Proposed Tech Specs,Permitting Loading of ATRIUM-9B Fuel in Plant Unit Core for Operational Modes 3,4 & 5.Modes Will Support Refueling Activities Such as Fuel Load,Vessel re- Assembly & Single Rod TimingML20138G332 |
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Quad Cities ![Constellation icon.png](/w/images/b/be/Constellation_icon.png) |
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04/29/1997 |
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COMMONWEALTH EDISON CO. |
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ML20138G326 |
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NUDOCS 9705060266 |
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Category:TECHNICAL SPECIFICATIONS
MONTHYEARML20217D7961999-10-12012 October 1999 Proposed Tech Specs Pages,Removing Turbine EHC Low Oil Pressure Trip from RPS Trip Function Requirements in TS Sections 2.2 & 3/4.1.A ML20210R8281999-08-13013 August 1999 Revised Bases Page B.3/4.9-6 to TS Section 3/4.9,providing Clarity & Consistency with Sys Design Description in UFSAR Sections 8.3.2.1 & 8.3.2.2 ML20209J2321999-07-16016 July 1999 Proposed Tech Specs 3/4.7.D Replacing Limit for Any One Msli Valve of Less than or Equal 11.5 Sfch with Aggregate Value of Less than or Equal 46 Scfh for All MSIVs ML20209C2951999-06-29029 June 1999 Proposed Tech Specs Section 3/4.3.C, Reactivity Control - Control Rod Operability ML20205L2631999-04-0505 April 1999 Tech Spec Page B 3/4.5-2 to TS Section 3/4.5, ECCS, to Clarify Requirement Discussed in ML20205J9321999-03-30030 March 1999 Proposed Tech Specs 3/4.6.E Changing SRs 4.6.E.2 to Allow one-time Extension of 18 Month Requirement to Pressure Test or Replace One Half of MSSVs to Interval of 24 Months ML20205J9911999-03-30030 March 1999 Proposed Tech Specs Allowing Alternative Methodology for Quantifying RCS Leakage When Normal RCS Leakage Detection Sys Is Inoperable ML20205J9741999-03-30030 March 1999 Proposed Tech Specs,Deleting Various License Conditions That Have Been Completed,Making Editorial Changes & Providing Clarifying Info ML20199L7741999-01-21021 January 1999 Proposed Tech Specs Bases for Sections 3/4.10.K & 3/4.10.L, Provides Description of Design & Operation of RHR SD Cooling Subsystem ML20199L6921999-01-21021 January 1999 Proposed Tech Specs Section 3/4.6.I,relocating from Chemistry TS Requirements to UFSAR ML20196H4571998-11-30030 November 1998 Proposed Tech Specs 3/4.8.J, Safe Shutdown Makeup Pump, Reducing Current AOT from 67 Days to 14 Days ML20196F6451998-11-30030 November 1998 Proposed Tech Specs 3/4.1.A,3/4.10.B & 3/4.12.B,proposing Changes to Relocate Requirement to Remove RPS Shorting Links Which Enable non-coincident Scram for Neutron Instrumentation,To Licensee Controlled Document ML20155D8091998-10-29029 October 1998 Proposed Tech Specs Bases Sections 3/4.2.D & 3/4.5.D, Providing Clarity & Consistency with Sys Design Description Contained in UFSAR Section 5.4.6.2 ML20151S7991998-08-31031 August 1998 Proposed Tech Specs,Increasing Max Allowable MSIV Leakage from 11.5 Scfh to 30 Scfh Per Valve When Tested at 25 Psig, IAW SR 4.7.D.6 ML20236W8401998-07-31031 July 1998 Proposed Tech Specs Bases 3/4.7.C & 3/4.7.12.C,clarifying Testing Requirements for Primary Containment Excess Flow Check Valves ML20247D7761998-05-0505 May 1998 Proposed Tech Specs Page B 3/4.4-1,changing Administrative Error.Bases for Net Quantity of Gallons for Solution Is Changed from 3254 (Correct Quantity) to 3245 ML20246Q3481998-04-29029 April 1998 TS Page B 3/4.5-3,reflecting Change to TS Bases for Section 3/4.5.C ML20217G1481998-03-27027 March 1998 Proposed Tech Specs Bases Section 3/4.5.A,reflecting Design Info Contained