (0 - 720 hr.)

Attachment 2 Page 52 of 54 Figure 7B: RADTRAD Model Developed to Calculate CREA Doses (SG PORV Release)

Emergency Mode Recirculation Filter 95% aerosol, elemental, &

organic (2 - 720 hours0.00833 days <br />0.2 hours <br />0.00119 weeks <br />2.7396e-4 months <br />) 19,125 cfm @ 0 to 720 hours0.00833 days <br />0.2 hours <br />0.00119 weeks <br />2.7396e-4 months <br />

__{JFP 5 Volume 1 FP 2 - 1,291 cfm filtered makeup (0 - 2 hr.)

1,265 cfm filtered makeup (2 - 720 hr.) - Volume 3 w

Source FP 1 Volume 2 95% aerosol, elemental, & organic @ 2 hours2.314815e-5 days <br />5.555556e-4 hours <br />3.306878e-6 weeks <br />7.61e-7 months <br /> Control Room Volume L.OE+10 cfm Habitability (arbitrary flow rate) Environment 3 Envelope 1.0+04 ft No filtration or holdup FP 3 - 217 cfm unfiltered inleakage (0 - 2 hr.) (226,040 ft3)

(arbitrary volume) 243 cfm unfiltered inleakage (2 - 720 hr.)

unfiltered inleakage/ damper leakage /

ingress/egress FP 4 - CR exhaust: makeup plus unfiltered inleakage 1,508 cfm (0 - 720 hr.)

Attachment 2 Page 53 of 54 4.9 Equipment Qualification and NUREG-0737 As previously noted, this request for an amendment to the licensing basis for the Virgil C.

Summer Nuclear Station that a full implementation application of an Alternative Source Term (AST) methodology with the following exceptions. The exceptions are that the current TID-14844 (Reference 9) accident source term will remain the licensing basis for equipment qualification, NUREG-0737 evaluations other than Control Room Habitability Envelope (CRHE) doses, and FSAR accidents not included in Regulatory Guide 1.183.

5.0 REGULATORY SAFETY ANALYSIS (10 CFR 50.92 Evaluation)

The 10 CFR 50.92 Evaluation is included as Attachment 9 of this submittal.

6.0 ENVIRONMENTAL CONSIDERATION

S (10 CFR 50.21 Evaluation)

The 10 CFR 50.21 Evaluation is included as Attachment 9 of this submittal.

7.0 REFERENCES

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

2. NUREG-0800, Section SRP 15.0.1, "Radiological Consequence Analyses Using Alternative Source Terms", Revision 0, July 2000.

3. USNRC Regulatory Guide 1.194, "Atmospheric Relative Concentrations For Control.

Room Radiological Habitability Assessments At Nuclear Power Plants", June 2003.

4. NUREG-0737, "Clarification of TMI Action Plan Requirements", November 1980.

5. J.V. Ramsdell and C.A. Simonen, "Atmospheric Relative Concentrations in Building Wakes", NUREG-633 1, Revision 1, USNRC, May 1997 (ARCON96 computer code).

6. USNRC, "Standard Review Plan For the Review of Safety Analysis Reports for Nuclear Power Plants", Chapter 6.4, "Control Room Habitability System," NUREG-0800, USNRC, 1987.

7. NRC Letter Virgil C. Summer Nuclear Station - NRC Receipt of response to Generic Letter 2003-01, "Control Room Habitability" (TAC NO. MB9860), from Robert E.

Martin to Jeffery Archie, 10/24/2006.

8. Westinghouse Electric Company, "Radiation Analysis Manual", for Virgil C. Summer (CGE/3-1), Revision 1, attached to letter CGE-98-036, dated October 14, 1998.

9. U.S. Atomic Energy Commission Technical Information Document TID-14844, "Calculation of Distance Factors for Power and Test Reactor Sites", dated March 23, 1962.

Attachment 2 Page 54 of 54

10. Jeffery B. Archie (SCE&G) letter, RC-05-0193 to Document Control Desk (NRC),

"Response to NRC Generic Letter 2003-01 Control Room Habitability", November 18, 2005.

11. General Test Procedure, GTP-006, "General Procedure for System Leakage Assessment."

12. Abnormal Operating Procedure, AOP- 112.2, "Steam Generator Tube Leak Not Requiring SI."

13. NUREG/CR-6604, RADTRAD: A Simplified Model for RADionuclide Transport and Removal And Dose Estimation, December 1997, Supplement 1, June 8, 1999 &

Supplement 2, October 2002.

Attachment 3 Regulatory Guide 1.183 Compliance Table

Attachment 3 Regulatory Guide 1.183 Compliance Table Page 1 of 41 Regulatory Guide 1.183 Compliance Regulatory Guide 1.183 Sections 3 through 7 and Appendices A, B, E, F, G and H provide methodologies and assumptions that are acceptable to the NRC staff related to design basis radiological analyses for Alternate Source Term. Compliance with Regulatory Guide 1.183 positions are discussed below:

Please note: the information provided in this table is based on the calculations provided in Attachment 10.

RG 1.183 Regulatory Guide 1.183 Position Basis of Compliance Section

3. ACCIDENT SOURCE TERM Conforms This section provides an AST that is acceptable to the NRC staff. The data in Regulatory Positions Accident terms source of were developed based on the guidance Section 3 of RG 1.183.

g 3.2 through 3.5 are fundamental to the definition of an AST. Once approved, the AST assumptions or parameters specified in these positions become part of the facility's design basis. Deviations from this guidance must be evaluated against Regulatory Position 2. After the NRC staff has approved an implementation of an AST, subsequent changes to the AST will require NRC staff review under 10 CFR 50.67.

Attachment 3 Regulatory Guide 1.183 Compliance Table Page 2 of 41 3.1 Fission Product Inventory Conforms The inventory of fission products in the reactor The inventory of fission products in the reactor core and available for release to the containment core and available for release to the should be based on the maximum full power operation of the core with, as a minimum, current containment was based on the maximum full licensed values for fuel enrichment, fuel burnup, and an assumed core power equal to the current power operation with a core thermal power of licensed rated thermal power times the ECCS evaluation uncertainty. Note: the uncertainty factor 2958 MWt [102% (ECCS evaluation used in determining the core inventory should be that value provided in Appendix K to 10 CFR uncertainty) of 2900 MWt],

Part 50, typically 1.02. The period of irradiation should be of sufficient duration to allow the activity of dose-significant radionuclides to reach equilibrium or to reach maximum values. Note that some radionuclides, such as Cs-137, equilibrium will not be reached prior to fuel offload. Thus, The maximum inventory at the end of life was the maximum inventory at the end of life should be used. The core inventory should be determined used.

using an appropriate isotope generation and depletion computer code such as ORIGEN 2 (Ref. 17) or ORIGEN-ARP (Ref. 18). Core inventory factors (Ci/MWt) provided in TID 14844 and used in some analysis computer codes werederived for low burnup, low enrichment fuel and should not be used with higher burnup and higher enrichment fuels.

For the DBA LOCA, all fuel assemblies in the core are assumed to be affected and the core average inventory should be used. For DBA 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, radial peaking factors from the facility's core operating limits report (COLR) or Technical Specifications should be applied in determining the inventory of the damaged rods.

No adjustment to the fission product inventory should be made for events postulated to occur during power operations at less than full rated power or those postulated to occur at the beginning of core life. For events postulated to occur while the facility is shutdown, e.g., a fuel handling accident, radioactive decay from the time of shutdown may be modeled.

Attachment 3 Regulatory Guide 1.183 Compliance Table Page 3 of 41 3.2 Release Fractions Conforms Note: the release fractions listed here have been determined to be acceptable for use with currently approved LWR fuel with a For the LOCA event, the core inventory release peak burnup of 62,000 MWD/MTU. The data in this section may not be applicable to cores containing mixed oxide (MOX) fractions, by radionuclide groups, for the gap fuel.

release and early in-vessel damage phases in The core inventory release fractions, by radionuclide groups, for the gap release and early in-vessel damage phases for DBA Table 2 were utilized.

LOCAs are listed in Table 2 for PWRs. These fractions are applied to the equilibrium core inventory described in Regulatory Position 3.1.

For non-LOCA events, the fractions of the core inventory assumed to be in the gap for the various radionuclides are given in Table 3. The release fractions from Table 3 are used in conjunction with the fission product inventory calculated with the maximum core radial peaking factor. For non-LOCA events, the fraction of the core Table 2 inventory assumed to be in the gap by PWR Core Inventory Fraction radionuclide group in Table 3 were utilized in Released Into Containment conjunction with the maximum core radial Gap Early Release In-Vessel peaking factor. The CREA was evaluated per Group Phase Phase Total Note 11 of RG 1.183 (the gap fractions are Noble Gases 0.05 0.95 1.00 assumed to be 10% for iodines and noble Halogens 0.05 0.35 0.40 gases).

Alkali Metals 0.05 0.25 0.30 Tellurium Metals 0.00 0.05 0.05 To account for possible variations in burnup Ba, Sr 0.00 0.02 0.02 Noble Metals 0.0025 0.0025 and rod power, the gap fractions in Table 3 0.00 Cerium Group 0.00 0.0005 0.0005 were increased by a factor of 2. This factor was Lanthanides 0.00 0.0002 0.0002 applied in the FHA, RCP LRA and CREA analyses.

For non-LOCA events, the fractions of the core inventory assumed to be in the gap for the various radionuclides are given in Table 3. The release fractions from Table 3 are used in conjunction with the fission product inventory calculated with the maximum core radial peaking factor.

Table 3 Non-LOCA Fraction of Fission Product Inventory in Gap Group Fraction 1-131 0.08 Kr-85 0.10 Other Noble Gases 0.05 Other Halogens 0.05 Alkali Metals 0.12 Note: Table 3 release fractions listed here have been determined to be acceptable for use with currently approved LWR fuel with a peak bumup up to 62,000 MWD/MTU provided that the maximum linear heat generation rate does not exceed 6.3 kw/ft peak rod average power for bumups exceeding 54 GWD/MTU. As an altemative, fission gas release calculations performed using NRC approved methodologies may be considered on a case-by-case basis. To be acceptable, these calculations must use a projected power history that will bound the limiting projected plant-specific power history for the specific fuel load. For the BWR rod drop accident and the PWR rod ejection accident, the gap fractions are assumed to be 10% for iodines and noble aases.

Attachment 3 Regulatory Guide 1.183 Compliance Table Page 4 of 41 3.3 Timing of Release Phases Conforms Table 4 onset and duration of each sequential Table 4 tabulates the onset and duration of each sequential release phase for DBA LOCAs at PWRs. release phase for the DBA LOCA were utilized The specified onset is the time following the initiation of the accident (i.e., time = 0). The early in the analysis in a linear manner.

in-vessel phase immediately follows the gap release phase. The activity released from the core during each release phase should be modeled as increasing in a linear fashion over the duration of the phase. For non-LOCA DBAs in which fuel damage is projected, the release from the fuel gap and the fuel pellet should be assumed to occur instantaneously with the onset of the projected damage.

Table 4 LOCA Release Phases Phase Onset Duration Gap Release 30 sec. 0.5 hr Early In-Vessel 0.5 hr 1.3 hr In lieu of treating the release in a linear ramp manner, the activity for each phase can be modeled as being released instantaneously at the start of that release phase, i.e., in step increases.

For facilities licensed with leak-before-break methodology, the onset of the gap release phase may be assumed to be 10 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.

Attachment 3 Regulatory Guide 1.183 Compliance Table Page 5 of 41 3.4 Radionuclide Composition Conforms Table 5 elements in each radionuclide group Table 5 lists the elements in each radionuclide group that should be considered in design basis were utilized in design basis analyses.

analyses.

Table 5 Radionuclide Groups Group Elements Noble Gases Xe, Kr Halogens I, Br Alkali Metals Cs, Rb Tellurium Group Te, Sb, Se, Ba, Sr Noble Metals Ru, Rh, Pd, Mo, Tc, Co Lanthanides La, Zr, Nd, Eu, Nb, Pm, Pr Sm, Y, Cm, Am Cerium Ce, Pu, Np 3.5 Chemical Form Conforms Of the radioiodines released from the reactor Of the radioiodines released from the reactor coolant system (RCS) to the containment in a coolant system (RCS) to the containment in a postulated accident, 95 percent of the iodine released should be assumed to be cesium iodide (CsI),

postulated accident, 95 percent of the iodine 4.85 percent elemental iodine, and 0.15 percent organic iodide. This includes releases from the gap released was assumed to be cesium iodide and the fuel pellets. With the exception of elemental and organic iodine and noble gases, fission products should be assumed to be in particulate form. The same chemical form is assumed in (CsI), 4.85 percent elemental iodine, and 0.15 releases from fuel pins in FHAs and from releases from the fuel pins through the RCS in DBAs percent organic iodide.

other than FHAs or LOCAs. However, the transport of these iodine species following release from With the exception of elemental and organic the fuel may affect these assumed fractions. The accident-specific appendices to this regulatory iodine and noble gases, fission products were guide provide additional details. assumed to be in particulate form. The same chemical form was assumed in releases from fuel pins in FHAs and from releases from the fuel pins through the RCS in DBAs other than FHAs or LOCAs.

Attachment 3 Regulatory Guide 1.183 Compliance Table Page 6 of 41 3.6 Fuel Damage in Non-LOCA DBAs Conforms The amount of fuel damage caused by non-The amount of fuel damage caused by non-LOCA design basis events should be analyzed to LOCA design basis events was analyzed to determine, for the case resulting in the highest radioactivity release, the fraction of the fuel that determine, for the case resulting in the highest reaches or exceeds the initiation temperature of fuel melt and the fraction of fuel elements for which radioactivity release, the fraction of the fuel the fuel clad is breached. Although the NRC staff has traditionally relied upon the departure from that reaches or exceeds the initiation nucleate boiling ratio (DNBR) as a fuel damage criterion, licensees may propose other methods to temperature of fuel melt and the fraction of fuel the NRC staff, such as those based upon enthalpy deposition, for estimating fuel damage for the elements for which the fuel clad is breached.

purpose of establishing radioactivity releases.

The amount of fuel damage caused by a FHA is addressed in Appendix B of this guide.

4.0 DOSE CALCULATION METHODOLOGY Conforms The DBA analyses, based on ASTs, utilized the The NRC staff has determined that there is an implied synergy between the ASTs and total effective dose calculation methodology of Section 4 of dose equivalent (TEDE) criteria, and between the TID- 14844 source terms and the whole body and RG 1.183.

thyroid dose criteria, and therefore, they do not expect to allow the TEDE criteria to be used with TID-14844 calculated results. The guidance of this section applies to all dose calculations performed with an AST pursuant to 10 CFR 50.67. Certain selective implementations may not require dose calculations as described in Regulatory Position 1.3 of this guide.

4.1 Offsite Dose Consequences Conforms The dose calculation methodology of in determining the TEDE for persons located at or The following assumptions should be used (EB):Section Secton 4.1 Raof 3 tlize o beyod tearebonday oftheexclsio 4.1 of RG 1. 183 was utilized to beyond the boundary of the exclusion area (EAB): calculate the offsite dose consequences.

4.1.1 The dose calculations should determine the TEDE. TEDE is the sum of the committed effective Conforms dose equivalent (CEDE) from inhalation and the deep dose equivalent (DDE) from external The dose calculations determine the TEDE and exposure. The calculation of these two components of the TEDE should consider all radionuclides, consider all radionuclides, including progeny including progeny from the decay of parent radionuclides that are significant with regard to dose from the decay of parent radionuclides that are consequences and the released radioactivity.

• ficay significant with of pardntto dose regard dionu enes .

consequences.

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

Attachment 3 Regulatory Guide 1.183 Compliance Table Page 7 of 41 4.1.2 The exposure-to-CEDE factors for inhalation of radioactive material should be derived from the Conforms data provided in ICRP Publication 30, "Limits for Intakes of Radionuclides by Workers" (Ref. 19). Conversion factors for isotopes other than the Table 2.1 of Federal Guidance Report 11, "Limiting Values of Radionuclide Intake and Air standard 60 isotopes of the RADTRAD Concentration and Dose Conversion Factors for Inhalation, Submersion, and Ingestion" (Ref. 20), computer program (default FGR A files)

D were "effective"tables provides yieldofdoses conversion factors acceptable corresponding to the NRC staff. The factors in the column headed to the CEDE. taken from 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."

4.1.3 For the first 8 hours9.259259e-5 days <br />0.00222 hours <br />1.322751e-5 weeks <br />3.044e-6 months <br />, the breathing rate of persons offsite should be assumed to be 3.5 x 10-4 cubic Conforms meters per second. 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 Breathing rates provided in Section 4.1.3 of to be 1.8 x 10-4 cubic meters per second. After that and until the end of the accident, the rate should rg re utilized to cacte the of be assumed to be 2.3 x 10-4 cubic meters per second. d s cons equ enes.

dose consequences.

4.1.4 The DDE should be calculated assuming submergence in semi-infinite cloud assumptions with Conforms appropriate credit for attenuation by body tissue. The DDE is nominally equivalent to the effective Conversion factors for isotopes other than the dose equivalent (EDE) from external exposure if the whole body is irradiated uniformly. Since this Condardion factors f the otran is a reasonable assumption for submergence exposure situations, EDE may be used in lieu of DDE standard 60 isotopes of the R 1ADTRAD in determining the contribution of external dose to the TEDE. Table HI.1 of Federal Guidance taken from Table 111.1 of Federal Guidance Report 12, "External Exposure to Radionuclides in Air, Water, and Soil" (Ref. 21), provides external EDE conversion factors acceptable to the NRC staff. The factors in the column headed Report 12, "External Exposure to "effective" yield doses corresponding to the EDE. Radionuclides in Air, Water, and Soil."

4.1.5 The TEDE should be determined for the most limiting person at the EAB. The maximum EAB Conforms TEDE for any two-hour period following the start of the radioactivity release should be determined The TEDE was determined for the most and used in determining compliance with the dose criteria in 10 CFR 50.67. The maximum two- limiting person at the EAB. The maximum hour TEDE should be determined by calculating the postulated dose for a series of small time two-hour TEDE was determined by calculating increments and performing a "sliding" sum over the increments for successive two-hour periods, the postulated dose for a series of small time The maximum TEDE obtained is submitted. The time increments should appropriately reflect the increments and performing a "sliding" sum progression of the accident to capture the peak dose interval between the start of the event and the increments formingcessive sur end of radioactivity release (see also Table 6). periods.

Note: With regard to the EAB TEDE, the maximum two-hour value is the basis for screening and evaluation under 10 CFR 50.59. Changes to doses outside of the two-hour window are only considered in the context of their impact on the maximum two-hour EAB TEDE.

Attachment 3 Regulatory Guide 1.183 Compliance Table Page 8 of 41 4.1.6 TEDE should be determined for the most limiting receptor at the outer boundary of the low Conforms population zone (LPZ) and should be used in determining compliance with the dose criteria in The TEDE was determined for the most 10 CFR 50.67. limiting receptor at the outer boundary of the low population zone (LPZ).

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

4.2 Control Room Dose Consequences Conforms The DBA analyses utilized the guidance of The following guidance should be used in determining the TEDE for persons located in the Control Section 4.2 of RG 1.183 to determining the Room: TEDE for persons located in the Control Room.

4.2.1 The TEDE analysis should consider all sources of radiation that will cause exposure to Control Conforms Room personnel. The applicable sources will vary from facility to facility, but typically will include:

The TEDE analysis considered all significant sources of radiation that will cause exposure to

• Contamination of the Control Room atmosphere by the intake or infiltration of the radioactive Control Room personnel.

material contained in the radioactive plume released from the facility.

• Contamination of the Control Room atmosphere by the intake or 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.

• Radiation shine from radioactive material in the reactor containment.

• Radiation shine from radioactive material in systems and components inside or external to the Control Room envelope, e.g., radioactive material buildup in recirculation filters.

The radioactive material releases and radiation levels used in the Control Room dose analysis Conforms should be determined using the same source term, transport, and release assumptions used for The radioactive material releases and radiation determining the EAB and the LPZ TEDE values, unless these assumptions would result in levels used in the Control Room dose analysis non-conservative results for the Control Room. were determined using the same source term, transport, and release assumptions used for determining the EAB and the LPZ TEDE values as appropriate.

Attachment 3 Regulatory Guide 1.183 Compliance Table Page 9 of 41 4.2.3 The models used to transport radioactive material into and through the Control Room, and the Conforms shielding models used to determine radiation dose rates from external sources, should be structured The models used to transport radioactive to provide suitably conservative estimates of the exposure to Control Room personnel. material into and through the Control Room, and the shielding models used to determine The iodine protection factor (IPF) methodology of Reference 22 may not be adequately conservative radiation dose rates from external sources, were dope p to rovi es tal conrve for all DBAs and Control Room arrangements since it models a steady-state Control Room condition. Since many analysis parameters change over the duration of the event, the IPF developed to provide suitably conservative methodology should only be used with caution. The NRC computer codes HABIT (Ref. 23) and estimates of the exposure to Control Room personnel.

RADTRAD (Ref. 24) incorporate suitable methodologies.

4.2.4 Credit for engineered safety features that mitigate airborne radioactive material within the Control Conforms Room may be assumed. Such features may include Control Room isolation or pressurization, or Cre intake or recirculation filtration. Refer to Section 6.5.1, "ESF Atmospheric Cleanup System," of the dit for engineered safety features that SRP (Ref. 3) and Regulatory Guide 1.52, "Design, Testing, and Maintenance Criteria for Post- mitigate airborne radioactive material within accident Engineered-Safety-Feature Atmosphere Cleanup System Air Filtration and Adsorption appropriate. Credit for engineered safety Units of Light-Water-Cooled Nuclear Power Plants" (Ref. 25), for guidance. The Control Roomfeatures varied for each of the analyzed DBAs.

design is often optimized for the DBA LOCA and the protection afforded for other accident sequences may not be as advantageous. In most designs, Control Room isolation is actuated by engineered safeguards feature (ESF) signals or radiation monitors (RMs). In some cases, the ESF signal is effective only for selected accidents, placing reliance on the RMs for the remaining accidents. Several aspects of RMs can delay the Control Room isolation, including the delay for activity to build up to concentrations equivalent to the alarm setpoint and the effects of different radionuclide accident isotopic mixes on monitor response.

4.2.5 Credit should generally not be taken for the use of personal protective equipment or prophylactic Conforms drugs. Deviations may be considered on a case-by-case basis. Credit was not taken for the use of personal protective equipment or prophylactic drugs.

Attachment 3 Regulatory Guide 1.183 Compliance Table Page 10 of 41 4.2.6 The dose receptor for these analyses is the hypothetical maximum exposed individual who is present Conforms in the Control 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 event, 60% of the time Occupancy factors and breathing rate of between 1 and 4 days, and 40% of the time from 4 days to 30 days. For the duration of the event, RG 1.183, Section 4.2.6 were utilized to the breathing rate of this individual should be assumed to be 3.5 x 10-4 cubic meters per second.

determine the doses to the hypothetical maximum exposed individual who is present in Note: This occupancy is modeled in the X/Q values determined in Reference 22 and should not be the Control Room.

credited twice. The ARCON96 Code (Ref. 26) does not incorporate these occupancy assumptions, making it necessary to apply this correction in the dose calculations. Control Room X/Q values were determined utilizing the ARCON96 computer code.

Occupancy factors were included in the RADTRAD computer code for dose evaluations.

4.2.7 Control Room doses should be calculated using dose conversion factors identified in Regulatory Conforms Position 4.1 above for use in offsite dose analyses. The DDE from photons may be corrected for the The DDE from photons was corrected for the difference between finite cloud geometry in the Control Room and the semi-infinite cloud difference between finite cloud geometry in the assumption used in calculating the dose conversion factors. The following expression may be used Control Room and the semi-infinite cloud to correct the semi-infinite cloud dose, DDE., to a finite cloud dose, DDEfmite, where the Control Room is modeled as a hemisphere that has a volume, V, in cubic feet, equivalent to that of the assumption used in calculating the dose conversion factors by Equation 1 as necessary.

Control Room (Ref. 22).

DDE_ V° 3 338 DDE finite = 1173 1173 Equation I 4.3 Other Dose Consequences Conforms The guidance provided in Regulatory Positions 4.1 and 4.2 should be used, as applicable, in re-assessing the radiological analyses identified in Regulatory Position 1.3.1, such as those in Exception - The current TID- 14844 accident NUREG-0737 (Ref. 2). Design envelope source terms provided in NUREG-0737 should be updated source term will remain the licensing basis for for consistency with the AST. In general, radiation exposures to plant personnel identified in equipment qualification, NUREG-0737 Regulatory Position 1.3.1 should be expressed in terms of TEDE. Integrated radiation exposure of evaluations other than Control Room plant equipment should be determined using the guidance of Appendix I of this guide. Habitability Envelope (CRHE) doses, and FSAR accidents not included in Regulatory Guide 1.183.

Attachment 3 Regulatory Guide 1.183 Compliance Table Page 11 of 41 4.4 Acceptance Criteria Conforms The radiological criteria for the EAB, the outer boundary of the LPZ, and for the Control Room are in 10 CFR The DBAs were updated for consistency with 50.67. These criteria are stated for evaluating reactor accidents of exceedingly low probability of occurrence the TEDE criterion in 10 CFR 50.67(b)(2)(iii).

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) items generally reference General Design Criteria 19 (GDC 19) from Appendix A to 10 CFR Part 50 or specify criteria derived from GDC-19. These criteria are generally specified in terms of whole body dose, or its equivalent to any body organ. For facilities applying for, or having received, approval for the use of an AST, the applicable criteria should be updated for consistency with the TEDE criterion in 10 CFR 50.67(b)(2)(iii).

Table 6 Accident Dose Criteria AB and LPZ Accident or Case Dose Criteria Analysis Release Duration LOCA 25 rem TEDE 30 days for containment and ECCS leakage PWR Steam Generator Tube Rupture Fuel Damage or Pre-incident Spike 25 rem TEDE Affected SG: time to isolate; Unaffected SG(s): until cold shutdown is established Coincident Iodine Spike 2.5 rem TEDE PWR Main Steam Line Break Fuel Damage or Pre-incident Spike 25 rem TEDE Until cold shutdown is established Coincident Iodine Spike 2.5 rem TEDE PWR Locked Rotor Accident 2.5 rem TEDE Until cold shutdown is established PWR Rod Ejection Accident 6.3 rem TEDE 30 days for containment pathway; until cold shutdown is established for secondary pathway Fuel Handling Accident 6.3 rem TEDE 2 hours2.314815e-5 days <br />5.555556e-4 hours <br />3.306878e-6 weeks <br />7.61e-7 months <br /> The column labeled "Analysis Release Duration" is a summary of the assumed radioactivity release durations identified in the individual appendices to this guide. Refer to these appendices for complete descriptions of the release pathways and durations.

Attachment 3 Regulatory Guide 1.183 Compliance Table Page 12 of 41 ANALYSIS ASSUMPTIONS AND METHODOLOGY General Considerations Analysis Quality Conforms Analyses performed per 10 CFR 50, The evaluations required by 10 CFR 50.67 are re-analyses of the design basis safety analyses and Appendix B and the guidance consistent with evaluations required by 10 CFR 50.34; they are considered to be a significant input to the RG 1.183.

evaluations required by 10 CFR 50.92 or 10 CFR 50.59. These analyses should be prepared, reviewed, and maintained in accordance with quality assurance programs that comply with Appendix B, "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 conservative set of assumptions to test the performance of one or more aspects of the facility design. Many physical processes and phenomena are represented by conservative, bounding assumptions rather than being modeled directly. The staff has selected assumptions and models that provide an appropriate and prudent safety margin against unpredicted events in the course of an accident and compensate for large uncertainties in facility parameters, accident progression, radioactive material transport, and atmospheric dispersion.

Licensees should exercise caution in proposing deviations based upon data from a specific accident sequence since the DBAs were never intended to represent any specific accident sequence - the proposed deviation may not be conservative for other accident sequences.

