PLA-6112, Revision to Proposed Amendment Nos. 281 and 251, Application for License Amendment and Related Technical Specification Changes to Implement Full-Scope Alternative Source Term in Accordance with 10 CFR 50.67
| ML062710360 | |
| Person / Time | |
|---|---|
| Site: | Susquehanna  |
| Issue date: | 09/15/2006 |
| From: | Mckinney B Susquehanna |
| To: | Document Control Desk, Office of Nuclear Reactor Regulation |
| References | |
| PLA-6112 | |
| Download: ML062710360 (49) | |
Text
Britt T. McKlnney Sr. Vice President & Chief Nuclear Officer SEP 15 2006 PPL Susquehanna, LLC 769 Salem Boulevard Berwick, PA 18603 Tel. 570.542.3149 Fax 570.542.1504 btmckinney@pplweb.com pp TM U. S. Nuclear Regulatory Commission Attn: Document Control Desk Mail Stop OPl-17 Washington, DC 20555 SUSQUEHANNA STEAM ELECTRIC STATION REVISION TO PROPOSED AMENDMENT NO. 281 TO LICENSE NPF-14 AND PROPOSED AMENDMENT NO. 251 TO LICENSE NPF-22: "APPLICATION FOR LICENSE AMENDMENT AND RELATED TECHNICAL SPECIFICATION CHANGES TO IMPLEMENT FULL-SCOPE ALTERNATIVE SOURCE TERM IN ACCORDANCE WITH 10 CFR 50.67" Docket Nos. 50-387 PLA-6112 and 50-388
References:
Proposed Amendment No. 281 to License NPF-14 and Proposed Amendment No. 251 to License NPF-22: "Application for License Amendment and Related Technical Specification Changes to Implement Full-Scope Alternative Source Term in Accordance with 10 CFR 50.67, " dated October 13, 2005.
Supplement to Proposed Amendment No. 281 to License NPF-14 and Proposed Amendment No. 251 to License NPF-22: "Application for License Amendment and Related Technical Specification Changes to Implement Full-Scope Alternative Source Term in Accordance with 10 CFR 50.67,"
dated May 18, 2006.
The purpose of this letter is to revise the Reference 1 request for an amendment to the licensing basis for the Susquehanna Steam Electric Station (SSES) Units 1 and 2 that supports a full implementation application of an Alternative Source Term (AST) methodology.
As discussed with the SSES Project Manager, PPL Susquehanna, LLC agrees to withdraw the following proposed changes delineated below as they appear in Reference 1:
- "Revised TS Section 3.7.3 concerning Control Room Emergency Outside Air Supply System (CREOASS) to reflect Improved Standard Technical Specifications Change Traveler (TSTF-448)."
14001
ýý03 Document Control Desk PLA-6112 "Add TS Section 5.5.13 concerning CREOASS to reflect Improved Standard Technical Specifications Change Traveler (TSTF-448) concerning Control Room Habitability Program."
As identified in Attachment 2, PPL commits to submit changes to address Control Room Habitability in the SSES Technical Specifications after the Notice of Availability for TSTF-448 is published in accordance with the Consolidated Line Item Improvement Process (CLIIP). to this letter contains markups of the revised Reference 1 affected pages.
Reference 1 identified that AST is required to support the extended power uprate (EPU) implementation which, at the time, was planned for implementation upon startup from the U2-13 RIO in Spring 2007. Given that EPU implementation has been delayed and to avoid outdoor construction involving extensive scaffold work required to relocate the Control Room Habitability Envelope Air Intake, PPL Susquehanna, LLC requests the AST amendment be conditioned to be effective no later than October 30, 2007 rather than being effective upon startup from the U2-13 RIO in Spring 2007, as originally requested in Reference 1.
If you have any questions regarding this submittal, please contact Mr. Michael H. Crowthers at (610) 774-7766.
I declare under penalty of perjury that the foregoing is true and correct.
Executed on:
Respectfully, Britt T. McKinney Attachments: - Markup of the Reference 1 Affected Pages - PPL Susquehanna List of Regulatory Commitments cc:
NRC Region I Mr. A. Blamey, NRC Sr. Resident Inspector Mr. R. V. Guzman, NRC Project Manager Mr. R. Janati, DEP/BRP
M to PLA-6112 Markup of the Reference 1 Affected Pages Document Control Desk PLA-5963 PPL proposes implementation of this proposed change through a change to the SSES licensing basis, including the TS and associated Bases. Upon approval, conforming changes will be made to the SSES Final Safety Analysis Report (FSAR) and submitted to the NRC staff in accordance with 10 CFR 50.71 as part of the regular FSAR update process.
Proposed changes in the licensing basis for SSES resulting from application of the AST include the following:
" New offsite and Control Room atmospheric dispersion factors (X/Qs) based on site specific meteorological data collected between 1999 and 2003, the new location of the CRHE air intake and Regulatory Guides 1.145 and 1.194 revised methodologies,
" Revised CR-E unfiltered inleakage from 10 cfm to 510 cfm.
" New AST analyses performed in accordance with the guidance in Regulatory Guide 1.183 for the four design basis accidents: loss of coolant accident, the main steam line break accident, the refueling accident, and the control rod drop accident.
" Revised TS Section 1.1 definition of Dose Equivalent 1-131.
" Revised TS Section 3.1.7 to credit use of the Standby Liquid Control (SLC) System to buffer suppression pool pH to prevent iodine re-evolution following a postulated desi basis loss of coolant accident (DBA LOCA).
" eie SSection.73concerning Cor tol Room E )rgency Outsio Air Sup, y Systen (CREOAS* to reflect Improvj*d Standard Tec1nical Speeifiton Ch nem Traveler (TSTF-
)./
Ad tTS Section *.5.13 concerning lROASS to rect Improve Standra Thnical Spe /icafions Chane/raveer TST
- 8) conceri g Contr' Room Habitability *ogram.
Table 5-1 of Attachment 5 provides a description of each proposed TS and TS Bases change.
In addi'on to revising he SSES licensing basis to adopt the AST, licensing basis changes are prop ed and justifi to respond to tC Generic Letter 2 3-01, "Contro oom Habitapility,"
datd June 12, 20 (Reference 12) and the Technical pecification T$
Force Impr ed I Stndard Techn al Specification6Change Traveler T8TF-448, Revisi, 2 (Referenc4 2.2).
The proposed S (Section 5.5.1 ), "Control Room J~abitability Progr~m", is providid in f this LAR.
In PPL Letter PLA-5916, dated 06/2812005, PPL identified that one commitment (provide dose consequence analysis using Regulatory Guide 1.183) will be rovided with the PPL AST submittal. This submittal serves to close that commitmenttgfi-as swI*, all acds' P*pý_
(c-tter 20e063en-r cornoete. No dew R gilat*
The current operating license allows SSES to operate at a maximum steady-state power level of 3489 megawatts thermal (MWt). PPL is also currently engaged in an Extended Power Uprate (EPU) project to increase the maximum licensed thermal power to 3952 MWt. Therefore, the AST analyses supporting this amendment request have been performed with the core isotopic values at EPU conditions and this application for license amendment is based on that bounding core isotopic inventory.
to PLA-5963 Description for the Alternative Source Term License Amendment to PLA-5963 Page 2 of 6 Upon issuance of a license amendment, conforming FSAR changes will be completed as required by PPL procedures and submitted to the NRC staff in accordance with the regular FSAR update process as required by 10 CFR 50.71. In lieu of providing the NRC staff with proposed FSAR changes at this time, the supporting DBA calculations are being provided in Attachments 10 and 11.