in Rev 4 to Ufsar,Dtd Apr 1997 ML20216C6381997-08-29029 August 1997 Proposed Tech Specs,Incorporating New Siemens' Methodologies That Will Enhance Operational Flexibility & Reducing Likelihood of Future Plant Derates ML20196G0271997-05-0101 May 1997 Proposed Tech Specs 4.9.A.8.b Revising Load Value for Diesel Generator to Be Equal to or Greater than Largest Single Load & Revising Frequency & Voltage Requirements During Performance of Test ML20138G3321997-04-29029 April 1997 Proposed Tech Specs,Permitting Loading of ATRIUM-9B Fuel in Plant Unit Core for Operational Modes 3,4 & 5.Modes Will Support Refueling Activities Such as Fuel Load,Vessel re- Assembly & Single Rod Timing ML20138B3231997-04-21021 April 1997 Proposed Tech Specs,Requesting That NRC Grant Exigent Amend to TS 2.1.B & 6.9.A.6.b to Support Plant Unit 2 Cycle 15 Operation Scheduled to Begin 970519 ML20137G3981997-03-26026 March 1997 Proposed Tech Specs 3/4.7.P Re Standby Gas Treatment & TS 5.2.C Re Secondary Containment ML20135F7321997-03-0303 March 1997 Proposed Tech Spec Bases 3/4.9.E,clarifying Purpose of SR 4.9.E ML20135D9461997-02-24024 February 1997 Proposed Tech Specs,Clarifying Bases for TS Surveillance 4.8.D.5.c ML20138L4011997-02-17017 February 1997 Proposed Tech Specs Section 2.1.B Re Thermal Power,Section 3/4.11 Re Power Distribution Limits,Section 3/4.6 Re Primary Sys Boundary,Section 5.3 Re Reactor Core & Section 6.9 Re Reporting Requirements ML20138L3701997-02-17017 February 1997 Proposed Tech Specs 4.9.A.8.h Re Diesel Generator Endurance Test Surveillance Requirements ML20134D2191997-01-27027 January 1997 Proposed Tech Specs Deleting marked-up Sentence from TS Bases for Section 3/4.7.K ML20129C2391996-10-16016 October 1996 Proposed Tech Specs for Dresden 2 & 3 & Quad Cities 1 & 2, marked-up to Show Transition Verbiage ML20129D3981996-09-20020 September 1996 Proposed Tech Specs 3/4.6.K,updating Pressure-Temp Curves to 22 Effective Full Power Yrs & TS Bases ML20113C3571996-06-25025 June 1996 Proposed Tech Specs Re Upgrade Program ML20113A7861996-06-10010 June 1996 Proposed Tech Specs,App A,To Reflect Transition of Fuel Supplier from General Electric to Siemens Power Corp ML20117D7121996-05-0606 May 1996 Proposed Tech Specs,Implementing New LCO & SR Re Revs to TS for 10CFR50,App J,Lrt ML20107A1881996-04-0404 April 1996 Proposed Tech Specs 3.4/4.4 Re Standby Liquid Control Sys ML20101H1381996-03-25025 March 1996 Complete Version of TS Upgrade Program Pages That Reflect Current Configuration of Plant & Specifies SRs That Will Not Be Current Upon Implementation of Tsup Project ML20097D9231996-02-0808 February 1996 Proposed Tech Specs,Upgrading Existing TS 3/4.5, Eccs ML20098A3821995-09-20020 September 1995 Proposed Tech Specs,Revising TS Upgrade Program & Improving Plant Submittals ML20086D4741995-06-30030 June 1995 Proposed Tech Specs Re TS Upgrade Program for Dresden Units 2 & 3 & Quad Cities Units 1 & 2 ML20087H8651995-05-0202 May 1995 Proposed Tech Specs Re TS Upgrade Program Section 3/4.10 ML20082H7481995-04-10010 April 1995 Proposed Tech Specs,Revising SR for HPCI & RCIC Sys ML20080K8171995-02-23023 February 1995 Proposed Tech Specs,Changing Name of Iige to Reflect Results of Merger Between Iige,Mid American Energy Co,Midwest Power Sys Inc & Midwest Resources Inc ML20078E2051995-01-20020 January 1995 Proposed Tech Specs Re Snubber Visual Insp Intervals ML20078Q6061994-12-12012 December 1994 Proposed TS Section 3.4/4.4 Re Standby Liquid Control Sys ML20064J3181994-03-11011 March 1994 Proposed Tech Specs Re Snubber Visual Insp Intervals ML20059A7321993-12-20020 December 1993 Proposed Tech Specs 1.1/2.1-1 Increasing MCPR Safety Limit from 