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Category:TECHNICAL SPECIFICATIONS
MONTHYEARML20210J1751999-07-30030 July 1999 Marked-up TS Pages for Proposed Changes Re Upper Temp Limit for UHS ML20209B7411999-06-30030 June 1999 Proposed Tech Specs Section 3.8.5, DC Sources - Shutdown, Correcting LCO & Braidwood TS Section 3.8, Electrical Power Systems, Deleting Various References to At&T Batteries ML20204H9781999-03-23023 March 1999 Proposed Tech Specs,Revising Sections 3.7.15,3.7.16,4.3.1 & 4.3.3 to Support Installation of New Boral high-density SFP Storage Racks at Byron & Braidwood Stations ML20204H4291999-03-22022 March 1999 Proposed Tech Specs 3.9.3,allowing Use of Gamma-Metrics post-accident Neutron Monitors to Provide Neutron Flux Info During Operational Mode 6 ML20198N2041998-12-29029 December 1998 Revised Tech Specs Change,Page 3/4 3-54,providing Early Implementation of Containment Floor Drain Sump Water Level Instrumentation Requirements ML20198N3471998-12-29029 December 1998 Proposed ITS Tables 3.3.1-1 & 3.3.2-1,revising Twelve Allowable Values ML20198K5841998-12-23023 December 1998 Revised Tech Spec Pages 3/4 3-53,3/4 3-53a,6-27 & 6-27a,for Rv LI Sys ML20198A0811998-12-14014 December 1998 Proposed Rev T to Improved Tech Specs Section 3.4, Reactor Coolant Sys ML20196G6611998-11-30030 November 1998 Proposed Rev to Improved Tech Specs Section 3.1 ML20196B4101998-11-25025 November 1998 Proposed Tech Specs Facilitating Replacement of 125 Vdc At&T Batteries with New 125 Vdc C&D Batteries While in Mode 1-4 ML20155J2051998-11-0505 November 1998 Proposed TS Converting to Its,Rev R ML20155J0041998-10-30030 October 1998 Proposed Tech Specs Section 5.6.2, Fuel Storage Drainage, to Identify Sf Pool Level Sufficient to Ensure SRP Acceptance Criteria ML20154S5011998-10-18018 October 1998 Proposed Rev N to Improved TS Section 3.7 ML20154M5281998-10-15015 October 1998 Revisions K,O & P of 961213 ITS Submittal ML20154A8881998-10-0202 October 1998 Proposed Rev L to Improved Tech Specs Section 3.8 Closeout ML20153G4331998-09-25025 September 1998 Revs J & M to Tech Specs Sections 3.6 & 5.0,converting to Improved Tech Specs (Its),Final Closeout Package ML20236W5851998-07-31031 July 1998 Proposed Rev G to Sections 3.1 & 3.2 of Improved Tech Specs ML20237B6391998-07-30030 July 1998 Proposed Rev H to Section 3.5 of Improved Tech Specs ML20237E9971998-07-21021 July 1998 Rev I to Proposed Improved Tss,Section 3.9 Re Final Closeout ML20237B7021998-07-0909 July 1998 Proposed Improved TS (ITS) Section 3.3 Issued as Result of Removing Generic Change Traveler TSTF-135,Rev E from ITS Submittal ML20236H6531998-07-0202 July 1998 Rev F to 961213 Improved TS Submittal,Containing Final Package Closeout for Improved TS Sections 1.0,2.0 & 3.0 ML20248M1491998-06-0101 June 1998 Proposed Tech Specs Bases Page B 3.8-58b,converting to Improved Tech Specs ML20248C5511998-05-29029 May 1998 Proposed Tech Specs Bases Section 3/4.4.4, Relief Valves, Specifically Crediting Automatic Function of PORVs to Provide Mitigation for Inadvertent Operation of ECCS at Power Accident ML20216D9431998-04-0909 April 1998 Modified Proposed TS Pages Re 980324 Request for Amends to Licenses NPF-37 & NPF-66 ML20217E1891998-03-24024 March 1998 Proposed Tech Specs Surveillance Sections & Bases Allowing Util to Defer 10CFR50,App J,Type a Testing of Byron Unit 2 Containment Until Next Refuel Outage in 1999 ML20217B2681998-02-14014 February 1998 Proposed Rev D to ITS ML20198L8811998-01-14014 January 1998 Proposed Tech Specs Pages,Revising TS Section 3/4.8.2 & Bases,To Allow