The license amendment would revise the following SSES licensing bases:
" New offsite and Control Room atmospheric dispersion factors (X/Qs) based on site specific meteorological data collected between 1999 and 2003, the new location of the CRHE air intake and Regulatory Guides 1.145 and 1.194 revised methodologies.
" Revised CRHE unfiltered inleakage from 10 cfm to 510 cfm.
- New AST analyses performed in accordance with the guidance in Regulatory Guide 1.183 for the four design basis accidents: loss of coolant accident, the main steam line break accident, the refueling accident, and the control rod drop accident.
- Revised TS Section 1.1 definition of Dose Equivalent 1-131.
- Revised TS Section 3.1.7 to credit use of the Standby Liquid Control (SLC) System to buffer suppression pool pH to prevent iodine re-evolution following a postulated design basis loss of coolant accident (DBA LOCA).
Revise TS Section 3.7.3 oncerning ControlVoom Emergency Outside Air Spply Syste (CREOASS) to eflect Improved St dard Technical cifications *'hange Trav ler (TSTF-448)
/
A" TS Section 5.13 concerning CP ASS to reflect proved Starnt d
Tehnical Specif ations Change Tra eler (TSTF-448) oncerning CoIol Room Habitability Pro am.
Implementation of the AST is scheduled for the spring 2007. To support this schedule, PPL requests approval of this proposed License Amendment by December, 2006, with the amendment conditioned to be effective upon startup from the U2-13RIO in spring 2007. Implementation of AST is required to support the extended power uprate (EPU) implementation for which the submittal is currently being prepared and is scheduled to be submitted to NRC in the spring of 2006.
2.0 REGULATORY BACKGROUND The current SSES 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.
.6 Attachment I to PLA-5963 Page 4 of 6
- 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 postulated accidental releases of radioactive material to the atmosphere.
These revised procedures are largely based on the NRC sponsored computer code, ARCON96.
TSTF Revision 2 BWOG-1 11,RO, "Technical Specificatiqn Task Force Improved Sta dard Technil Specifictions Change T veler,"
devel ped proposed hanges to Te hnica Spe ifications to rep ace the diffe ential pressuresurveillance 1kith a trace/ gas surveillanc and to institute Sa C ntrol Room -I6itability Program that ll-ensure that Control Room habitability is maintained.
On August 15, 1995, the NRC staff issued amendment 121 to Facility Operating License No. NPF-22 and amendment 151 for Facility Operating License No. NPF-14, to increase the allowable main steam isolation valve (MSIV) leakage rate and to delete the MSIV Leakage Control Systems. These amendments permitted SSES Units 1 and 2 to take credit for the Isolated Condenser Treatment Method (ICTM) for reducing the radiological consequences of MSIV leakage for a DBA LOCA. The ICTM uses the main steam drain lines to direct any MSIV leakage to the main condenser, as an alternative method for MSIV leakage treatment and the removal of the MSIV leakage control system (MSIVLCS). This drain path takes advantage of the large volume of the main steam lines (MSLs) and condenser to provide holdup and plate-out of fission products that may leak through the closed MSIVs. PPL performed evaluations and seismic verification walk downs to demonstrate that the main steam system piping and components which comprise the ICTM system were seismically rugged and are able to perform the safety function of an MSIV leakage treatment system. The seismic ruggedness evaluation was performed to demonstrate the seismic adequacy of the Turbine Building which houses the ICTM system.
The structural integrity of the Turbine Building is an important consideration to the adequacy of the alternate MSIV leakage path because a non-seismically designed Turbine Building should be capable of withstanding the earthquake without degrading the capability of the ICTM system.
to PLA-5963 AST Safety. Assessment Report to PLA-5963 Page 6 of 60
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 2), provides guidance to licensees on performing evaluations, and reanalyses as required to adopt an alternative source term (AST).
SSES has performed radiological consequence analyses of the four applicable boiling water reactor (BWR) DBAs identified in RG 1.183. These DBAs are a Loss of Coolant Accident (LOCA), a Fuel Handling Accident (FHA), a Control Rod Drop Accident (CRDA) and a Main Steam Line Break (MSLB). 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 3). The analyses were prepared, reviewed, and approved in accordance with the PPL 10 CFR 50, Appendix B Quality Assurance Program. Comparison with the guidance contained in RG 1.183 and RG 1.194 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 four BWR DBAs.
" Application of the deposition and removal mechanisms.
- Evaluation of suppression pool pH to ensure that the particulate iodine deposited into the suppression pool during a DBA LOCA does not re-evolve and become airborne as elemental iodine.
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 5) requirements and operation of the SLC System.
The radiological dose analyses have been performed assuming reactor operation at 4032 MWt (102% of the EPU rated power level of 3952 MWt, conservatively rounded high). This results in a conservative estimate of fission product releases for operation at current licensed power of 3489 MWt.
In additio to revising the SS licensing basito adopt the ASia, licensing basis c anges are propped and justified t respond to NROGeneric Letter )603-01, "Controloom Habital Jlity", dated June 1*, 2003.
L to PLA-5963 a
Page 7 of 60 Th* rooed TS (Sectio /5.5.13), "Control oom. Habita ity Program", s provid/eA in Atc:et6of this LA 7
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.
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 PPL 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 PPL's SSES current design basis source term acceptance criteria. The AST analyses were performed for the four BWR DBAs identified in RG 1.183 that could potentially result in Control Room and offsite doses. These include the loss of coolant accident, the main steam line break accident, the refueling accident, and the control rod drop accident.
Inadditin t s LAR provides e bases for resoling the nonco formiance issue with the current -esign and licens' g basis for the C OASS. TS 3.3 (Control R mi Emerge* yOutside Air S ply System) wa vised and TS.5.13 was ad d in I respo e to TSTF-448, evision 2, BWO 11, RO, 'Tehnical Speci fia, ion Task For
- Improved Stan ard Technical Secifications Cnge Traveler' (Reference 36).
to PLA-5963 Page 59 of 60
7.0 REFERENCES
- Continued
- 26.
SSES Amendment 252 to License NPF-14 and 217 to License NPF-22, "Application for Technical Specification Improvement to Eliminate Requirements for Post Accident Sampling Stations for Boiling Water Reactors Using the Consolidated Line Item Improvemrient Process," dated 3/3/2003.
- 27.
NRC approved Industry/Technical Specification Task Force Standard Technical Specification Change Traveler, TSTF-413, "Elimination of Requirements for Post Accident Sampling System (PASS)."
- 28.
SSES Amendment 151 to License NPF-14 and 121 to License NPF-22, "Susquehanna Steam Electric Station, Units 1 and 2 (TAC Nos. M91013 and M91014)," dated 8/15/1995.
- 29.
