ML20138G332

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

ML20138G332
Person / Time
Site:	Quad Cities
Issue date:	04/29/1997
From:	COMMONWEALTH EDISON CO.
To:
Shared Package
ML20138G326	List: ML20138G323 ML20138G332
References
NUDOCS 9705060266
Download: ML20138G332 (6)

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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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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is only allowed in the reactor core in Operational Aiodes 3. 4 and 5, and with no i more than one control rod withdrawn.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

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ADMINISTRATIVE CONTROLS

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(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.

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L QUAD CITIES - UNITS 1 & 2 6-16 Amendment Nos. 171 s 167

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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:

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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.

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