Replacement of 125 Volt Dc At&T Batteries W/New Charter Power Sys,Inc (C&D) Batteries ML20198L8131998-01-14014 January 1998 Proposed Tech Specs Pages Revising TS 3.4.8, Specific Activity, Figure 3.4-1,Table 4.4-4 & TS Bases 3.4.8 ML20198C3181997-12-30030 December 1997 Proposed Tech Specs 3.7.1.3 Re Condensate Storage Tank ML20203M5921997-12-17017 December 1997 Proposed Tech Specs,Rev C Changes Improved TSs 3.0,3.3,3.7, 3.8 & 5.0 as Result of Removing Generic Change Traveler TSTF-115 from Improved TS Submittal ML20203D0361997-12-0909 December 1997 Proposed Tech Specs Pages Correcting Errors Discovered in Current TS W/Regards to Total RCS Volume & Correction to Increase in RCS Volume Associated W/Unit 1 Replacement SGs Accounting for Hot Conditions ML20199A4751997-11-0707 November 1997 Proposed Tech Specs Pages Revising TS Surveillance Sections 4.6.1.1.c,4.6.1.2.a,4.6.1.2.c & Bases to Allow Performance of 10CFR50 App J,Type a Testing ML20217K4461997-10-21021 October 1997 Proposed Tech Specs Re Boron Credit in SFP ML20202F4561997-10-10010 October 1997 Proposed Tech Specs,Deleting Lower Flow Rate Requirement Associated W/Nonaccessible Area Exhaust Filter Plenum & Fuel Handling Bldg Ventilation Sys ML20211D9421997-09-24024 September 1997 Proposed Tech Specs Revising Allowable Time Interval for Performing Turbine Throttle Valve & Turbine Valve SRs Requirements from Monthly to Quarterly ML20216G8541997-09-0808 September 1997 Proposed Tech Specs Change to TS 4.5.2.b & Associated Bases Bringing Byron Unit 1 & Braidwood Unit 1 Requirement in Conformance W/Unit 2 Requirements Approved by NRC in ML20216F1351997-09-0202 September 1997 Proposed Tech Specs 3.4.8 Re Specific Activity ML20217H6071997-08-0707 August 1997 Proposed Tech Specs Pages,Revising Bases for Proposed Improved TS SR 3.8.6.1 & 3.8.6.3,to Indicate That Correction for Level Is Not Required When Battery Charging Current Is Less than 2 Amps for Gould & Less than 3 Amps for C&D ML20148P7721997-06-30030 June 1997 Proposed Tech Specs,Revising TS 3.9.11,5.6.1.1 & 6.9.10 to Allow Util to Permanently Take Credit for Soluble B in Spent Fuel Storage Pool Water to Maintain Acceptable Margin of Subcriticality ML20141F3081997-06-24024 June 1997 Proposed Tech Specs,Changing TS for ECCS Venting ML20141B7781997-06-17017 June 1997 Proposed Tech Specs Revising TS Sections 3/4.6.1.6,4.6.1.2, 6.8.4 & 6.9.1.11 to Support New Requirements in 10CFR50.55a, Which Requires Utils to Update Existing Containment Vessel Structural Integrity Programs ML20148J3231997-06-0909 June 1997 Proposed TS Reflecting Latest Rev of Waste Gas Decay Tank Rupture Accident Dose Calculation ML20140D0081997-05-31031 May 1997 Proposed Tech Specs,Revising TS Surveillance Requirement Re ECCS Pump Casings & Discharge Piping High Points Outside of Containment ML20141K8991997-05-24024 May 1997 Proposed Tech Specs Revising TS Surveillance Requirement 4.5.2.b to Encompass non-operating ECCS Pumps & Discharge Piping Which Are Provided W/High Point Vent Valves ML20141K8931997-05-23023 May 1997 Proposed Tech Specs Revising TS Surveillance Requirement 4.5.2.b.1 for Unit 1 as It Relates to Requirement to Vent ECCS Pump Casings & Discharge Piping High Points Outside Containment ML20141K3381997-05-23023 May 1997 Proposed Tech Specs Requesting Enforcement Discretion from Compliance W/Ts 4.5.2.b.1 Requirements of Venting of Emergency Core Cooling Sys Pump Casings & Discharge Piping High Points Outside of Containment ML20148D6861997-05-23023 May 1997 Proposed Tech Specs Revising TS Surveillance Requirement 4.5.2.b.1 for Unit 1 as It Relates