USNRC Regulatory Guide 1.3, "Assumption Used for Evaluating the Potential Radiological Consequences of a Loss of Coolant Accident for Boiling Water Reactors,"
Revision 2, June 1974.
- 30.
USNRC Regulatory Guide 1.5 (Safety Guide 5), "Assumptions Used for Evaluating the Potential Radiological Consequences of a Steam Line Break Accident for Boiling Water Reactors," Revision 0, 3110/71.
- 31.
USNRC Regulatory Guide 1.25 (Safety Guide.25), "Assumptions Used for Evaluating the Potential Radiological Consequences of a Fuel Handling Accident in the Fuel Handling and Storage facility for Boiling and Pressurized Water Reactors," Revision 0, 3/23/72.
- 32.
PPL Calculation EC-RADN-1038, "Radioactive Material Source Term Evaluation for Normal Conditions with Hydrogen Water Chemistry," Revision 0.
- 33.
USNRC Regulatory Guide 1.98, "Assumptions Used For Evaluating The Potential Radiological Consequences Of A Radioactive Offgas System Failure In A Boiling Water Reactor," March 1976.
- 34.
SSES Technical Specification 3.7.5, "Main Condenser Off Gas," Amendments 151 and 178, Bases for Improved Specification B 37.5.
- 35.
PPL Calculation EC-RADN-1 134, "Impact of AST on Current NUREG-0737 Radiological Evaluations that use TID-14844 DBA-LOCA Releases," Revision 0.
- 36.
Tfr48, 1d ison 2, T
G-hi l p11,R*
echnicah pecificatie
'ask F
'e -
,1ipro~ved Sandard T;Inical Specif 'ations Chý ge Travel~
to PLA-5963 Page 60 of 60
7.0 REFERENCES
- Continued
- 37.
PAVAN, "An Atmospheric Dispersion Program for Evaluating Design Basis Accidental Releases of Radioactive Materials from Nuclear Power Stations," NUREG/CR-2858, November, 1982.
- 38.
USNRC Regulatory Guide 1.49, 'Tower Levels of Nuclear Power Plants,"
December, 1973.
P.
-5916, SSES Finoresolution Generic letter 2 =-01 Control Roj Habit.it ket Nos. 50-387 nd 50-388,4 ne 28 2005.
to PLA-5963 Safety Assessment for the Proposed Technical Specification And Bases Changes - Units 1 & 2 to PLA-5963 Page 3 of 12 Description and Safety Assessment for Specific Changes to TS and TS Bases Change
- 3 Change
- 3 Current Technical Specification:
Unit 1 & 2, Section 3.3.6.1 Table 3.3.6.1-1 (page 5 of 6)
Item 5.e, SLC System Initiation APPLICABLE MODES OR OTHER SPECIFIED CONDITIONS = 1, 2 Proposed Change:
Unit 1 & 2, Section 3.3.6.1 Table 3.3.6.1-1 (page 5 of 6)
Item 5.e, SLC System Initiation APPLICABLE MODES OR OTHER SPECIFIED CONDITIONS = 1, 2,3 Basis / Safety Assessment:
Boron injection from the SLC system is required for suppression pool pH control during a DBA LOCA. The maintenance of a suppression pool pH level above 7.0 is important to prevent re-evolution of iodine from the suppression pool water. Consequently, operation of the SLC system was revised to address reactor modes during a DBA LOCA.
Change
- 4 rrent Technical Specification:
Unit
& 2, Section 3.7.3 LCO 3. 3 ACTION CONDITION A: One CREOAS subsystem in rable ACTIONS: CO ITION B: Two CREOAS subsystems inope le due to inoperable Control Room habitability en ope boundary in MODES 1, 2, and 3.
ACTIONS: REQUIRED A ION: B.I Restore Control Room habitability en ope boundary to OPERABLE status.
ACTIONS: COMPLETION TIME:
.1:24 hours ACTIONS: CONDITION F: Two AS subsystems inoperable during movement irradiated fuel assemblies in the secondary containment, during CORE ALTERATIS, oS during OPDRVs.
SR 3.7.3.4, SURVEILLANCE :
rify each CREOAS subsystem can mai in a positive pressure of> 0.125 inches ater gauge relative to the outside atmosphere ring the pressurization/filtra n mode of operation at a flow rate _< 5810 m.
SR 3.7.3.4, QUENCY: 24 months on a STAGG D TEST BASIS Proposed Change:
Unit 1 & 2, Section 3.7.3 LCO 3.7.3 ACTIONS: CONDITION A:
subsystem inoperable for
-s other than Condition B ACTIONS: CONDIT B: One or more CREOAS subsyst inoperable due to inoperable Co ol Room habitability envelope boundary i ODE 1, 2, or 3 ACTIO
- REQUIRED ACTION: B.1 Im ment mitigating actions AND B.2 store Control Room habitability envelope boundary to OPERABLE status ACTIONS: COMPLETION TIME:
B.1:Immediately B.2:24 hours ACTIONS: CONDITION F: Two CREOAS susystems inoperable during movement of diated fuel assemblies in the secondary con nment, during CORE ALTERATIONS, or during PDRVs. OR Required Action and associate ompletion Time of Condition B not met dur movement of irradiated fuel assemblies in s ondary containment, during CORE ALTERA ONS, or during OPDRVs.
SR 3.7.3.4, SURVE CE: Verify Control Room boundary integri in accordance with the Control Room habitab Hty Program.
SR 3.7.3.4, FREQUENCY: I ccordance with the ControltRoom habitability,oram.
to PLA-5963 Page 4 of 12 Description and Safety Assessment for Specific Changes to TS and TS Bases Change Basis Safety Ass ent:
- 4 In NRC Generic Letter 003-01, Licensees were alerted tof ings at facilities that existing Technical Specifications surveillanc requirements for the Control oom Emergency Filtration System (CREFS) may not be adequiate. Spec cally, the results of trac gas tests at facilities indicated that the differential pressure surveillan is not a reliable thod for demonstrating Control Room integrity.
The Technical Specificion Task e
e Nuclear Energy Institute Control Room Habitability Task Force have developed proposed ges to the Improved Standard Technical Specifications (NUREGs 1430 through 1434) to r ace ie differential pressure surveillance with a tracer gas surveillance and to institute a C ol Room bitability Program that will ensure that Control Room Habitability is maintained.
These changes were in rporated into TSTF-448, R ision 2, Technical Specification Task Force -
Improved Standard echnical Specifications Change veler (Reference 12.2). As a result of this Traveler, TS Section 3.7.3 was added.
Nc~ to PLA-5963 Page 5 of 12 i
Description and Safety Assessmldt for Specific Change 1
,¢ urrent Technical Specification:
Unit & 2, Section 5.5 The 1 item number is 5.5.12, "Primary Cont~ain nt Leakage Rate Testing Program".
A new sectio 5.5.13 was added.