5.1.2 Credit for Engineered Safeguard Features Conforms Credit may be taken for accident mitigation features that are classified as safety-related, are required Credit was taken for Engineered Safeguard to be operable by Technical Specifications, are powered by emergency power sources, and are either Features with failure assumptions to maximize automatically actuated or, in limited cases, have actuation requirements explicitly addressed in the calculated doses. Assumptions regarding the most emergency operating procedures. The single active component failure that results in limiting radiological consequences should be assumed. Assumptions regarding the occurrence and power were also selected with the objective of timing of a loss of offsite power should be selected with the objective of maximizing the postulated maximizing the postulated radiological radiological consequences. consequences

Attachment 3 Regulatory Guide 1.183 Compliance Table Page 13 of 41 5.1.3 Assignment of Numeric Input Values Conforms The numeric values that were chosen as inputs The numeric values that are chosen as inputs to the analyses required by 10 CFR 50.67 should be to the analyses required by 10 CFR 50.67 were selected with the objective of determining a conservative postulated dose. In some instances, a selected with the objective of determining a particular parameter may be conservative in one portion of an analysis but be non-conservative in conservative postulated dose.

another portion of the same analysis. For example, assuming minimum containment system spray flow is usually conservative for estimating iodine scrubbing, but in many cases may be non- For a range of values, the value that resulted in a conservative postulated dose was used.

conservative when determining sump pH. Sensitivity analyses may be needed to determine the appropriate value to use. As a conservative alternative, the limiting value applicable to each portion of the analysis may be used in the evaluation of that portion. A single value may not be applicable for a parameter for the duration of the event, particularly for parameters affected by changes in density. For parameters addressed by Technical Specifications, the value used in the analysis should be that specified in the technical specifications. If a range of values or a tolerance band is specified, the value that would result in a conservative postulated dose should be used. If the parameter is based on the results of less frequent surveillance testing, e.g., steam generator nondestructive testing (NDT), consideration should be given to the degradation that may occur between periodic tests in establishing the analysis value.

Note that for some parameters, the Technical Specification value may be adjusted for analysis purposes by factors provided in other regulatory guidance. For example, ESF filter efficiencies are based on the guidance in Regulatory Guide 1.52 (Ref. 25) and in Generic Letter 99-02 (Ref. 27) rather than the surveillance test criteria in the Technical Specifications. Generally, these adjustments address potential changes in the parameter between scheduled surveillance tests. 4-

-~ -.

5.1.4 Applicability of Prior Licensing Basis Conforms The NRC staff considers the implementation of an AST to be a significant change to the design Licensee has ensured that analysis assumptions basis of the facility that is voluntarily initiated by the licensee. In order to issue a license and methods are compatible with the AST and amendment authorizing the use of an AST and the TEDE dose criteria, the NRC staff must make a the TEDE criteria.

current finding of compliance with regulations applicable to the amendment. The characteristics of the ASTs and the revised dose calculational methodology may be incompatible with many of the analysis assumptions and methods currently reflected in the facility's design basis analyses. The NRC staff may find that new or unreviewed issues are created by a particular site-specific implementation of the AST, warranting review of staff positions approved subsequent to the initial issuance of the license. This is not considered a backfit as defined by 10 CFR 50.109, "Backfitting." However, prior design bases that are unrelated to the use of the AST, or are unaffected by the AST, may continue as the facility's design basis. Licensees should ensure that analysis assumptions and methods are compatible with the ASTs and the TEDE criteria.

Attachment 3 Regulatory Guide 1.183 Compliance Table Page 14 of 41 5.2 Accident-Specific Assumptions Conforms Licensee analyzed the DBAs that are affected The appendices to this regulatory guide provide accident-specific assumptions that are acceptable to by the specific proposed applications of an the staff for performing analyses that are required by 10 CFR 50.67. The DBAs addressed in these AST, utilizing the guidance provided in the attachments were selected from accidents that may involve damage to irradiated fuel. This guide appendices of RG 1.183.

does not address DBAs with radiological consequences based on technical specification reactor or secondary coolant-specific activities only. The inclusion or exclusion of a particular DBA in this guide should not be interpreted as indicating that an analysis of that DBA is required or not required.

Licensees should analyze the DBAs that are affected by the specific proposed applications of an AST.

The NRC staff has determined that the analysis assumptions in the appendices to this guide provide an integrated approach to performing the individual analyses and generally expects licensees to address each assumption or propose acceptable alternatives. Such alternatives may be justifiable on the basis of plant-specific considerations, updated technical analyses, or, in some cases, a previously approved licensing basis consideration. The assumptions in the appendices are deemed consistent with the AST identified in Regulatory Position 3 and internally consistent with each other.

Although licensees are free to propose alternatives to these assumptions for consideration by the NRC staff, licensees should avoid use of previously approved staff positions that would adversely affect this consistency.

The NRC is committed to using probabilistic risk analysis (PRA) insights in its regulatory activities and will consider licensee proposals for changes in analysis assumptions based upon risk insights.

The staff will not approve proposals that would reduce the defense in depth deemed necessary to provide adequate protection for public health and safety. In some cases, this defense in depth compensates for uncertainties in the PRA analyses and addresses accident considerations not adequately addressed by the core damage frequency (CDF) and large early release frequency (LERF) surrogate indicators of overall risk.

Attachment 3 Regulatory Guide 1.183 Compliance Table Page 15 of 41 5.3 Meteorology Assumptions Conforms Atmospheric dispersion values (X/Q) for the Atmospheric dispersion values (x/Q) for the EAB, the LPZ, and the Control Room that were EAB and the LPZ, that were approved by the approved by the staff during initial facility licensing or in subsequent licensing proceedings may be staff during initial facility licensing or in used in performing the radiological analyses identified by this guide. Methodologies that have been subsequent licensing proceedings are used in used for determining X/Q values are documented in Regulatory Guides 1.3 and 1.4, Regulatory performing the radiological analyses identified Guide 1.145, "Atmospheric Dispersion Models for Potential Accident Consequence Assessments at by this guide.

Nuclear Power Plants," and the paper, "Nuclear Power Plant Control Room Ventilation System Design for Meeting General Criterion 19" (Refs. 6, 7, 22, and 28).

New x/Qs for the Control Room were References 22 and 28 should be used if the FSAR X/Q values are to be revised or if values are to be developed based on more recent meteorological determined for new release points or receptor distances. Fumigation should be considered where data, utilizing the guidance of RG 1.194.

applicable for the EAB and LPZ. For the EAB, the assumed fumigation period should be timed to be included in the worst 2-hour exposure period. The NRC computer code PAVAN (Ref. 29) implements Regulatory Guide 1.145 (Ref. 28) and its use is acceptable to the NRC staff. The methodology of the NRC computer code ARCON96 (Ref. 26) is generally acceptable to the NRC staff for use in determining Control Room X/Q values. Meteorological data collected in accordance with the site-specific meteorological measurements program described in the facility FSAR should be used in generating accident X/Q values. Additional guidance is provided in Regulatory Guide 1.23, "Onsite Meteorological Programs" (Ref. 30). All changes in X/Q analysis methodology should be reviewed by the NRC staff.

6.0 ASSUMPTIONS FOR EVALUATING THE RADIATION DOSES FOR EQUIPMENT N/A QUALIFICATION An AST assessment was not performed for equipment qualification. TID- 14844, The assumptions in Appendix I to this guide are acceptable to the NRC staff for performing "Calculation of Distance Factors for Power and radiological assessments associated with equipment qualification. The assumptions in Appendix I Test Reactor Sites," will continue to be used as will supersede Regulatory Positions 2.c(l) and 2.c(2) and Appendix D of Revision 1 of Regulatory the radiation dose basis for equipment Guide 1.89, "Environmental Qualification of Certain Electric Equipment Important to Safety for qualification and radiation zone maps/shielding Nuclear Power Plants" (Ref. 11), for operating reactors that have amended their licensing basis to calculations.

use an alternative source term. Except as stated in Appendix I, all other assumptions, methods, and provisions of Revision I of Regulatory Guide 1.89 remain effective. The NRC staff is assessing the effect of increased cesium releases on EQ doses to determine whether licensee action is warranted.

Until such time as this generic issue is resolved, licensees may use either the AST or the TID 14844 assumptions for performing the required EQ analyses. However, no plant modifications are required to address the impact of the difference in source term characteristics (i.e., AST vs.

TID14844) on EQ doses pending the outcome of the evaluation of the generic issue.

Attachment 3 Regulatory Guide 1.183 Compliance Table Page 16 of 41 D. IMPLEMENTATION Conforms The AST analysis utilized the guidance of The purpose of this section is to provide information to applicants and licensees regarding the NRC RG 1.183 in the AST evaluations and did not staff's plans for using this regulatory guide. Except in those cases in which an applicant or licensee use alternative method for complying with the proposes an acceptable alternative method for complying with the specified portions of the NRC's specified portions of the NRC's regulations.

regulations, the methods described in this guide will be used in the evaluation of submittals related to the use of ASTs in radiological consequence analyses at operating power reactors.

Attachment 3 Regulatory Guide 1.183 Compliance Table Page 17 of 41 Appendix ASSUMPTIONS FOR EVALUATING THE RADIOLOGICAL CONSEQUENCES OF A Conforms A LWR LOSS-OF-COOLANT ACCIDENT An analysis was performed utilizing the guidance of Appendix A and appropriate The assumptions in this appendix are acceptable to the NRC staff for evaluating the radiological sections in the main body of RG 1.183 to consequences of loss-of-coolant accidents (LOCAs) at light water reactors (LWRs). These evaluate a LOCA.

assumptions supplement the guidance provided in the main body of this guide.

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 design basis accidents. (DBAs), 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 ECCS performance. 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 and for evaluating the proposed siting of a facility.

SOURCE TERM ASSUMPTIONS Acceptable assumptions regarding core inventory and the release of radionuclides from the fuel are Conforms provided in Regulatory Position 3 of this guide. Assumptions regarding core inventory and the release of radionuclides from the fuel were per Regulatory Position 3 of RG 1.183.

2 If the sump or suppression pool pH is controlled at values of 7 or greater, the chemical form of Conforms radioiodine released to the containment should be assumed to be 95% cesium iodide (CsI), 4.85 percent elemental iodine, and 0.15 percent organic iodide. Iodine species, including those from Thesi r t o iodine re-evolution, for sump or suppression pool pH values less than 7 will be evaluated on a case-by-case basis. Evaluations of pH should consider the effect of acids and bases created during the LOCA event, e.g., radiolysis products. With the exception of elemental and organic iodine and noble gases, fission products should be assumed to be in particulate form.

ASSUMPTIONS ON TRANSPORT IN PRIMARY CONTAINMENT 3 Acceptable assumptions related to the transport, reduction, and release of radioactive material in and Conforms from the primary containment in PWRs are as follows: The analysis utilized the assumptions related to the transport, reduction, and release of radioactive material in and from the primary containment in PWRs per RG 1.183.

Attachment 3 Regulatory Guide 1.183 Compliance Table Page 18 of 41 3.1 The radioactivity released from the fuel should be assumed to mix instantaneously and Conforms homogeneously throughout the free air volume of the primary containment in PWRs as it is released. This distribution should be adjusted if there are internal compartments that have limited The radioactivity released from the fuel was ventilation exchange. The release into the primary containment should be assumed to terminate at homogened tomisanout the free air volume the end of the early in-vessel phase. of the primary containment as it was released.

This distribution was not adjusted because there is adequate ventilation exchange.

3.2 Reduction in airborne radioactivity in the containment by natural deposition within the containment Conforms may be credited. Acceptable models for removal of iodine and aerosols are described in Chapter 6.5.2, "Containment Spray as a Fission Product Cleanup "System,"

in18,NREGCR- of the Standard Siplifed Review ode ofAersol Tecontprcnti l wes Decontamination Model wasAeroso l conservatively Plan (SRP), NUREG-0800 (Ref. A- 1) and in NUREG/CR-6189, "A Simplified Model of Aerosoldeposition.

Removal by Natural Processes in Reactor Containments" (Ref. A-2). The latter model is incorporated into the analysis code RADTRAD (Ref. A-3). The prior practice of deterministically assuming that a 50% plateout of iodine is released from the fuel is no longer acceptable to the NRC staff as it is inconsistent with the characteristics of the revised source terms.

Attachment 3 Regulatory Guide 1.183 Compliance Table Page 19 of 41 3.3 Reduction in airborne radioactivity in the containment by containment spray systems that have been Conforms designed and are maintained in accordance with Chapter 6.5.2 of the SRP (Ref. A- 1) may be The VCSNS containment building atmosphere credited. Acceptable models for the removal of iodine and aerosols are described in Chapter 6.5.2 is considered a single, well-mixed volume of the SRP and NUREG/CR-5966, "A Simplified Model of Aerosol Removal by Containment since the spray covers at least 90% of the Sprays"I (Ref. A-4). This simplified model is incorporated into the analysis code RADTRAD volume and adequate mixing of unsprayed (Refs. A-I to A-3).

compartments is provided in the design.

The elemental iodine spray removal coefficient The evaluation of the containment sprays should address areas within the primary containment that is 20 hr-1 and the particulate iodine spray are not covered by the spray drops. The mixing rate attributed to natural convection between removal coefficient is 5.68 hrl. The removal sprayed and unsprayed regions of the containment building, provided that adequate flow exists rate for the particulate iodines is reduced by a between these regions, is assumed to be two turnovers of the unsprayed regions per hour, unless factor of 10 (0.568 hr-1) when a DF of 50 is other rates are justified. The containment building atmosphere may be considered a single, well- reached. There is no specified maximum DF mixed volume if the spray covers at least 90% of the volume and if adequate mixing of unsprayed for aerosols removed by sprays. No credit is compartments can be shown. taken for the removal of organic iodine by the spray system.

The SRP sets forth a maximum decontamination factor (DF) 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 time after decontamination. The SRP also states that the particulate iodine removal rate should be reduced by a factor of 10 when a DF of 50 is reached. The reduction in the removal rate is not required if the removal rate is based on the calculated time-dependent airborne aerosol mass. There is no specified maximum DF for aerosol removal by sprays. The maximum activity to be used in determining the DF is defined as the iodine activity in the columns labeled "Total" in Tables 1 and 2 of this guide multiplied by 0.05 for elemental iodine and by 0.95 for particulate iodine (i.e., aerosol treated as particulate in SRP methodology).

This document describes statistical formulations with differing levels of uncertainty. The removal rate constants selected for use in design basis calculations should be those that will maximize the dose consequences.

Attachment 3 Regulatory Guide 1.183 Compliance Table Page 20 of 41 3.4 Reduction in airborne radioactivity in the containment by in-containmentrecirculationfilter systems Conforms.

may be credited if these systems meet the guidance of Regulatory Guide 1.52 and Generic Letter 99-02 (Refs. A-5 and A-6). The filter media loading caused by the increased aerosol release The Reactor Building Cooling System associated with the revised source term should be addressed. recirculation flow rate is rated at 60, 270 +/- 10%

acfm. The minimum allowable system flow is 54,243 acfm. 54, 200 cfm is used in the RADTRAD analyses. The recirculation HEPA efficiency is assumed to be 90 percent for the removal of iodine particulates only. This is conservative since Regulatory-Guide 1.52 permits a 99 percent removal efficiency for particulates in accident dose evaluations.

3.5 Reduction in airborne radioactivity in the containment by suppression pool scrubbing in BWRs NA should generally not be credited. However, the staff may consider such reduction on an individual case basis. The evaluation should consider the relative timing of the blowdown and the fission product release from the fuel, the force driving the release through the pool, and the potential for any bypass of the suppression pool (Ref. 7). Analyses should consider iodine re-evolution if the suppression pool liquid pH is not maintained greater than 7.

3.6 Reduction in airborne radioactivity in the containment by retention in ice condensers, or other N/A engineering safety features not addressed above, should be evaluated on an individual case basis.

See Section 6.5.4 of the SRP (Ref. A- 1).

No credit is taken in this analysis for reduction of airborne radioactivity in the containment by retention in ice condensers, or other engineering safety features not addressed above.

3.7 The primary containment should be assumed to leak at the peak pressure Technical Specification Conforms leak rate for the first 24 hours2.777778e-4 days <br />0.00667 hours <br />3.968254e-5 weeks <br />9.132e-6 months <br />. For PWRs, the leak rate may be reduced after the first 24 hours2.777778e-4 days <br />0.00667 hours <br />3.968254e-5 weeks <br />9.132e-6 months <br /> to The primary containment peak pressure leak rate 50% of the Technical Specification leak rate. Leakage from sub-atmospheric containments is is defined as 0.2% by weight of containment air.

assumed to terminate when. the containment is brought to and maintained at a sub-atmospheric This leak rate is reduced to 0.1 % after the first condition as defined by Technical Specifications. 24 hours2.777778e-4 days <br />0.00667 hours <br />3.968254e-5 weeks <br />9.132e-6 months <br /> of the accident.

Attachment 3 Regulatory Guide 1.183 Compliance Table Page 21 of 41 3.8 If the primary containment is routinely purged during power operations, releases via the purge N/A system prior to containment isolation should be analyzed and the resulting doses summed with the postulated doses from other release paths. The purge release evaluation should assume that 100% of The primary containment is not routinely the radionuclide inventory in the reactor coolant system liquid is released to the containment at the purged during power operations.

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 For facilities with dual containment systems, the acceptable assumptions related to the transport, NA reduction, and release of radioactive material in and from the secondary containment or enclosure buildings are as follows:

4.1 Leakage from the primary containment should be considered to be collected, processed by NA engineered safety feature (ESF) filters, if any, and released to the environment via the secondary containment exhaust system during periods in which the secondary containment has a negative pressure as defined in Technical Specifications. Credit for an elevated release should be assumed only if the point of physical release is more than two and one-half times the height of any adjacent structure.

NA 4.2 Leakage from the primary containment is assumed to be released directly to the environment as a ground-level release during any period in which the secondary containment does not have a negative pressure as defined in Technical Specifications.

NA 4.3 The effect of high wind speeds on the ability of the secondary containment to maintain a negative pressure should be evaluated on an individual case basis. The wind speed to be assumed is the I-hour average value that is exceeded only 5% of the total number of hours in the data set. Ambient temperatures used in these assessments should be the 1-hour average value that is exceeded only 5%

or 95% of the total numbers of hours in the data set, whichever is conservative for the intended use (e.g., if high temperatures are limiting, use those exceeded only 5%).

Credit for dilution in the secondary containment may be allowed when adequate means to cause NA mixing can be demonstrated. Otherwise, the leakage from the primary containment should be assumed to be transported directly to exhaust systems without mixing. Credit for mixing, if found to be appropriate, should generally be limited to 50%. This evaluation should consider the magnitude of the containment leakage in relation to contiguous building volume or exhaust rate, the location of exhaust plenums relative to projected release locations, the recirculation ventilation systems, and internal walls and floors that impede stream flow between the release and the exhaust.

Attachment 3 Regulatory Guide 1.183 Compliance Table Page 22 of 41 NA 4.5 Primary containment leakage that bypasses the secondary containment should be evaluated at the bypass leak rate incorporated in the Technical Specifications. If the bypass leakage is through water, e.g., via a filled piping run that is maintained full, credit for retention of iodine and aerosols may be considered on a case-by-case basis. Similarly, deposition of aerosol radioactivity in gas-filled lines may be considered on a case-by-case basis.

4.6 Reduction in the amount of radioactive material released from the secondary containment because NA of ESF filter systems may be taken into account provided that these systems meet the guidance of Regulatory Guide 1.52 (Ref. A-5) and Generic Letter 99-02 (Ref. A-6).

ASSUMPTIONS ON ESF SYSTEM LEAKAGE 5 ESF systems that recirculate sump water outside of the primary containment are assumed to leak Conforms during their intended operation. This release source includes leakage through valve packing glands, This analysis utilized the ESF leakage pump shaft seals, flanged connections, and other similar components. This release source may also assumptions in Section 5 of RG 1.183.

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

5.1 With the exception of noble gases, all the fission products released from the fuel to the containment Conforms (as defined in Table 1 of this guide) should be assumed to instantaneously and homogeneously mix With the exception of noble gases, all the in the suppression pool (in BWRs) at the time of release from the core. In lieu of this deterministic approach, suitably conservative mechanistic models for the transport of airborne activity in containment to the sump water may be used. Note that many of the parameters that make spray and containment (as defined in Table 1 of deposition models conservative with regard to containment airborne leakage are non-conservativee homogeneously and with regard to the buildup of sump activity. and homogeneously mix in the primary Icontainment sump water for this analysis.

Attachment 3 Regulatory Guide 1.183 Compliance Table Page 23 of 41 5.2 The leakage should be taken as two times the sum of the simultaneous leakage from all components Conforms in the ESF recirculation systems above which the Technical Specifications, or licensee The VCSNS technical specifications do not commitments to Item Im.D. 1.1 of NUREG-0737 (Ref. A-8), would require declaring such systems provide a specific limit for operational leakage inoperable. The leakage should be assumed to start at the earliest time the recirculation flow occurs in these systems and end at the latest time the releases from these systems are terminated.

that is allowed within the recirculation loop.

Administrative limits, however, ensure that Consideration should also be given to design leakage through valves isolating ESF recirculation operational leakage is adequately controlled.

systems from tanks vented to atmosphere, e.g., emergency core cooling system (ECCS) pump miniflow return to the refueling water storage tank.

A post LOCA recirculation leakage of 12,000 cc/hr (4.238E-01 cfh or 7.063E-03 cfm) is used as input to the dose calculations. This is twice the operational limit that is used in plant procedures for system leakage assessments (GTP-006). In the event total recirculation loop leakage exceeds 6,000 cc/hr, a Condition Evaluation Report is generated to facilitate a licensing basis impact assessment and an operability determination.

Leakage through the RWST and NaOH Tank is neglected in this calculation based on plant procedures (EOPs) that require closure of the 20" RWST outlet valve (6700) and closure of the 3" NAOH outlet valve (3012) following the transition to CL recirculation. This results in 3 valve isolation and a minimum of 2 valve isolation in the long term with a single failure.

As outlined in Attachment 8, EOP changes will be made prior to AST implementation to support this analysis assumption.

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

Attachment 3 Regulatory Guide 1.183 Compliance Table Page 24 of 41 5.4 If the temperature of the leakage exceeds 212'F, the fraction of total iodine in the liquid that N/A becomes airborne should be assumed equal to the fraction of the leakage that flashes to vapor. This The temperature of the leakage does not exceed flash fraction, FF, should be determined using a constant enthalpy, h, process, based on the 212age maximum time-dependent temperature of the sump water circulating outside the containment:

FFh -h hf Where: hfj is the enthalpy of liquid at system design temperature and pressure; h2 is the enthalpy of liquid at saturation conditions (14.7 psia, 212'F); and hfg is the heat of vaporization at 212'F.

5.5 If the temperature of the leakage is less than 212'F or the calculated flash fraction is less than 10%, Conforms the amount of iodine that becomes airborne should be assumed to be 10% of the total iodine activity The temperature of the leakage does not exceed in the leaked fluid, unless a smaller amount can be justified based on the actual sump pH history and 212'F and a flash fraction of 10% was assumed area ventilation rates. for the iodine.

5.6 The radioiodine that is postulated to be available for release to the environment is assumed to be Conforms 97% elemental and 3% organic. Reduction inrelease activity by dilution or holdup within The radioiodine that is postulated to be buildings, or by ESF ventilation filtration systems, may be credited where applicable. Filter systems available for release to the environment was used in these applications should be evaluated against the guidance of Regulatory Guide 1.52 assumed to be 97% elemental and 3% organic.

(Ref. A-5) and Generic Let ter 99-02 (Ref. A-6). No reductions due to dilution, holdup, or by ESF ventilation filtration systems were assumed.

ASSUMPTIONS ON MAIN STEAM ISOLATION VALVE LEAKAGE IN BWRS 6 For BWRs, the main steam isolation valves (MSIVs) have design leakage that may result in a NA radioactivity release. The radiological consequences from postulated MSIV leakage should be analyzed and combined with consequences postulated for other fission product release paths to determine the total calculated radiological consequences from the LOCA. The following assumptions are acceptable for evaluating the consequences of MSIV leakage.

NA 6.1 For the purpose of this analysis, the activity available for release via MSIV leakage should be assumed to be that activity determined to be in the drywell for evaluating containment leakage (see Regulatory Position 3). No credit should be assumed for activity reduction by the steam separators or by iodine partitioning in the reactor vessel.

Attachment 3 Regulatory Guide 1.183 Compliance Table Page 25 of 41 NA 6.2 All the MSIVs should be assumed to leak at the maximum leak rate above which the Technical Specifications would require declaring the MSIVs inoperable. The leakage should be assumed to continue for the duration of the accident. Postulated leakage may be reduced after the first 24 hours2.777778e-4 days <br />0.00667 hours <br />3.968254e-5 weeks <br />9.132e-6 months <br />, if supported by site-specific analyses, to a value not less than 50% of the maximum leak rate.

NA 6.3 Reduction of the amount of released radioactivity by deposition and plateout on steam system piping upstream of the outboard MSIVs may be credited, but the amount of reduction in concentration allowed will be evaluated on an individual case basis. Generally, the model should be based on the assumption of well-mixed volumes, but other models such as slug flow may be used if justified.

NA 6.4 In the absence of collection and treatment of releases by ESFs such as the MSIV leakage control system, or as described in paragraph 6.5 below, the MSIV leakage should be assumed to be released to the environment as an unprocessed, ground- level release. Holdup and dilution in the Turbine Building should not be assumed.

NA 6.5 A reduction in MSIV releases that is due to holdup and deposition in main steam piping downstream of the MSIVs and in the main condenser, including the treatment of air ejector effluent by offgas systems, may be credited if the components and piping systems used in the release path are capable of performing their safety function during and following a safe shutdown earthquake (SSE). The amount of reduction allowed will be evaluated on an individual case basis. References A-9 and A-10 provide guidance on acceptable models.

ASSUMPTION ON CONTAINMENT PURGING 7 The radiological consequences from post-LOCA primary containment purging as a combustible gas N/A or pressure control measure should be analyzed. If the installed containment purging capabilities are maintained for purposes of severe accident management and are not credited in any design basis analysis, radiological consequences need not be evaluated. If the primary containment purging is required within 30 days of the LOCA, the results of this analysis should be combined with consequences postulated for other fission product release paths to determine the total calculated radiological consequences from the LOCA. Reduction in the amount of radioactive material released via ESF filter systems may be taken into account provided that these systems meet the guidance in Regulatory Guide 1.52 (Ref. A-5) and Generic Letter 99-02 (Ref. A-6).

Attachment 3 Regulatory Guide 1.183 Compliance Table Page 26 of 41 Appendix ASSUMPTIONS FOR EVALUATING THE RADIOLOGICAL CONSEQUENCES OF A Conforms B FUEL HANDLING ACCIDENT An analysis was performed utilizing the This appendix provides assumptions acceptable to the staff for evaluating the radiological guidance of Appendix B and appropriate consequences of a fuel handling accident at light water reactors. These assumptions supplement the sections in the main body of RG 1.183 to guidance provided in the main body of this guide. evaluate a fuel and an equipment accident.

SOURCE TERM Conforms Acceptable assumptions regarding core inventory and the release of radionuclides from the fuel are Assumptions regarding core inventory and the provided in Regulatory Position 3 of this guide. The following assumptions also apply. release of radionuclides from the fuel are taken from Regulatory Position 3 of RG 1.183.

1.1 The number of fuel rods damaged during the accident should be based on a conservative analysis Conforms that considers the most limiting case. This analysis should consider parameters such as the weight For the postulated FHA inside Containment of the dropped heavy load or the weight of a dropped fuel assembly (plus any attached handling and inside the FHB, a total of 314 pins are grapples), the height of the drop, and the compression, torsion, and shear stresses on the irradiated assumed to be damaged as a result of this fuel rods. Damage to adjacent fuel assemblies, if applicable shouldll (e.g., events over the reactor vessel),

considered.spnt event. All 264 pins in the dropped spent fuel 64ben ue should be considered. assembly and 50 pins in the impacted assembly are assumed to be rupture. These assumptions are consistent with the CLB analysis.

1.2 The fission product release from the breached fuel is based on Regulatory Position 3.2 of this guide Conforms and the estimate of the number of fuel rods breached. All the gap activity in the damaged rods is The fission product release from the breached assumed to be instantaneously released. Radionuclides that should be considered include xenons, fuel is based on Regulatory Position 3.2 of kryptons, halogens, cesiums, and rubidiums. RG 1.183 (see response to RG 1.183, Section 3.2 of Attachment 3) and the estimate of the number of fuel rods breached. All the gap activity in the damaged rods is assumed to be instantaneously released. Radionuclides that were considered include xenons, kryptons, halogens, cesiums, and rubidiums. Please note, RG 1.183, Appendix B, Section 3, the pool DF for particulates which includes cesiums, and rubidiums is infinite. Therefore, they are neglected from further consideration in the analysis.