to Requirement to Vent ECCS Pump Casings & Discharge Piping High Points Outside Containment ML20141K0011997-05-21021 May 1997 Proposed Tech Specs Relocating Reactor Vessel Surveillance Program Capsule Withdrawal Schedules IAW GL 91-01 ML20148B6151997-05-0606 May 1997 Proposed Tech Specs,Revising TS 3/4.7.5, Ultimate Heat Sink & Associated Bases to Support SG Replacement & Incorporate Recent UHS Design Evaluations ML20196G0501997-04-25025 April 1997 Proposed Tech Specs Revising Primary Containment & Reactor Coolant Sys Volume Associated W/Unit 1 Steam Generator Replacement 1999-07-30
[Table view] Category:TECHNICAL SPECIFICATIONS & TEST REPORTS
MONTHYEARML20210J1751999-07-30030 July 1999 Marked-up TS Pages for Proposed Changes Re Upper Temp Limit for UHS ML20209B7411999-06-30030 June 1999 Proposed Tech Specs Section 3.8.5, DC Sources - Shutdown, Correcting LCO & Braidwood TS Section 3.8, Electrical Power Systems, Deleting Various References to At&T Batteries ML20211C3311999-04-30030 April 1999 Rev 2.0 to Generic ODCM for Dresden,Quad Cities,Zion, Lasalle,Byron & Braidwood ML20204H9781999-03-23023 March 1999 Proposed Tech Specs,Revising Sections 3.7.15,3.7.16,4.3.1 & 4.3.3 to Support Installation of New Boral high-density SFP Storage Racks at Byron & Braidwood Stations ML20204H4291999-03-22022 March 1999 Proposed Tech Specs 3.9.3,allowing Use of Gamma-Metrics post-accident Neutron Monitors to Provide Neutron Flux Info During Operational Mode 6 ML20202G9361999-01-30030 January 1999 Rev 1.4 to Chapter 10, Radioactive Effluent Treatment & Monitoring, Rev 1.6 to Chapter 11, Radiological Environ Program & Rev 1.6 to Chapter 12, Radioactive Effluent Technical Standards (Rets), for Odcm,Byron Annex ML20198N2041998-12-29029 December 1998 Revised Tech Specs Change,Page 3/4 3-54,providing Early Implementation of Containment Floor Drain Sump Water Level Instrumentation Requirements ML20198N3471998-12-29029 December 1998 Proposed ITS Tables 3.3.1-1 & 3.3.2-1,revising Twelve Allowable Values ML20198K5841998-12-23023 December 1998 Revised Tech Spec Pages 3/4 3-53,3/4 3-53a,6-27 & 6-27a,for Rv LI Sys ML20198A0811998-12-14014 December 1998 Proposed Rev T to Improved Tech Specs Section 3.4, Reactor Coolant Sys ML20196G6611998-11-30030 November 1998 Proposed Rev to Improved Tech Specs Section 3.1 ML20196B4101998-11-25025 November 1998 Proposed Tech Specs Facilitating Replacement of 125 Vdc At&T Batteries with New 125 Vdc C&D Batteries While in Mode 1-4 ML20155J2051998-11-0505 November 1998 Proposed TS Converting to Its,Rev R ML20155J0041998-10-30030 October 1998 Proposed Tech Specs Section 5.6.2, Fuel Storage Drainage, to Identify Sf Pool Level Sufficient to Ensure SRP Acceptance Criteria ML20154S5011998-10-18018 October 1998 Proposed Rev N to Improved TS Section 3.7 ML20154M5281998-10-15015 October 1998 Revisions K,O & P of 961213 ITS Submittal ML20154A8881998-10-0202 October 1998 Proposed Rev L to Improved Tech Specs Section 3.8 Closeout ML20153G4331998-09-25025 September 1998 Revs J & M to Tech Specs Sections 3.6 & 5.0,converting to Improved Tech Specs (Its),Final Closeout Package ML20153C0351998-08-31031 August 1998 Revs to ODCM for Plant,Including Rev 2 to Chapters 10 & 11, Rev 4 to Chapter 12 and Rev 3 to App F ML20238F8221998-08-25025 August 1998 Rev 2 to Braidwood Station Units 1 & 2 Second Interval ISI Program Plan ML20236Y5481998-08-0303 August 1998 Rev 1 to Braidwood Station Units 1 & 2 Second Interval ISI Program Plan ML20236W5851998-07-31031 July 1998 Proposed Rev G to Sections 3.1 & 3.2 of Improved Tech Specs ML20237B6391998-07-30030 