Proposed Change:
Unit 1 & 2, Section 5.5 5.5.13 Control Room Habitabilitv Proaram A Control Room Habitab-ty Program shall be established and imple ted to ensure that Control Room habita ity Is maintained such that, with an OPE BLE CREOAS System, Control Room oc pants can control the reactor safely u er normal conditions and maintain it in safe condition following a radiological ent, hazardous chemical release, or a smoke ailenge from outside the Control Room env lope. The program shall ensure that adequa radiation protection is provided to permi access and occupancy of the Control Roo under accident conditions without pe innel receiving radiation exposures in e cess of 5 rem total effective dose equivalent EDE) for the duration of the accident. The program shall include the following elements:
- a.
The definition of the Control Room envelope and the Control Room boundary; Requirements for maintaining Control Room boundary integrity, cluding configuration control, mi agement of breaches, and pre ntive maintenance.
- c.
Reqm ements for assessing Control Room h bitability at the frequencies specified Regulatory Guide 1.197 "Demonst ring Control Room Envelope Inte
.ty at Nuclear Power Reactors" Revion 0, May 2003.
- d. Requirements for termining the unfiltered air inlea e past the Control Room bounda into the Control Room envelope accordance with the testin ethods and at the frequencies speci d in Regulatory Guide 1.197, Revi n 0, May 2003.
Continued next page
i
~-& Se to PLA-5963 Page 6 of 12 Description and Safety Assessmelit for Specific Changand TS Bases/
I Change
- 5 C
ent Technical Specification:
Conti ued
/
Proposed Chang:ý Continued
- e.
Measurement of the ontrol Room envelope positive essure relative to outside atmosp re during the pressurizatio9 ode of operation by one subsyst of the CREOAS System every 24 fonths on a STAGGERED TEST SIS. The results shall be trended and compared to the positive presdure measurements taken or to be ta/n during the Control Room jhleakage testing. These evaluations shall be used as part of an assessment of Control Room boundary integrity between Control Room inleakage tests.
- f.
The quantitative limits on unfiltered air inleakage past the Control Room boundary into the Control Room envelope. These limits shall be stated in a manner to allow direct comparison to the unfiltered air inleakage measured by the testing described In paragraph d.
The unfiltered air inleakage limits must demonstrate that radiation dose and hazardous chemical exposure to the Control Room occupants will be within the assumptions in the licensing basis.
- g.
Limitations on the use of compensatory measures to consider the CREOAS System OPERABLE when there are degraded or nonconforming conditions t result in the unfiltered air i eakage through the Control Room ho dary Into the Control Room enve e greater than the unfiltered lnlea eassumed in the licensing basis analyses. Compensatory measures are interim ac ons used to maintain OPERABI Y of the CREOAS System until U qualification of the Control Room oundary is restored.
Degraded or non onforming conditions affecting the Cont I Room boundary integrity should be solved in a time frame commensurat with the safety significance of the con ition.
I
r~v'-)~~
0 attachment 5 to PLA-5963 Page 7 of 12 DescrItion and Safety Assessmen 1 1or
_eciic Changes to TS Basg Change
- 5 Change
- 5 Current Techni Specification:
Continued
/
Proposed Change:
Continued The program place additional limits on the of compensatory measures clh address a degraded or nonco rmlng Control Room barrier that r ts in unfiltered air inleakage in e Control Room envelope greater the unfiltered air inleakage assumed in the licensing basis analysis for the following two conditions:
- 1. When such compensatory measures may adversely affect the ability of the Control Room occupants to respond to an accident (including, but not limited to, the use of personal air filtration or bottled air systems),
their use may be credited to support OPERABILITY of the CREAOS System until the next entry Into MODE 2 following a fueling outage or for a maximum of 12 m ths, whichever is greater; and
- 2. Wh such compensatory measures may comp te the response of the Control Room occupan to an accident (including, but not limited to, e use of potassium iodine, temporary m configurations, or manual actions), their may be credited to support OPERABILIT of the CREAOS System for a maximum of nths.
The provision of SR 3.0.2 is applica e to the Control Room Inleakage testing frequencies.
J
jAttachment 5 to PLA-5963 Page 8 of 12 Description and Safety Assess ent for Specific Changes to TS and TS Bases Change Basis I Safety Assessment:
- censees were alerted to ndings at facilities that existing technical specifications surveillance require mts for the. Control ioon Emergency Filtration System (CREFS) may not be adeq'ate. Specifically the esults of tracer as tests at facilities indicated that the differential pressure surveillance is not reliable od for demonstrating Control Room integrity.
The Technical Specification Task force an'ct th uclear Energy Institute Control Room habitability task Force have developed proposed changeo the Improved Standard Technical Specifications (NUREGs 1430 through 1434) to replace e li ferential pressure surveillance with a tracer gas surveillance and to institute a Control Rin Hab'i ability Program that will ensure that Control Room habitability is maintained.
/
These changes were incorporatecinto TSTF-448, Revis n 2, Technical Specification Task Force -
Improved Standard Technical tecifications Change Travy er (Reference 12.2). As a result of this Traveler, TS Section 5.5.13, s added. Please note, that P is aware that this Traveler may be revised in the near future an( require additional revisions to Section 5.5.13.
to PLA-5963 Page 11 of 12 Description and Safety Assessment for Specific Changes to TS and TS Bases Change Basis / Safety Assessment:
- 8 Changes were made to the TS Bases for clarity and to conform to the changes made to the associated TS. The revisions to the TS bases incorporate supporting information for the proposed TS changes.
Bases do not establish actual requirements, and as such, do not change technical requirements of the TS. The Bases changes are therefore acceptable, since they administratively document the reasons and provide additional understanding for the associated TS requirements.
Change Current Technical Specification:
Proposed Change:
- 9 The TS Bases provide an explanation and rationale Associated changes to the TS Bases were made for associated TS requirements, and in some cases, to provide some relief for establishing secondary how they are to be implemented. The current TS containment drawdown pressure and still provide Bases of Section B 3.6.4.1, SURVEILLANCE significant margin with DBA LOCA drawdown REQUIREMENTS, provides the maximum time requirements. The maximum drawdown drawdown time required by the SGTS to establish time was increased from 125 and 117 seconds and maintain the secondary containment to for Zones I, IL & MI and Zones I & mI
> 0.25 inches of vacuum water gauge. Per respectively, to 300 seconds for both cases. of this LAR, a drawdown time of 600 seconds was utilized in the DBA LOCA This change is administrative in nature.
analysis.
Change Basis / Safety Assessment:
- 9 Changes to the TS Bases were made to provide some relief for-establishing secondary containment drawdown pressure and still provide significant margin with DBA LOCA drawdown time requirements. The surveillance requirement establishes a time of 300 seconds for the maximum drawdown time. The DBA LOCA analysis assumes a maximum drawdown time of 10 minutes for the unfiltered release to the environs. Consequently, the change in the allowable maximum drawdown time does not represent an increase in the calculated Control Room, EAB, or LPZ doses.
Changes were made to the TS Bases for clarity and to conform to the changes made to the associated TS. The revisions to the TS bases incorporate supporting information for the proposed TS changes.
Bases do not establish actual requirements, and as such, do not change technical requirements of the TS. The Bases changes are therefore acceptable, since they administratively document the reasons and provide additional understanding for the associated TS requirements.
Change C
nt Technical Specification:
Proposed Change:.