Attachment 3 Regulatory Guide 1.183 Compliance Table Page 27 of 41 1.3 The chemical form of radioiodine released from the fuel to the spent fuel pool should be assumed to Conforms be 95% cesium iodide (CsI), 4.85 percent elemental iodine, and 0.15 percent organic iodide. The All iodine released to the spent fuel pool CsI released from the fuel is assumed to completely dissociate in the pool water. Because of the dissociates and re-evolves as elemental iodine low pH of the pool water, the iodine re-evolves as elemental iodine. This is assumed to occur instantaneously. The NRC staff will consider, on a case-by-case basis, justifiable mechanistic instantaneously.

treatment of the iodine release from the pool.

2 WATER DEPTH Conforms The minimum water depth over the reactor core If the depth of water above the damaged fuel is 23 feet or greater, the decontamination factors for when handling fuel and over the spent fuel in the elemental and organic species are 500 and 1, respectively, giving an overall effective the FHB is 23 feet decontamination factor of 200 (i.e., 99.5% of the total 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. If the depth of water is not 23 feet, the decontamination factor will have to be determined on a case-by-case method (Ref. B-1).

NOBLE GASES Conform The retention of noble gases in the water in the fuel pool or reactor cavity is negligible (i.e., Noble gas DF =

decontamination factor of 1). Particulate radionuclides are assumed to be retained by the water in Particulate radionuclides are assumed to be the fuel pool or reactor cavity (i.e., infinite decontamination factor). retained by the water in the fuel pool or reactor cavity (i.e., infinite decontamination factor).

FUEL HANDLING ACCIDENTS WITHIN THE FUEL BUILDING Conforms postulated to occur within the fuel building, the following assumptions For fuel handling accidentsare The fuel handling accident was evaluated ccepableto he NC stff.within the Fuel Handling Building outside are acceptable to the NRC staff. containment.

The radioactive material that escapes from the fuel pool to the fuel building is assumed to be Conforms released to the environment over a 2-hour time period. The radioactive material that escapes from the fuel pool to the Fuel Handling Building is assumed to be released to the environment over a 2-hour time period.

Attachment 3 Regulatory Guide 1.183 Compliance Table Page 28 of 41 4.2 A reduction in the amount of radioactivematerialreleased from the fuel pool by engineered safety NA feature (ESF) filter systems may be taken into account provided these systems meet the guidance of Consistent with the proposed change to VCSNS Regulatory Guide 1.52 and Generic Letter 99-02 (Refs. B-2, B-3). Delays in radiation detection, Technical Specification 3/4.7.11 SPENT FUEL actuation of the ESF filtration system, or diversion of ventilation flow to the ESF filtration system POOL VENTILATION SYSTEM, no credit is should be determined and accounted for in the radioactivity release analyses taken for filtration in the FHA analysis Note: These analyses should consider the time for the radioactivity concentration to reach levels corresponding to the monitor setpoint, instrument line sampling time, detector response time, diversion damper alignment time, and filter system actuation, as applicable.

4.3 The radioactivity release from the fuel pool should be assumed to be drawn into the ESF filtration Conforms system without mixing or dilution in the fuel building. If mixing can be demonstrated, credit for mixing and dilution may be considered on a case-by-case basis. This evaluation should consider the No credit is taken for mixing or dilution in the magnitude of the building volume and exhaust rate, the potential for bypass to the environment, the fuel building.

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

5 FUEL HANDLING ACCIDENTS WITHIN CONTAINMENT Conforms For fuel handling accidents postulated to occur within the containment, the following assumptions The fuel handling accident was evaluated are acceptable to the NRC staff. within the Containment Building.

5.1 If the containment is isolated during fuel handling operations, no radiological consequences need to NA be analyzed. Consistent with the proposed change to Note: Containment isolation does not imply containment integrity as defined by Technical VCSNS REACTOR Technical Specification 3/4.9.4 BUILDING PENETRATIONS, no Specifications for non-shutdown modes. The term isolation is used here collectively to encompass credit is taken for containment isolation in the both containment integrity and containment closure, typically in place during shutdown periods. To FHA analysis.

FHA analysis.

be credited in the analysis, the appropriate form of isolation should be addressed in Technical Specifications.

5.2 If the containment is open during fuel handling operations, but designed to automatically isolate in NA the event of a fuel handling accident, the release duration should be based on delays in radiation detection and completion of containment isolation. If it can be shown that containment isolation No credit is taken for containment isolation in occurs before radioactivity is released to the environment, no radiological consequences need to be the FHA analysis.

analyzed.

Attachment 3 Regulatory Guide 1.183 Compliance Table Page 29 of 41 5.3 If the containment is open during fuel handling operations (e.g., personnel air lock or equipment Conforms hatch is open), the radioactive material that escapes from the reactor cavity pool to the containment The containment is assumed open during the is released to the environment over a 2-hour time period. postulated FHA and the activity is released Note: The staff will generally require that technical specifications allowing such operations include over a 2-hour period.

administrative controls to close the airlock, hatch, or open penetrations within 30 minutes. Such administrative controls will generally require that a dedicated individual be present, with necessary equipment available, to restore containment closure should a fuel handling accident occur.

Radiological analyses should generally not credit this manual isolation.

5.4 A reduction in the amount of radioactive material released from the containment by ESF filter NA systems may be taken into account provided that these systems meet the guidance of Regulatory No credit is taken for reduction in the amount Guide 1.52 and Generic Letter 99-02 (Refs. B-2 and B-3). Delays in radiation detection, actuation of radioactive material released from the of the ESF filtration system, or diversion of ventilation flow to the ESF filtration system should be containment by ESF filter systems.

determined and accounted for in the radioactivity release analyses.

5.5 Credit for dilution or mixing of the activity released from the reactor cavity by natural or forced NA convection inside the containment may be considered on a case-by-case basis. Such credit is generally limited to 50% of the containment free volume. This evaluation should consider the No credit is taken for dilution or mixing in magnitude of the containment volume and exhaust rate, the potential for bypass to the environment, containment prior to release of activity to the the location of exhaust plenums relative to the surface of the reactor cavity, recirculation ventilation environment.

systems, and internal walls and floors that impede stream flow between the surface of the reactor cavity and the exhaust plenums.

Attachment 3 Regulatory Guide 1.183 Compliance Table Page 30 of 41 ASSUMPTIONS FOR EVALUATING THE RADIOLOGICAL CONSEQUENCES OF A Appendix PWR MAIN STEAM LINE BREAK ACCIDENT E

This appendix provides assumptions acceptable to the NRC staff for evaluating the radiological consequences of a main steam line break accident at PWR light water reactors. These assumptions supplement the guidance provided in the main body of this guide.

SOURCE TERMS Conforms Assumptions acceptable to the NRC staff regarding core inventory and the release of radionuclides No fuel damage is postulated to occur during from the fuel are provided in Regulatory Position 3 of this regulatory guide. The release from the the MSLB.

breached fuel is based on Regulatory Position 3.2 of this guide and the estimate of the number of fuel rods breached. The fuel damage estimate should assume that the highest worth control rod is stuck at its fully withdrawn position.

2 If no or minimal fuel damage is postulated for the limiting event, the activity released should be the Conforms maximum coolant activity allowed by the technical specifications. Two cases of iodine spiking should be assumed. There is no fuel damage for the VCSNS MSLB. Two cases of iodine spiking are used in the analysis.

2.1 A reactor transient has occurred prior to the postulated main steam line break (MSLB) Conforms and has raised the primary coolant iodine concentration to the maximum value (typically 60 The VCSNS analysis is based on the maximum p Ci/gm DE 1-13 1) permitted by the technical specifications (i.e., a preaccident iodine spike case). value of 60 pCiSgm DE 1-131 permitted by the technical specifications.

2.2 The primary system transient associated with the MSLB causes an iodine spike in the Conforms primary system. The increase in primary coolant iodine concentration is estimated using a spiking model The VCSNS analysis uses a spiking model that that assumes that the iodine release rate from the fuel rods to the primary coolant (expressed in curies assumes that the iodine release rate from the fuel per unit time) increases to a value 500 times greater than the release rate corresponding to the iodine rods to the primary coolant (expressed in curies concentration at the equilibrium value (typically 1.0 ViCi/gm DE 1-131) specified in technical per unit time) increases to a value 500 times specifications (i.e., concurrent iodine spike case). A concurrent iodine spike need not be perun tie incease t e 500 ti 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 />. the iodine concentration at the equilibrium Shorter spike durations may be considered on a case-by-case basis if it can be shown that the activity value (typically 1.0 pCi/gm DE 1-131) specified released by the 8- hour spike exceeds that available for release from the fuel gap of all fuel pins. in technical specifications. The assumed iodine spike duration is 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 should be assumed to be released instantaneously and Conforms homogeneously through the primary coolant.

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

Attachment 3 Regulatory Guide 1.183 Compliance Table Page 31 of 41 The chemical form of radioiodine released from the fuel should be assumed to be 95% cesium iodide Conforms (CsI), 4.85 percent elemental iodine, and 0.15 percent organic iodide. Iodine releases from the steam generators to the environment should be assumed to be 97% elemental and 3% organic. These Iodine chemical form is in accordance with this fractions apply to iodine released as a result of fuel damage and to iodine released during normal guidance, 97% elemental and 3% organic.

operations, including iodine spiking.

TRANSPORT 5 Conforms Assumptions acceptable to the NRC staff related to the transport, reduction, and release of The noble gases are released to the radioactive material to the environment are as follows. enonme gases environment without reduto the reduction or mitigation.

5.1 For facilities that have not implemented alternative repair criteria (see Ref. E- 1, DG- Conforms 1074), the primary-to-secondary 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 Per Technical Specification Bases 3/4.4.5, a apportioned between affected and unaffected doseis steam generators in such a manner that the calculated mximied.steam total primary to secondary generators of 1 gpm leakage is used infortheall three dose is maximized. evaluation of design basis accidents. Prior to the event this leakage is assumed to be distributed throughout the three steam generators. Recognizing that an extended plant cooldown may be required under natural circulation conditions, the MSLB is conservatively analyzed using 24 hours2.777778e-4 days <br />0.00667 hours <br />3.968254e-5 weeks <br />9.132e-6 months <br /> for the cooldown time. The activity associated with the I gpm leak is assumed to be'released to the environment via the faulted steam generator at a rate of 0.35 gpm for the 24 hour2.777778e-4 days <br />0.00667 hours <br />3.968254e-5 weeks <br />9.132e-6 months <br /> duration of the event with no credit taken for any reduction or mitigation, i.e. a partition factor of 1.0. This is conservative in that the actual maximum value allowed by Technical Specification 3.4.6.2.c for any one SG is 150 gpd (- 0.104 gpm). The remaining 0.65 gpm is released to the environment via the two intact steam generators for the 24 hour2.777778e-4 days <br />0.00667 hours <br />3.968254e-5 weeks <br />9.132e-6 months <br /> duration of the event crediting a partition factor of 100.

5.2 The density used in converting volumetric leak rates (e.g., gpm) to mass leak rates (e.g., lbm/hr) should Conforms be consistent with the basis of the parameter being converted. The ARC leak rate correlations are generally based on the collection of cooled liquid. Surveillance tests and facility instrumentation used to show compliance with leak rate technical specifications are typically based on cooled liquid. In most cases, the density should be assumed to be 1.0 gm/cc (62.4 lbm/ft3).

Attachment 3 Regulatory Guide 1.183 Compliance Table Page 32 of 41 5.3 The primary-to-secondary leakage should be assumed to continue until the primary system Conforms pressure is less than the secondary system pressure, or until the temperature of the leakage is less than I00°C (2 12'F). The release of radioactivity from unaffected steam generators should be assumed to See item 5.1.

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

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

The noble gases are released to the environment without reduction or mitigation.

5.5 and particulate releases from The transport model described in this section should be utilized for iodine below: Conforms the steam generators. This model is shown in Figure E- I and summarized See item 5.1.

Figue E- I TransportModel Steam Spaoe Firntfary Bulk Watere 5.5.1 A portion of the primary-to-secondary leakage will flash to vapor, based onthe thermodynamic Conforms conditions in the reactor and secondary coolant.

• During periods of steam generator dryout, all of the primary-to-secondary leakage is assumed to See item 5.1.

flash to vapor and be released to the environment with no mitigation.

• With regard to the unaffected steam generators used for plant cooldown, the primary-to-secondary leakage can be assumed to mix with the secondary water without flashing during periods of total tube submergence.

5.5.2 The leakage that immediately flashes to vapor will rise through the bulk water of the steam Conforms generator and enter the steam space. Credit may be taken for scrubbing in the generator, using the models in NUREG-0409, "Iodine Behavior in a PWR Cooling System Following a Postulated Steam See item 5.1.

Generator Tube Rupture Accident" (Ref. E-2), during periods of total submergence of the tubes.

5.5.3 The leakage that does not immediately flash is assumed to mix with the bulk water. Conforms See item 5.1.

Attachment 3 Regulatory Guide 1.183 Compliance Table Page 33 of 41 5.5.4 The radioactivity in the bulk water is assumed to become vapor at a rate that is the function of the Conforms steaming rate and the partition coefficient. A partition coefficient for elemental iodine of 100 may be assumed. The retention of particulate radionuclides in the steam generators is limited by the See item 5.1.

moisture carryover from the steam generators.

5.6 Operating experience and analyses have shown that for some steam generator designs, tube uncovery NA may occur for a 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 evaluated.

Attachment 3 Regulatory Guide 1.183 Compliance Table Page 34 of 41 ASSUMPTIONS FOR EVALUATING THE RADIOLOGICAL CONSEQUENCES OF A Appendix PWR STEAM GENERATOR TUBE RUPTURE ACCIDENT F

This appendix provides assumptions acceptable to the NRC staff for evaluating the radiological consequences of a steam generator tube rupture accident at PWR light-water reactors. These assumptions supplement the guidance provided in the main body of this guide.

SOURCE TERMS Conforms Assumptions acceptable to the NRC staff regarding core inventory and the release of radionuclides No fuel damage is postulated to occur during from the fuel are in Regulatory Position 3 of this 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.

If no or minimal fuel damage is postulated for the limiting event, the activity released should be the Conforms maximum coolant activity allowed by technical specification. Two cases of iodine spiking should be assumed. There is no fuel damage for the VCSNS SGTR.

Two cases of iodine spiking are used in the The activity assumed in the analysis should be based on the activity associated with the projected fuel analysis.

damage or the maximum technical specification values, whichever maximizes the radiological consequences. In determining dose equivalent 1-131 (DE 1131), only the radioiodine associated with normal operations or iodine spikes should be included. Activity from projected fuel damage should not be included.

2.1 A reactor transient has occurred prior to the postulated steam generator tube rupture (SGTR) and has Conforms raised the primary coolant iodine concentration to the maximum value (typically 60 VCi/gm DE 1-131) permitted by the technical specifications (i.e., a preaccident iodine spike case). The VCSNS analysis is based on the maximum value of 60 pCi/gm DE 1-131 permitted by the technical specifications.

2.2 The primary system transient associated with the SGTR causes an iodine spike in the primary system. Conforms The increase in primary coolant iodine concentration is estimated using a spiking model that assumes that the iodine release rate from the fuel rods to the primary coolant (expressed in curies per The VCSNS analysis uses a spiking model that unit time) increases to a value 335 times greater than the release rate corresponding to the iodine assumes that the iodine release rate from the fuel concentration at the equilibrium value (typically 1.0 pCi/gm DE 1-131) specified in technical rods to the primary coolant (expressed in curies specifications (i.e., concurrent iodine spike case). A concurrent iodine spike need not be per unit time) increases to a value 335 times 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 />. greater than the release rate corresponding to Shorter spike durations may be considered on a case-by-case basis if it can be shown that the activity the iodine concentration at the equilibrium released by the 8- hour spike exceeds that available for release from the fuel gap of all fuel pins. value (typically 1.0 pCi/gm DE 1-131) specified in technical specifications. The assumed iodine spike duration is 8 hours9.259259e-5 days <br />0.00222 hours <br />1.322751e-5 weeks <br />3.044e-6 months <br />.

Attachment 3 Regulatory Guide 1.183 Compliance Table Page 35 of 41 The activity released from the fuel, if any, should be assumed to be released instantaneously and Conforms homogeneously through the primary coolant.

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

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

Iodine chemical form is in accordance with this guidance, 97% elemental and 3% organic.

TRANSPORT 5 Conforms Assumptions acceptable to the NRC staff related to the transport, reduction, and release of The VCSNS analysis is in accordance with this radioactive material to the environment are as follows: guidance.

5.1 The primary-to-secondary leak rate in the steam generators should be assumed to be the leak rate Conforms limiting condition for operation specified in the technical specifications. The leakage should be apportioned between affected and unaffected steam generators in such a manner that the calculated dose is maximized. The SGTR analysis accounts for a bounding primary-to-secondary leakage rate equal to 1 gpm and the leakage rate associated with a double-ended rupture of a single tube. Leakage through the ruptured tube is the dominate contributor to dose releases. Since contaminated fluid in the ruptured steam generator is only briefly released to the atmosphere as steam via the main steam safety valves, the entire 1 gpm primary-to-secondary leakage is conservatively assumed to occur in the intact steam generators where it can be released during the subsequent cooldown of the plant. Per Technical Specification 3.4.6.2.c, the maximum value in any one steam generator is limited to 150 gpd (- 0.104 gpm). This release is therefore conservative in that the actual maximum value allowed by TS for the two intact SGs would be -0.208 gpm.

Attachment 3 Regulatory Guide 1.183 Compliance Table Page 36 of 41 5.2 The density used in converting volumetric leak rates (e.g., gpm) to mass leak rates (e.g., lbm/hr) should Conforms be consistent with the basis of surveillance tests used to show compliance with leak rate technical specifications. These tests are typically based on cool liquid. Facility instrumentation used to 3 determine leakage is typically located on lines containing cool liquids. In most cases, the density should be assumed to be 1.0 gm/lcc (62.4 Ibm/ft3 ).

5.3 The primary-to-secondary leakage should be assumed to continue until the primary system Conforms pressure is less than the secondary system pressure, or until the temperature of the leakage is less than 100'C (2 12'F). The release of radioactivity from the unaffected steam generators should be The activity associated with the I gpm assumed to continue until shutdown cooling is in operation and releases from the steam generators primary-to-secondary leak is conservatively have been terminated. assumed to be released to the environment via the intact steam generator for the 24 hour2.777778e-4 days <br />0.00667 hours <br />3.968254e-5 weeks <br />9.132e-6 months <br /> duration of the event.

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

A coincident loss of offsite power is assumed.

5.5 All noble gas radionuclides released from the primary system are assumed to be released to the Conforms environment without reduction or mitigation.

The noble gases are released to the environment without reduction or mitigation.

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

The transport model described in Regulatory Positions 5.5 and 5.6 of Appendix E for iodine and particulates is considered as appropriate in the SGTR.

Attachment 3 Regulatory Guide 1.183 Compliance Table Page 37 of 41 ASSUMPTIONS FOR EVALUATING THE RADIOLOGICAL CONSEQUENCES OF A Appendix PWR LOCKED ROTOR ACCIDENT G

This appendix provides assumptions acceptable to the NRC 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 body of this guide.

SOURCE TERMS Conforms Assumptions acceptable to the NRC staff regarding core inventory and the release of radionuclides from See the responses to RG 1.183, Sections 3.1 the fuel are in Regulatory Position 3 of this regulatory guide. The release from the breached fuel is and 3.2 of Attachment 3.

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

2 If no fuel damage is postulated for the limiting event, a radiological analysis is not required as the consequences of this event are bounded by the consequences projected for the main steam line break outside containment. Analysis uses 15% fuel damage in the core.

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

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

The chemical form of radioiodine released from the fuel should be assumed to be 95% cesium iodide Conforms (CsI), 4.85 percent elemental iodine, and 0.15 percent organic iodide. Iodine releases from the steam generators to the environment should be assumed to be 97% elemental and 3% organic. These fractions apply to iodine released as a result of fuel damage and to iodine released during normal guine %eem en an3organic.

operations, including iodine spiking. guidance, 97% elemental and 3% organic.

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

5.1 The primary-to-secondary limiting leak rate condition for operation in theinsteam specified generators the technical should be assumed specifications. to beshould The leakage the leak-rate-be Conforms apportioned between the steam generators in such a manner that the calculated dose is maximized. Per the VCSNS Technical Specification Bases, for use in the accident analysis the total leakage for all three steam generators is I gpm.

5.2 The density used in converting volumetric leak rates (e.g., gpm) to mass leak rates (e.g., lbm/hr) should Conforms be consistent with the basis of surveillance tests used to show compliance with leak rate technical specifications. These tests are typically based on cool liquid. The density used is 1.0 gm/cc (62.4 lbm/ft3).

Facility instrumentation used to determine leakage is typically located on lines containing cool liquids. In most cases, the density should be assumed to be 1.0 gm/cc (62.4 lbm/ft3).

Attachment 3 Regulatory Guide 1.183 Compliance Table Page 38 of 41 5.3 The primary-to-secondary leakage should be assumed to continue until the primary Conforms system pressure is less than the secondary system pressure, or until the temperature of the leakage is less than 100'C (2 12'F). The release of radioactivity should be assumed to continue until shutdown The activity is released for a duration of 24 cooling is in operation and releases from the steam generators have been terminated. hours.

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

A coincident loss of offsite power is assumed.

All noble gas radionuclides released from the primary system are assumed to be released to the Conforms environment without reduction or mitigation.

The noble gases are released to the environment without reduction or mitigation.

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

The transport model described in Regulatory Positions 5.5 and 5.6 of Appendix E for iodine and particulates is considered as appropriate in the RCP LRA analysis.

Attachment 3 Regulatory Guide 1.183 Compliance Table Page 39 of 41 ASSUMPTIONS FOR EVALUATING THE RADIOLOGICAL CONSEQUENCES OF A Appendix PWR ROD EJECTION ACCIDENT H

This appendix provides assumptions acceptable to the NRC staff for evaluating the radiological consequences of a rod ejection accident at PWR light-water reactors. These assumptions supplement the guidance provided in the main body of this guide.

SOURCETERM Conforms Assumptions acceptable to the NRC staff regarding core inventory are in Regulatory Position 3 of See the responses to RG 1.183 Sections 3.1 and this guide. For the rod ejection accident, the release from the breached fuel is based on the estimate of 3.2 of Attachment 3.

the number of fuel rods breached and the assumption that 10% of the core inventory of the noble gases and iodines is in the fuel gap. The release attributed to fuel melting is based on the fraction of the fuel that reaches or exceeds the initiation temperature for fuel melting and the assumption that 100% of the noble gases and 25% of the iodines contained in that fraction are available for release from containment. For the secondary system release pathway, 100% of the noble gases and 50% of the iodines in that fraction are released to the reactor coolant.

2 If no fuel damage is postulated for the limiting event, a radiological analysis is not required as the consequences of this event are bounded by the consequences projected for the loss-of-coolant accident (LOCA), main steam line break, and steam generator tube rupture. Analysis uses 10% fuel damage and 0.25% fuel melt in the core.

Two release cases are to be considered. In the first, 100% of the activity released from the fuel should Conforms be assumed to be released instantaneously and homogeneously through the containment atmosphere.

In the second, 100% of the activity released from the fuel should be assumed to be completely Two release cases are considered. 100% of the dissolved in the primary coolant and available for release to the secondary system. activity released from the fuel is released instantaneously and homogeneously through the containment atmosphere. 100% of the activity released from the fuel is assumed to be completely dissolved in the primary coolant and available for release to the secondary system.

The chemical form of radioiodine released to the containment atmosphere should be assumed to be 95% cesium iodide (CsI), 4.85% elemental iodine, and 0.15% organic iodide. If containment sprays do not actuate or are terminated prior to accumulating sump water, or if the containment sump pH is not The chemical form of radioiodine released to the controlled at values of 7 or greater, the iodine species should be evaluated on an individual case containment atmosphere is assumed to be 95%

basis. Evaluations of pH should consider the effect of acids created during the rod ejection accident cesium iodide (CsI), 4.85% elemental iodine, event, e.g., pyrolysis and radiolysis products. With the exception of elemental and organic iodine and and 0.15% organic iodide.

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

Attachment 3 Regulatory Guide 1.183 Compliance Table Page 40 of 41 Iodine releases from the steam generators to the environment should be assumed to be Conforms 97% elemental and 3% organic.

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

6 TRANSPORT FROM CONTAINMENT 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 leakage from the containment that is due to natural deposition, containment sprays, recirculating filter systems, dual Conforms containments, or other engineered safety features may be taken into account. Refer to Appendix A to Credit is taken for natural deposition removal this guide for guidance on acceptable methods and assumptions for evaluating these mechanisms. of aerosols using the RADTRAD 10% Power's model in this analysis. No credit is taken for containment sprays or the containment reactor building cooling unit recirculating filter system.

6.2 The containment should be assumed to leak at the leak rate incorporated in the technical specifications at peak accident pressure for the first 24 hours2.777778e-4 days <br />0.00667 hours <br />3.968254e-5 weeks <br />9.132e-6 months <br />, and at 50% of this leak rate for the Conforms remaining duration of the accident. Peak accident pressure is the maximum pressure defined in the technical specifications for containment leak testing. Leakage from subatmospheric The primary containment peak pressure leak containments is assumed to be terminated when the containment is brought to a subatmospheric rate is definedair.as This containment 0.2% leak by weight rate is of reduced to condition as defined in technical specifications. 0.1% after the first 24 hours2.777778e-4 days <br />0.00667 hours <br />3.968254e-5 weeks <br />9.132e-6 months <br /> of the accident.

TRANSPORT FROM SECONDARY SYSTEM Conforms 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 A leak rate equivalent to the primary-to-secondary leak rate limiting condition for operation specified Conforms in the technical specifications should be assumed to exist until shutdown cooling is in operation and releases from the steam generators have been terminated. Per the VCSNS Technical Specification Bases, for use in accident analyses the total leakage for all three steam generators is 1 gpm. The activity is released for a duration of 24 hours2.777778e-4 days <br />0.00667 hours <br />3.968254e-5 weeks <br />9.132e-6 months <br />.

7.2 The density used in converting volumetric leak rates (e.g., gpm) to mass leak rates (e.g., Conforms lbm/hr) should be consistent with the basis of surveillance tests used to show compliance with leak rate technical specifications. These tests typically are based on cooled liquid. The facility's instrumentation used to determine leakage typically is located on lines containing cool liquids. In The density used is 1.0 gm/cc (62.4 lbm/ft').

most cases, the density should be assumed to be 1.0 gm/cc (62.4 lbm/ft ). 3 II

Attachment 3 Regulatory Guide 1.183 Compliance Table Page 41 of 41 All noble gas radionuclides released to the secondary system are assumed to be released Conforms to the environment without reduction or mitigation.

The noble gases are released to the environment without reduction or mitigation.

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

The transport model described in Regulatory Positions 5.5 and 5.6 of Appendix E for iodine and particulates is considered as appropriate in the CREA analysis.

Attachment 4 Regulatory Guide 1.194 Compliance Table

Attachment 4 Regulatory Guide 1.194 Compliance Table Page 1 of 33 Regulatory Guide 1.194 Compliance Regulatory Guide 1.194 Sections 3 through 7 and Table A-2 provide methodologies and assumptions that are acceptable to the NRC staff related to atmospheric relative concentrations for Control Room radiological habitability assessments at nuclear power plants. Compliance with Regulatory Guide 1.194 positions are discussed below:

Please Note: The information provided in this table is based on the calculations provided in Attachment 10.

RG 1.194 Regulatory Guide 1.194 Position Basis of Compliance Section

3. CALCULATION OF x/Q USING ARCON96 Conforms The ARCON96 code is maintained under a Twe A lity asne program tha This section addresses the use of the ARCON96 code for calculating X/Q values for design basis Control Room radiological habitability assessments. The ARCON96 code should be software quality assurance program that obtained and maintained under an appropriate software quality assurance program that Part 50 and applicable industry consensus standards.

complies with the applicable criteria of Appendix B, "Quality Assurance Criteria for standards.