July 1998 Proposed Rev H to Section 3.5 of Improved Tech Specs ML20237E9971998-07-21021 July 1998 Rev I to Proposed Improved Tss,Section 3.9 Re Final Closeout ML20237B7021998-07-0909 July 1998 Proposed Improved TS (ITS) Section 3.3 Issued as Result of Removing Generic Change Traveler TSTF-135,Rev E from ITS Submittal ML20236H6531998-07-0202 July 1998 Rev F to 961213 Improved TS Submittal,Containing Final Package Closeout for Improved TS Sections 1.0,2.0 & 3.0 ML20248M1491998-06-0101 June 1998 Proposed Tech Specs Bases Page B 3.8-58b,converting to Improved Tech Specs ML20248K7361998-05-31031 May 1998 Commonwealth Edison Bnps Unit 1 Cycle 9 Startup Rept ML20248C5511998-05-29029 May 1998 Proposed Tech Specs Bases Section 3/4.4.4, Relief Valves, Specifically Crediting Automatic Function of PORVs to Provide Mitigation for Inadvertent Operation of ECCS at Power Accident ML20217Q8521998-05-0101 May 1998 Rev 9 to Bzp 310-2, Nuclear Accident Reporting Sys Form (Primary Responsibility - Station Director). W/Notes & Comments ML20216D9431998-04-0909 April 1998 Modified Proposed TS Pages Re 980324 Request for Amends to Licenses NPF-37 & NPF-66 ML20216C1011998-03-26026 March 1998 Revs to ODCM for Braidwood,Including Rev 1.9 to Chapter 10 & Rev 3 to Chapter 12 ML20217E1891998-03-24024 March 1998 Proposed Tech Specs Surveillance Sections & Bases Allowing Util to Defer 10CFR50,App J,Type a Testing of Byron Unit 2 Containment Until Next Refuel Outage in 1999 ML20217B2681998-02-14014 February 1998 Proposed Rev D to ITS ML20217C3011998-01-31031 January 1998 Rev 0 to Inservice Testing Program Plan Pumps & Valves Braidwood Nuclear Generating Station,Units 1 & 2 ML20198L8131998-01-14014 January 1998 Proposed Tech Specs Pages Revising TS 3.4.8, Specific Activity, Figure 3.4-1,Table 4.4-4 & TS Bases 3.4.8 ML20198L8811998-01-14014 January 1998 Proposed Tech Specs Pages,Revising TS Section 3/4.8.2 & Bases,To Allow Replacement of 125 Volt Dc At&T Batteries W/New Charter Power Sys,Inc (C&D) Batteries ML20199J7581997-12-31031 December 1997 Rev 1 to IST Plan Pumps & Valves Byron Nuclear Generating Station,Units 1 & 2 ML20216H4321997-12-31031 December 1997 Revs to OCDM for Braidwood,Including Rev 1.8 to Chapter 10, Rev 1.9 to Chapter 11,rev 2 to Chapter 12 & Rev 2 to App F ML20198C3181997-12-30030 December 1997 Proposed Tech Specs 3.7.1.3 Re Condensate Storage Tank ML20203M5921997-12-17017 December 1997 Proposed Tech Specs,Rev C Changes Improved TSs 3.0,3.3,3.7, 3.8 & 5.0 as Result of Removing Generic Change Traveler TSTF-115 from Improved TS Submittal ML20203D0361997-12-0909 December 1997 Proposed Tech Specs Pages Correcting Errors Discovered in Current TS W/Regards to Total RCS Volume & Correction to Increase in RCS Volume Associated W/Unit 1 Replacement SGs Accounting for Hot Conditions ML20199A4751997-11-0707 November 1997 Proposed Tech Specs Pages Revising TS Surveillance Sections 4.6.1.1.c,4.6.1.2.a,4.6.1.2.c & Bases to Allow Performance of 10CFR50 App J,Type a Testing ML20198M2621997-10-31031 October 1997 Revs to Offsite Dose Calculation Manual,Consisting of Rev 1.5 to Chapter 11 & Rev 1.5 to Chapter 12 ML20216H8241997-10-31031 October 1997 Revs to OCDM for Byron Station,Including Rev 1.3 to Chapter 10,rev 1.5 to Chapters 11 & 12 & Rev 1.3 to App F ML20217K4461997-10-21021 October 1997 Proposed Tech Specs Re Boron Credit in SFP ML20198P1601997-10-20020 October 1997 Rev 1.5 to Odcm,Chapter 12 ML20202F4561997-10-10010 October 1997 Proposed Tech Specs,Deleting Lower Flow Rate Requirement Associated W/Nonaccessible Area Exhaust Filter Plenum & Fuel Handling Bldg Ventilation Sys ML20211D9421997-09-24024 September 1997 Proposed Tech Specs Revising Allowable Time Interval for Performing Turbine Throttle Valve & Turbine Valve SRs Requirements from Monthly to Quarterly ML20216H7011997-09-10010 September 1997 Revised Procedures,Including Rev 2 to Bwzp 2000-18, Post- Accident Sampling Sys (Primary Responsibility - Chemistry Director) & Rev 2 to Bwzp 2000-18A1, PASS Sample Collection Procedures 1999-07-30