1.06 to 1.07 for Units 1 & 2 ML20059A8301993-10-21021 October 1993 Proposed Tech Specs Deleting Requirements for Demonstrating Operability of Redundant Equipment When ECCS Equipment Is Found Inoperable or Made Inoperable for Maint ML20125D6381992-12-0808 December 1992 Proposed Tech Specs 3/4.1 Re Reactor Protection Sys ML20116J7091992-11-0606 November 1992 Corrected Proposed TS 3.2/4.2-8 Re Administrative Changes ML20106A6221992-09-15015 September 1992 Proposed TS 2.0, Safety Limits & Limiting Safety Sys Settings & 3/4.11, Power Distribution Limits ML20099D7791992-07-29029 July 1992 Proposed Tech Specs Sections 1.0, Definitions, 3/4.0, Applicability & 3/4.3, Reactivity 1999-08-13
[Table view] Category:TECHNICAL SPECIFICATIONS & TEST REPORTS
MONTHYEARML20217D7961999-10-12012 October 1999 Proposed Tech Specs Pages,Removing Turbine EHC Low Oil Pressure Trip from RPS Trip Function Requirements in TS Sections 2.2 & 3/4.1.A ML20210R8281999-08-13013 August 1999 Revised Bases Page B.3/4.9-6 to TS Section 3/4.9,providing Clarity & Consistency with Sys Design Description in UFSAR Sections 8.3.2.1 & 8.3.2.2 ML20209J2321999-07-16016 July 1999 Proposed Tech Specs 3/4.7.D Replacing Limit for Any One Msli Valve of Less than or Equal 11.5 Sfch with Aggregate Value of Less than or Equal 46 Scfh for All MSIVs ML20196K1941999-06-30030 June 1999 Rev 2.0 to Chapter 11 of Quad Cities Offsite Dose Calculation Manual ML20209C2951999-06-29029 June 1999 Proposed Tech Specs Section 3/4.3.C, Reactivity Control - Control Rod Operability ML20211C3311999-04-30030 April 1999 Rev 2.0 to Generic ODCM for Dresden,Quad Cities,Zion, Lasalle,Byron & Braidwood ML20205L2631999-04-0505 April 1999 Tech Spec Page B 3/4.5-2 to TS Section 3/4.5, ECCS, to Clarify Requirement Discussed in ML20205J9741999-03-30030 March 1999 Proposed Tech Specs,Deleting Various License Conditions That Have Been Completed,Making Editorial Changes & Providing Clarifying Info ML20205J9321999-03-30030 March 1999 Proposed Tech Specs 3/4.6.E Changing SRs 4.6.E.2 to Allow one-time Extension of 18 Month Requirement to Pressure Test or Replace One Half of MSSVs to Interval of 24 Months ML20205J9911999-03-30030 March 1999 Proposed Tech Specs Allowing Alternative Methodology for Quantifying RCS Leakage When Normal RCS Leakage Detection Sys Is Inoperable ML20199L6921999-01-21021 January 1999 Proposed Tech Specs Section 3/4.6.I,relocating from Chemistry TS Requirements to UFSAR ML20199L7741999-01-21021 January 1999 Proposed Tech Specs Bases for Sections 3/4.10.K & 3/4.10.L, Provides Description of Design & Operation of RHR SD Cooling Subsystem ML20196H4571998-11-30030 November 1998 Proposed Tech Specs 3/4.8.J, Safe Shutdown Makeup Pump, Reducing Current AOT from 67 Days to 14 Days ML20196F6451998-11-30030 November 1998 Proposed Tech Specs 3/4.1.A,3/4.10.B & 3/4.12.B,proposing Changes to Relocate Requirement to Remove RPS Shorting Links Which Enable non-coincident Scram for Neutron Instrumentation,To Licensee Controlled Document ML20196K5861998-11-0505 November 1998 Rev 3 to Qcap 0280-01, Process Control Program for Processing of Radioactive Wet Wastes at Quad Cities Nuclear Power Station ML20155D8091998-10-29029 October 1998 Proposed Tech Specs Bases Sections 3/4.2.D & 3/4.5.D, Providing Clarity & Consistency with Sys Design Description Contained in UFSAR Section 5.4.6.2 ML20195J9041998-09-24024 September 1998 Rev 0 to TR-VQ1500-02, Clean ECCS Suction Strainer Head Loss Test Rept ML20151S7991998-08-31031 August 1998 Proposed Tech Specs,Increasing Max Allowable MSIV Leakage from 11.5 Scfh to 30 Scfh Per Valve When Tested at 25 Psig, IAW SR 4.7.D.6 ML20236W8401998-07-31031 July 1998 Proposed Tech