Nuclear Power Plants and Fuel Reprocessing Plants," to 10 CFR Part 50 and applicable industry consensus standards to which the licensee has committed.

3.1 Meteorological Data Input Conforms The meteorological data needed for x/Q calculations include wind speed, wind direction, The meteorological -data includes wind and a measure of atmospheric stability. These data should be obtained from an onsite speed, wind direction, and a measure of meteorological measurement program based on the guidance of Safety Guide 23, "Onsite atmospheric stability. These data were Meteorological Programs" (Ref. 12), that includes quality assurance provisions consistent measuem program base onte with Appendix B to 10 CFR Part 50. The meteorological data set used in these meance o graGase m onth e assessments should represent hourly averages as defined in Safety Guide 23. Data should quality assurance provisions consistent with be representative of the overall site conditions and be free from local effects such as Alit a po consistenttwith building and cooling tower wakes, brush and vegetation, or terrain. Collected data should Appendix B to 10 CFR Part 50.

be reviewed to identify instrumentation problems and missing or anomalous observations (see Ref. 13). The size of the data set used in the X/Q assessments should be sufficiently

Attachment 4 Regulatory Guide 1.194 Compliance Table Page 2 of 33 large such that it is representative of long-term meteorological trends at the site. The NRC The meteorological data consists of five staff considers 5 years of hourly observations to be representative of long-term trends at years of hourly data, covering the years most sites. With sufficient justification of its representativeness, however, the minimum meteorological data set is one complete year (including all four seasons) of hourly from 2002 to 2006. Each record of the hourly data contains a location identifier, observations.

Julian day (1-366), hour (0 to 23), low-level direction, low-level speed, stability class Wind direction should be expressed as the direction from which the wind is blowing (I=A to 7=G), upper level direction, and (i.e., the upwind direction from the center of the site) referenced from true north.

upper level speed. Wind speeds are entered in tenths of a reporting unit with no Atmospheric stability should be determined by the vertical temperature difference (AT) decimal. Wind speed is entered in units of measured over the difference in height appropriate for the projected release height meters/second.

(including plume rise as applicable). A table of AT values in units of degrees Centigrade per 100 meters (0 C/100m) versus stability class is given in Safety Guide 23 (Ref. 12). If Wind directions are from 1 to 360 in other well-documented methodologies are used to estimate atmospheric stability (with degrees.

appropriate justification), the models described in this guide may require modification. A Atmospheric stability was determined by well-documented methodology is one that is substantiated by diffusion data for conditions the vertical temperature difference (AT) similar to those at the nuclear power plant site involved. measured over the difference in height appropriate for the projected release height Appendix A provides information on the structure and content of the meteorological data per Safety Guide 23.

set and input parameters used by the ARCON96 code.

3.2 Determination of Release Point (Source) Characteristics Conforms A total of 10 potential release points were A 95th-percentile X/Q value should be determined for each identified source-receptor evaluated, based on distance, direction, combination. However, it may be possible to identify bounding combinations in order to release mode, and height of the various reduce the needed calculational effort. In determining the bounding combinations, it will release points to the environment in relation be necessary to consider the distance, direction, release mode, and height of the various to the Control Room intake.

release points to the environment in relation to the various Control Room intakes.

Additional parameters, such as those used in establishing plume rise, may need to be considered in determining the bounding combination.

For cases involving two or more release pathways associated with a single release source, a There were no cases crediting two or more calculated composite value of X/Q may be considered on a case-by-case basis if the licensee release pathways associated with a single can demonstrate an acceptable modeling approach and justify the conservatism of any release source. The most limiting release assumed weighting factors. path was always used.

Attachment 4 Regulatory Guide 1.194 Compliance Table Page 3 of 33 Changes in associated parameters that could occur as a result of differences between normal operation and accident conditions, differences between accidents, differences that occur over the duration of the accident, single failure considerations, and considerations of loss of offsite power, consistent with accident sequences and descriptions, must all be considered in the characterization of the release points.

The ARCON96 code provides options that allow an analyst to model ground-level, elevated stack, and vent-point source releases. In addition, the analyst can model diffuse area sources as a sub mode of the ground-level release type. These modes and limitations on their use are discussed in the positions that follow.

3.2.1 Ground-Level Releases Conforms All release points are assumed to be ground The ground-level release mode is appropriate for the majority of Control Room X/Q level releases.

related assessments. If the release type is ground level, ARCON96 ignores all user inputs to release velocity and radius. Release height is used to establish the plume slant path.

3.2.2 Elevated (Stack) Releases N/A The stack release mode is appropriate for releases from a freestanding, vertical, uncapped See Section 3.2.1.

stack that is outside the directionally dependent zone of influence of adjacent structures.

Such a stack should be more than 2-1/2 times the height of the adjacent structures or be located:

• more than 5L downwind of the trailing edge of upwind buildings, and

• more than 2L upwind of the leading edge of downwind buildings, and

• more than 0.5L crosswind of the closest edge of crosswind buildings Where L is the lesser of the height or width of the building creating the downwind, upwind, or crosswind wake. Since L will be dependent on wind direction for most building clusters, it will generally be necessary to assess the zone of influence for all directions within the 900 wind direction sector centered on the line of sight between the stack and the Control Room intake. If multiple intakes are involved such that upwind, downwind, and crosswind

Attachment 4 Regulatory Guide 1.194 Compliance Table Page 4 of 33 orientations are confounded, 5L could be used for each orientation. Plume rise from buoyancy or mechanicaljet effects are not calculated by ARCON96. The analyst may determine plume rise and add the amount of rise to the physical height of the stack to obtain an effective plume height as described in Regulatory Position 6 of this guide (Note: The plume rise may not be added to the physical height of the stack for the purpose of meeting the 2-1/2 times height criterion). Although ARCON96 does not determine plume rise, the input values of stack flow, radius, and vertical velocity are used by ARCON96 to assess downwash and to estimate a limiting X/Q value.

If the Control Room intake is located close to the base of a tall stack, the elevated release model in ARCON96 generates negligibly low X/Q values. Although perhaps numerically correct, these model results may not be sufficiently conservative for a design basis assessment since the model does not adequately address meteorological conditions that could result in higher x/Q values. Although the staff has previously suggested that licensees model fumigation as a mechanism to address this situation, the fumigation model did not appear to adequately estimate the effluent concentrations at the bases of industrial stacks.

Concentrations greater than those predicted by ARCON96 could result from diurnal wind direction changes, meander, or stagnation. Therefore, the following procedure should be used to assess whether a particular stack-intake configuration is subject to this concern and to determine the appropriate x/Q values.

In addition to running ARCON96 to determine the elevated stack x/Q values for the Control Room assessment, the analyst should calculate the maximum elevated stack X/Q value (non-fumigation) using the methodology of Regulatory Guide 1.145 (Ref. 9) to determine the maximum X/Q value at ground level for the 0-2 hour interval and for the 24-96 and 96-720 hour intervals. The NRC-sponsored code, PAVAN (Ref. 14), is acceptable to the staff for this assessment. For this assessment, the input parameters should be adjusted such that the effective release height is measured from the elevation of the Control Room outside air intake rather than plant grade. The same release point characterization and meteorological data sets used in ARCON96 should be used to determine the X/Q values for several distances in each wind direction sector with the objective of identifying the maximum X/Q value.

Figure A.4 of Reference 15 may be useful in this regard. The maximum x/Q value obtained for the 0-2 hour interval should be compared to the corresponding X/Q value generated by ARCON96 and the higher value used in habitability assessments. The x/Q values generated bv ARCON96 for the 2-8 and the 8-24 hour intervals may be used without adjustment.

Attachment 4 Regulatory Guide 1.194 Compliance Table Page 5 of 33 For the 24-96 hour and 96-720 hour intervals, the following expressions may be used to determine the effective X/Q. This deterministic approach assumes that the stack plume reverses direction for 1 hour1.157407e-5 days <br />2.777778e-4 hours <br />1.653439e-6 weeks <br />3.805e-7 months <br /> of each day for the duration of the event. The plume is assumed to fold over itself such that the ground level concentration is at its maximum value at the Control Room intake.

X PAVAN kQ) 24--)96 HRS

+ K3*( ARCON96 Q24 )96 H-RS (1)

Q)24---)96 HRS 24 (X)PAVAN

• X )ARCON96

'Q 96-4720hrs KQ)96--720hrs (2)

(XQ)9->72hrs 24

.1 4 3.2.3 Vent Releases N/A See Section 3.2.1.

The ARCON96 calculation of vent releases includes an algorithm to model mixed-mode releases as described in Regulatory Guide 1.111 (Ref. 10), which addresses X/Q values used in the assessment of routine effluent releases. The development of this algorithm was based in part on limited field experiments. Given the limited experiment set, the results obtained with this algorithm may not be sufficiently conservative for accident evaluations.

For this reason, the vent release mode should not be used in design basis assessments. This position is consistent with the guidance of Regulatory Guide 1.145 (Ref. 9) for offsite X/Q values. These releases should be treated as a ground level release (Section 3.2.1) or as an elevated release (Section 3.2.2).

Attachment 4 Regulatory Guide 1.194 Compliance Table Page 6 of 33 3.2.4 Diffuse Area Sources N/A The diffusion models are based on point-The diffusion models in ARCON96 are based on point-source formulations. However, source formulations.

some release sources may be better characterized as area sources. Examples of possible area sources are postulated releases from the surface of a reactor or a secondary containment building. Typical assessments for loss-of-coolant accidents (LOCAs) have conservatively assumed that the containment structure could leak anywhere on the exposed surface. As such, these assessments typically used the shortest distance between the building surface and the Control Room intake and have treated the building as a point source. This approach may be unnecessarily conservative. A more reasonable approach, while still maintaining adequate conservatism, would be to model the building surface as a vertical planar area source. This approach is not intended to address dispersion resulting from building-induced turbulence. Treatment of a release as a diffuse source will be acceptable for design basis calculations if the guidance herein is followed. The staff may consider deviations from this guidance on a case-by-case basis.

3.2.4.1 Diffuse source modeling should be used only for those situations in which the activity N/A being released is homogeneously distributed throughout the building and when the assumed release rate from the building surface would be reasonably constant over the The ffusion sa surface of the building. For example, steam releases within a Turbine Building with roof ventilators or louvered walls would generally not be suitable for modeling as a diffuse source. (See Regulatory Positions 3.2.4.7 and 3.2.4.8.).

3.2.4.2 Since leakage is more likely to occur at a penetration, analysts must consider the potential N/A impact of building penetrations exposed to the environment within this modeled area. If the penetration release would be more limiting, the diffuse area source model should not be The ffusion sa used. Releases from personnel air locks and equipment hatches exposed to the environment, or containment purge releases prior to cOntainment isolation, may need to be treated differently. It may be necessary to consider several cases to ensure that the x/Q value for the most limiting location is identified.

Attachment 4 Regulatory Guide 1.194 Compliance Table Page 7 of 33 Note: Penetrations that are enclosed within safety-related structures need not be considered in this evaluation if the release would be captured and released via a plant ventilation system, as ventilation system releases should have already been addressed as a separate release point.

3.2.4.3 The total release rate (e.g., Ci.s- 1) from the building atmosphere is to be used in N/A conjunction with the diffuse area source X/Q in assessments. This release rate is assumed The diffusion models are based on point-to be equally distributed over the entire diffuse source area from which the radioactivity source formulations.

release can enter the environment. For freestanding containments, this would be the entire periphery above grade or above a building that surrounds the lower elevations of the containment. When a licensee can justify assuming collection of a portion of the release from the containment within the surrounding building, the total release from the containment may be apportioned between the exposed and enclosed building surfaces.

Similarly, if the building atmosphere release is modeled through more than one simultaneous pathway (e.g., drywell leakage and main steam safety valve leakage in a BWR), only that portion of the total release released through the building surface should be used with the diffuse area X/Q. The release rate should not be averaged or otherwise apportioned over the surface area of the building. For example, reducing the release rate by 50 percent because only 50 percent of the surface faces the Control Room intake would be inappropriate.

3.2.4.4 ARCON96 uses two initial diffusion coefficients entered by the user to represent the area N/A source. There are insufficient field measurements to mechanistically model these initial The diffusion models are based on point-diffusion coefficients. The following deterministic equations should be used in the source formulations.

absence of site-specific empirical data.

Note: See Regulatory Position 7 regarding the use of site-specific empirical measurements.

Widtharea source (3)

Yo= 6 Height area source cZ0 - 6 (4)

Attachment 4 Regulatory Guide 1.194 Compliance Table Page 8 of 33 3.2.4.5 The height and width of the area source (e.g., the building surface) are taken as the N/A maximum vertical and horizontal dimensions of the above-grade building cross-sectional area perpendicular to the line of sight from the building center to the Control Room intake. The ffusion sa These dimensions are projected onto a vertical plane perpendicular to the line of sight and located at the closest point on the building surface to the Control Room intake. The release height is set at the vertical center of the projected plane. The source-to-receptor distance (slant path) is measured from this point to the Control Room intake.

3.2.4.6 Intentional releases from a secondary containment (e.g., standby gas treatment systems N/A (SGTS) at BWR reactors) or annulus ventilation systems in dual containment structures should be treated as a ground-level release or an elevated stack release, as appropriate. The source formulations.

diffuse area source model may be appropriate for time intervals for which the secondary containment or annulus ventilation system is not capable of maintaining the requisite negative pressure differential specified in Technical Specifications or in the FSAR.

Secondary containment bypass leakage (i.e., leakage from the primary containment that bypasses the secondary containment and is not collected by the SGTS) should be treated as a ground-level release or an elevated stack release, as appropriate.

3.2.4.7 A second possible application of the diffuse area source model is determining a x/Q value N/A for multiple (i.e., 3 or more) roof vents. This treatment would be appropriate for configurations in which (1) the vents are in a close arrangement, (2) no individual vent is The ffusion sa significantly closer to the Control Room intake than the center of the area source, (3) the release rate from each vent is approximately the same, and (4) no credit is taken for plume rise. The distance to the receptor is measured from the closest point on the perimeter of the assumed area source. For assumed areas that are not circular, the area width is measured perpendicular to the line of sight from the center of the assumed source to the Control Room intake. The initial diffusion coefficient Tyo is found by Equation 3; GYzo is assumed to be 0.0.

Note: The degree of significance will depend on the radius or width of the assumed area and the proximity of the vent cluster to the Control Room intake. As the radius decreases or the distance from the cluster to the Control Room intake increases, the less significance the position of any one vent has.

Attachment 4 Regulatory Guide 1.194 Compliance Table Page 9 of 33 A third possible application of the diffuse area source model is determining a X/Q value for 3.2.4.8 N/A large louvered panels or large openings (e.g., railway doors on BWR Mark I plants) on vertical walls. This treatment would be appropriate for a louvered panel or opening when The diffusion models are based on point-(1) the release rate from the building interior is essentially equally dispersed over the entire source formulations.

surface of the panel or opening and (2) assumptions of mixing, dilution, and transport within the building necessary to meet condition 1 are supported by the interior building arrangement. The staff has traditionally not allowed credit for mixing and holdup in Turbine Buildings because of the buoyant nature of steam releases and the typical presence of high volume roof exhaust ventilators. The distance to the receptor and the release height is measured from the center of the louvered panel or opening. Initial diffusion coefficients are found using Equations 3 and 4 assuming the width and height is that of the panel or opening rather than that of the building. If the area source and the intake are on the same building surface such that wind flows along the building surface would transport the release to the intake, the initial dispersion coefficient will need to be adjusted. If the included angle between the source-receptor line of sight and the vertical axis of the assumed source is less than 45 degrees, aoy should be set to 0.0. If the included angle between the source-receptor line of sight and the horizontal axis of the assumed source is less than 45 degrees, cyzo should be set to 0.0.

3.3 Determination of Control Room Intakes (Receptors) Conforms The receptors considered in the calculation This section of the guide provides guidance to the meteorological analyst in applying are the two Control Room intakes. The 'A' models for determining X/Q values that are appropriate for the as-built configuration of train intake is at 484' and the 'B' train is at Control Room intakes. Radioactive materials released during an accident can enter the 507'-9" above grade. Ground level for the Control Room envelope via several potential pathways. These pathways may be VC Summer site is 436'. Therefore, the intentional (e.g., ventilation system outside air intakes) and unintentional infiltration paths intake height for Intake 'A' will be 48' or (e.g., doorways, envelope penetrations, leakage in ventilation system components). The 14.6m and Intake 'B' 71.75' or 21.9m. All applicable pathways will vary from site to site depending on the arrangement of the elevations are measured from the same Control Room envelope in relation to other site buildings, the pressure differentials reference point and the elevation difference between these buildings and the Control Room, the configuration of Control Room is zero.

ventilation systems, and the classification of the Control Room dose control (e.g., zone isolation with filtered pressurization, zone isolation with no pressurization). It may be necessary to determine X/Q values for each potential pathway. However, the selection of one or more bounding intakes for the X/Q evaluation may be sufficient to establish compliance with regulatory guidelines.

Attachment 4 Regulatory Guide 1.194 Compliance Table Page 10 of 33 3.3.1 Ventilation System Outside Air Intakes Conforms All Control Room ventilation systems draw makeup air from the environment during Control Room ventilation system normal operations and many draw air from the environment for the purpose of supplying configuration for normal and emergency modes was considered when identifying the filtered pressurization air. The configuration of these systems may change between normal Conto Roo ouside air intafor whi and emergency modes. In some configurations, normal ventilation outside air intakes isolate and different intakes open to supply pressurization air. Some intake dampers may x/Q values should be calculated.

have failure modes related to loss of ac power or single failures. These considerations should be evaluated in identifying the Control Room outside air intakes for which X/Q values should be calculated.

3.3.2 Dual Ventilation Outside Air Intakes N/A This section applies to Control Room ventilation system configurations that have two No credit is taken in the radiological outside air intakes, each of which meets applicable design criteria of an engineered consequence analyses for use of alternate safeguards feature (ESF), including single-failure criterion, missile protection, seismic intakes during postulated accidents.

criteria, and operability under loss-of-offsite AC power conditions. Operability requirements should be provided in Technical Specifications. The outside air intakes should be located with the intent of providing a low contamination intake regardless of wind direction. The assurance of a low contamination outside air intake depends on release point configuration, building wake effects, terrain, and the possibility of wind stagnation or wind direction reversals. The two intakes should not be within the same wind direction window, defined as a wedge centered on the line of sight between the source and the receptor with the vertex located on the release point. If ARCON96 is used, the wedge angle is 900 (i.e., 45 degrees on either side of the line of sight). If the methods of Regulatory Position 4 are used, the size of the wedge is as given in Table 2. Figure 3 illustrates four examples of the interplay between Control Room intakes, release points, and wind direction windows. In addition, the analyst should consider X/Q values for infiltration pathways as discussed in Regulatory Position 3.3.3.

The methods of this regulatory position involve identification of the limiting and favorable intakes with regard to their X/Q value. Because of the interplay of building wake, plume rise, wind direction frequency, intake flow rate, and other parameters, it may not be possible to identify the limiting or favorable intake by observation. In these situations, X/Q values should be calculated for each release point-intake combination and the limiting and favorable intakes identified on the basis of these values.

Attachment 4 Regulatory Guide 1.194 Compliance Table Page 11 of 33 3.3.2.1 If both of the dual intakes are located within the same wind direction window, both intakes N/A could be contaminated (See Figure 3(a)). In this case, the X/Q values for each air intake No credit is taken in the radiological should be calculated using ARCON96 as described in other sections of this guide and an consequence analyses for use of alternate effective X/Q value calculated. Equation 5a should be used if the intake flow rates are intakes during postulated accidents.

equal. If the intake flow rates are not equal, but the imbalance does not shift between intakes, Equation 5b should be used. If the flow rate imbalance can shift between intakes, Equation 5c should be used. This calculation is repeated for each averaging time interval.

X/Q = 0.5[(X/Q), + (X/Q) 2] (5a)

X/Q - F(XQ),2 (X/Q)2 (5b)

F, +F 2 X/Q= max (F,, F2 )

• max [(X/Q) ,, (X/Q) 2 ] + min F1 +F 2 (F,, F2 )

• min [(X/Q),, (X/Q) 2]

(5c)

Where:

X/Q = Effective X/Q, s m-3 (X/QI, (X/Q) 2 = X/Q value for outside air intakes 1 and 2, s m-3 F1 , F2 = Flow rate for outside air intakes l and 2, cfm 3.3.2.2 If the dual outside air intakes are not in the same wind direction window but cannot be N/A isolated by design, the x/Q values for the limiting outside air intake should be calculated No credit is taken in the radiological for each time interval as described elsewhere in this guide. Equation 6a should be used if consequence analyses for use of alternate the intake flow rates are equal. If the intake flow rates are not equal, but the imbalance intakes during postulated accidents.

does not shift between intakes, Equation 6b should be used. If the flow rate imbalance can shift between intakes, Equation 6c should be used.

Attachment 4 Regulatory Guide 1.194 Compliance Table Page 12 of 33 X/Q = 0.5 max [(X/Q) 1 , (X/Q) 2 ] (6a) max[(F, (X/Q)l, F2 (X/Q) 2] (6b)

F, +F2 max (F*,F 2 ) max i(X/Q) 1 , (X/Q) !I 2

X/Q= F1 +F 2 (6c) 3.3.2.3 If the ventilation system design allows the operator to manually select the least N/A contaminated outside air intake as a source of outside air makeup and close the other No credit is taken in the radiological intake, the X/Q values for each of the outside air intakes should be calculated for each time consequence analyses for use of alternate interval as described elsewhere in this guide. The x/Q value for the limiting intake should intakes during postulated accidents.

be used for the time interval prior to intake isolation. This x/Q value may be reduced by a factor of 2 to account for dilution by the flow from the other intake (see Equation 6a). The x/Q values for the favorable intake are used for the subsequent time intervals. The x/Q values for the favorable intake may be reduced by a factor of 4 to account for the dual inlet and the expectation that the operator will make the proper intake selection. This protocol should be used only if the dual intakes are in different wind direction windows and if there are redundant, ESF-grade radiation monitors within each intake, with Control Room indication and alarm, to monitor the intakes. The requisite steps to select the least contaminated outside air intake, and provisions for monitoring to ensure the least contaminated intake is in use throughout the event, should be addressed in procedures and in operator training.

A conservative delay time should be assumed for the operator to complete the necessary actions. This delay period should consider: (1) the time for the operator to recognize the radiation monitor alarm and determine its validity (as provided for in the alarm response procedure), (2) delays associated with other accident response actions competing for the operator's attention, (3) the time needed to complete the actions, and (4) diesel generator sequencing time, if applicable. If actions are required outside the Control Room, delays associated with transit to the local control stations (including those delays caused by worker radiological protection controls associated with accident dose rates), and the availability of personnel should be considered.

Attachment 4 Regulatory Guide 1.194 Compliance Table Page 13 of 33 Note: The adjustment protocol and the numeric factors of this section are deterministic in nature and are expected to be conservative for most sites. Different factors may be considered on a case-by-case basis with sufficient justification.

3.3.2.4 If the ventilation system design provides for automatic selection of the least contaminated N/A outside air intake, the X/Q values for the favorable intake should be calculated for each No credit is taken in the radiological time interval as described elsewhere in this guide. The X/Q values may be reduced by a consequence analyses for use of alternate factor of 10 to account for the ability to automatically select a "clean" intake. This intakes during postulated accidents.

protocol should be used only if the dual intakes are in different wind direction windows, there are redundant ESF-grade radiation monitors within each intake and an ESF-grade control logic and actuation circuitry is provided for the automatic selection of a clean intake throughout the event.

3.3.3 Infiltration Pathways Conforms Infiltration of contaminated air to a Control Room can be minimized by proper design and maintenance of the Control Room envelope (CRE). However, infiltration is always a possibility and the location and significance of these leakage pathways may warrant determination of X/Q values. An unfiltered inleakage path of 100 cfm can admit the same quantity of radioactive material as a pressurization air intake having a flow of 2000 cfm through a 95 percent efficient filter. The situation can be further compounded if the X/Q for the unfiltered pathway is more limiting than that for the Control Room outside air intake.

The infiltration paths actually applicable to a particular facility will be identified via inleakage testing or CRE inspections and surveillances. Refer to Table H-1, "Determination of Vulnerability Susceptibility," of NEI 99-03, "Control Room Habitability Guidance" (Ref. 16), for further guidance on infiltration pathways.

A 95w-percentile x/Q value should be determined for each time interval for any infiltration path that could result in a significant intake of contaminated air into the CRE. Because of the interplay of source-to-receptor distance and direction, infiltration path flow rate, whether the path is filtered or unfiltered, and other considerations, it may not be possible to

Attachment 4 Regulatory Guide 1.194 Compliance Table Page 14 of 33 identify the potential impact of an infiltration path by observation. In these situations, X/Q values should be calculated for each pathway and the limiting X/Q value(s) identified. If there is sufficient margin available, it may be possible to calculate X/Q values assuming the shortest distance between the release point and any identified point of infiltration on the outside of the CRE.

+

3.4 Determination of Source-Receptor Distances and Directions Conforms When the combinations of release points and intakes have been identified, the direction and distance between the release point and the intake should be determined. Wind direction Appropriate wind directions and source-data are recorded as the direction from which the wind blows (e.g., a north wind blows receptor distances were input into from the north; a wind blowing out of the west is recorded with a direction of 270 degrees). ARCON96 for determination of the X/Qs for The direction input to ARCON96 is the wind direction that would carry the plume from the each of the accidents analyzed. No taut release point to the intake. For example, an analyst standing at the intake facing west to string distances were used in the X/Q the release point, would enter 270 degrees; an analyst facing north, would enter 360 determination.

degrees, etc.

The source-to-receptor distance is the shortest horizontal distance between the release point All distances to the CRHE air intakes are >

and the intake. ARCON96 will use this distance and the elevations of the source and 10m.

receptor to calculate the slant path. For an area source such as building surface, the shortest horizontal distance from the building surface to the Control Room intake is used as the source-to-receptor distance. For releases within building complexes, the shortest horizontal distance between the release point and the intake could be through intervening buildings. In these cases, it is acceptable to take the length of the shortest path (e.g., "taut string length") around or over the intervening building as the source-to-receptor distance.

If the distance to the receptor is less than about 10 meters, the ARCON96 code and the procedures in Regulatory Position 4 should not be used to assess x/Q values. These situations will need to be addressed on a case-by-case basis.

Note: The site meteorological tower wind direction sensors are generally calibrated with reference to true north (360 degrees). Analysts should use caution in measuring directions on site engineering drawings since these drawings typically incorporate a plant grid and a plant "north" that may not align with true north. The source-to-receptor directions input to ARCON96 must use the same north reference as the wind direction observations.

__________ I ___________________________________________________________________________________ C _________________________________________

Attachment 4 Regulatory Guide 1.194 Compliance Table Page 15 of 33 4.0 ALTERNATIVE PROCEDURES FOR GROUND-LEVEL RELEASES N/A This regulatory position addresses alternative methods for determining X/Q values for All ground level releases were determined Control Room radiological habitability assessments. The methods in Regulatory per the preceding methodology.

Positions 4.1 to 4.3 are based on Murphy-Campe (Ref. 2) and the Standard Review Plan Chapter 6.4 (Ref. 3).

4.1 Point Source-Point Receptor N/A The 0-8 hour 95%-percentile (Note: The Murphy-Campe document identified this as the 5 All ground level releases were determined percentile /Q value.) X/Q value for a single point source on the surface of the containment per the preceding ARCON96 methodology.

or other building and a single point receptor with a difference in elevation less than 30 percent of the building height may be estimated using Equation 7.

X 1 S- (7)

Where:

x/Q = Relative concentration at plume centerline for time interval 0-8 hours, s m3 3 = Wake factor U = Wind speed at 10 meters, m s-I (y,= Standard deviation, in meters, of the gas concentration in the horizontal and (FYI vertical cross wind directions evaluated at distance x and by stability class 4.2 Diffuse Source-Point Receptor N/A Equation 8 may be used when the activity is assumed to leak from many points on the All ground level releases were determined surface of a building such as the containment in conjunction with a single point receptor. and all sources were considered point This equation is also appropriate for point source-point receptors where the difference in elevation between the source and the receptor is greater than 30 percent of the height of the sources.

Attachment 4 Regulatory Guide 1.194 Compliance Table Page 16 of 33 upwind building, typically the containment, which creates the most significant building wake impact. The equation is also applicable to a point source and volume receptor (e.g.,

an isolated Control Room with infiltration occurring at many locations).