[Table view] |
Text
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ATTACHMENT B-1 PROPOSED CHANGES TO APPENDIX A, TECHNICAL SPECIFICATIONS, <
OF FACILITY OPERATING LICENSES NPF-37 AND NPF-66 BYRON NUCLEAR POWER STATION, UNITS 1 AND 2 i
i 1
Affected page: 5-4 2
=
)
s 4
l J
4 9509200359 950914 PDR ADOCK 05000454
-P., _ _ .
PDR
, DESIGN FEATURES i 5.3 REACTOR CORE.
of Qre.vious\ oobq l i FUEL ASSEMBLIES ci e
or pgL0 32 iR'.0, l 5.3.1 The core all contain 193 fuel assemb ies with ach fuel assembly containing 264 1 rods clad with Zircaloy-45 excep t limited substitution i of fuel rods b filler rods consisting of Zircaloy-4 r stainless steel or by
- vacancies may, made if dustified by a cycle specific reload analysis. Each i
fuel rod sha 'I have a nominal active fuel length.of 144 inches. The initial i
core loadin shall have a maximum enrichment of less than 3.20 weight percent U-235.
i R l'oad fuel shall be similar in physical design to the initial core loading.d. The enrichment of any reload fuel design.shall be determined to be 1 acceptable for storage in either the spent fuel pool 'or the new fuel vault.
Such acceptance criteria shall be based on the results of the CRITICALITY j
ANALYSIS OF BYRON AND BRAIDWOOD STATION FUEL STORAGE RACKS. -
j CONTROL R00 ASSEMILIES i
! . 5. 3. 2 The core shall contain 53 full-length and no part-length control rod i cssemblies. The full-length control rod assemblies shall contain a nominal 142 inches of absorber material.
indium-cadmium, or a mixture of both types.All control rods shall be hafnium, silver-All control rods shall be clad
(
l with stainless steel tubing. )
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j i
5.4. REACTOR COOLANT SYSTEM i
DESIGN PRESSURE AND TEMPERATURE i
! 5.4.1 The Reactor Coolant System is designed and shall ne maintained: ,
a.
i- Ingaccordance with the Code requirements specified in Section 5.2 of th'eUF)AR, with allowance for nomal degradation pursuant to the i
ap'pr14 table Surveillance Requirements,
- b. For a pressure of 2485 psig, and i- c.
j For a temperature of 650*F, except for the pressurizar which is I 680'F.
l VOLUME i
5.4.2 12,257 The cubictotal feetwater at a and steam volume of the' Reactor Coolant System is nominal !
T,,, of 588.4*F.
5.5 METEOROLOGICAL TOWER LOCATION 5.5.1
' The meteorological tower shall be located as shown on Figure 5.1-1.
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- BYRON - UNITS 1 & 2 5-4 AMENDMENTNO.,36 l
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- ATTACHMENT B-2 PROPOSED CHANGES TO APPENDIX A, TECHNICAL SPECIFICATIONS,
) OF FACILITY OPERATING LICENSES i NPF-72 AND NPF-77
' BRAIDWOOD NUCLEAR POWER STATION, UNITS 1 AND 2
- Affected page: 5-4 i
d II a
4 d
k DESIGN. FEATURES '
- 5. 3 REACTOR 00RE or fmle45 cycle _
FUEL ASSEMBLIES
'N or EIRLO JIRL@
5.3.1 The core s all contain 193 fuel assembli s with each fuel assembly l containing 264 ft 1 rods clad with Zircaloy- except tha limited substitu-i tion of fuel rod by filler rods consisting of Zircaloy-4 or stainless steel .or by vacancies may be made if justified by a cycle specific reload analysis.