- 10N The T ases provide an e lanation and rationale sociated changes to e TS Bases were made for associ ed TS requirement and in some cases, to a ress the requiremen of TSTF-448, how they ar 0 be implemented.
he current TS Revisi 2, Technical Spec ation Task Force -
Bases of Secti B 3.7.3 discusses e Control Improve Standard Technical ifications Room emergenc utside air supply (
EOAS)
Change Tr eler and incorporate e Control syste m,
,,R oom H abita ility Program.
This change is administrative in nature.
-Tjj to PLA-5963 Page 12 of 12 Description and Safety Assessmen for Specific Changes to TS and TS Bases Change Basis / Safety Assessment:
- 10 Changes were made to the T Bases for clarity and to conform to the changes made to the associated TS. The revisions to the TS b'es incorporate supporting information for the propokd TS changes.
Bases do not establish actual req *ements, and as such, do not c ange technical requ*reraents of the TS. The Bases changes are therefo acceptable, since they admi *stratively documenth4e reasons and provide additional understanding for e associated TS requirement NRC Generic Letter 2003-01, Licens were alerted to findings at cilities that existingýtechnical s
ifications surveillance requirements f the Control Room Emergen Filtration System"(CREFS) may ot be adequate. Specifically, the resu. of tracer gas tests at facilitie indicated that the \\
differ tial pressure surveillance is not a relia le method for demonstrating ntrol Room integt.
The Tec
'cal Specification Task force and the uclear Energy Institute Contro oom Habitabilit Task Force ave developed proposed changes to t Improved Standard Technic pecifications0 (NUREGs 1,0 through 1434) to replace the differ tial pressure surveillance with tracer gas surveillance an to institute a Control Room Habitabhty Program that will ensure th Control Room habitability is ma tained.
These changes were corporated into TSTF-448, Revision, Technical Specification T Force -
Improved Standard T nical Specifications Change Travel (Reference 12.2). As a result Iof this Traveler, TS Section 3.7. was revised. Please note, that PPL i aware that this Traveler may be revised in the near future require additional revisions to TS Se tion 3.7.3.
Change Current Technical Specificati Proposed Change:
- 11 The TS Bases provide an explanation and rationale Associated changes to the TS Bases were made for associated TS requirements, and in some cases, to update the FHA to reflect RG 1.183 fuel rod how they are to be implemented. The current TS gap release fractions. The original analysis Bases of Section B 3.9.6, APPLICABLE SAFETY assumes that 10% of the total fuel rod iodine ANALYSES, provides the fuel rod gap release inventory in the gap is available for release. Per fractions per RG 1.25 for a FHA. Per Attachment RG 1.183 requirements, 8% of the 1-13 1, and 5%
2 of this LAR, the FHA was revised to reflect the of the 1-132, 1-133, 1-134, & 1-135 inventory is new release fractions of RG 1.183.
available for release from the gap.
This change is administrative in nature.
Change Basis / Safety Assessment:
- 11 Changes were made to the TS Bases for clarity and to conform to the changes made to the associated TS. The revisions to the TS bases incorporate supporting information for the proposed TS changes.
Bases do not establish actual requirements, and as such, do not change technical requirements of the TS. The Bases changes are therefore acceptable, since they administratively document the reasons and provide additional understanding for the associated TS requirements.
to PLA-5963 Proposed Technical Specification Changes Units 1 & 2 Mark-ups to PLA-5963 Page 1 of 1 Table 6-1: List of Proposed Technical Specification Changes (Marked ups)
Units s
& 2 Sections Title 1.1 Definitions 3.1.7 Standby Liquid Control (SLC) System 3.3.6.1 Primary Containment Isolation Instrumentation C totrol
ý~mEer~ncy O ide irf Sup"t'
- M*
e2.Y 2*sa
- /
PPL Rev. O CREOAS System 3.7.3 LCO 3 I.3IT APPLICABILITY:
Control Room Emergency Outside Air Supply (CREOAS) Syste.
Two CREOAS subsystems shall be OPERABLE.
NOTE-- 7 The control room habitability envelope boundary be opened Intermittently under administrative control.
"S 1, 2, and 3, movement of Irradiated fuel a mblies in the secondary containment, CORE ALTERATIONS, operations with a potential draining the reactor vessel (OPDRVs).
ICREOAS subsystems o,*e noperable due to inoperabl control room habitability enveloj; bouar in
~/
status.
C. Required Action a associated Comn etion Time of Condition A B not met in MODE 1, 2or 3.
C.1 Be in MODE 12 hours1.388889e-4 days <br />0.00333 hours <br />1.984127e-5 weeks <br />4.566e-6 months <br /> AND 36 hours4.166667e-4 days <br />0.01 hours <br />5.952381e-5 weeks <br />1.3698e-5 months <br /> SUSQUEHANNA - UNIT 1 "TS / 3.7-6 Amendment 203
PPL Rev. 0 CREOAS System 3.7.3 ACTIONS (continued)
CONDITION.
REQUIRED ACTION COMPLETION TIME F. Two CREOAS subsystems inoperable during movement of irradiated fuel assemblies in the secondary containment, during CORE ALTERATIONS, or durp OPDRVs.
/
NOTE LCO 3.0.3 Is not applicable.
F.1 Suspend movement of Irradiated fuel assemblies In the secondary containment.
AND F.2 Suspend CORE ALTERATIONS.
AND F.3 Initiate action to suspend OPDRVs.
Immediately Immediately Immediatey
(
I SURVEILLANCE REQUIREMENTS SURVEILLANCE FREQUENCY SR 3.7.3.1 Operate each CREOAS filter train for k 10 31 days continuous hours with the heaters operable.
SR 3.7.3.2 Perform required CREOAS filter testing in In accordance with the VFTP accordance with the Ventilation Filter Testing Program (VFTP).
SR 3.7.3.3 Verify each CREOAS subsystem actuates on an 24 months actual or simulated initiation signal.
(continued)
SUSQUEHANNA - UNIT 1 TS / 3.7-8 Amendment 203
PPL Rev. 0 CREOAS System 3.7.3 SUSQUEHANNA - UNIT 1 3.7-9 Amendment 178
Programs and Manuals 5.5 5.5.11 Safety Function Determination Pro-gram (SFDP) (continued)
The SFDP identifies where a loss of safety function exists. If a loss of Safety furttihonls determined to exist by this program, the appropriate Conditions and Required Actions of the LCO in which the loss of safety function exists are required to be entered. When a loss of safety function is caused by the inoperability of a single Technical Specification support system, the appropriate Conditions and Required Actions to enter are those of the support system.
5.5.12 Primary Containment Leakaqe Rate Testina Proaram A program shall be established, implemented, and maintained to comply with the leakage rate testing of the containment as required by 10 CFR 50.54(o) and 10 CFR 50, Appendix J, Option B, as modified by approved exemptions. This program shall be in accordance with the guidelines contained in Regulatory Guide 1.163, "Performance-Based Containment Leak-Test Program", dated September 1995, as modified by the following exception:
- a. NEI 94-01-1995, Section 9,2.3: The first Type A test performed after the May 4, 1992 Type A test shall be performed no later than May 3, 2007.
The peak calculated containment internal pressure for the design basis loss of coolant accident, Pa, is 45.0 psig.