Specs Bases 3/4.7.C & 3/4.7.12.C,clarifying Testing Requirements for Primary Containment Excess Flow Check Valves ML20247D7761998-05-0505 May 1998 Proposed Tech Specs Page B 3/4.4-1,changing Administrative Error.Bases for Net Quantity of Gallons for Solution Is Changed from 3254 (Correct Quantity) to 3245 ML20246Q3481998-04-29029 April 1998 TS Page B 3/4.5-3,reflecting Change to TS Bases for Section 3/4.5.C ML20217G1481998-03-27027 March 1998 Proposed Tech Specs Bases Section 3/4.5.A,reflecting Design Info Contained in Rev 4 to Ufsar,Dtd Apr 1997 ML20216C6381997-08-29029 August 1997 Proposed Tech Specs,Incorporating New Siemens' Methodologies That Will Enhance Operational Flexibility & Reducing Likelihood of Future Plant Derates ML20196G0271997-05-0101 May 1997 Proposed Tech Specs 4.9.A.8.b Revising Load Value for Diesel Generator to Be Equal to or Greater than Largest Single Load & Revising Frequency & Voltage Requirements During Performance of Test ML20138G3321997-04-29029 April 1997 Proposed Tech Specs,Permitting Loading of ATRIUM-9B Fuel in Plant Unit Core for Operational Modes 3,4 & 5.Modes Will Support Refueling Activities Such as Fuel Load,Vessel re- Assembly & Single Rod Timing ML20138B3231997-04-21021 April 1997 Proposed Tech Specs,Requesting That NRC Grant Exigent Amend to TS 2.1.B & 6.9.A.6.b to Support Plant Unit 2 Cycle 15 Operation Scheduled to Begin 970519 ML20137G3981997-03-26026 March 1997 Proposed Tech Specs 3/4.7.P Re Standby Gas Treatment & TS 5.2.C Re Secondary Containment ML20135F7321997-03-0303 March 1997 Proposed Tech Spec Bases 3/4.9.E,clarifying Purpose of SR 4.9.E ML20135D9461997-02-24024 February 1997 Proposed Tech Specs,Clarifying Bases for TS Surveillance 4.8.D.5.c ML20138L4011997-02-17017 February 1997 Proposed Tech Specs Section 2.1.B Re Thermal Power,Section 3/4.11 Re Power Distribution Limits,Section 3/4.6 Re Primary Sys Boundary,Section 5.3 Re Reactor Core & Section 6.9 Re Reporting Requirements ML20138L3701997-02-17017 February 1997 Proposed Tech Specs 4.9.A.8.h Re Diesel Generator Endurance Test Surveillance Requirements ML20134D2191997-01-27027 January 1997 Proposed Tech Specs Deleting marked-up Sentence from TS Bases for Section 3/4.7.K ML20129K3321996-10-18018 October 1996 Cycle 15 Startup Test Results ML20129C2391996-10-16016 October 1996 Proposed Tech Specs for Dresden 2 & 3 & Quad Cities 1 & 2, marked-up to Show Transition Verbiage ML20129D3981996-09-20020 September 1996 Proposed Tech Specs 3/4.6.K,updating Pressure-Temp Curves to 22 Effective Full Power Yrs & TS Bases ML20216H8841996-06-30030 June 1996 Revs to ODCM for Quad Cities,Including Rev 1.8 to Chapters 10,11,12 & App F ML20116F3971996-06-30030 June 1996 Rev 1.8 to ODCM, Annex,Chapters 10,11,12 & App F ML20113C3571996-06-25025 June 1996 Proposed Tech Specs Re Upgrade Program ML20113A7861996-06-10010 June 1996 Proposed Tech Specs,App A,To Reflect Transition of Fuel Supplier from General Electric to Siemens Power Corp ML20117D7121996-05-0606 May 1996 Proposed Tech Specs,Implementing New LCO & SR Re Revs to TS for 10CFR50,App J,Lrt ML20107A1881996-04-0404 April 1996 Proposed Tech Specs 3.4/4.4 Re Standby Liquid Control Sys ML20101H1381996-03-25025 March 1996 Complete Version of TS Upgrade Program Pages That Reflect Current Configuration of Plant & Specifies SRs That Will Not Be Current Upon Implementation of Tsup Project ML20097D9231996-02-0808 February 1996 Proposed Tech Specs,Upgrading Existing TS 3/4.5, Eccs ML20100C0441996-01-24024 January 1996 Secondary Containment Leak Test Summary ML20093K7721995-10-12012 October 1995 Quad-Cities Nuclear Power Station Unit 2 Cycle 14 Startup Test Results Summary ML20098A3821995-09-20020 