U + A/TZ (8)

Where:

x/Q =Relative concentration at plume centerline for time interval 0-8 hours, s m3 U = Wind speed at 10 meters, m s1 uy, oz = Standard deviation, in meters, of the gas concentration in the horizontal and vertical cross wind directions evaluated at distance x and by stability class 3

K= *(s /d)1.

s= Shortest distance between building surface and receptor location, m d= Diameter or width of building, m A= Cross-section area of building, m2 The reference to "building" in the definitions of s, d, and A is to the diffuse source (e.g.,

containment). If the equation is used with a point source, the reference is to the building that has the greatest impact on the building wake. The values of the parameters Ty, (Fz and U should be determined on the basis of the values of the site meteorological data. Some early analyses may have been based on generic meteorology conditions (e.g., F stability with wind speeds of 1.0 m-sl). If these early analyses are to be updated, the staff recommends that the ARCON96 code be used. If the ARCON96 code is not used, site-

Attachment 4 Regulatory Guide 1.194 Compliance Table Page 17 of 33 specific hourly meteorological data should be used to determine the 95th-percentile x/Q value. Figures 4 and 5 provide sigma values by stability category for distances greater than 10 meters. The data on these graphs should not be extrapolated for distances less than 10 meters.

4.3 Point or Diffuse Source with Two Alternative Receptors N/A Equations 7 and 8 of this guide may be used in conjunction with the procedures in All ground level releases were determined Regulatory Position 3.3.2 to determine X/Q values for Control Room designs having two or per the preceding ARCON96 methodology.

more Control Room outside air intakes, each of which meets the requirements of an engineered safety feature (ESF) including, as applicable, single-failure criteria for active components, seismic criteria, and missile criteria. If Equation 8 of this guide is used, the parameter K should be set to 0.0. In a change from previous practice, the staff no longer finds Equation 7 of Reference 2 to be acceptable for use in new applications.

4.4 Determination of X/Q Values for Other Time Intervals N/A Equations 7 and 8 are used to determine x/Q values for the first time interval of 0-8 hours. All ground level releases were determined The x/Q values for other time intervals are obtained by adjusting for long-term per the preceding ARCON96 methodology.

meteorological averaging of wind speed and wind direction. This is accomplished by multiplying the 0-8 hour time interval X/Q value by a correction factor for wind speed and a correction factor for wind direction.

Note: Previous guidance also provided for including a factor to account for personnel occupancy factors. Since typical radiological analysis codes provide the capability to enter these factors separately, the staff recommends that the factors not be included in the x/Q value to avoid inadvertent double crediting.

4.4.1 X/Q Correction for Wind Speed Averaging N/A This correction is defined as the ratio of the wind speed used to determine the 0-8 hour X/Q All ground level releases were determined value to the wind speed appropriate for each of the other time intervals. Column 2 of per the preceding ARCON96 methodology Table 1 tabulates the wind speed percentiles that correspond to each of these intervals. The utilizing standard time intervals; hourly data should be arranged in order of increasing wind speed and the wind speed consequently, no X/Q correction is required percentiles determined (i.e., the lowest wind speeds associated with the lowest percentiles). per wind speed averaging.

Attachment 4 Regulatory Guide 1.194 Compliance Table Page 18 of 33 I

Include only the wind speed data associated with wind directions from sectors that result in receptor contamination. Table 2 tabulates the size of the minimum wind direction window to be used. From this ranking, identify the wind speed value for each interval that is not exceeded more than the stated percentage of the time. Divide this wind speed value into the 5th-percentile wind speed used to determine the 0-8 hour x/Q to obtain the X/Q correction factor for wind speed. The values shown in Column 1 of Table 1 are representative correction factors that may be used if hourly observation meteorological data are not available.

Table 1 X/Q Correction for Wind Speed Averaging Column 1 Column 2 Representative Corresponding Time Interval X/Q Factors Wind Speed Percentile 0-8 hours 1.0 5 8-24 hours 0.67 10 1-4 days 0.50 20 4-30 days 0.33 40 Table 2 Wind Direction Sectors Minimum Window (Note: Centered on the s/d Ratio source-to-receptor direction.)

>2.5 680 1.25-2.5 900 0.8- 1.25 1130 0.6-0.8 1350 0.5-0.6 1580 0.35 - 0.5 1800

<0.35 2250 I

Attachment 4 Regulatory Guide 1.194 Compliance Table Page 19 of 33 The s/d is defined as:

s Shortest distance between building surface and receptor location, m d Diameter or Width of building, m The reference to "building" in Equation 9 is to the diffuse source (e.g., containment). If the equation is used with a point source, the reference is to the building that has the greatest impact on the building wake.

_______ I +

4.4.2 X/Q Correction for Wind Direction Averaging N/A All ground level releases were determined The average wind direction frequency F is obtained by summing the annual average wind per the preceding ARCON96 methodology direction frequencies within the minimum window. Table 2 tabulates the size of the utilizing standard time intervals; minimum wind direction window to be used. Column 2 of Table 3 is used to determine the consequently, no X/Q correction is required X/Q correction factor for wind direction for each time interval. Column 1 is used when F per wind direction averaging.

has not been determined.

Table 3 Wind Direction Averaging Correction Column 1 Column 2 Representative Equations for Time Interval x/Q Factors X/Q Factors 0-8 hours 1.0 1.0 8-24 hours 0.88 0.75 + F/4 1-4 days 0.75 0.50 + F/2 4-30 days 0.5 F

Attachment 4 Regulatory Guide 1.194 Compliance Table Page 20 of 33 5.0 INSTANTANEOUS PUFF RELEASES N/A All ground level releases were determined The alternative method in this section may be used to model the release to the environment per the preceding ARCON96 methodology.

as an instantaneous puff release. One hundred percent of the radionuclides must be No puff releases are assumed.

released directly to the environment over a period no longer than about 1 minute for a release to qualify as a puff release. Releases to enclosed buildings, intermittent releases that occur over a period longer than about 1 minute (e.g., releases from relief valves, atmospheric dumps), and releases that occur over a period longer than about 1 minute should be treated as continuous point source releases. The diffusion equation for an*

instantaneous puff ground level release, with no puff rise and no crosswind offset (i.e.,

center of puff is assumed to pass over Control Room intake), integrated over the duration of the puff passage is:

Where:

fr 2

((x, Jo k) +U;)1 (2.T) 3 ~ y~(x, k) + a;)

- (x,u,k,h)

=

Q kp 2-(O'2Y (x'k) +oý) + (O2 (xk) + 2) F(t)dt

Attachment 4 Regulatory Guide 1.194 Compliance Table Page 21 of 33 X-(x, u, x, h) = Effective puff relative concentration, s m-'

Q 3 X = Integrated concentration at control intake, Ci m- s-'

Qi Relative quantity, for nuclide i, Ci x = Release point to receptor distance, m u = Wind speed, n/sec. Assume 1.0 m s-1 k = Stability Class. Assume F.

h = Difference in elevation between the physical release point and the control room intake, m. If the control room intake is at a higher elevation than the release point and the puff is bouyant, assume h = 0.

T = Time for trailing edge of puff to pass control room intake, sec.

x + 3 [LYoy (x, k) +oi]

U F = Control room total intake flow rate, cfm. (If the control room intake flow rate is constant over the period 0 to T seconds, the F(t) terms can be omitted from Equation 10.

yx,y (x, k) = Standard deviation, m, of the puff in the horizontal along the wind direction and cross - wind directions at the receptor locations. Use Figure 4 with the distance x and Stability Class K to determine 0,,y at the receptor, e.g., ox, y = cy.

Yz (x, k) = Standard deviation, m, of the puff in the vertical cross - wind direction at the receptor location. Use Figure 5 with .the distance x and Stability Class K to determine a. at the receptor.

_____________ ~1~

Attachment 4 Regulatory Guide 1.194 Compliance Table Page 22 of 33 Gi = Initial standard deviation, m

=[2V V =Initial puff volume (expanded to standard atmospheric conditions), m3 (The puff dimensions that would exist when the puff is at the control room intake are assumed to exist during the entire puff transit.

Equation 10 provides the effective relative concentration for the puff. This value can be input to dose assessment codes such as RADTRAD or HABIT as any value of X/Q would be if the intake flows, release duration, and release rates are modeled consistent with the inputs to Equation 10.

6.0 PLUME RISE N/A Plume rise was not considered in the X/Q An applicant or licensee may propose adjustments to the release height for plume rise that determinations.

are due to buoyancy or mechanical jet on a case-by-case basis. In order to credit these adjustments, the applicant or licensee must be able to demonstrate that the assumed buoyancy or vertical velocity of the effluent plumes will be maintained throughout the time intervals that plume rise is credited. Such justifications need to consider the availability of AC power, failure modes of dampers and ductwork, time-dependent release stream temperatures and pressures, and 95%-percentile wind speeds and ambient temperatures.

(Note: As used here, 95t-percentile wind speed is that wind speed that is not exceeded more than 5 percent of the time. A 95Ih-percentile ambient temperature is that temperature that is not exceeded more than 5 percent of the time). Plume rise may be considered for freestanding stacks and for vents located on plant buildings. However, plume rise may not be used in demonstrating that a particular stack meets the 2-1/2 times the adjacent structure height criterion in Regulatory Position 3.2.2. A mixed-mode release model, such as that in Regulatory Guide 1.111 (Ref. 10), should not be used for design basis assessments.

The plume rise may be determined through the use of the following set of equations (Ref. 17). The plume rise for plant vents is determined using Equation 11. The distance x is entered as the horizontal distance between the vent and the Control Room outside air intake.

Attachment 4 Regulatory Guide 1.194 Compliance Table Page 23 of 33 The plume rise for isolated, free-standing stacks is calculated using Equations 11, 12, and

13. The distance x in Equation 11 should be based on the downwind location corresponding to the maximum X/Q value. See Regulatory Position 3.2.2. The plume rises calculated using Equations 12 and 13 should be compared and the larger plume rise identified. The result of this comparison is then compared to the plume rise determined using Equation 11 and the smaller plume rise selected for use.

Ah= 3*

• Fb *22 11/3 (11)

P2 u2 20 2 3

)1/

Ah 2.6 (Fb (12) 1/4 Fm Ah =2.44 (13)

_____________ .1~

Attachment 4 Regulatory Guide 1.194 Compliance Table Page 24 of 33 Where:

Ah = Plume rise, m 4

Fm = Momentum flux parameter, M s-2 p0 V0 w0 71pa

= Dimensionless entrainment constant for momentum -0.6 U = Wind speed at release height, m s-x = Distance from release point to receptor, m 4 3 Fb = Bouyancy flux parameter, m s-g(pa - p 0 )Vo 7Pa w0 = Effluent exit velocity, ms 3

V 0 = Volumetric release rate, m s-1 P0 = Effluent density after expansion to atmospheric pressure, kg m-3 kg m-3 Pa = Density of air, s = 0.0001 S-2 for A, B, C, and D stability; 0.00049 s- 2 for E stability; 0.0013 S-2 for F stability; 0.002 S-2 for G stability g = Gravitational acceleration, 9.8 m S-2 Although ARCON96 processes ambient meteorological conditions on an hour-by-hour basis, the code cannot vary the other parameters that enter into a plume rise determination.

For example, wind speed and stability class are varied hour by hour, but the density of air, the density of the effluent stream, and the vertical velocity are not varied hour-by-hour. As such, the analyst should ensure that these parameters are bounding for the entire period of the X/Q assessment or use individual time intervals to model the time-variant parameters.

An alternative approach would be to calculate the plume rise for each hour independently of ARCON96 and to select a plume rise that is exceeded more than 95 percent of the time.

Attachment 4 Regulatory Guide 1.194 Compliance Table Page 25 of 33 This rise is then added to the stack height as input to ARCON96.

In lieu of mechanistically addressing the amount of buoyant plume rise associated with energetic releases from steam relief valves or atmospheric dump valves, the ground level X/Q value calculated with ARCON96 (on the basis of the physical height of the release point) may be reduced. (Note: This adjustment factor and the associated velocity ratio criterion are deterministic in nature and their selection was based on sensitivity analyses performed for typical steam release points at LWRs.) The adjustment factor should not be ratioed for different vertical velocity ratios) by a factor of 5. This reduction may be taken only if (1) the release point is uncapped and vertically oriented and (2) the time-dependent vertical velocity exceeds the 95h-percentile wind speed. [Note: As used here, 95th-percentile wind speed is that wind speed that is not exceeded more than 5 percent of the time. A 95thpercentile ambient temperature is that temperature that is not exceeded more than 5 percent of the time (at the release point height) by a factor of 5].

7.0 USE OF SITE-SPECIFIC EXPERIMENTAL DATA N/A No experimental data was utilized to The methods and parameters provided in this guide are acceptable for use for design basis calculate the X/Qs.

Control Room habitability radiological assessments provided that all stated prerequisites and conditions are met. The staff believes that use of the guidance in this guide will result in X/Q values that are acceptably conservative. However, there may be circumstances in which these methods and parameters may not be advantageous for a particular plant configuration and site meteorological regimes and may lead to results that are deemed to be unnecessarily conservative. Licensees and applicants may opt to propose alternative methods and parameters such as those that are based in part on data obtained from site-specific experimental measurements. Data based on wind tunnel tests should be accompanied with an evaluation of the representativeness of the experiment results to the particular plant configuration and site meteorological regimes. These proposed alternatives, with supporting data, will be considered by the staff on a case-by-case basis.

The staff recommends that licensees considering an experimental program request a meeting with the staff in advance of starting the program. The intent of this recommendation is to allow the staff and the licensee (or applicant) to discuss the proposed program, prior to resource expenditure, and for the staff to provide a preliminary assessment of the proposal. The staff's approval of the proposed alternative methods and

Attachment 4 Regulatory Guide 1.194 Compliance Table Page 26 of 33 parameters will not be granted, however, until the licensee or applicant completes the experimental program and dockets the proposal with supporting analyses and data for formal staff review.

An acceptable experimental program should incorporate the following standards:

7.1 The experimental program should be appropriately structured so as to provide data of N/A appropriate quantity and quality to support data analysis and conclusions drawn from that data. The program should be developed by personnel who have educational and work No experimental data was utilized to experience credentials in air dispersion meteorology and modeling, calculate the x/Qs.

7.2 The experimental program should encompass a sufficient range of meteorological N/A conditions applicable to the particular site so as to ensure that the data obtained address the site-specific meteorological regimes and the site-specific release point/receptor No-experimental data was utilized to configurations that impact the Control Room X/Q values. Meteorological conditions calculate the X/Qs.

observed at the particular site with a frequency of 5 percent or greater in a year should be addressed. Parameters derived from statistical analyses on the experimental data should represent the 95t-percentile confidence level.

7.3 The experimental program, including data reduction and analysis, should incorporate N/A applicable quality control criteria of Appendix B to 10 CFR Part 50. The products of the experimental program should be verified and validated. No experimental data was utilized to calculate the X/Qs.

Attachment 4 Regulatory Guide 1.194 Compliance Table Page 27 of 33 Table A-2 ARCON96 INPUT PARAMETERS FOR DESIGN BASIS ASSESSMENTS Parameter/Discussion/ Acceptable Input Basis of Compliance Lower Measurement Height, meters Conforms height instrumetatilowed.

The value of this parameter is used by ARCON96 to adjust wind speeds for differences between the heights of the instrumentation and the release. heiact instrumetatilower e Use the actual instrumentation height when known. Otherwise, assume 10 meters.

Upper Measurement Height, meters Conforms The actual instrumentation upper measurement The value of this parameter is used by ARCON96 to adjust wind speeds for differences between height of 61 meters was utilized.

the heights of the instrumentation and the release.

Use the actual instrumentation height when known. Otherwise, use the height of the containment or the stack height, as appropriate. If wind speed measurements are available at more than two elevations, the instrumentation at the height closest to the release height should be used.

Wind Speed Units Conforms ARCON96 requires that wind speed be entered as miles per hour, mes-1, or knots. Wind speed was entered as meters per second, Use the wind speed units that correspond to the units of the wind speeds in the meteorological which corresponds to the units of the wind data file. speeds in the meteorological data files.

Attachment 4 Regulatory Guide 1.194 Compliance Table Page 28 of 33 Release Height, meters Conforms For the ten source-receptor locations, the The value of the release height is used for three purposes in ARCON96: (1) to adjust wind actual release height was utilized or the release speeds for differences between the heights of the instrumentation and the release, (2) to height was conservatively assumed to be 0 determine slant path for ground level releases, (3) to correct off-centerline data for elevated (assume release height equals intake height).

releases.

Plume rise from buoyancy and mechanical jet Use the actual release heights whenever available. Plume rise from buoyancy and mechanical jet effects was not considered in establishing the effects may be considered in establishing the release height if the analyst can demonstrate with release height.

reasonable assurance that the vertical velocity of the release will be maintained during the course of the accident. If actual release height is not available, set release height equal to intake height.

Building Area, meters2 Conforms The actual building vertical cross-sectional ARCON96 uses the value of the building area in the high speed wind speed adjustment for area perpendicular to the wind direction was ground-level and vent release models.

conservatively calculated. A value of 1,740 m2 was utilized.

Use the actual building vertical cross-sectional area perpendicular to the wind direction. Use default of 2000 m2 if the area is not readily available. Do not enter zero. Use 0.01 m2 if a zero entry is desired.

Note: This building area is for the building(s) that has the largest impact on the building wake within the wind direction window. This is usually, but need not always be, the reactor containment. With regardto the diffuse areasource option, the building area entered here may be differentfrom that used to establish the diffuse source.

Attachment 4 Regulatory Guide 1.194 Compliance Table Page 29 of 33 Vertical Velocity, meters/second N/A Vent and stack release models were not In ARCON96, the value of the vertical velocity is used only in vent and stack release models. It utilized. All releases were ground level.

is used for the downwash calculation. In the vent release model the velocity is used in the mixed-mode calculation.

If the vertical velocity is set to zero, the maximum downwash will be calculated and the release height will be reduced by an amount equal to six times the stack radius.

Note: The vent release model should not be usedfor DBA accident calculations.

For stack release calculations only, use the actual vertical velocity if the licensee can demonstrate with reasonable assurance that the value will be maintained during the course of the accident (e.g., addressed by technical specifications), otherwise, enter zero. If the vertical velocity is set to zero, ARCON96 will reduce the stack height by 6 times the stack radius for all wind speeds. If this reduction is not desired, the stack radius should also be set to zero.

Stack Flow, meters3/second N/A Vent and stack release models were not ARCON96 uses the value of the stack flow in X/Q calculations for all 3 release types to ensure utilized. All releases were ground level.

that the near field concentrations are no greater than the concentration at the release point. The impact diminishes with increasing distance.

Use actual flow if it can be demonstrated with reasonable assurance that the value will be maintained during the course of the accident (e.g., addressed by Technical Specifications).

Otherwise, enter zero.

The flow is used in both elevated and ground-level release modes to establish a maximum X/Q value. This value is significant only if the flow is large and the distance from the release point to the receptor is small.

Attachment 4 Regulatory Guide 1.194 Compliance Table Page 30 of 33 Stack Radius, meters N/A ARCON96 uses the value of the stack radius in downwash calculations in the vent and stack Vent and stack release models were not release modes. utilized. All releases were ground level.

Use the actual stack internal radius when both the stack radius and vertical velocity are available.

If the stack flow is zero, the radius should be set to zero.

Distance to Receptor, meters Conforms The actual straight-line horizontal distances g

The value of horizontal distance to the receptor from the release point is used in ARCON96 for between the release point and the ContrOl for ground level releases and the off-centerline correction factors for Room are utlze (nt anC es calculating the slant rangestac reeasemodls.Room are utilized (no taut string distances stack release models.

even for releases in the building complex) to calculate the X/Qs.

Use the actual straight-line horizontal distance between the release point and the Control Room intake. All source to receptor distances are greater than 10 meters.

For ground-level releases, it may be appropriate to consider flow around an intervening building if the building is sufficiently tall that it is unrealistic to expect flow from the release point to go over the building.

Note: If the distance to receptor is less than about 10 meters, ARCON96 should not be used to assess relative concentrations.

Intake Height, meters Conforms The actual Control Room intake heights were Tie toalCutthe R iksw The value of the intake height is used in ARCON96 for calculating the slant range for ground utilized to calculate the X/Qs.

level releases and the off-centerline correction factors for stack release models.

Use the actual intake height. If the intake height is not available for ground level releases, assume the inthke height is equal to the release height. For elevated releases, assume the height of the tallest site building.

Attachment 4 Regulatory Guide 1.194 Compliance Table Page 31 of 33 Elevation Difference, meters Conforms Actual release heights and the intake heights the intake an relea tiondference were hti The value of this parameter is used by ARCON96 to normalize the release heights andthe release grade" with different grades for heights when the two heights are specified as "above X/Qs.

and the other as appropriate to determine point and intake height, or when one measurement is referenced to "above grade" to "above sea level."

Use zero unless it is known that the release heights are reported relative to different grades or reference data.

Direction to Source, degrees Conforms Wind direction from the intakes back to the ARCON96 uses the value of this parameter and the Wind Direction Window to establish which release point was utilized to calculate the X/Qs.

range of wind directions should be included in the assessment of the X/Q.

Plant north and true north are considered the Use the direction FROM the intake back TO the release point. (Wind directions are reported as same at VCSNS.

the direction from which the wind is blowing. Thus, if the direction from the intake to the release point is north, a north wind will carry the plume from the release point to the intake).

No analyses considered ground-level releases Note: Some facilities have a "plant north" shown on site arrangement drawings that is different that flow around a building rather than over it.

from "true north." The direction entered must have the same point of reference as the wind directions reported in the meteorological data.

For ground-level releases, if the plume is assumed to flow around a building rather than over it, the direction may need to be modified to account for the redirected flow. In this case, the X/Q should be calculated assuming flow around and flow over (through) the building and the higher of the two X/Q s should be used.

Attachment 4 Regulatory Guide 1.194 Compliance Table Page 32 of 33 Surface Roughness Length, meters Conforms w dA surface roughness length of 0.2 in lieu of the ARCON96 uses the value of this parameter in adjusting wind speeds to account for differences default value of 0. 1 was utilized to calculate in meteorological instrumentation height and release height. the X/Qs.

Use a value of 0.2 in lieu of the default value of 0.1 for most sites. (Reasonable values range from 0.1 for sites with low surface vegetation to 0.5 for forest-covered sites).

Wind Direction Window, degrees Conforms Used the default values.

Code Default ARCON96 uses the value of this parameter and the Direction to Source to establish which range of wind directions should be included in the assessment of the X/Q.

Use the default window of 90 degrees (45 degrees on either side of line of sight from the source to the receptor).

Minimum Wind Speed, meters/second Conforms Code Default Used the default values.

ARCON96 uses the value of this parameter to identify calm conditions.

Use the default wind speed of 0.5 m-s-1 (regardless of the wind speed units entered earlier),

unless there is some indication that the anemometer threshold is greater than 0.6 m-s-'.

Attachment 4 Regulatory Guide 1.194 Compliance Table Page 33 of 33 Averaging Sector Width Constant Conforms Code Default An averaging sector width constant of 4.3 (the preferred value) was utilized to calculate the and X/Qs.

ARCON96 uses the value of this parameter to prevent inconsistency between the centerline sector average X/Q s for wide plumes. Has largest effect on ground level plumes.

Although the default value is 4, a value of 4.3 is preferred. (A future revision to ARCON96 will change the default to 4.3).

Initial Diffusion Coefficients, meters Conforms The initial diffusion coefficients are set to zero ARCON96 uses these parameters in modeling a diffuse source. since only point sources were utilized in the is being used, see evaluation.

These values will normally be set to zero. If the diffuse source option Regulatory Position 2.2.4.

Hours in Averages Conforms Used the default values.

Code Default The values of this parameter were selected to provide results for desired periods and to provide a smooth X/Q curve.

Use the default values.

Minimum Number of Hours Conforms Code Default Used the default values.

The default values of this parameter will allow processing with up to 10% missing data.

Use the default values.

I Attachment 5 Safety Assessment for the Proposed Technical Specification And Bases Changes

Attachment 5 Page 1 of 10 Proposed Technical Specification and Bases Changes A description of each proposed TS change and the associated basis/safety assessment are included in Table 5-1.

The majority of the Technical Specification changes are revised in accordance with the Improved Standard Technical Specifications Change Traveler (TSTF-51, Revision 2) which permits removal of the Technical Specification requirements for ESF features to be OPERABLE after sufficient radioactive decay has occurred to ensure off-site doses remain below the SRP limits.

Associated with this change is the deletion of OPERABILITY requirements during CORE ALTERATIONS for ESF mitigation features. This change allows flexibility to move personnel and equipment and perform work which would affect containment OPERABILITY during the handling of irradiated fuel.

Following reactor shutdown, decay of the short-lived fission products greatly reduces the fission product inventory present in irradiated fuel. The proposed changes are based on performing analyses assuming a longer decay period to take advantage of the reduced radionuclide inventory available for release in the event of a fuel handling accident. Following sufficient decay, the primary success path for mitigating the fuel handling accident no longer includes the functioning of the active containment systems. Therefore, the OPERABILITY requirements of the Technical Specifications are modified to reflect that water level (23') and decay time (72 hours8.333333e-4 days <br />0.02 hours <br />1.190476e-4 weeks <br />2.7396e-5 months <br /> after shutdown) are the primary elements for the success path for mitigating a fuel handling accident.

To support this change in requirements during the handling of irradiated fuel, the OPERABILITY requirements during CORE ALTERATIONS for ESF mitigation features are deleted. The accidents postulated to occur during core alterations, in addition to fuel handling accidents, are: inadvertent criticality (due to a control rod removal error or continuous control rod withdrawal error during refueling or boron dilution) and the inadvertent loading of, and subsequent operation with, a fuel assembly in an improper location. These events are not postulated to result in fuel cladding integrity damage.

Also, the Technical Specifications only allow the handling of irradiated fuel in the reactor vessel when the water level in the reactor cavity is at the high water level. Therefore, the proposed changes only affect containment requirements during periods of relatively low shutdown risk during refueling outages. Therefore, the proposed changes do not significantly increase the shutdown risk.

Other Technical Specification revisions reflect the update of the accident source term and associated design basis accidents utilizing the guidance provided in USNRC Regulatory Guide 1.183 and the associated control room and offsite dose requirements of 10 CFR 50.67.

The preceding discussion in Attachment 1, the Safety Assessment in Attachment 2, and the calculations in Attachment 10 support these changes.

Attachment 5 Page 2 of 10 Technical Specification changes to the following sections and tables are proposed:

Index Table 3.3-6 Table 4.3-3 Section 3/4.7.11 Section 3/4.9.4 Section 3/4.9.8 BASES 3/4.4.5 BASES 3/4.4.6.2 BASES 3/4.6.1.1 BASES 3/4.7.1.4 BASES 3/4.7.10 BASES 3/4.7.11 BASES 3/4.9.4 BASES 3/4.9.8 BASES 3/4.9.9 ADMINISTRATIVE CONTROLS 6.8.4.1

Attachment 5 Page 3 of 10 Table 5-1: Proposed Technical Specification and Bases Chanimes Description and Safety Assessment for Specific Changes to TS and TS Bases Change Current Technical Specification: Proposed Change:

1. 1 INDEX ENTRIES MAIN SECTION ENTRIES: 3/4.7.11 SPENT INDEX, MAIN SECTION ENTRIES: 3/4.7.11 deleted, FUEL POOL VENTILATION SYSTEM, 3/4.9.4 3/4.9.4 deleted and 3/4.9.8 deleted.

REACTOR BUILDING PENETRATIONS and 3/4.9.8, REACTOR BUILDING PURGE AND EXHAUST ISOLATION SYSTEM BASES SECTION ENTRIES: 3/4.7.11 SPENT INDEX, BASES SECTION ENTRIES: 3/4.7.11 deleted, FUEL POOL VENTILATION SYSTEM, 3/4.9.4 3/4.9.4 deleted and 3/4.9.8 deleted.

REACTOR BUILDING PENETRATIONS and 3/4.9.8, REACTOR BUILDING PURGE AND EXHAUST ISOLATION SYSTEM Change Basis / Safety Assessment:

1. 1 The index changes are administrative and reflect changes to the VCSNS Limiting Conditions for Operations and Surveillance Requirements and their associated Bases that are supported by this submittal. The basis for each technical change is discussed below.