The maximum allowable primary containment leakage rate, La, at Pa, shall be 1% of the primary containment air weight per day.
Leakage Rate Acceptance Criteria are:
- a. Primary Containment leakage rate acceptance criterion is < 1.0 La. During each unit startup following testing in accordance with this program, the leakage rate acceptance criteria are :5 0.60 La for Type B and Type C tests and 5 0.75 La for Type A tests:
- b. Air lock testing acceptance criteria are:
- 1) Overall air lock leakage rate Is < 0.05 La when tested at 2: Pa.
- 2) For each door, leakage rate Is _5 5 scfh when pressurized to ?- 10 psig.
The provisions of SR 3.0.2 do not apply to the test frequencies specified In the Primary Containment Leakage Rate Testing Program.
The provisions of SR 3.0.3 are applicable to the Primary Containment Leakage Rate Testing Pram SUSQUEHANNA - UNIT 1 TS /5.0-18 Amendment 209
Insert 2:
5.5.13 Control Room Habitability Progm A
ontrol Room Habitability Program shall be established and imple nted to ensu that control room habitability is maintained such that, with PERABLE CRE S System, control room occupants can control the reactor afely under no maa onditions and maintain it in a safe condition following radiological event, h ardous chemical release, or a smoke challenge fro utside the control room. enve pe..The program shall ensure that adequate ation protection is provided to t access and occupancy of the control
. m under accident conditions wi out personnel receiving radiation exposs in excess of 5 rem total effective d e equivalent (TEDE) for the duratio of the accident. The program shall inc de the following elements:
- a. The definition of control room envelope d the control room boundary;
- b. Requirements for m ntaining control room undary integrity, including configuration control, anagement of b hes, and preventive maintenance.
- c. Requirements for asses i g control roo abitability at the frequencies specified in Regulatory
- de 1.197 monstrating Control Room Envelope Integrity at Nuclear Power eactors,"
evision 0, May 2003.
- d. Requirements for determinin the filtered air inleakage past the control room boundary into the conrrm envelope in accordance with the testing methods and at the frequencies
'fled in Regulatory Guide 1.197, Revision 0, May 2003.
- e. Measurement of the control m e velope-positive pressure relative to outside atmosphere durin e press *zation mode of operation by one subsystem of the CREO System ev 24 months on a STAGGERED TEST BASIS. The resits shall be tred and compared to the positive pressure measureme taken or to be talc during the control room inleakage testing. These cvleations shall be used as of an assessment of control room boundary i erity between control roorinleakage tests.
- f. The quantitativ on unfiltered air inleaka past the control room boundary into e control room envelope. These mits shall be stated in a manner to all w direct comparison to the unfilte ý inleakage measured by the testing cribed in paragraph d. The unfiltered r inleakage limits must demons that radiation dose and hazardous chemic exposure to the control r m occupants will be within the assumptions i the licensing basis.
- g. Limita ons on the use of compensatory measures to cons er the CREOAS Syst OPERABLE when there are degraded or nonconfo ing conditions th result in the unfiltered air inleakage through the control m boundary i o the control room envelope greater than the unfiltered inle age assumed i the licensing basis analyses. Compensatory measures are in tm actions used to maintain OPERABILITY of the CREOAS System until qualification of the control room boundary is restored. Degraded or nonconforming conditions affecting the control room boundary integrity
should be res ved in a time frame commensurate wi e safety significance of the conditio The program shall place additionaltmits on the use of" compensatory m asures which address a degrad r nonconforming control room barrier that ults in unfiltered air inleaae" into the control room envelope greater th the unfiltered air inleak e assumed in the licensing basis analysis for the lowing two conditis I. When such co nsatory mess may adversely affect the ability of the control room upants to spond to an accident (including, but not limited to, the us of nal air filtration or bottled air systems),
their use may be credi t support OPERABITY of the CREAOS System until the next en into MODE 2 following a refueling outage or for a maximum of 12 nths, whichever is greater, and
- 2. When such compensat asures may complicate the response of the control room occ ants t an accident (including, but not limited to, the use of potass iodine temporary system configurations, or manual actions),
eir use may credited to support OPERABILITY of the CREAOS ystem for a mum of 36 months.
The provision of SR 3.. is applicable to the c trol room inleakage testing frequencies.
(
PPL Rev. 0 CREOAS System 3.7.3 3.7 PLANT SYSTEMS 3.7.3 LCO 3.7.3, Control Room Emergency Outside Air Supply (CREOAS)
Two CREOAS subsystems shall be OPERABLE.
/
The control room habitability envelope bounda ay be opened intermittently under administrative control.
MODES 1, 2, and 3, During movement of irradiated fuel During CORE ALTERATIONS, During operations with a potentialf(
in the secondary containment, 4Ehour status.
C. Required Action a associated Corn etion Time of Condition A B not met in MODE 1, 2 or 3.
12 hours1.388889e-4 days <br />0.00333 hours <br />1.984127e-5 weeks <br />4.566e-6 months <br /> 36 hours SUSQUEHANA-UNIT 2 TS /13.7-6 Amendment 177
PPLRev. 0 CREOAS System 3.7.3 SURVEILLANCE REQUIREMENTS SURVEILLANCE FREQUENCY SR 3.7.3.1 Operate each CREOAS filter train for > 10 31 days continuous hours with the heaters operable.
SR 3.7.3.2 Perform required CREOAS filter testing in In accordance with the VFTP accordance with the Ventilation Filter Testing Program (VFTP).
SR 3.7.3.3 Verify each CREOAS subsystem actuates on an 24 months actual or simulated initiation signal.
(continued)
SUSQUEHANA - UNIT 2 "TS / 3.7-8 Amendment 177
IL t
C)~~~~f (A~w5e
~~~o PPL Rev.0 CREOAS System 3.7.3 SUSQUEHANA - UNIT 2 TS / 3.7-9 Amendment 151
Programs and Manuals 5.5 5.5 Programs and Manuals 5.5.11 Safety Function Determination Program (SFDP) (continued)
The SFDP identifies where a loss of safety function exists. If a loss of safety function is determined to exist by this program, the appropriate Conditions and Required Actions of the LCO in which the loss of safety function exists are required to be entered. When a loss of safety function is caused by the inoperability of a single Technical Specification support system, the appropriate Conditions and Required Actions to enter are those of the support system.
5.5.12 Primary Containment Leakaqe Rate Testing Prooram A program shall be established, implemented, and maintained to comply with the leakage iate testing of the containment as required by 10 CFR 50.54(o) and 10 CFR 50, Appendix J, Option B, as modified by approved exemptions. This program shall be in accordance with the guidelines contained in Regulatory Guide 1.163, "Performance-Based Containment Leak-Test Programs, dated September 1995, as modified by the following exception:
- a. NEI 94-01-1995, Section 9.2.3: The first Type A test performed after the October 31, 1992 Type A test shall be performed no later than October 30, 2007.
The peak calculated containment Internal pressure for the design basis loss of coolant accident, Pa, is 45.0 psig.
The maximum allowable primary containment leakage rate, La, at Pa. shall be 1% of the primary containment air weight per day.