September 1995 Proposed Tech Specs,Revising TS Upgrade Program & Improving Plant Submittals ML20086D4741995-06-30030 June 1995 Proposed Tech Specs Re TS Upgrade Program for Dresden Units 2 & 3 & Quad Cities Units 1 & 2 ML20087H8651995-05-0202 May 1995 Proposed Tech Specs Re TS Upgrade Program Section 3/4.10 ML20082H7481995-04-10010 April 1995 Proposed Tech Specs,Revising SR for HPCI & RCIC Sys ML20080K8171995-02-23023 February 1995 Proposed Tech Specs,Changing Name of Iige to Reflect Results of Merger Between Iige,Mid American Energy Co,Midwest Power Sys Inc & Midwest Resources Inc 1999-08-13
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l- ATTACHMENT E l PROPOSED CHANGES TO THE ;
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TECHNICAL SPECIFICATIONS '
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- LICENSES DPR-29 and DPR-30 i
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J Remove Insert 5-5 5-5 1
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i 9705060266 970429 PDR ADOCK 05000254
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REACTOR CORE 5.3 j .
- ~, 5.0 DESIGN FEATURES
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5.3 REACTOR CORE glies e>' C##e S' c, e ow W l
, fhe O pu r O Fuel Assemblies 5.3.A The reactor core shall contain 724 fue' assemblie Each assembly consists of a
! matrix of Zircaloy clad fuel rods with an 'nitial composition of natural or slightly
- enriched uranium dioxide as fuel materia end wet
- ::d:. Umited substitutions of i o r- zircenium :"ey, in accordance with NRC approved applications of fuel rod 3'."" g , configurations, may be used. Fuel assemblies shall be limited to those fuel designs g.)RLO that have been analyzed with applicable NRC staff approved codes an ethods, and shown by tests or analyses to comply with all fuel safety design base . A limited number of lead test assemblies that have not completed representative testing may be
. placed in non-limiting core regions.
! Control Rod Assemblies 5.3.8 The reactor core shall contain 177 cruciform shaped control rod assemblies. The control material shall be boron carbide powder (B C)'and/or hafnium metal. The i control rod assembly shall have a nominal axial absorber length of 143 inches.
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t 1 A TRIUAl-9Bfuel is only allowed in the reactor core in Operational Aiodes 3. 4 and 5, and with no i more than one control rod withdrawn.
2 The design bases applicable to ATR1Uhi-9Bfuel are those which are applicable to Operational hiodes 3. 4 and 5.
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QUAD CITIES - UNITS 1 & 2 5-5 Amendment Nos. tri a 167
Reporting Requirements 6.9
~
ADMINISTRATIVE CONTROLS
,[
(3) Commonwealth Edison Topical Report NFSR-OOB5, Supplement 1,
" Benchmark of BWR Nuclear Design Methods - Quad Cities Gamma Scan I Comparisons," (latest approved revision). l I
(4) Commonwealth Edison Topical Report NFSR-OO85, Supplement 2,
" Benchmark of BWR Nuclear Design Methods - Neutronic Licensing j 3 1 Analyses," (latest approved revision). i l AJScTT >
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- c. The core operating limits shall be determined so that all applicable limits (e.g., fuel '
thermal-mechanical limits, core thermal-hydraulic limits, ECCS limits, nuclear limits such as shutdown margin, and transient and accident analysis limits) of the safety analysis are met. The CORE OPERATING LIMITS REPORT, including any mid-cycle revisions or supplements thereto, shall be provided upon issuance, for each reload cycle, to the NRC Document Control Desk with copies to the Regional Administrator and Resident inspector.