J Each fuel rod sh til have a nominal active fuel length of 144 inches. The j
initini core lo ding shall have a maximum enrichment of less than 3.20 weight
- percent U-235. Reload fuel shall be similar in physical design to the initial j core loading The enrichment of any reload fuel design shall be determined to
- be acceptable for storage in either the spent fuel pool or the new fuel vault. Such acceptance criteria shall be based on the results of the CRITICALITY ANALYSIS OF BYRON AND BRAIDWOOD STATION FUEL STORAGE RACKS. .
CONTROL ROD ASSEMBLIES .
' 5.3.2 The core shall contain 53 full-length and no part-length control rod assemblies. The full-length control rod assemblies shall contain a nominal 142 inches of absorber material. All control rods shall be hafnium, silver-
[
, indium-cadmium, or a mixture of both types. All control rods shall be clad ;
! with stainless steel tubing. <
, 5.4 REACTOR COOLANT SYSTEM
~
DESIGN PRESSURE AND TEMPERATURE
! 5.4.1 The Reactor Coolant System is designed and shall be maintained:
- a. In S cordance with the Code requirements specified in Section 5.2 of thtuPSAR, with allowance for normal degradation pursuant to the .
} apiiNeable Surveillance Requirements,
- b. For a pressure of 2485 psig, and i c. For a temperature of 650*F, except for the pressurizer which is
- 680'F.
VOLUME .
5.4.2 The total water and steam volume of the Reactor Coolant System is i
12,257 cubic feet at a noisinal T,,g of 588.4*F.
5.5 METEOROLOGICAL TOWER LOCATION 5.5.1 The meteorological tower shall be located as shown on Figure 5.1-1.
i BRAIDWOOD - UNITS 1 & 2 5-4 Amendment No. ) f
ATTACHMENT C EVALUATION OF SIGNIFICANT HAZARDS CONSIDERATIONS OF PROPOSED CHANGES 4
TO APPENDIX A, TECHNICAL SPECIFICATIONS, OF FACILITY OPERATING LICENSES NPF-37, NPF-66, NP'/-72 AND NPF-77 Commonwealth Edison Company (Comed) has evaluated this proposed amendment and determined that it involves no significant hazards considerations. According to Title 10, Code of Federal Regulations, Part 50, Section 92, Paragraph c [10 CFR 50.92(c)], a proposed amendment to an operating license involves no significant hazards 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 different kind of accident from any accident previously evaluated; or
- 3. Involve a significant reduction in a margin of safety.
Comed proposes to revise Technical Specification 5.3.1, Fuel Assemblies, to allow use of an alternate zirconium based fuel cladding, ZIRLO. Limited substitution of fuel rods by ZIRLO filler rods would also be permitted. Editorial changes are also proposed.
- 1. The proposed changes do not involve a significant increase in the probability or consequences of an accident previously evaluated.
The methodologies used in the accident analyses remain unchanged. The proposed changes do not change or alter the design assumptions for the systems or components used to mitigate the consequences of an accident. Use of ZlRLO fuel cladding does not adversely affect fuel performance or impact nuclear design methodology. Therefore, accident analysis results are not impacted.
I
x, 4"
The operating limits will not be chang J and the analysis methods to
~
I demonstrate operation within the limits will remain in accordance with NRC-approved methodologies. Other than the changes to the fuel assemblies, there are no physical changes to the plant associated with this Technical Specification change. A safety analysis will continue to be performed for each cycle to demonstrate compliance with all fuel safety design bases.
VANTAGE 5 fuel assemblies with ZIRLO clad fuel rods meet the same fuel assembly and fuel rod design bases as other VANTAGE 5 fuel assemblies. In addition, the 10 CFR 50.46 criteria are applied to the ZIRLO clad fuel rods.
The use of these fuel assemblies will not result in a change to the reload design and safety analysis limits. Since the original design criteria are met, the ZIRLO clad fuel rods will not be an initiator for any new accident. The clad material is similar in chemical composition and has similar physical and mechanical properties as Zircaloy-4. Thus, the cladding integrity is maintained and the structural integrity of the fuel assembly is not affected. ZIRLO cladding improves corrosion performance and dimensional stability. No concerns have been identified with respect to the use of an assembly containing a combination of Zircaloy-4 and ZIRLO clad fuel rods. Since the dose predictions in th; safety analyses are not sensitive to the fuel rod cladding material used, the radiological consequences of accidents previously evaluated in the safety analysis remain valid.