Leakage Rate Acceptance Criteria are:
- a. Primary Containment leakage rate acceptance criterion is < 1.0 La. During each unit startup following testing in accordance with this program, the leakage rate acceptance criteria are _ 0.60 La for Type B and Type C tests and < 0.75 La for Type A tests;
- b. Air lock testing acceptance criteria are:
- 1) Overall air lock leakage rate is < 0.05 La when tested at;- Pa,
- 2) For each door, leakage rate is <_ 5 scfh when pressurized to > 10 psig.
The provisions of SR 3.0.2 do not apply to the test frequencies specified in the Primary Containment Leakage Rate Testing Program.
The provisions of SR 3.0.3 are applicable to the Primary Containment Leakage Rate TetnPoram.
SUSQUEHANNA - ýUNIT2 TS / 5.0-18 Amendment 1A1, 1 183
Insert 2:
5.5.13 Contr I Room Habitability Program A Cont Room Habitability Program shall be established and implemen to ensure tha ontrol room habitability is maintained such that, with an OP LE CREOAS S tem, control room occupants can control the reactor safe under normal condit ns and maintain it in a safe condition following a radogical event, hazardou hemical release, or a smoke challenge from outsi e the control room envelope.
e program shall ensure that adequate radiation tection is provided to permit cess and occupancy of the control room unr accident conditions without onnel receiving radiation exposures in cess of 5 rem total effective dose eq ivalent (TEDE) for the duration of th accident. The program shall include th following elements:
- a. The definition of the co I room envelope and the ontrol room boundary;
- b. Requirements for maintaii ng control room bound integrity, including configuration control, man ement of breaches, d preventive maintenance.
- c. Requirements for assessing c ntrol room habita ity at the frequencies specified in Regulatory Guide.197 "Demons ting Control Room Envelope Integrity at Nuclear Power Rea ors," Revisi 0, May 2003.
- d. Requirements for determining th unfiltered inleakage past the control room boundary into the control r envel pe in accordance with the testing methods and at the frequencies spe ied i Regulatory Guide 1.197, Revision 0, May 2003.
- e. Measurement of the control room enve positive pressure relative to outside atmosphere during the s
on mode of operation by one subsystem of the CRErAS Sys e ry months on a STAGGERED TEST BASIS. The results shall be ded d compared to the positive pressure measurements taken or t tkn-ng the control room inleakage testing. These evaluations shall us p
of an assessment of control room boundary integrity betw control room leakage tests.
- f. The quantitative limits on u te air inleakag past the control room boundary into the control ro envelope. These ts shall be stated in a manner to allow direct co arison to the unfilteredr inleakage measured by the testing described in p rph d. The unfiltered r inleakage limits must demonstrate that radiati dose and hazardous chemi exposure to the control room occupan will beswitin the assumtionsAn the licensing basis.
- g. Limitations on the us of compensatory measu to con 'der the CREOAS System OPERAB hen there are degraded or nonco n ng conditions that result in the un Itered air inleakage through the contro oom boundary into the control ro m envelope greater than the unfiltered inl age assumed in the licensing is analyses. Compensatory measures are in im actions used to mainta OPERABILITY of the CREOAS System until 11 qualification f the control room boundary is restored. Degraded r nonconfo ng conditions affecting the control room bound in ty
should be res ved in a time frame co surate with the safety significance of the conditio The program shall pla additional limits on the use of compensatory mi ures which addres degraded or nonconforming control room barrier that-r sults in unfiltereai inleakage into the control room envelope greater th the unfil air inleakage assumed in the licensing basis analysis for the ollowing o conditions:
- 1. When such con s
measures may adversely affect the ability of the control room c
ants to respond to an accident (including, but not limited to, the e of personal air filtration or bottled air systems),
their use may be eted to support OPERABILITY of the CREAOS System until the ext try into MODE 2 following a refueling outage or for a maxim in of 1 onths, whichever is greater; and
- 2. When such c pensato easures may complicate the response of the control r om occupan o an accident (including, but not limited to, the use f potassium iodi, temporary system configurations, or manual aqons), their use ma bcredited to support OPERABILITY of the C OS System for a mum of 36 months.
The provision of R 3.0.2 is applicable to the ontrol room inleakage testing frequencies.
to PLA-5963 For Information -
Proposed Technical Specification Bases Changes Units 1 & 2 Mark-ups to PLA-5963 to PLA--5963 Page 1 of 1 Table 7-1: List of Proposed Technical Specification Bases Chanees (Marked ups)
Units 1 & 2 Sections Title B 2.1.1 Reactor Core SLs B 2.1.2 Reactor Coolant System (RCS) Pressure SL B 3.1.7 Standby Liquid Control (SLC) System B 3.1.8 Scram Discharge Volume (SDV) Vent and Drain Valves B 3.2.3 Linear Heat Generation Rate (LHGR)
B 3.3.6.1 Primary Containment Isolation Instrumentation B 3.3.6.2 Secondary Containment Isolation Instrumentation B 3.3.7.1 Control Room Emergency Outside Air Supply (CREOAS)
System Instrumentation B 3.4.7 Reactor Coolant Specific Activity B 3.6.1.1 Primary Containment B 3.6.1.3 Primary Containment Isolation Valves (PCIVs)
B 3.6.4.1 Se da ty_~.ainment..
B 3.7.5 Main Condenser Offgas B 3.7.7 Spent Fuel Storage Pool Water Level B 3.9.6 Reactor Pressure Vessel (RPV) Water Level to PLA-5963 Activities to be Completed Before AST Implementation to PLA-5963 Page 2 of 5
- 2.
Applicable Sections of the TS and Bases were revised to reflect changes associated wit entation of the proposedks he:
I abitab' awsasio 5.4,13.,".
There are new manual operator actions associated with the SLC System reqtiired as part of this LAR that are not currently considered in the SSES design basis and must be directed by new Emergency Operation procedures that will be written and approved before SSES AST implementation.
The operator actions assumed in the proposed DBA LOCA AST dose consequence analyses is the initiation of the SLC system for boron injection to maintain the suppression pool water pH above 7.0, precluding iodine re-evolution. TS Sections 3.1.7, "Standby Liquid Control (SLC) System and 3.3.6.1, "Primary Containment Isolation Instrumentation" and their Bases were also revised to address this change in the SLC system requirements (see Attachments 5 through 7).
No hardware changes are necessary to use SLC in this new functional mode.
Applicable procedure(s) will be reviewed/revised as necessary to ensure the operation of the SLC System during a DBA LOCA. See #4.
Applicable sections of the TS Matrix shall be revised.
As a result of the revision to the TS Bases concerning the definition of Dose Equivalent 1-131, Plant Chemistry will evaluate revising appropriate software, counting system library (data file), and chemistry (CH) and emergency plan position specific (EP-PS) procedures. This assumes that the limiting values for coolant concentrations of 0.2 and 4.0 uCi/g DE 1-131 are not re-evaluated. If Chemistry chooses to re-evaluate the limits, other changes would also be required.
A markup of the TS and Bases impacted by implementation of the AST is provided in Attachments 6 and 7 respectively.
Applicable procedure(s) requirements and TS Matrix revisions shall be completed prior to implementation of the AST License Amendments for Units I & 2.
