ML20092H326

Proposed Tech Specs 5.3.1, Fuel Assemblies, Allowing Use of Alternate Zirconium Based Fuel Cladding,Zirlo & 5.4.1.a, Referring to UFSAR Rather than FSAR

ML20092H326
Person / Time
Site:	Byron, Braidwood
Issue date:	09/14/1995
From:	COMMONWEALTH EDISON CO.
To:
Shared Package
ML20092H317	List: ML20092H315 ML20092H326
References
NUDOCS 9509200359
Download: ML20092H326 (8)

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

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Affected page: 5-4 2

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4 9509200359 950914 PDR ADOCK 05000454

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, 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

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l with stainless steel tubing. )

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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

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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-

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, indium-cadmium, or a mixture of both types. All control rods shall be clad  ;

! with stainless steel tubing. <

, 5.4 REACTOR COOLANT SYSTEM

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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.

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The operating limits will not be chang J and the analysis methods to

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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.

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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

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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.

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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.

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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.

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