6.9.B Special Reports Special reports shall be submitted to the Regional Administrator of the NRC Regional Office within the time period specified for each report.
l
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L QUAD CITIES - UNITS 1 & 2 6-16 Amendment Nos. 171 s 167
INSERT l (for Page 6-16) j l
l (5) Advanced Nuclear Fuels Methodology for Boiling Water Reactors, XN-NF-80-19 (P) (A), l Volume 1, Supplement 3, Supplement 3 Appendix F, and Supplement 4, Advanced Nuclear '
Fuels Corporation, November 1990.
(6) Commonwealth Edison Topical Report NFSR-0091,"Benclunark of CASMO/MICROBURN BWR Nuclear Design Methods", Revisior 0, Supplements 1 and 2, December 1991, March 1992, and May 1992, respectively; SER letter dated March 22,1993.
(7) Generic Mechanical Design Criteria for BWR Fuel Designs, ANF-89-98(P)(A) Revision 1, and Revision 1 Supplement 1, Advanced Nuclear Fuels Corporation, May 1995.
(8) Advanced Nuclear Fuels Corporation Generic Mechanical Design for Advanced Nuclear Fuels 9X9-IX and 9X9-9X BWR Reload Fucl. ANF-89-014(P)(A), Revision 1 and Supplements 1 and 2, Advanced Nuclear Fuels Corporation, October 1991.
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ATTACIIMENT C SIGNIFICANT IIAZARDS CONSIDERATION 4
4 The Commission has prosided standards for determining whether a no significant hazards consideration exists as stated in 3 10CFR50.92(c). A proposed amendment to an operating license involves a no significant hazards consideration if operation of the facility in accordance with the proposed amendment would not: (1) involve a significant increase in the probability or consequences of an accident previously evaluated; or (2) create the possibility of a new or difTerent kind of accident from 4
any accident presiously evaluated; or (3) involve a significant reduction in a margin of safety.
I Comed has evaluated the proposed License Amendment and determined that it does not represent a significant hazards
} consideration. Based on the criteria for defining a significant hazards consideration established in 10 CFR 50.92, operation of Quad Cities Units 1 and 2 in accordance with the proposed amendment will not:
5 I) Involve a significant increase in the probability or consequences of an accident previously evaluated because of the following:
The description of a fuel assembly (section 5.3.1) is resised to reflect the fact that ATRIUM-9B contains a central water box. The change is administrative in nature and serves to describe the ATRIUM-9B fuel design terminology. The mechanical aspects of the ATRIUM-9B fuel design have been resiewed and accepted by the NRC.
A notation has been added to allow ATRIUM-98 fuel in the reactor core in Operational modes 3,4 and 5. Due to the mode limitation of this proposed change, only a subset of the accident events analyzed in the FSAR needed to be addressed. The j addition of ATRIUM-9B fuel to the reactor core in Operational Conditions 3,4, or 5 does not increase the probability or consequences of an accident previously evaluated. The events considered are described below.
The fuel equipment handling accidents were considered. Comed has evaluated the bundle drop accident for an ATRIUM-9B fuel assembly and has determined that it is bounded by the results of the fuel handling accident presented in the FSAR.
The grappling of the ATRIUM-9B fuel is similar to that of GE fuel due to the comparable bail handle dimensions and assembly weights. Therefore, ATRIUM-98 fuel is completely compatible with the refueling platform main grapple.
Because the assembly weights of the ATRIUM-9B fuel and the GE fuel are essentially the same, the capacity of the refueling platform main hoist will be sufficient to handle the ATRIUM-9B fuel. Also, the ATRIUM-9B fuel uses a fuel channel design with mechanical and structural characteristics similar to the GE fuel. Therefore the ATRIUM-9B fuelis compatible with, and can be safely inserted /placed into the reactor core.