Replacing the reference to the Final Safety Analysis Report (FSAR) with a reference to the Updated Final safety Analysis Report (UFSAR) is an editorial change to reflect the current document. Adding that reload fuel shall be similar in physical design to the initial core loading or previous cycle loading is a clarification. A reload analysis is completed for each cycle, in accordance with USNRC-approved methodologies.
Therefore, the proposed changes do not involve a significant increase in the probability or consequences of an accident previously evaluated.
- 2. The proposed changes do not create the possibility of a new or different kind of accident from any accident previously evaluated.
VANTAGE 5 fuel assemblies with ZIRLO clad fuel rods satisfy the same design bases as those used for other VANTAGE 5 fuel assemblies. All design and performance criteria continue to be met and no new failure mechanisms have been identified. The ZIRLO cladding material offers improved corrosion resistance and structural integrity.
2
- . . - - - . -. - - = . -. .- __- . . ...
f
- L ..
s t .
The proposed changes do not affect the design or operation of any system or
- component in the plant. The safety functions of the related structures, systems,
- or components are not changed in any manner, nor is the reliability of any structure, system, or component reduced. The changes do not affect the
+
manner by which the facility is operated and do not change any. facility design
' feature, stmeture, or system. No new or different type of equipment will be installed. Since there is no change to the facility or operating procedures, and the safety functions and reliability of structures, systems, or components are not '
affected, the proposed changes do not create the possibility of a new or different kind of accident from any accident previously evaluated.
l
- 3. The proposed changes do not involve a significant reduction in a margin of safety. ,
i The use of Zircaloy-4, ZIRLO, or stainless steel filler rods in fuel assemblies ;
will not involve a significant reduction in the margin of safety because analyses 2 using NRC-approved methodology will be performed for each configuration to
. demonstrate continued operation within the limits that assure acceptable plant response to accidents and transients. These analyses will be performed using -
NRC. approved methods that have been approved for application to the fuel
- configuration.
l Use of ZIRLO cladding material does not change the VANTAGE 5 reload
- design and safety analysis limits. The use of these fuel assemblies will take into consideration the normal core operating conditions allowed in the Technical Specifications. For each cycle reload core, the fuel assemblies will be evaluated using NRC-approved reload design methods, including consideration of the core physics analysis peaking factors and cc'e average i linear heat rate effects.
Therefore, based on the above evaluation, Comed has concluded that these changes do not involve significant hazards considerations.
a 3
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ATTACHMENT D ENVIRONMENTAL ASSESSMENT OF PROPOSED CHANGES TO APPENDIX A, TECHNICAL SPECIFICATIONS, OF FACILITY OPERATING LICENSES NPF-37, NPF-66, NPF-72 AND NPF-77 Commonwealth Edison Company (Comed) has evaluated the proposed amendment against the criteria for and identification of licensing and regulatory actions requiring environmental assessment in accordance with Title 10, Code of Federal Regulations, Part 51, Section 21 (10 CFR 51.21). It has been determined that the proposed changes m.:t the criteria for a categorical exclusion as provided for under 10 CFR 51.22(c)(9).
This determination is based on the fact that this change is being proposed as an amendment to a license issued pursuant to 10 CFR 50 that changes a requirement with respect to a facility component located within a restricted area, and the amendment meets the following specific criteria:
(i) the amendment involves no significant hazards considerations, As demonstrated in Attachment C, this proposed amendment does not involve any significant hazards considerations.
(ii) there is no significant change in the types or significant increase in the amounts of any effluents that may be released offsite, and As documented in Attachment A, there will be no change in the types or significant increase in the amounts of any effluents released offsite.
(iii) there is no significant increase in individual or cumulative occupational radiation exposure.
The proposed changes will not result in changes in the operation or configuration of the facility. Core design will continue to meet all core design criteria, and reactor operation will not be impacted. There will be no change in the level of controls or methodology used for processing of radioactive effluents or handling of solid radioactive waste, nor will the proposal result in any change in the normal radiation levels within the plant. Therefore, there will be no increase in individual or cumulative occupational radiation exposure resulting from this change.
I