Change Current Technical Specification: Proposed Change:

1. 2 TABLE 3.3-6, RADIATION MONITORING Delete these entries from the Table.

INSTRUMENTATION, INSTRUMENT, 1, AREA MONITORS, b. Reactor Building Manipulator Crane Area (RM-G17A or RM-G17B) and associated Action 28 Change Basis / Safety Assessment:

1. 2 Calculation DC00040-102, "Fuel Handling Accidents - AST" (see Attachment 10). This new dose analysis, performed in accordance with Regulatory Guide 1.183, assumes (1) no filtration or radionuclide removal for the release and (2) the containment is open for the duration of the event. Under these assumptions, the resulting Control Room and offsite doses are within the regulatory limits of 10CFR50.67. Other accidents considered during Mode 6 are inadvertent criticality due to boron dilution and inadvertent loading of a fuel assembly in an improper location. Neither of these accidents are postulated to result in fuel cladding integrity damage. Since the only accident postulated to occur during Mode 6 that results in a significant radioactive release is the Fuel Handling Accident, the proposed Technical Specification change omitting the use of RM-G17A or RM-G17B to initiate automatic isolation of the Containment Purge and Exhaust is justified. This change is consistent with TSTF-51 which allows deletion of OPERABILITY requirements during CORE ALTERATIONS for ESF mitigation features previously credited for the FHA.

Attachment 5 Page 4 of 10 Description and Safety Assessment for Specific Changes to TS and TS Bases Change Current Technical Specification: Proposed Change:

1. 3 TABLE 3.3-6, RADIATION MONITORING Delete these entries from the Table.

INSTRUMENTATION, INSTRUMENT, 2, PROCESS MONITORS, a. Spent Fuel Pool Exhaust - Ventilation System (RM-A6),

i. Gaseous Activity, ii. Particulate Activity and associated Action 27 Change Basis / Safety Assessment:

1. 3 Calculation DC00040-102, "Fuel Handling Accidents - AST" (see Attachment 10). This new dose analysis, performed in accordance with Regulatory Guide 1.183, assumes no filtration or radionuclide removal for the Fuel Handling Building release, and the resulting Control Room and offsite doses are within the regulatory limits of 10CFR50.67. Other than the Fuel Handling Accident, no other accident involving movement of irradiated fuel in the spent fuel pool and during crane operation with loads over the pool are predicted/postulated to result in a significant radioactive release. Therefore, removal of the OPERABILITY requirements for the Spent Fuel Pool Exhaust Ventilation System radiation monitor (RM-A6) is justified. This is consistent with intent of TSTF-51 which allows deletion of OPERABILITY requirements during CORE ALTERATIONS for ESF mitigation features previously credited for the FHA.

Change Current Technical Specification: Proposed Change:

1. 4 TABLE 3.3-6, RADIATION MONITORING Delete this entry from the Table.

INSTRUMENTATION, INSTRUMENT, 2, PROCESS MONITORS, b. Containment, i.

Gaseous Activity - Purge & Exhaust Isolation (RM-A4)

Change Basis / Safety Assessment:

1. 4 New dose analyses (see Attachment 10) for the six analyzed accidents are enclosed and proposed for use in Chapter 15 of the FSAR. The new analyses are performed utilizing the guidance of Regulatory Guide 1.183 to meet the 10 CFR 50.67 regulatory requirements. During the FHA, the containment is assumed to be open for the duration of the event. Other accidents considered during Mode 6 are inadvertent criticality due to boron dilution and inadvertent loading of a fuel assembly in an improper location. Neither of these accidents are postulated to result in fuel cladding integrity damage. Since the only accident postulated to occur during Mode 6 that results in a significant radioactive release is the Fuel Handling Accident, the proposed Technical Specification change omitting the use of RM-A4 to initiate automatic isolation of the Containment Purge and Exhaust is justified. This change is consistent with TSTF-51 which allows deletion of OPERABILITY requirements during CORE ALTERATIONS for ESF mitigation features previously credited for the FHA.

Attachment 5 Page 5 of 10 Description and Safety Assessment for Specific Changes to TS and TS Bases Change Current Technical Specification: Proposed Change:

1. 5 TABLE 4.3-3, RADIATION MONITORING Delete these entries from the Table.

INSTRUMENTATION SURVEILLANCE REQUIREMENTS, INSTRUMENT 1, AREA Monitors, b. Reactor Building Manipulator Crane Area (RM-G17A or RM-G17B)

Change Basis / Safety Assessment:

1. 5 Change 2 above deletes the operational requirements of process radiation monitors RM-G17A and RM-G17B since automatic isolation of the Containment Purge and Exhaust is no longer required for mitigation of the Fuel Handling Accident. For the same reasons. documented under Change 2, the associated surveillance requirements for RM-G17A and RM-G17B in Table 4.3-3 are no longer required.

Change Current Technical Specification: Proposed Change:

1. 6 TABLE 4.3-3, RADIATION MONITORING Delete these entries from the Table.

INSTRUMENTATION SURVEILLANCE REQUIREMENTS, INSTRUMENT, 2, PROCESS MONITORS, a. Spent Fuel Pool Exhaust - Ventilation System (RM-A6),

i. Gaseous Activity and ii. Particulate Activity Change Basis / Safety Assessment:

1. 6 Change 3 above deletes the operational requirements of process radiation monitor RM-A6 since filtration or radionuclide removal for the Fuel Handling Building release is no longer required for mitigation of the Fuel Handling Accident. For the same reasons documented under Change 3, the associated surveillance requirements for RM-A6 in Table 4.3-3 are no longer required.

Change Current Technical Specification: Proposed Change:

1. 7 TABLE 4.3-3, RADIATION MONITORING Delete this entry from the Table.

INSTRUMENTATION SURVEILLANCE REQUIREMENTS, INSTRUMENT, 2, PROCESS MONITORS, b. Containment, i. Gaseous Activity -

Purge & Exhaust Isolation (RM-A4)

Change Basis / Safety Assessment:

1. 7 Change 4 above deletes the operational requirements of process radiation monitor RM-A4 since automatic isolation of the Containment Purge and Exhaust is no longer required for mitigation of the Fuel Handling Accident. For the same reasons documented under Change 4, the associated Mode 6 surveillance requirements for RM-A4 in Table 4.3-3 are no longer required.

Attachment 5 Page 6 of 10 Description and Safety Assessment for Specific Changes to TS and TS Bases Change Current Technical Specification: Proposed Change:

1. 8 3/4.7.11 SPENT FUEL POOL VENTILATION Delete this section in its entirety.

SYSTEM, LIMITING CONDITIONS FOR OPERATION and SURVEILLANCE REQUIREMENTS Change Basis / Safety Assessment:

1. 8 Calculation DC00040-102, "Fuel Handling Accidents - AST" (see Attachment 10). This new dose analysis, performed in accordance with Regulatory Guide 1.183, assumes no filtration or radionuclide removal for the Fuel Handling Building release, and the resulting Control Room and offsite doses are within the regulatory limits of 10CFR50.67. Other than the Fuel Handling Accident, no other accident involving movement of irradiated fuel in the spent fuel pool and during crane operation with loads over the pool are predicted/postulated to result in a significant radioactive release. Therefore, removal of the OPERABILITY requirements for the Spent Fuel Pool Ventilation Sub-Systems is justified. This is consistent with intent of TSTF-51 which allows deletion of OPERABILITY requirements during CORE ALTERATIONS for ESF mitigation features previously credited for the FHA.

Change Current Technical Specification: Proposed Change:

1. 9 REFUELING OPERATIONS, 3/4.9.4 REACTOR Delete this section in its entirety.

BUILDING PENETRATIONS Change Basis / Safety Assessment:

Calculation DC00040-102, "Fuel Handling Accidents - AST" (see Attachment 10). This new dose analysis,

1. 9 performed in accordance with Regulatory Guide 1.183, assumes no filtration, holdup, or radionuclide removal for the Reactor Building release, and the resulting Control Room and offsite doses are within the regulatory limits of 10CFR50.67. Other accidents considered during CORE ALTERATIONS or movement of irradiated fuel within the reactor building are inadvertent criticality due to boron dilution and inadvertent loading of a fuel assembly in an improper location. Neither of these accidents is postulated to result in fuel cladding integrity damage. Since the only accident postulated to occur that results in a significant radioactive release is the Fuel Handling Accident, the proposed Technical Specification change omitting OPERABILITY requirements for Reactor Building Penetrations is justified. This change is consistent with TSTF-5 1 which allows deletion of OPERABILITY requirements during CORE ALTERATIONS for ESF mitigation features previously credited for the FHA.

Attachment 5 Page 7 of 10 Descrintion and Safety Assessment for Snecitic Changes to TS and TS Bases Change Current Technical Specification: Proposed Change:

1. 10 REFUELING OPERATIONS, 3/4.9.8 REACTOR Delete this section in its entirety.

BUILDING PURGE SUPPLY AND EXHAUST ISOLATION SYSTEM Change Basis / Safety Assessment:

Calculation DC00040-102, "Fuel Handling Accidents - AST" (see Attachment 10). This new dose analysis,

1. 10 performed in accordance with Regulatory Guide 1.183, assumes no filtration, holdup, or radionuclide removal for the Reactor Building release, and the resulting Control Room and offsite doses are within the regulatory limits of 10CFR50.67. Other accidents considered during CORE ALTERATIONS or movement of irradiated fuel within containment are inadvertent criticality due to boron dilution and inadvertent loading of a fuel assembly in an improper location. Neither of these accidents is postulated to result in fuel cladding integrity damage. Since the only accident postulated to occur that results in a significant radioactive release is the Fuel Handling Accident, the proposed Technical Specification change omitting OPERABILITY requirements for the Reactor Building Purge Supply and Exhaust Isolation System is justified. This change is consistent with TSTF-51 which allows deletion of OPERABILITY requirements during CORE ALTERATIONS for ESF mitigation features previously credited for the FHA.

Change Current Technical Specification: Proposed Change:

1. 11 REACTOR COOLANT SYSTEM, BASES, REACTOR COOLANT SYSTEM, BASES, STEAM GENERATOR TUBE INTEGRITY, STEAM GENERATOR TUBE INTEGRITY, Applicable Safety Analyses, 2 nd paragraph, "The Applicable Safety Analyses, 2 nd paragraph, "The dose consequences of these events are within the dose consequences of these events are within the limits of GDC 19 (Reference 2), 10 CFR 100 limits of GDC 19 (Reference 2), 10 CFR 50.67 (Reference 3) or the NRC approved licensing basis (Reference 3) or the NRC approved licensing (e.g., a small fraction of these limits)." basis (e.g., a small fraction of these limits)."

Change Basis / Safety Assessment:

The enclosed steam generator tube rupture dose (see Attachment 10, Calculation DC00040-98, "Steam

1. 11 Generator Tube Rupture - AST") analysis is performed utilizing the guidance of Regulatory Guide 1.183 to meet the 10 CFR 50.67 regulatory requirements. This change updates the BASES section to reflect the appropriate regulatory requirements.

Change Current Technical Specification: Proposed Change:

1. 12 REACTOR COOLANT SYSTEM, BASES, REACTOR COOLANT SYSTEM, BASES, STEAM STEAM GENERATOR TUBE INTEGRITY, GENERATOR TUBE INTEGRITY, References, 3, 10 References, 3, 10 CFR 100, "Reactor Site Criteria" CFR 50.67, "Accident Source Term" Change Basis / Safety Assessment:

The enclosed steam generator tube rupture dose (see Attachment 10, Calculation DC00040-98, "Steam

1. 12 Generator Tube Rupture - AST") analysis is performed utilizing the guidance of Regulatory Guide 1.183 to meet the 10 CFR 50.67 regulatory requirements. This change supports change #11 above by updating Reference 3 to reflect the appropriate regulatory requirements.

Attachment 5 Page 8 of 10 Description and Safety Assessment for Specific Changes to TS and TS Bases Change Current Technical Specification: Proposed Change:

1. 13 REACTOR COOLANT SYSTEM, BASES, REACTOR COOLANT SYSTEM, BASES, OPERATIONAL LEAKAGE, Applicable Safety OPERATIONAL LEAKAGE, Applicable Safety Analyses, 4th paragraph, "The dose consequences Analyses, 4h paragraph, "The dose consequences resulting from the SLB accident are well within resulting from the SLB accident are well within the limits the limits defined in 10 CFR 100, or the staff defined in 10 CFR 50.67, or the staff approved licensing approved licensing basis (i.e., a small fraction of basis (i.e., a small fraction of these limits)."

these limits)."

Change Basis / Safety Assessment:

The enclosed main stream line break dose (see Attachment 10, Calculation DC00040-99, "Main Steam Line

1. 13 Break - AST") analysis is performed utilizing the guidance of Regulatory Guide 1.183 to meet the 10 CFR 50.67 regulatory requirements. This change updates the BASES section to reflect the appropriate regulatory requirements.

Change Current Technical Specification: Proposed Change:

1. 14 CONTAINMENT SYSTEMS, BASES, 3/4.6.1.1 CONTAINMENT SYSTEMS, BASES, 3/4.6.1.1 CONTAINMENT INTEGRITY, "This restriction, CONTAINMENT INTEGRITY, "This restriction, in in conjunction with the leak rate limitation, will conjunction with the leak rate limitation, will limit site limit site boundary radiation doses to within the boundary radiation doses to within the limits of 10 CFR limits of 10 CFR 100 during accident conditions." 50.67 during accident conditions."

Change Basis / Safety Assessment:

New dose analyses (see Attachment 10) for the six analyzed accidents are enclosed and proposed for use in

1. 14 Chapter 15 of the FSAR. The new analyses are performed utilizing the guidance of Regulatory Guide 1.183 to meet the 10 CFR 50.67 regulatory requirements. This change updates the BASES section to reflect the appropriate regulatory requirements.

Change Current Technical Specification: Proposed Change:

1. 15 PLANT SYSTEMS, BASES, 3/4.7.1.4 PLANT SYSTEMS, BASES, 3/4.7.1.4 ACTIVITY, "The ACTIVITY, "The limitations on secondary system limitations on secondary system specific activity ensure specific activity ensure that the resultant offsite that the resultant offsite radiation dose will be limited to radiation dose will be limited to a small fraction of 10 CFR 50.67 limits in the event of a steam line rupture."

10 CFR 100 limits in the event of a steam line rupture."

Change Basis / Safety Assessment:

New dose analyses (see Attachment 10) for the six analyzed accidents are enclosed and proposed for use in

1. 15 Chapter 15 of the FSAR. The new analyses are performed utilizing the guidance of Regulatory Guide 1.183 to meet the 10 CFR 50.67 regulatory requirements. This change updates the BASES section to reflect the appropriate regulatory requirements.

Attachment 5 Page 9 of 10 Descrintion and Safety As~e~sment for Snecific Chanees to TS and TS Bases Change Current Technical Specification: Proposed Change:

1. 1 PLANT SYSTEMS, BASES, 3/4.7. 10 WATER PLANT SYSTEMS, BASES, 3/4.7. 10 WATER LEVEL -

LEVEL - SPENT FUEL POOL, "The restrictions SPENT FUEL POOL, "The restrictions on minimum on minimum water level ensure that sufficient water level ensure that sufficient water depth is available water depth is available to remove 99% of the to remove 99.5% of the assumed 16% 1-131 and 10%

assumed 10% iodine gap activity released from the other halogens gap activity released from the rupture of rupture of an irradiated fuel assembly. The an irradiated fuel assembly. The minimum water depth is minimum water depth is consistent with the consistent with the assumptions of the accident analysis."

assumptions of the accident analysis."

Change Basis / Safety Assessment:

1. 6 A new dose analysis (see Attachment 10, DCOOO40-102, "Fuel Handling Accidents - AST") for the Fuel

1. 6 Handling Accident is enclosed and proposed for use in Chapter 15 of the FSAR. The analysis is performed utilizing the guidance of Regulatory Guide 1.183 to meet the 10 CFR 50.67 regulatory requirements. The analysis utilizes iodine removal factors and gap activities that differ from the CLB assumptions currently shown in BASES Section 3/4.7. 10. This change updates the BASES section to reflect the revised analysis assumptions.

Change Current Technical Specification: Proposed Change:

1. 17 PLANT SYSTEMS, BASES, 3/4.7.11 SPENT Delete this section in its entirety.

FUEL POOL VENTILATION SYSTEM Change Basis / Safety Assessment:

97 Change #8 deletes all Limiting Conditions for Operation and Surveillance Requirements for the Spent Fuel

1. 7 Pool Ventilation System since they are no longer needed for mitigation of the Fuel Handling Accident. This change deletes the associated BASES section (3/4.7.11) for this specification.

Change Current Technical Specification: Proposed Change:

1. 18 REFUELING OPERATIONS, BASES, 3/4.9.4 Delete this section in its entirety.

REACTOR BUILDING PENETRATIONS Change Basis / Safety Assessment:

1. 8 Change #9 deletes all Limiting Conditions for Operation and Surveillance Requirements for Reactor Building

1. 8 Penetrations since they are no longer needed for mitigation of the Fuel Handling Accident. This change deletes the associated BASES section (3/4.9.4). __________________________

Change Current Technical Specification: Proposed Change:

1. 19 REFUELING OPERATIONS, BASES, 3/4.9.8 Delete this section in its entirety.

REACTOR BUILDING PURGE SUPPLY AND EXHAUST ISOLATION SYSTEM_________________

Change Basis / Safety Assessment:

1. 9 Change #10 deletes all Limiting Conditions for Operation and Surveillance Requirements for the Reactor

1. 9 Building Purge Supply and Exhaust Isolation System since the isolation function is no longer needed for

____mitigation of the Fuel Handling Accident. This change deletes the associated BASES section (3/4.9.8).

Attachment 5 Page 10 of 10 Deserintion and Safetv Assessment for Snecific Chana es to TS and TS Bases Change Current Technical Specification: Proposed Change:

1. 20 REFUELING OPERATIONS, BASES, 3/4.9.9 REFUELING OPERATIONS, BASES, 3/4.9.9 WATER WATER LEVEL - REACTOR VESSEL, "The LEVEL - REACTOR VESSEL, "The restrictions on restrictions on minimum water level ensure that minimum water level ensure that sufficient water depth is sufficient water depth is available to remove 99% available to remove 99.5% of the assumed 16% 1-131 and of the assumed 10% iodine gap activity released 10% other halogens gap activity released from the rupture from the rupture of an irradiated fuel assembly. of an irradiated fuel assembly. The minimum water depth The minimum water depth is consistent with the is consistent with the assumptions of the accident assumptions of the accident analysis." analysis."

Change Basis / Safety Assessment:

A new dose analysis (see Attachment 10, DC00040-102, "Fuel Handling Accidents - AST") for the Fuel

1. 20 Handling Accident is enclosed and proposed for use in Chapter 15 of the FSAR. The analysis is performed utilizing the guidance of Regulatory Guide 1.183 to meet the 10 CFR 50.67 regulatory requirements. The analysis utilizes iodine removal factors and gap activities that differ from the CLB assumptions currently shown in BASES Section 3/4.9.9. This change updates the BASES section to reflect the revised analysis assumptions.

Change Current Technical Specification: Proposed Change:

1. 21 ADMINISTRATIVE CONTROLS, 6.8.4.1 The Spent Fuel Ventilation System is removed from the Ventilation Filter Testing Program (VFTP) scope of the VFTP.

Change Basis / Safety Assessment:

Change #8 deletes all Limiting Conditions for Operation and Surveillance Requirements for the Spent Fuel

1. 21 Pool Ventilation System since they are no longer needed for mitigation of the Fuel Handling Accident. This change deletes the Spent Fuel Pool Ventilation System from the VFTP defined by ADMININSTRATIVE CONTROL 6.8.4.1. Removal of current inplace and laboratory filter tests requirements for the Spent Fuel Pool Ventilation Sub-Systems is, likewise, justified since the system is no longer needed for mitigation of the Fuel Handling Accident.

Attachment 6 Proposed Technical Specification Changes (Mark-ups)

Attachment 6 Page 1 of 13 Table 6-1: List of Proposed Technical Specification Changes (Marked ups)

Table 5-4 Change # Sections Title 1 TOC Index 2,3 & 4 Table 3.3-6 RADIATION MONITORING INSTRUMENTATION 5,6 & 7 Table 4.3-3 RADIATION MONITORING INSTRUMENTATION SURVEILLANCE REQUIREMENTS 8 3/4.7.11 SPENT FUEL POOL VENTILATION SYSTEM 9 3/4.9.4 REFUELING OPERATIONS REACTOR BUILDING PENETRATIONS 10 3/4.9.8 REACTOR BUILDING PURGE SUPPLY AND EXHAUST.

ISOLATION SYSTEM 21 6.8.4.1 ADMINISTRATIVE CONTROLS -Ventilation Filter Testing I I_Program (VFTP)

Attachment 6 Page 2 of 13 INDEX LIMITING CONDITIONS FOR OPERATION AND SURVEILLANCE REQUIREMENTS SECTION PG 34,7 PLANT SYSTEMS 3/4.7.1 TURBINE CYCLE Safety Valves . ...... . ........ .......... ........... 3t47.1 Emergency Feedwater System.......... ........ ........... 3/474 Condensate Storage Tank ......... ................. 3/4 7.6 Activity ................................... ..................... ............. .... ..... 3/4 7-7 Main Steam Line Isolation Valves ............... ............... 3/47-9 Feedwater Isolation Valves ................................ 3/4 7-ga 3/4.7T2 STEAM GENERATOR PRESSURE/TEMPERATURE LIMITATION..... 3/4 7-10 314173 COMPONENT COOLING WATER SYSTEM ....... ........... 34 7-11 3/4.7.4 SERVICE WATER SYSTEM .... .......... ........................ 3/4 7.12 3/4,7.5 ULTIMATE HEAT SINK ......... ......................... 3/47-13 3/4.7.6 CONTROL ROOM EMERGENCY FILTRATION SYSTEM (CREFS).... 3/4 7-14 3/4.7.7 SNUBBERS .......................... 3.4 7-16 3/4.7.8 SEALED SOURCE CONTAMINATION........................ ......... 3/47-23 314.7,9 AREA TEMPERATURE MONITORING ..... ................ 34 7-37 3/4.7A10 WATER LEVEL - SPENT FUEL POOL ........................... 34 7-39 3/4.7,11 S NT E 0 XENT 110n Ys m_. ./ .. ... .

3/41.712 SPENT FUEL ASSEMBLY STORAGE-..__......-....34 2 314,7.13 SPENT FUEL POOL BORON CONCENTRATION ......... .......... 314 7-44

' ,*,4_4~4 SUMMER - UNIT I Vil Amnendment No. 2 , 0;

Attachment 6 Page 3 of 13 INDEX UMITING CONDITIONS FOR OPERATION AND SURVEILLANCE REQUIREMENTS SECTION PAGE 3/4,9 REFUELING OPERATIONS 3/4.9.1 BORON CONCENTRATION ............................................................. 34 9-1 3/4.9.2 INSTRUMENTATION ...................................... 3/4 9-2 3/4.9.3 DECAY TIME ................................................................................. . 3/49-3 3/4.9.5 COMMUNICATIONS ............... . .................................... 3/4*9*-5 3/4.9.6 MANIPULATOR CRANE .................................................................... 3/4 9-6 3/4.9.7 RESIDUAL HEAT REMOVAL AND COOLANT CIRCULATION High W ater Level ....................................... ..................... .... I................ 3/4 9-7 Low W ater Level ...................................................... ............................ 34 314.9.9 WATER LEVEL -REFUELING CAVITY AND FUEL TRANSFER CANAL. 3149-10 3/4.10 SPECIAL TEST EXCEPTIONS 3/4.10 1 SHUTDOWN MARGIN ................................. . ................................... 3/4 10-1 3(4.10.2 GROUP HEIGHT, INSERTION AND' POWER DISTRIBUTION UMITS 3(4,10-2 3(4.10.3 PHYSICS TESTS ............................................................. ............... 3/41 04 I 3/4.10.4 REACTOR COOLANT LOOPS .........................................

. 3/410-4 314.10.5 POSITION INDICATION SYSTEM - SHUTDOWN ................. I.......3/4110-5 SUMMER - UNIT I X Amendment No. 4% 160

Attachment 6 Page 4 of 13 INDEX BASES SECTION PAGE 314.7 PLANT SYSTEMS 314.7.1 TURBINE CYCLE ............ ................ ...................... B 3/4 7-1 3/4.7.2 STEAM GENERATOR PRESSURE/TEMPERATURE LIMITATION..... 8 3/4 7-3 Y'4.7.3 COMPONENT COOLING WATER SYSTEM ........ -. B 3/4 7-3 314J7.4 SERVICE WATER SYSTEM ............. ................... B 3/4 7-3 3/417.5 ULTIMATE HEAT SINK ........................................ ...... ....... . 8 3Y4 7-3 314.7.6 CONTROL ROOM EMERGENCY FILTRATION SYSTEM (CREFS) .... B 3/4 7-4 3)4.7.7 SNUBBERS ............................................. . B 314 7-40 3/4,7.8 SEALED SOURCE CONTAMINATION ........ ................. B 314 7-6 3/417.9 AREA TEMPERATURE MONITORING ........................ 8......

34 7-6 3/4.7.10 WATER LEVEL - SPENT FUEL POOL ........................... . ........ B 3/47-6 314.7,11 S NT UE ENT...TI SY EM47 31437.12 SPENT FUEL ASSEMBLY STORAGE.............................B 314 7-7 3/4.7.13 SPENT FUEL POOL BORON CONCENTRATION ................ B 3/4 7-7 3/4.8 ELECTRICAL POWER SYSMS 3/4.8.1, 314.8.2 and 314.8.3 A.C. SOURCES, D.C. SOURCES AND ONSITE POWER DISTRIBUTION SYSTEMS .....-.............................. 83148-1 3/4.8A4 ELECTRICAL EQUIPMENT PROTECTION DEVICES.. ......... B 34 8-4 SUMMER - UNIT I XIV Amendment No. 34Oa48O.

1-7T180

Attachment 6 Page 5 of 13 INDEX BASES SECTION 3/4.9 REFUELING OPERATIONS 3/4.9.1 BORON CONCENTRATION ................................................................ B 3(4 9-1 3/4.9.2 INSTRUMENTATION, ........................................................................ B 314 9-1 3/4.9.3 DECAY TIME ........................................................................ .............. B3149-1 3/49.4 R TO UILP&tG PE T N...........

s ...... 3 9-3/4.9.5 COMMUNICATIONS ........................................ ................................ 3/49-1 3/4.9.6 MANIPULATOR CRANE .................................................................... B 3(49-2 3/4.9.7 RESIDUAL HEAT REMOVAL AND COOLANT CIRCULATION ........... 8 34 9-2 3/4.9.9 WATER LEVEL - REACTOR VESSEL ........................................................ B 3/4 9-2 3/4.10 SPEAL TEST EXCEPTIONS 3/4.10.1 SHUTDOWN MARGIN .................................................................. B 3/410-1 3/4.102 . GROUP HEIGHT, INSERTION AND POWER DISTRIBUTION LIMITS B 3/4 10-1 3/4.10.3 PHYSICS TESTS ...............-........... ......... B 3/4 10-1 3/4.10.4 REACTOR COOLANT LOOPS ... ............. .. .... B 3/410-1 3/4.10.5 POSmON INDICATION SYSTEM - SHUTDOWN ....... ........... B 3/410-1 SUMMER - UNIT I xv Amendment No. 4G, 160

TABLE 3.3-6 RADIATION MONITORING INSIRU4ENTAT[ON V-,

MINIMUM INSTRUMENT CHANNELS APPLICABLE ALARM4/TRIP MEASUREMENT OPERABLE 4DES O* SETPOINT RAGE ,, ACTION

1. AREA MONITORS I *

• 15 m/hr 10 Io4 mR/hr 25

a. Spent Fuel Pool Area (RWGS)

_ _IU b.

a.

I.