The SDM for Quad Cities Unit 2 Cycle 15 was determined by Comed using the NRC approved methodology identified in References (c) and (f). The Quad Citics Unit 2 Cycle 15 minimum calculated SDM is 1.88 % AK. This value occurs at beginning of Cycle 15. The SDM at other Cycle 15 exposures is greater than this value. Additionally, at BOC any moderator temperature increase above 68'F will increase SDM.
Per Sections 3.3. A/4.3. A of the Quad Citics Technical Specifications, and noting that the strongest worth control rod is analytically determined, the required SDM for Quad Cities Unit 2 Cycle 15 is 0.38 % AK +R. R accounts for: a) any decrease in SDM over the cycle relative to the BOC determined value, and b) the potential SDM loss assuming full B4C settling in all inverted control blade poison tubes present in the core. Since the SDM is a minimum at BOC 15, and the potential SDM loss assuming full B4C settling in all inverted control blade poison tubes present in the core is 0.05 % AK, the required SDM from the Technical Specifications is 0.38 % AK + 0.00 % AK + 0.05 % AK = 0.43 % AK. Therefore, the calculated SDM of 1.88 % AK is significantly greater than the required Technical Specification value of 0.43 % AK.
Based on the foregoing, the proposed action does not involve a significant increase in the probability or consequences of an accident presiously evaluated.
- 2) Create the possibility of a new or different kind of accident from any accident previously evaluated because:
_.. ._ _ _. ._____ _ _ _ _ ._ - .m _ - _ _ _ _ . .___
T l ATTACIIMENT C
! SIGNIFICANT IIAZARDS CONSIDERATION Creation of the possibility of a new or difTerent kind of accident would require the creation of one or more new precursors of I
that accident. New accident precursors may be created by modifications of the plant configuration, including changes in allowable modes of cperation. This Technical Specification submittal does not involve any modifications of the plant 4
I configuration or allowable modes of operation. The changes to the Technical Specifications to allow loading of ATRIUM-9B fuel into the Unit 2 reactor core do not require any physical plant modifications (other than loading of the 4
ATRIUM-9B assemblies), physically affect any plant components, or entail changes in plant operations. ATRIUM-9B fuel assemblics have approximately the same weight, outer dimensions, and the same basic bail tiandic design as GE fuel ;
assemblics and are handled with the same refueling equipment.
l Based on the foregoing, the proposed action does not create the possibility of a new or difTerent kind of accident from any accident previously evaluated.
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- 3) Involve a significant reduction in the margin of safety because: '
l No modifications of the plant configuration other than the loading of ATRIUM-9B fuel into the Unit 2 reactor core is being I made. The consequences of the Fuel Handling Accidents and the plant systems ability to respond are not afTected. The calculated SDM of 1.88 % AK is significant!y greater than the required Technical Specification value of 0.43 % AK required SDM for Quad Citics Unit 2 Cycle 15. The margin of safety is maintained with ATRIUM-9B fuel loaded in the i reactor core and in Operational modes 3,4, or 5.
Guidance has been provided in " Final Procedures and Standards on No Significant Hazard Considerations," Final Rule, 51 FR 7744, for the application of standards to license change requests for determination of the existence of significant hazards considerations. This document provides examples of amendments which a~e and are not considered hkely to involve significant hazards considerations.
This proposed amendment does not involve a significant relaxation of the criteria used to establish safety limits, a significant relaxation of the bases for the limiting safety system settings or a significant relaxation of the bases for the limiting conditions for operations. Therefore, based on the guidance provided in the Federal Register and the criteria established in 10 CFR 50.92(c), the proposed change does not constitute a significant hazards consideration.
ENVIRONMENTAL ASSESSMENT Comed has evaluated the proposed amendment against the criteria for identification oflicensing and regulatory actions requiring environmental assessment in accordance with 10 CFR 51.21. It has been determined that the proposed changes meet the criteria for a categorical exclusion as provided under 10 CFR 51.22 (c)(9). This conclusion has been determined because the changes requested do not pose significant hazards consideration and do not involve a significant increase in the amounts, and no significant changes in the types, of any effluents that may be released off-site. Additionally, this request does not involve a significant increase in individual or cumulative occupational radiation exposure.
!