The necessary software, data file, and procedural changes required to reflect the change in the DE 1-131 definition shall be completed prior to implementation of the AST License Amendments for Units I & 2.
- 3.
Nuclear Fuels Engineering Technical Instruction NF-202 The procedural changes required shall be modified to incorporate limits on LHGR for to reflect compliance with burnups exceeding 54 GWD/MTU per Footnote 11 of Footnote 11 of RG 1.183 shall be RG 1.183 and core average burnup of 39 GWd/MTU.
completed prior to implementation of the AST License Amendments for Units 1 & 2.
to PLA-5963 Page 5 of 5
- 7.
The x/Qs calculated at the CRHE outside air intake are based on a new location, located on the roof of the Unit 2 Reactor Building (at column lines U and 36). A preliminary evaluation of the new location determined that seismic and security concerns were found to be acceptable. A more thorough evaluation of the acceptability of the new CRUE outside air intake location, including the impact of hazardous chemical and smoke on CRHE operators will be conducted.
Appropriate drawings, station modification package(s),
10 CFR 50.59s, hazardous chemical and smoke evaluations, and other activities, as deemed appropriate, will be completed.
Generate applicable TS changes to allow for the shutdown of the CREOASS to allow for connection of the existing CRHE ductwork to the new air intake.
The documentation and evaluation of the new CR-E outside air intake shall be completed prior to implementation of the AST License Amendments for Units 1 & 2.
Appropriate drawings, station modification package(s),
10 CFR 50.59s, hazardous chemical and smoke evaluations, TS changes and other activities, as deemed appropriate, shall be completed prior to implementation of the AST License Amendments for Units 1 & 2.
- 8.
The following assumptions were utilized in the AST analysis, based on projected EPU values:
- For the MSLB accident, the mass releases were increased by 20%.
For the DBA LOCA, a 50% reduction of primary containment leakage, secondary containment bypass leakage, and MSIV leakage at 24 hours2.777778e-4 days <br />0.00667 hours <br />3.968254e-5 weeks <br />9.132e-6 months <br />.
These assumptions shall be verified when the information becomes available, but prior to AST implementation.
The evaluations shall be completed prior to implementation of the AST License Amendments for Units 1 & 2.
- 9.
The Emergency Plan and implementing procedures will be The reviews shall be completed reviewed and updated as appropriate to reflect TEDE.
prior to implementation of the AST License Amendments for Units I & 2.
- 10. The C ntrol Room Hab/ ability Programshal 'be developed.
evelopment of Pe Control
/oom H~abitab)Kty Progra hall be completefprior to implemeotion of the T
LicensyAmendment or/
Units 4 &2.
-T, r-rIJ07
, '--J to PLA-5963 No Significant Hazards Consideration Determination & Environmental Consideration for the Proposed Changes
£Attachment 9 to PLA-5963 Page 1 of 6 NO SIGNIFICANT HAZARDS CONSIDERATION DETERMINATION Description of Amendment Request PPL Susquehanna, LLC (PPL) is proposing to amend the operating license for Susquehanna Steam Electric Station (SSES) Units 1 and 2, by revising the Technical Specifications (TS) and incorporating an *itemative 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 Altemativq Source Terms." The four BWR limiting design basis accidents (DBAs) identified in RG 1.183 considered were the Control Rod Drop Accident, the Refueling Accident, the Loss of Coolant Accident, and the Main Steam Line Break Accident. As a result of the application of a revised accident source term, changes are proposed to the TS which revise the definition of dose equivalent 1-13 1, and the operation of the SLC.
The AST analyses are based on new offsite and CRHE atmospheric dispersion coefficients (X/Qs) based on site specific meteorological data determined based on Regulatory Guides 1.145 and 1.194.
I-nn addition torevising the SSES icensing basis to adopt the AST, licensing basis changes are proposed justified to respo to NRC Generic LeIr 2003-01, Con I Room HabitabE y",
dated Jun 12, 2003 (Refere 6e 12. 1). These prop ed changes are pnuant to the Tec 1cal Specifi tion Task Forceie proved Standard Teefnical Specificat ns Change TravqMir TS 8, Revision 2 (Reference 12.2).
Basis for No Significant Hazards Determination:
Pursuant to 10 CFR 50.92, SSES 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.
to PLA-5963 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.
doption of the AST and pursuant TS changes a-tQoe'TS'jreo d&
qR e
er 20-3901,Ref en and the changes to e atmosp enc spersion 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 in the probability of an accident previously evaluated of a DBA discussed 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 revised 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 a DBA 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 well within acceptable limits. This guidance is presented in 10 CFR 50.67, RG 1.183, and Standard Review Plan Section 15.0.1.
Therefore, the proposed amendment does not result in a significant increase in the consequences of any previously evaluated accident.
A to PLA-5963 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. 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.
Licensing basis changes are proposed and justified to credit use of the SLC System to buffer suppression pool pH to prevent iodine re-evolution following a postulated design basis loss of coolant accident. There are new required manual operator actions associated with the SLC System that are not currently considered in the SSES design basis. Operator training will be updated to reflect the new manual operator actions for the pH control function of the SLC System as defined in the TS Section 3.1.7. These changes are not significant because the operators are already trained for the operation of the SLC System. Procedural changes are mostly limited to the timing of SLC initiation and termination. In addition, no new hardware changes are necessary to use SLC in this new functional mode.
Ka result offese changespno new foiure modes4re created'by these clhanges' Therefore, the proposed license amendment will not create the possibility of a new or different kind of accident from any accident previously evaluated.
to PLA-5963 Page 4 of 6 3.0 Does the proposed change involve a significant reduction in a margin of safety?
Response: No.
The results of the accident analyses revised in support of the proposed change 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.
Changes to the SLC System to credit use of the Standby LiQuid Control ISLC) System to buffer suppression pool pH to prevent iodine re-evolutio and-e
ýe
ýA
,;-l*RC(* qapc Le.Wr 2003,l and Pimprove the margin of safety.
New offsite and Control Room atmospheric dispersion factors (x/Qs) based on site specific meteorological data, calculated in accordance with the guidance of RGs 1.145 and 1.194, utilizes more recent data and improved calculational methodologies.
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, SSES 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.
2 to PLA-5963 References 2 to PLA-5963 Page 1 of 1
12.0 REFERENCES
- 1.
NR
'Generic Letter 2W3-01, "Control Ropi Habitability;:-ted June 1,003.
- 2.
T--F 8,Revision 2,)VOG-11, R0, "Technic Specificati'a Task Fo 7 - Impyed St rd Technical sification§ Change Trave*r."
70 to PLA-6112 PPL Susquehanna List of Regulatory Commitments
Rqw to PLA-6112 Page 1 of 1 List of Regulatory Commitments The following table identifies those actions committed to by PPL Susquehanna in this document. Any other statements in this submittal are provided for information purposes and are not considered to be regulatory commitments. Please direct questions regarding these commitments to Michael H. Crowthers.
REGULATORY COMM1TT{ENT DUE DATE.
Submit changes to the SSES Technical After the Notice for Availability is Specifications in accordance with TSTF-448 published in accordance with the to address Control Room Habitability.