2. ii.~

b. Containment 0*

I , Ga6soiw AGW 5x back ou% y 10W6 It. Particulate and Gaseous Activity (RH-AZ) - 1 1. 2, 3 & 4 N/A 10 - 106Cpm 26 RCS Leakage Detection

c. Control Room Isolation (RH-AI) 1 ALL MODES S 2 x background 10 - 106cpm 29

... _ Wth fuel inutestorage tor ool buildi*n 00~ ~prtIcO;ia ~~I'aual~taial.t. u~a ~~-is~~i' sgizz-~pstp"twl t -.1J

Attachment 6 Page 7 of 13 f4STRUI4ENATION TABLE 3.3-6 (Continued)

ACTION STATEMENTS EL CITED ACTION 25 - With the number of channels OPERABLE less than required by the Minioun Channels OPERABLE requiremnt, perform area surveys of the monitored area with portable monitoring instrumentation at least once per 24 hours2.777778e-4 days <br />0.00667 hours <br />3.968254e-5 weeks <br />9.132e-6 months <br />.

ACTION 26 - With the number of channels OPERABLE less than required by the Minimum Channels OPERABLE requirement, comply with the ACTION requirements of Specification-3.4.6.l.

ACTION 27 lith nh~umber oca%;nnls WeRUE less an eul b ACTION 28 - iiith the-Number of-iftuAnels 6PEAIjLaýr s than ýre04ýd by ýth

%F1.4 een%' r-"ý'.5~n nfl -7 -. ~ *f -

ACTION 2ý9 - With the number of channels OPERABLE less than required by the Minimum Channels OPERABLE requirement, within I hour initiate and maintain operation of the control room emergency ventilation I

system in the emergency mode of operation.

ACTION 30- With the number of OPERABLE channels less than required by the Minimum Channels OPERABLE requirement, either restore the inoperable Channel(s) to OPERABLE Status within 72 hours8.333333e-4 days <br />0.02 hours <br />1.190476e-4 weeks <br />2.7396e-5 months <br />, or:

1) Initiate the preplanned alternate method of monitoring the appropriate parameter(s), and

2) Prepare and submit a Special Report to the Commission pursuant to Specification 6.9.2 within 14 days following the event outlining the action taken, the cause of the inoperability and the plans and schedule for restoring the system to OPERABLE status.

IMER - UNIT 1 3/4 3-44

RADIATION MONITORING INSTRUNENTATIOM SURVEILLANCE REQUIREMENTS ANALOG CHANNEL MODES FOR WHICH CHANNEL CHANNEL OPERATIONAL SURVEILLANCE

-41 INSTRUMENT CH=C CALIBRATION TEST I$ REQUIRED 1.

i ,, ,,

2.

G ýLEcT JEr ii. rarticuiate and taseous Activity - RCS Leakage S K 1, 2. 3 & 4 Detection (RM-A2) K

c. Control Room Isolation (JRNAI) S R All MODES

d. Noble Gas Effluent Monitors (High Range) t4

i. Main Plant Vent (RW-A13) S R 1. 2. 3 & 4 Ii. Main Stem Lines (R"-019A. 8. C) S R 1, 2, 3 & 4

!iM.Reactor building Purge Supply & Exhaust 0

System (RM-A14) S R 1 1. 2. 3 & 4 CDo I-. _ - uel stre pooI or 1bdt*

4 00

Attachment 6 Page 9 of 13 PLANT SYSTEMS 3/4.7JI N ULP ENIAINS E LMIIN JCONDITION FO/ PRTO 3.7.1) Two independe spentfuelpwvbOEALwt al t, .one sub-syste inoeain IC enever irradiated *eis being movedinth spent fuel pool and d rane operation loads over the

a. ith one spent fuel I ventilation sub-sys inoperable, fuel m ment within the spent fue pool or crone operatio ith loads over the s nt fuel pool may proceed pry 'od the OPERABLE s nt fuel pool ventiatio ub-system is capable of being owered from an OPE BLE emergency p r source and is inoperation a discharging through at ast one lrain of HEP filters and charcoal ads" ers.

b. With no sp t fuel pool ventilation b-system OPERABL , suspend all' operation olvlng movement of uel within the spent f I pool or crane operatio with loads over the sp t fuel pool.

c' The p vIsions of Specnicatpio 3.0.4are notsapplica e.

SURVEILLANO REQUIREMNT 4.7.11 The OPRB.....

a ove required spent fu

.. ,o:~,o pool ventilation sub ystems shalt be de

.. ,o,,

At least once per 3 days by initiating, fro the control room, ow through the nslrated I

HEPA tilers and rcoal adsorbers an verifying that each ub-system operates for at I st 15 minutes.

b. By performin equired filter testing accordance with Ventilation Filter Testing P am JVFTP).

C. At least o e per 18 months by:

1. rifyirng that on a to* of oftsite power t I ignal. the sse

• ~automatically starts./

Z Verifying that the ystem main~tains I spent fue pool are al

~~negative pressu* greater than or e Itto I/ inhe s WerGug relative to te tsd atmosphere/ rn SUMMER - UNIT 1 3/4 7-40 Amendment No. 42,46, 4 "180

Attachment 6 Page 10 of 13 REFUELING OPERATIONS 3/.. RE O BUILOI PNTAIN LIMITIN CONDITIONi/R/ OPERATION!

3. The retrbuilding nerations *all be in floigsatus:

a. T e equipment do closed an held in pia by a minimum of four

b. A minimum of ne door in ach airlock s closed, and Each penet ation provi ng direct a ess from the actor building atmosphe to the out de atmosph eshall be eith r:

1. C sed by an i olation val , blind flange, or mnual valve, or

2. Be capable being cls by an OPRBEatmtcRatr I Bilding and Exh stisolation va e.

APPLICABI TY: During ORE ALTERAT S or movement o irradiated fuel wi in the Ea o r buildih vi w the itn10 ours or spet at ion tdsatI*ioed, per da $

u Ilin eallthe ALTpe aions S invol/ng CORE raaddate or movement of d rector buildi SURVEI LL C EUIE4 4/.9.o Each of" the ov required reac rbiding penetr ins shall be 9det!* or capable of be g

, nred t°obe ei {herRE in its closed/ solated conditiono c 1,d by an OP Eautomatic Reac br Building Purge nd Exhaust isola o Ive wihi O00 ours prior to th atof and at I toncepe7ds ur;COEA T ION rmvmno irradiated fue in the rectruilding

/by: ao /

a. Ve ifying the penetra ons are in their losed/isolated Testing the Purge a B of Spcfcion Exhaust iso3a per SUMb. the 1ac applicable poions 4.6.4.2.

SUMMER -- UNIT 1 3/4 9-4

Attachment 6 Page 11 of 13 REFUELING OPERATION$

3. .8 The R ctor Bull Ing Purge SSup y and Exh ust Isol tion System s 11 be SERABLE.

APPLICAB ITY: Our CORE ALTE IONS or ement of irradiated f within the co mnt.

Wi the Rea or Building P rge Supply ridExha t Isolation Sy t inoperable, cl-se each the Purge a Exhaust tratio providing di access from a reacto!Ionbuilding pacific 3.0.4 ara notsphere to a. outsi applic atmosphere. T provisions of 4.9.8 The Rea or Buildi Supply and haust Isolatio System shall demonstrated ERASLE wi I hours prior o the start of and at least rce per 7 days ring CORE ALT RATION$ by veri ng that React Building Pur Supply and haust on man I initiation, n a high radi tion on occurs byla test sig from each of the Contanment adiation monit Pig iistrume tion channel . and by verify ng that isolatio occurs on the 6-inch lines f the Purge ly and Exh4 t Isolation Syst on a high ra atlon test s nal from the r actor building anipulator crane area channels.

ih',, 56Cr./ 4/&

SUMMER - UNIT 1 3/4 9-9 Amendment No. 19

Attachment 6 Page 12 of 13 ADMINISTRATIVE CONTROLS location of flaws to which the tubes may be susceptible and, based on this assessment, to determine which inspection methods need to be employed and at what locations.,

a) Inspect 100% of the tubes in each SG during the first refueling outage following SG replacement.

b) Inspect 100% of the tubes at sequential periods of 144, 108, 72 and thereafter, 60 effective full power months. The first sequential period shall be considered to begin after the first inservice inspection of the SGs. In addition, inspect 50% of the tubes by the refueling outage nearest the midpoint of the period and the remaining 50% by the refueling outage nearest the end of the period. No SG shall operate for more than 72 effective full power months or three refueling outages (whichever is less) without being inspected.

c) If crack indications are found In any SG tube, then the next inspection for each SG for the degradation mechanism that caused the crack indication shall not exceed 24 effective full power months or one refueling outage (whichever Is less). If definitive information, such as from examination of a pulled tube, diagnostic non-destructive testing, or engineering evaluation indicates that a crack-like indication is not associated with a crack(s), then the indication need not be treated as a crack.

5. Provisions for monitoring operational primary-to-secondary leakage.

Ventilation Filter Testing ProQram (VFTP)

A program shall be established to implement the following required testing of Engineered Safety Feature (ESF) filter ventilation systems at the frequencies specified in accordance with Regulatory Guide 1.52, Revision 2 and ASME N51 0-1989.

1. Demonstrate for each of the ESF systems that an inplace test of the high efficiency particulate air (HEPA) filters shows a penetration and system bypass

< 0.05% when tested in accordance with Regulatory Guide 1.52, Revision 2, and ASME N510-1989 at the system flowrate specified below t 10%.

ESF Ventilation System Flowrate Control Room Emergency Filtration System 21,270 SCFM Reactor Building Cooling Units 60,270 ACFM SUMMER - UNIT I 6-12f Amendment No. 4744 80

Attachment 6 Page 13 of 13 ADMINISTRATIVE CONTROLS

2. Demonstrate for each of the ESF systems that an inplace test of the charcoal adsorber shows a penetration and system bypass < 005% when tested in accordance with Regulatory Guide 1,52, Revision 2, and ASME N510-1989 at the system flowrate specified below

• 10%.

ESF Ventilation System Flowrate Control Room Emergency Filtration System 21,270 SCFM

- e Fuel P..o.e ...V i ,,l,,

i.... v , .r. A" 4V1 3, Demonstrate for each of the ESF systems that a laboratory test of a sample of the charcoal adsorbef, when obtained as described in Regulatory Guide 1.52, Revision 2, shows the methyl iodide penetration less than the value specified below when tested in accordance with ASTM D3803-1989 at a temperature of 30'C (86*F) and the relative humidity specified below.

ESF Ventilation System Penetration RH Face Velocity (fps)

Control Room <2.5% 70% 0,667

4. Demonstrate for each of the ESF systems that the pressure drop across the combined HEPA filters, the prefilters, and the charcoal adsorbers is less than the value specified below when tested in accordance with Regulatory Guide 1.52, Revision 2, and ASME N510-1989 at the system flowrate specified below t 10%.

ESF Ventilation System Delta P Flowrate Control Room <6 in. W.G. 21,270 SCFM

• spe~ ruc rii I, W.G. 30,000

• _FMD6 (LCT6)

Reactor Building Cooling Units <3 in. WG. 60,270 ACFM The provisions of SR 4,0.2 and SR 4.0.3 are applicable to the VFTP test frequencies.

m, Control Room Envelope Habitability Program A Control Room Envelope (CRE) Habitability Program shall be established and implemented to ensure that CRE habitability is maintained such that, with an OPERABLE Control Room Emergency Filtration System (CREFS). CRE occupants can control the reactor safely under normal conditions and maintain it in a safe condition following a radiological event, hazardous chemical release, or a smoke challenge. The program shall ensure that adequate radiation protection is provided to permit access and occupancy of the CRE under design basis accident (OBA) conditions without personnel receiving radiation exposures in excess of 5 rem whole body or its equivalent to any part of the body for the duration of the accident The program shall include the following elements:

SUMMER - UNIT 1 6-12g Amendment No. 180

Attachment 7 For Information -

Proposed Technical Specification Bases Changes Mark-ups

Attachment 7 Page 1 of 9 Table 7-1: List of Proposed Technical Specification Bases Changes (Marked ups)

Table 5-1 Change # Sections Title 11 & 12 3/4.4.5- STEAM GENERATOR TUBE INTEGRITY 13 3/4.4.6.2 OPERATIONAL LEAKAGE 14 3/4.6.1.1 CONTAINMENT INTEGRITY 15 3/4.7.1.4 ACTIVITY 16 3/4.7.10 WATER LEVEL SPENT FUEL POOL 17 3/4.7.11 SPENT FUEL POOL VENTILATION SYSTEM 18 3/4.9.4 REACTOR BUILDING PENETRATIONS REACTOR BUILDING PURGE SUPPLY AND 1EXHAUST ISOLATION SYSTEM 20 3/4.9.9 WATER LEVEL - REACTOR VESSEL

Attachment 7 Page 2 of 9 REACTOR COOLANT SYSTEM BASES STEAM GENERATOR TUBE INTEGRITY (Continued)

Aoplicable Safety Analyses The steam generator tube rupture (SGTR) accident is the limiting design basisevent for SG tubes and avoiding a SGTR is the basis for this Specification. The accident analysis for a SGTR event accounts for a bounding primary-to-secondary leakage rate equal to 1 gpm and the leakage rate associated with a double-ended rupture of a single tube. Contaminated fluid in a ruptured steam generator is only briefly released to the atmosphere as steam via the main steam safety valves. To maximize its contribution to the dose releases, the entire 1 gpm primary-to-secondary leakage is assumed to occur in the intact steam generators where it can be released during the subsequent cooldown of the plant.

The analyses for design basis accidents and transients other than a SGTR assume the SG tubes retain their structural integrity (i.e., they are assumed not to rupture). In these analyses the steam discharge to the atmosphere is based on the total primary-to-secondary leakage from all SGs of 1 gpm, or is assumed to increase to 1 gpm as a result of accident induced conditions. For accidents that do not involve fuel damage, the primary coolant activity level of DOSE EQUIVALENT 1-131 is assumed to be greater than or equal to the limits in LCO 3.4,8, "Reactor Coolant System, Specific Activity." For accidents that assume fuel damage, the primary coolant activity is a function of the amount of activity released from the damaged fuel. The dose consequences of these events are within the limits of GDC 19 (Reference 2), 10 CFR(fReference 3) or the NRC approved licensing basis (e.g., a small fraction of these li*,)*

Steam generator tube integrity satisfies Criterion 2 of 10 CFR 50.36(c)(2)(ii).

Limiting Condition for Operation (LCO)

The LCO requires that SG tube integrity be maintained. The LCO also requires that all SG tubes that satisfy the repair criteria be plugged in accordance with the Steam Generator Program.

During a SG inspection, any inspected tube that satisfies the Steam Generator Program repair criteria is removed from service by plugging. If a tube was determined to satisfy the repair criteria but was not plugged, the tube may still have tube integrity. Refer to Action a. below.

In the context of this Specification, a SG tube is defined as the entire length of the tube, including the tube wall between the tube-lo-tubesheet weld at the tube inlet and the tube-to-tubesheet weld at the tube outlet. The tube-to-tubesheet weld is not considered part of the tube.

A SG tube has tube integrity when it satisfies the SG performance criteria. The SG performance criteria are defined in Specification 6.8.4,k and describe acceptable SG tube performance. The Steam Generator Program also provides the evaluation process for determining conformance with the SG performance criteria.

SUMMER - UNIT 1 B 314 4-3a Amendment No. BRN-07-001

Attachment 7 Page 3 of 9 REACTOR COOLANT SYSTEM BASES STEAM GENERATOR TUBE INTEGRITY (Continued)

Surveillance Requirements (Continued) 4.4.5.2 During a SG inspection, any inspected tube that satisfies the Steam Generator Program repair criteria is removed from service by plugging. The tube repair criteria delineated in Specification 6.8.4.k are intended to ensure that tubes accepted for continued service satisfy the SG performance criteria with allowance for error in the flaw size measurement and for future flaw growth. In addition, the tube repair criteria, in conjunction with other elements of the Steam Generator Program, ensure that the SG performance criteria will continue to be met until the next inspection of the subject tube(s). Reference 1 provides guidance for performing operational assessments to verify that the tubes remaining in service will continue to meet the SG performance criteria.

The frequency of "Prior to entering MODE 4 following a SG inspection" ensures that the Surveillance has been completed and all tubes meeting the repair criteria are plugged prior to subjecting the SG tubes to significant primary-to-secondary pressure differential.

References

1. NEI 97-06, "Steam Generator Program Guidelines"

2. 10 CFR 50, Appendix A, GDC 19, "Control Room"

3. 10 CF 1 ,Reaifor reCer-de -J-~ - ccdel -*oe 4, ASME Boiler and Pressure Vessel Code, Section III, Subsection NB

5. Draft Regulatory Guide 1.121, "Basis for Plugging Degraded Steam Generator Tubes," August 1976

6. EPRI TR-107569, "Pressurized Water Reactor Steam Generator Examination Guidelines" SUMMER - UNIT 1 B 3/4 4-3e Amendment No, BRN-07-001

Attachment 7 Page 4 of 9 REACTOR COOLANT SYSTEM BASES OPERATIONAL LEAKAGE (Continued)

.Appllcable Safety Analyses*

Except for primary-to-secondary leakage, the safety analyses do not address operational leakage. However, other operational leakage is related to the safety analyses for a LOCA; the amount of leakage can affect the probability of such an event. The safety analysis for an event resulting in steam discharge to the atmosphere assumes that primary-to-secondary leakage from all steam generators is I gpm or increases to I gpm as a result of accident induced conditions. The LCO requirement to limit primary-to-secondary leakage through any one steam generator to less than or equal to 150 gallons per day is significantly less than the conditions assumed in the safety analysis.

Primary-to-secondary leakage is a factor in the dose releases outside containment resulting from a steam line break (SLB) accident. To a lesser extent, other accidents or transients involve secondary steam release to the atmosphere, such as a steam generator tube rupture (SGTR). The leakage contaminates the secondary fluid.

The FSAR analysis for SGTR accounts for a bounding primary-to-secondary leakage rate equal to 1 gpm and the leakage rate associated with a double-ended rupture of a single tube.

Leakage through the ruptured tube is the dominate contributor to dose releases. Since contaminated fluid in the ruptured steam generator is only briefly released to the atmosphere as steam via the main steam safety valves, the entire 1 gpm primary-to-secondary leakage is assumed to occur in the intact steam generators where it can be released during the subsequent cooldown of the plant. Overall, this pathway is a small contributor to dose releases.

The SLB is more limiting for site radiation releases. The safety analysis for the SLB accident assumes the entire 1 gpm primary-to-secondary leakage is through the effected steam generator as an initial condition. The dose consequences resulting from the SLB accident are well within the limits defined in 10 CFRt r the staff approved licensing basis (i.e., a small fraction of these limits). e , -c The RCS operational leakage satisfies Criterion 2 of 10 CFR 50,36(c)(2)(ii).

Limiting Condition for Operation (LCO)

Reactor Coolant System operational leakage shall be limited to:

a. PRESSURE BOUNDARY LEAKAGE No PRESSURE BOUNDARY LEAKAGE is allowed, being indicative of material deterioration. Leakage of this type is unacceptable as the leak itself could cause further deterioration, resulting in higher leakage. Violation of this LCO could result in continued degradation of the Reactor Coolant Pressure Boundary. Leakage past seals and gaskets is not PRESSURE BOUNDARY LEAKAGE.

SUMMER - UNIT I B 3/4 4-4a Amendment No. BRN-07-001

Attachment 7 Page 5 of 9

/

3/4.6 CONTAINMENT SYSTEMS BASES 3/4.6.1 PRIMARY CONTAINMENT 3/4.6.1.1 CONTAINMENT INTEGRITY Primary CONTAINMENT INTEGRITY ensures that the release of radioactive materials from the containment atmosphere will be restricted to those lealaqe paths and associated leak rates assumed in the accident analyes This restric-tion, in conjunction with the leakage rate limitation, wiRlmit the site boundary radiation doses to within the limits of 10 CF4 uring accident 3/4,6.12 CONTAInWENT LEAKAGE The limitations on containment leakage rates (including those used in demonstrating a 30 day water seal) ensure that the total containment volume will not exceed the value assumed in the accident analyses at the pea accident presure, Pa. As an added conswvatiam, the measured overall inte-grated lakag rate is further limited to less than or equal to 0.75 aduring performance of the peri'odic test to account for possible degradation of the containment leakage barriers between leakage test&s The surveillance testing for measuring leakage rates is consistent with the Containment Leakage Rate Testing Program.

3/4.6.1.8 REACTOR BUILDING AIR.,LOCKS The limitations on closure for the reactor building air locks are r to meet the restrictions on CONTANMDENT INTEGRITY and containment leak rate.

Surveillance testing of the air lock seals provide assurance that the overall air lock leakage will not become excessive due to seal damage during the intervals between air lock leakage tests.

SUMMER - UNIT 1 B 3/4 6-1 Amendment No. 44., n326-,

135

Attachment 7 Page 6 of 9 PLANT SYSTEMS BASES 34.7!*.2 EMERGENCY FEEDWATER SYSTEM The OPERABILITY of the emergency feedwater system ensures that the Reactor Coolant System can be cooled down to less than 350OF from normal operating conditions in the event of a total loss of off-site power.

Each emergency feedwater pump is capable of delivering a total feedwater flow of 380 gpm at a pressure of 1211 psig to the entrance of two out of three steam generators. This capacity is sufficient to ensure that adequate feedwater flow is available to remove decay heat and reduce the Reactor Coolant System temperature to less than 3501F at which point the Residual Heat Removal System may be placed into operation.

Also, each Emergency Feedwater (EFW) pump is capable of supplying 400 gpm to all 3 steam generators while the steam generators are pressurized to 1211 psig. This capacity is sufficient to ensure that the pressurizer does not overfill during a loss of normal feedwater event. The total head criteria of 3800 feet for the motor driven EFW pumps and 3140 feet for the turbine driven EFW pump includes margin that allows for a maximum EFW flow control valve leakage of 5 gpm for any one of 6 EFW flow control valves.

/47.1.3 CONDENSATE STORAGE TANK The OPERABILITY of the condensate storage tank with the minimum water volume ensures that sufficient water is available to maintain the RCS at HOT STANDBY conditions for 11 hours1.273148e-4 days <br />0.00306 hours <br />1.818783e-5 weeks <br />4.1855e-6 months <br /> with steam discharge to the atmosphere concurrent with total loss of offsite power. The contained water volume limit Includes an allowance for water not usable because of tank discharge line location or other physical characteristics.

3/4,7.1.4 ACTIVITY The limitations on secondary system specific activity ensu that the resultant offsite radiation dose will be limited to a small fraction of 10 CFR Pa t mits in the event of a steam line rupture. This dose also includes the effects of a coident 1.0 GPM primary to secondary tube leak in the steam generator of the affected ste line. These values are consistent with the assumptions used in the accident analyses SUMMER - UNIT I B 314 7-2 Amendment No. 40T.

• by-L e~/ 197?3

Attachment 7 Page 7 of 9 PLANT SYSTEMS BASES 3/4.7.8 SEALED SOURCE CONTAMINATION The limitations on removable contamination for sources requiring leak testing, including alpha emitters, is based on 10 CFR 70.39(c) limits for plutonium. This limitation will ensure that leakage from byproduct, source, and special nuclear material sources will not exceed allowable intake values. Sealed sources are classified into three groups according to their use, with surveillance requirements commensurate with the probability of damage to a source in that group. Those sources which are frequently handled are required to be tested more often than those which are not. Sealed sources which are continuously enclosed within a shielded mechanism (i.e. seated sources within radiation monitoring or boron measuring devices) are considered to be stored and need not be tested unless they are removed from the shielded mechanism.

3/4.*.9 AREA TEMPERATURE ONITORING The area temperature limitations ensure that safety-related equipment will not be subjected to temperatures in excess of their environmental qualification temperatures.

Exposure to excessive temperatures may degrade equipment and can cause a loss of its OPERABILITY. The temperature limits include an allowance for instrument error of 20F.

"*3/4.7. 10 WATE LEE-SPENT f:FUQ* POOL *r

=, The res

• n minimum wate e ensure that sufficient water depth is vaiable to re of the assumed gap activity released from the nrpture of an irradiated fuel assembly. The minimu water depth is consistent with the assumptions of the accident analysis. i Z6 7matedreleased f, a id" fuelasse proh y will be f red the HEPA ters

[and larcoal adrerpio e oe l . The OERABILI"Y *this eop \

cident an ,is. _..

SUMMER - UNIT 1 16 3/4 7-6 Amendment No..74, 160

Attachment 7 Page 8 of 9 314.9 REFUELING OPERATIONS BASES 3(4.9.1 BORON CONCENTRATION The limitations on reactivity conditions during REFUELING ensure that 1) the reactor will remain subcritical during CORE ALTERATIONS, and 2) a uniform boron concentration is maintained for reactivity control in the water volume having direct access to the reactor vessel. These limitations are consistent with the initial conditions assumed for the boron dilution incident in the accident analyses. The value of 0.95 or less for K,#Includes a 1 percent delta kIk conservative allowance for uncertainties. Similarly, the boron concentration value of 2000 ppm or greater includes a conservative uncertainty allowance of 50 ppm boron. Valves in the reactor makeup system are required to be closed to minimize the possibility of a boron dilution accident.

3X4.9.2 INSTfRUMENTATION The OPERA3IUTY of the, source range neutron flux moniors ensures that redundant monitoring capability is available to detect changes In the reactivity condition of the core.

3/4.9.3 DECAY TIME The minimum time of 72 hours8.333333e-4 days <br />0.02 hours <br />1.190476e-4 weeks <br />2.7396e-5 months <br /> for reactor subcrtlcellty prior to movement of Irradiated fuel assemblies In the reactor pressure vessel ensures that sufficient time has elapsed to allow tMe radioactive decay of the short-lived fission pwducts. The minimum decay time of 72 hours8.333333e-4 days <br />0.02 hours <br />1.190476e-4 weeks <br />2.7396e-5 months <br /> Is consistent with the a used In the accident analysis The tabulated hold times associted with Component Cooling Water (CCW) temperature ensure that the spent fuel heat load Is reduced sufficiently to allow the spent fuel pool cooling system to maintain the bulk pool tempereture below 170F. These hold times ensure that adequate cooling is provided to the Spent Fuel Pool under the highest possible heat load conditions. The hold times are based on the performance of the cooling system, which is dependent upon CCW temperature and recognizes that the spent fuel pool cooling system Is capable of increased flow rates up to 2400 gpm during single loop operation. This higher flow rate may be required when only a single cooling loop is operable during a refueling outage.

The COW temperature limits deffned In Figure 3.9-1 are adjusted for uncertainty in the Implementing procedure.

e* ,equir ents o eactor pILIetr OPn ens

/"that a gease of *oc materiy

• 'n taifn w~ill beiest ~drom I),aetthe

/envir dment. OPE BIUTY dc  ::sere are ufficientorst ra* ,

3/.9.45 COMMUNICATIONS p u"'poten while i he RE IN The requirement for communications capability ensures that refueling station nernonnel can be onmmtlv informed of significant changes in the facility status or core reactivity conditions during CORE ALTERATIONS.

SUMMER - UNIT I B 3/49-1 Amendment No. 160

Attachment 7 Page 9 of 9 REFUELING 9OERATIONS BASES S/4.9.6 MANIPULATOR CRANE The OPERABILITY requirements for the manipulator cranes ensure that

1) manipulator cranes will be used for movement of control rods and fuel assemblies,

2) each crane has sufficient load capacity to lift a control rod and fuel assembly, and

3) the core internals and pressure vessel are protected from excessive lifting force in the event they are inadvertently engaged during lifting operations.

3(4.9.7 RESIDUAL HEAT REMOVAL AND COOLANT CIRCULATION The requirement that at least one residual heat removal loop be In operation ensures that 1) sufficient cooling capacity is available to remove decay heat and maintain the water in the reactor pressure vessel below 140 0F as required during the REFUELING MODE, and 2) sufficient coolant circulation Is maintained thru the reactor core to minimize the effects of a boron dilution Incident and prevent boron stratification.

The requirement to have two RHR loops OPERABLE when there is less than 23 feet of water above the reactor pressure vesse flange ensures that a single faikure of the operating RHR loop wg not result in a complete loss of residuat heat removal capability.

WIth the reactor vessel head removed and at least 23 feet of water above the reaftr pressure vessel flange, a large heat sink is available for core cooling. Thus, In the event of a failure of the operating RHR loop, adequate ti to initiate 3/4,9.8RG SEr-OPLY AN XHAU_* ISOLLA46I"N

• The *PE T, of syt rsa h r clor bulng vernt opd u r 9- OPE ~~ILITY ofthis sys rm is req o tit h s mae from reactor ilding a spher the e 3/4.9.9 WATER LEVEL - REACTOR VESSEL

( C& "AI esIctanwson minmum water level a re that sufficient water depth is QE-a aifableto of the assumed activity released from he nrpture of an irradiated fuel assembly. The minimum ter depth is consistent with the assumptions of fth accident analysis.

ofI SUMMER - UNIT 1 B 314 9-2 Amendment No. 160

Attachment 8 Procedure Changes to be Completed Before AST Implementation

Attachment 8 Page 1 of 1 Procedure Changes to be Completed Before AST Implementation Change Basis Revise EOP-2.2, "Transfer to Cold Leg Leakage through the RWST and NAOH is Recirculation", to further isolate the RWST neglected in the AST calculations for and NAOH tanks following the transition LOCA based on the assumption that the the cold leg recirculation. EOPs will require closure of the 20" RWST outlet valve (6700) and closure of the 3" NAOH outlet valve (3012) following the transition to cold leg recirculation. This results in 3 valve isolation and a minimum of 2 valves isolation in the long term with a single failure.

Revise EOPs and other plants procedures The LOCA AST analysis credits the iodine as appropriate to require 4-hours of Spray removal capabilities of the RB Spray for a Operation following a LOCA. period of 4-hours.

Revise EOPs and other plant procedures as Placing the CR ventilation in the appropriate to ensure the CR ventilation is emergency mode has been credited within run in the emergency mode when required 2-hours for non-LOCA events (LRA &

for non-LOCA events that do not result in a CREA) that may not result in a SI.

SI. Although this is expected to occur automatically with high radiation on RMA-1, this feature has not been credited due to lack of redundancy. Consequently, as a backup for RMA-1,the operator will be asked to assess the need for protection actions based on existing radiation indications and atmospheric releases.

Attachment 9 No Significant Hazards Consideration Determination & Environmental Consideration for the Proposed Changes

Attachment 9 Page 1 of 6 NO SIGNIFICANT HAZARDS CONSIDERATION DETERMINATION Description of Amendment Request South Carolina Gas and Electric Co. (SCE&G) is proposing to amend the operating license for Virgil C. Summer Nuclear Station (VCSNS), by revising the Technical Specifications (TS) and incorporating an alternative source term (AST) methodology into the facility's licensing basis.

The proposed license amendment involves a full implementation of an AST methodology by revising the current accident source term and replacing it with an AST, as prescribed in 10 CFR 50.67.

AST analyses were performed using the guidance provided by Regulatory Guide 1.183, "Alternative Source Terms for Evaluating Design Basis Accidents at Nuclear Power Reactors,"

dated July 2000, and Standard Review Plan Section 15.0.1, "Radiological Consequences Analyses Using Alternative Source Terms." The six PWR limiting design basis accidents (DBAs) identified in RG 1.183 considered were the loss of coolant accident (LOCA), the main steam line break accident (MSLB), the refueling accident (FHA), the steam generator tube rupture (SGTR), reactor coolant pump locked rotor (RCP LRA) and the control rod ejection accident (CREA). As a result of the application of a revised accident source term, changes are proposed to the TS to implement Improved Standard Technical Specifications Change Traveler (Reference 12.1 - TSTF-5 1, Revision 2) which permits removal of the Technical Specification requirements for engineered safety features (ESF) features to be OPERABLE after sufficient radioactive decay has occurred to ensure off-site doses remain below the SRP limits. Other Technical Specification revisions reflect the update of the accident source term and associated design basis accidents utilizing the guidance provided in USNRC Regulatory Guide 1.183 and the associated control room and offsite dose requirements of 10 CFR 50.67.

The AST analyses are based on new control room habitability (CRHE) atmospheric dispersion coefficients (x/Qs) based on site specific meteorological data in accordance with Regulatory Guides 1.194.

Basis for No Significant Hazards Determination:

Pursuant to 10 CFR 50.92, VCSNS has reviewed the proposed change and concludes that the change does not involve a significant hazards consideration, since the proposed change satisfies the criteria in 10 CFR 50.92(c). These criteria require that the operation of the facility in accordance with the proposed amendment will not: (1) involve a significant increase in the probability or consequences of an accident previously evaluated, (2) create the possibility of a new or different kind of accident from any accident previously evaluated, or (3) involve a significant reduction in a margin of safety. The discussion below addresses each of these criteria and demonstrates that the proposed amendment does not constitute a significant hazard.

Attachment 9 Page 2 of 6 1.0 Does the proposed change involve a significant increase in the probability of occurrence or consequences of an accident previously evaluated?

Response: No.

Adoptions of the AST and pursuant TS changes and the changes to the atmospheric dispersion factors have no impact to the initiation of DBAs. Once the occurrence of an accident has been postulated, the new accident source term and atmospheric dispersion factors are an input to analyses that evaluate the radiological consequences. Some of the proposed changes do affect the design or manner in which the facility is operated following an accident; however, the proposed changes do not involve a revision to the design or manner in which the facility is operated that could increase the probability of an accident previously evaluated in Chapter 15 of the FSAR.

Therefore, the proposed change does not involve an increase in the probability of an accident previously evaluated.

The structures, systems and components affected by the proposed changes act as mitigators to the consequences of accidents. Based on the AST analyses, the proposed changes do revise certain performance requirements; however, the proposed changes do not involve a revision to the parameters or conditions that could contribute to the initiation of an accident previously discussed in Chapter 15 of the FSAR.

Plant-specific radiological analyses have been performed using the AST methodology and new atmospheric dispersion factors. Based on the results of these analyses, it has been demonstrated that the CRHE dose consequences of the limiting events considered in the analyses meet the regulatory guidance provided for use with the AST, and the offsite doses are within acceptable limits. This guidance is presented in 10 CFR 50.67, RG 1.183, and Standard Review Plan Section (SRP) 15.0.1.

Therefore, the proposed amendment does not result in a significant increase in the consequences of any previously evaluated accident.

Attachment 9 Page 3 of 6 2.0 Does the proposed change create the possibility of a new or different kind of accident from any accident previously evaluated?

Response: No.

Implementation of AST and the associated proposed TS changes and new atmospheric dispersion factors do not alter or involve any design basis accident initiators. With the exception of the fuel handling accident, these changes do not affect the design function or mode of operations of structures, systems and components in the facility prior to a postulated accident. Since structures, systems and components are operated essentially no differently after the AST implementation, no new failure modes are created by this proposed change. The alternative source term change itself does not have the capability to initiate accidents.

For the fuel handling accident, the Improved Standard Technical Specifications Change Traveler (TSTF-5 1, Revision 2) permits removal of the Technical Specification requirements for ESF features to be OPERABLE after sufficient radioactive decay has occurred to ensure off-site doses remain below the SRP limits. As noted in this submittal no credit is taken for the accident mitigation of the ESF features associated with the fuel handling accidents to meet these limits. Since these are not associated with accident initiators the proposed license amendment will not create the possibility of a new or different kind of accident from any accident previously evaluated.

3.0 Does the proposed change involve a significant reduction in a margin of safety?

Response: No.

The results of the AST analyses are subject to the acceptance criteria in 10 CFR 50.67. The analyzed events have been carefully selected, and the analyses supporting these changes have been performed using approved methodologies to ensure that analyzed events are bounding and safety margin has not been reduced. The dose consequences of these limiting events are within the acceptance criteria presented in 10 CFR 50.67, RG 1. 183, and SRP 15.0.1. Thus, by meeting the applicable regulatory limits for AST, there is no significant reduction in a margin of safety.

New Control Room atmospheric dispersion factors (X/Qs) based on site specific meteorological data, calculated in accordance with the guidance of RG 1.194, utilizes more recent data and improved calculational methodologies.

For the fuel handling accident, the Improved Standard Technical Specifications Change Traveler (TSTF-5 1, Revision 2) permits removal of the Technical Specification requirements for ESF features to be OPERABLE after sufficient radioactive decay has occurred to ensure off-site doses remain below the SRP limits. Following sufficient decay, the primary success path for mitigating the fuel handling accident no longer includes the functioning of the active containment or fuel handling building systems. With the proposed changes, the OPERABILITY requirements of the Technical Specifications will reflect that water level

Attachment 9 Page 4 of 6 (23') and decay time (72 hours8.333333e-4 days <br />0.02 hours <br />1.190476e-4 weeks <br />2.7396e-5 months <br /> after shutdown) are the primary success path for mitigating a fuel handling accident.

Therefore, because the proposed changes continue to result in dose consequences within the applicable regulatory limits, the changes are considered to not result in a significant reduction in a margin of safety.

Conclusion On the basis of the above, VCSNS has determined that operation of the facility in accordance with the proposed change does not involve a significant hazards consideration as defined in 10 CFR 50.92(C), in that it: (1) does not involve a significant increase in the probability or consequences of an accident previously evaluated; (2) does not create the possibility of a new or different kind of accident from any accident previously evaluated; and (3) does not involve a significant reduction in a margin of safety.

Attachment 9 Page 5 of 6 ENVIRONMENTAL CONSIDERATION VCSNS has evaluated the proposed amendment against the criteria for identification of licensing and regulatory actions requiring environmental assessment in accordance with 10 CFR 51.21.

VCSNS has determined that the proposed change meets the criteria for categorical exclusion as provided for under 10 CFR 51.22(c)(9). 10 CFR 51.22(c)(9) identifies certain licensing and regulatory actions, which are eligible for categorical exclusion from the requirement to perform an environmental assessment. A proposed amendment to an operating license for a facility does not require an environmental assessment if operation of the facility in accordance with the proposed amendment would not: (1) involve a significant hazards consideration; (2) result in a significant change in the types or significant increase in the amounts of any effluents that may be released offsite; or (3) result in a significant increase in individual or cumulative occupational radiation exposure. VCSNS has evaluated the proposed change and has determined that the proposed change meets the eligibility criteria for categorical exclusion set forth in 10 CFR 51.22(c)(9). Accordingly, pursuant to 10 CFR 51.22(b), no environmental impact statement or environmental assessment needs to be prepared in connection with issuance of the amendment.

The basis for this determination, using the above criteria, follows:

Basis As demonstrated in the No Significant Hazards Consideration Evaluation, the proposed amendment does not involve a significant hazards consideration.

There is no significant change in the types or significant increase in the amounts of any effluents that may be released offsite. Therefore, there is no significant increase in individual or cumulative occupational radiation exposure. Furthermore, the proposed change does not involve any unreviewed safety questions concerning the physical alteration of the plant or change in methods governing normal plant operation except for the fuel handling accident. For the fuel handling accident, the Improved Standard Technical Specifications Change Traveler (TSTF-5 1, Revision 2) permits removal of the Technical Specification requirements for ESF features to be OPERABLE after sufficient radioactive decay has occurred to ensure off-site doses remain below the SRP limits. Following sufficient decay, the primary success path for mitigating the fuel handling accident no longer includes the functioning of the active containment or fuel handling building systems. With the proposed changes, the OPERABILITY requirements of the Technical Specifications will reflect that water level (23') and decay time (72 hours8.333333e-4 days <br />0.02 hours <br />1.190476e-4 weeks <br />2.7396e-5 months <br /> after shutdown) are the primary success path for mitigating a fuel handling accident.

Conclusion The alternative source term does not affect the design or operation of the facility; rather, once the occurrence of an accident has been postulated, the alternative source term is an input to evaluate the consequences of accidents. The implementation of the alternative source term has been evaluated in AST analyses of the limiting design basis accidents at VCSNS (loss of coolant accident, the main steam line break accident, the refueling accident, the steam generator tube

Attachment 9 Page 6 of 6 rupture, reactor coolant pump locked rotor and the control rod ejection accident). Based upon the results of these analyses it has been demonstrated that, with the requested changes, the dose consequences are within NRC regulatory limits for alternative source term (i.e., 10 CFR 50.67 and 10 CFR 50, Appendix A, General Design Criterion 19).

On the basis of the above, VCSNS has determined that operation of the facility in accordance with the proposed change does not involve an environmental consideration as defined in 10 CFR 51.22(c)(9), in that it does not: (1) involve a significant hazards consideration; (2) result in a significant change in the types or significant increase in the amounts of any effluents that may be released offsite; or (3) result in a significant increase in individual or cumulative occupational radiation exposure.

Attachment 10 CD of Non-Proprietary Versions of Supporting Calculations and Meteorological Data Used to Determine New Control Room Z/Qs I

The following non-proprietary Supporting Calculations are included in the "Supporting Calculation" directory of the Attachment 10 CD.

1. VCSNS Calculation DC00040-079, "Atmospheric Dispersion Coefficients for Control Room", Revision 1.

2. VCSNS Calculation DC00040-097, "Loss of Coolant Accident - AST", Revision 0.

3. VCSNS Calculation DC00040-098, "Steam Generator Tube Rupture - AST", Revision 0.

4. VCSNS Calculation DC00040-099, "Main Steam Line Break - AST", Revision 0.

5. VCSNS Calculation DC00040-100, "Reactor Coolant Pump Locked Rotor - AST",

Revision 0.

6. VCSNS Calculation DC00040-101, "Rod Ejection - AST", Revision 0.

7. VCSNS Calculation DC00040-102, "Fuel Handling Accidents - AST", Revision 0.

The following non-proprietary information is included in the "Meteorological Data" directory of the Attachment 10 CD.

1. VCSNS Calculation DC00040-080, "Post DBA CR Dose - Met Data Input to ARCON96",

Revision 2.

2. ARCON96 Input File (VCSR2.MET).

3. 6-EXCEL Files Containing Raw Met Tower Data from 2002 to 2006.

Attachment 11 Proposed Technical Specification Changes (Re-Typed)

INDEX LIMITING CONDITIONS FOR OPERATION AND SURVEILLANCE REQUIREMENTS SECTION PAGE 3/4.7 PLANT SYSTEMS 3/4.7.1 TURBINE CYCLE S afety Valves ....................................................................................... 3/4 7-1 Emergency Feedwater System ............................................................. 3/4 7-4 Condensate Storage Tank .................................................................... 3/4 7-6 A ctivity .................................................................................................. 3/4 7 -7 Main Steam Line Isolation Valves ......................................................... 3/4 7-9 Feedwater Isolation Valves .................................................................. 3/4 7-9a 3/4.7.2 STEAM GENERATOR PRESSURE/TEMPERATURE LIMITATION ..... 3/4 7-10 3/4.7.3 COMPONENT COOLING WATER SYSTEM ....................................... 3/47-11 3/4.7.4 SERVICE WATER SYSTEM ................................................................ 3/4 7-12 3/4.7.5 ULTIMATE HEAT SINK ....................................................................... 3/4 7-13 3/4.7.6 CONTROL ROOM EMERGENCY FILTRATION SYSTEM (CREFS) .... 3/4 7-14 3/4.7.7 SN U BB ERS ......................................................................................... 3/4 7-16 3/4.7.8 SEALED SOURCE CONTAMINATION ................................................ 3/4 7-23 3/4.7.9 AREA TEMPERATURE MONITORING ............................................... 3/4 7-37 3/4.7.10 WATER LEVEL - SPENT FUEL POOL ................................................ 3/4 7-39 3/4.7.11 D E LET E D ............................................................................................. 3/4 7-40 3/4.7.12 SPENT FUEL ASSEMBLY STORAGE ................................................. 3/4 7-42 3/4.7.13 SPENT FUEL POOL BORON CONCENTRATION ............................... 3/4 7-44 SUMMER - UNIT 1 Vill Amendment No. 23,7, 7160, 180,

INDEX LIMITING CONDITIONS FOR OPERATION AND SURVEILLANCE REQUIREMENTS SECTION PAGE 3/4.9 REFUELING OPERATIONS 3/4.9.1 BORON CONCENTRATION ................................................................ 3/4 9-1 3/4.9.2 INSTRUMENTATION .......................................................................... 3/4 9-2 3/4.9.3 D EC AY T IME ....................................................................................... 3/4 9-3 3/4.9.4 D ELET ED ............................................................................................. 3/4 9-4 3/4.9.5 COMMUNICATIONS ............................................................................ 3/4 9-5 3/4.9.6 MANIPULATOR CRANE ...................................................................... 3/4 9-6 3/4.9.7 RESIDUAL HEAT REMOVAL AND COOLANT CIRCULATION H igh W ater Level .................................................................................. 3/4 9-7 Low W ater Level ................................................................................... 3/4 9-8 3/4 .9.8 D E LE T E D .............................................................................................. ..... 3/4 9-9 3/4.9.9 WATER LEVEL - REFUELING CAVITY AND FUEL TRANSFER CANAL. 3/4 9-10 3/4.10 SPECIAL TEST EXCEPTIONS 3/4.10.1 SHUTDOWN MARGIN ......................................................................... 3/410-1 3/4.10.2 GROUP HEIGHT, INSERTION AND POWER DISTRIBUTION LIMITS 3/4 10-2 3/4.10.3 PHYSICS TESTS ................................................................................. 3/4 10-3 3/4.10.4 REACTOR COOLANT LOOPS ............................................................ 3/4 10-4 3/4.10.5 POSITION INDICATION SYSTEM - SHUTDOWN ............................... 3/4 10-5 49 60 SUMMER - UNIT 1 X Amendment No. ,1

INDEX BASES SECTION PAGE 3/4.7 PLANT SYSTEMS 3/4.7.1 TURBINE CYCLE ................................................................................ B 3/4 7-1 3/4.7.2 STEAM GENERATOR PRESSURE/TEMPERATURE LIMITATION ..... B 3/4 7-3 3/4.7.3 COMPONENT COOLING WATER SYSTEM ....................................... B 3/4 7-3 3/4.7.4 SERVICE W ATER SYSTEM ................................................................ B 3/4 7-3 3/4.7.5 ULTIMATE HEAT SINK ........................................................................ B 3/4 7-3 3/4.7.6 CONTROL ROOM EMERGENCY FILTRATION SYSTEM (CREFS) .... B 3/4 7-4 3/4.7.7 S NUBBER S ......................................................................................... B 3/4 7-4f 3/4.7.8 SEALED SOURCE CONTAMINATION ................................................ B 3/4 7-6 3/4.7.9 AREA TEMPERATURE MONITORING ................................................ B 3/4 7-6 3/4.7.10 WATER LEVEL - SPENT FUEL POOL ................................................ B 3/4 7-6 3/4.7.11 D ELET ED ............................................................................................. B 3/4 7-6 3/4.7.12 SPENT FUEL ASSEMBLY STORAGE ................................................. B 3/4 7-7 3/4.7.13 SPENT FUEL POOL BORON CONCENTRATION ............................... B 3/4 7-7 3/4.8 ELECTRICAL POWER SYSTEMS 3/4.8.1, 3/4.8.2 and 3/4.8.3 A.C. SOURCES, D.C. SOURCES AND ONSITE POW ER DISTRIBUTION SYSTEMS .................................................... B 3/4 8-1 3/4.8.4 ELECTRICAL EQUIPMENT PROTECTION DEVICES ......................... B 3/4 8-4 SUMMER - UNIT 1 XIV Amendment No. 38, 63, 79,160, 4784,8O

INDEX BASES SECTION PAGE 3/4.9 REFUELING OPERATIONS 3/4.9.1 BORON CONCENTRATION ................................................................ B 3/4 9-1 3/4.9.2 INSTRUM ENTATIO N .......................................................................... B 3/4 9-1 3/4.9.3 D ECAY TIM E ....................................................................................... B 3/4 9-1 3/4.9.4 D E LET ED ............................................................................................ B 3/4 9-1 3/4.9.5 CO MMUNICATIO NS ........................................................................... B 3/4 9-1 3/4.9.6 MANIPULATOR CRANE ..................................................................... B 3/4 9-2 3/4.9.7 RESIDUAL HEAT REMOVAL AND COOLANT CIRCULATION ........... B 3/4 9-2 3/4.9.8 DELETED ............................................ B 3/4 9-2 3/4.9.9 WATER LEVEL - REACTOR VESSEL ........................................................ B 3/4 9-2 3/4.10 SPECIAL TEST EXCEPTIONS 3/4.10.1 SHUTDOW N MARG IN ........................................................................ B 3/4 10-1 3/4.10.2 GROUP HEIGHT, INSERTION AND POWER DISTRIBUTION LIMITS B 3/4 10-1 3/4.10.3 PHYSICS TESTS ................................................................................ B 3/4 10-1 3/4.10.4 REACTOR COOLANT LOOPS ............................................................ B 3/4 10-1 3/4.10.5 POSITION INDICATION SYSTEM - SHUTDOWN .................. I............ B 3/4 10-1 XV Amendment No. 49 ,1 6 0, SUMMER - UNIT 1

CO TABLE 3.3-6 RADIATION MONITORING INSTRUMENTATION z MINIMUM CHANNELS APPLICABLE ALARM/TRIP MEASUREMENT INSTRUMENT OPERABLE MODES SETPOINT RANGE ACTION

1. AREA MONITORS

a. Spent Fuel Pool Area (RM-G8) 1 < 15 mR/hr 101 - 104 mR/hr 25

b. Deleted G3 2. PROCESS MONITORS 4ý1

a. Deleted NO

b. Containment

i. Deleted ii. Particulate and Gaseous 1 1,2, 3& 4 N/A 10 - 106 cpm 26 Activity (RM-A2) -

RCS Leakage Detection

c. Control Room Isolation (RM-A1) 1 ALL MODES < 2 x background 10 - 106 cpm 29 3

CD

• With fuel in the storage pool or building z

0

INSTRUMENTATION TABLE 3.3-6 (Continued)

ACTION STATEMENTS ACTION 25 - With the number of channels OPERABLE less than required by the Minimum Channels OPERABLE requirement, perform area surveys of the monitored area with portable monitoring instrumentation at least once per 24 hours2.777778e-4 days <br />0.00667 hours <br />3.968254e-5 weeks <br />9.132e-6 months <br />.

ACTION 26 - With the number of channels OPERABLE less than required by the Minimum Channels OPERABLE requirement, comply with the ACTION requirements of Specification 3.4.6.1.

ACTION 27 - Deleted ACTION 28 - Deleted ACTION 29 - With the number of channels OPERABLE less than required by the Minimum Channels OPERABLE requirement, within 1 hour1.157407e-5 days <br />2.777778e-4 hours <br />1.653439e-6 weeks <br />3.805e-7 months <br /> initiate and maintain operation of the control room emergency ventilation system in the emergency mode of operation.

ACTION 30 - With the number of OPERABLE channels less than required by the Minimum Channels OPERABLE requirement, either restore the inoperable Channel(s) to OPERABLE status within 72 hours8.333333e-4 days <br />0.02 hours <br />1.190476e-4 weeks <br />2.7396e-5 months <br />, or:

1) Initiate the preplanned alternate method of monitoring the appropriate parameter(s), and

2) Prepare and submit a Special Report to the Commission pursuant to Specification 6.9.2 within 14 days following the event outlining the action taken, the cause of the inoperability and the plans and schedule for restoring the system to OPERABLE status.

SUMMER - UNIT 1 3/4 3-44 Amendment No.

U)

TABLE 4.3-3 RADIATION MONITORING INSTRUMENTATION SURVEILLANCE REQUIREMENTS z ANALOG CHANNEL MODES FOR WHICH CHANNEL CHANNEL OPERATIONAL SURVEILLANCE INSTRUMENT CHECK CALIBRATION TEST IS REQUIRED

1. AREA MONITORS

a. Spent Fuel Pool Area (RM-G8) S R M *

b. Deleted

2. PROCESS MONITORS

a. Deleted I

b. Containment

i. Deleted ii. Particulate and Gaseous Activity - RCS Leakage S R M 1,2,3&4 Detection (RM-A2)

c. Control Room Isolation (RM-A1) S R M ALL MODES

d. Noble Gas Effluent Monitors Cn (High Range)

i. Main Plant Vent (RM-A1 3) S R M 1,2,3&4 ii. Main Steam Lines (RM-G19A, B, C) S R M 1,2, 3& 4 iii. Reactor Building Purge CD Supply & Exhaust System (RM-A14) S R M 1,2,3 & 4 z

0

• With fuel in the storage pool or building

THIS PAGE INTENTIONALLY LEFT BLANK SUMMER - UNIT 1 3/4 7-40 Amendment No. 4 2,159 ,

160,1!90,

THIS PAGE INTENTIONALLY LEFT BLANK SUMMER - UNIT 1 3/4 9-4 Amendment No.

THIS PAGE INTENTIONALLY LEFT BLANK SUMMER - UNIT 1 3/4 9-9 Amendment No. 4 ADMINISTRATIVE CONTROLS location of flaws to which the tubes may be susceptible and, based on this assessment, to determine which inspection methods need to be employed and at what locations.

a) Inspect 100% of the tubes in each SG during the first refueling outage following SG replacement.

b) Inspect 100% of the tubes at sequential periods of 144, 108, 72 and thereafter, 60 effective full power months. The first sequential period shall be considered to begin after the first inservice inspection of the SGs. In addition, inspect 50% of the tubes by the refueling outage nearest the midpoint of the period and the remaining 50% by the refueling outage nearest the end of the period. No SG shall operate for more than 72 effective full power months or three refueling outages (whichever is less) without being inspected.

c) If crack indications are found in any SG tube, then the next inspection for each SG for the degradation mechanism that caused the crack indication shall not exceed 24 effective full power months or one refueling outage (whichever is less). If definitive information, such as from examination of a pulled tube, diagnostic non-destructive testing, or engineering evaluation indicates that a crack-like indication is not associated with a crack(s), then the indication need not be treated as a crack.

5. Provisions for monitoring operational primary-to-secondary leakage.

Ventilation Filter Testing Program (VFTP)

A program shall be established to implement the following required testing of Engineered Safety Feature (ESF) filter ventilation systems at the frequencies specified in accordance with Regulatory Guide 1.52, Revision 2, and ASME N510-1989.

1. Demonstrate for each of the ESF systems that an inplace test of the high efficiency particulate air (HEPA) filters shows a penetration and system bypass

< 0.05% when tested in accordance with Regulatory Guide 1.52, Revision 2, and ASME N510-1989 at the system flowrate specified below +/- 10%.

ESF Ventilation System Flowrate Control Room Emergency Filtration System 21,270 SCFM Reactor Building Cooling Units 60,270 ACFM SUMMER - UNIT 1 6-12f Amendment No. 179,-180,

ADMINISTRATIVE CONTROLS

2. Demonstrate for each of the ESF systems that an inplace test of the charcoal adsorber shows a penetration and system bypass < 0.05% when tested in accordance with Regulatory Guide 1.52, Revision 2, and ASME N510-1989 at the system flowrate specified below +/- 10%.

ESF Ventilation System Flowrate Control Room Emergency Filtration System 21,270 SCFM

3. Demonstrate for each of the ESF systems that a laboratory test of a sample of the charcoal adsorber, when obtained as described in Regulatory Guide 1.52, Revision 2, shows the methyl iodide penetration less than the value specified below when tested in accordance with ASTM D3803-1989 at a temperature of 30 0 C (860 F) and the relative humidity specified below.

ESF Ventilation System Penetration RH Face Velocity (fps)

Control Room <2.5% 70% 0.667

4. Demonstrate for each of the ESF systems that the pressure drop across the combined HEPA filters, the prefilters, and the charcoal adsorbers is less than the value specified below when tested in accordance with Regulatory Guide 1.52, Revision 2, and ASME N510-1989 at the system flowrate specified below

+ 10%.

ESF Ventilation System Delta P Flowrate Control Room <6 in. W.G. 21,270 SCFM Reactor Building Cooling Units <3 in. W.G. 60,270 ACFM The provisions of SR 4.0.2 and SR 4.0.3 are applicable to the VFTP test frequencies.

m. Control Room Envelope Habitability Program A Control Room Envelope (CRE) Habitability Program shall be established and implemented to ensure that CRE habitability is maintained such that, with an OPERABLE Control Room Emergency Filtration System (CREFS), CRE occupants can control the reactor safely under normal conditions and maintain it in a safe condition following a radiological event, hazardous chemical release, or a smoke challenge. The program shall ensure that adequate radiation protection is provided to permit access and occupancy of the CRE under design basis accident (DBA) conditions without personnel receiving radiation exposures in excess of 5 rem whole body or its equivalent to any part of the body for the duration of the accident. The program shall include the following elements:

SUMMER - UNIT 1 6-12g Amendment No. 4W0

w Attachment 12 Regulatory Commitments There are no commitments contained in this submittal.