View source for RS-05-010, Application for Amendments to Renew Licenses DPR-19, DPR-25, DPR-29 & DPR-30, Supporting Transition to Westinghouse SVEA-96 Optima2 Fuel

{{Adams
| number = ML050800371
| issue date = 01/20/2005
| title = Application for Amendments to Renew Licenses DPR-19, DPR-25, DPR-29 & DPR-30, Supporting Transition to Westinghouse SVEA-96 Optima2 Fuel
| author name = Simpson P
| author affiliation = Exelon Generation Co, LLC, Exelon Nuclear
| addressee name = 
| addressee affiliation = NRC/Document Control Desk, NRC/NRR
| docket = 05000237, 05000249, 05000254, 05000265
| license number = DPR-019, DPR-025, DPR-029, DPR-030
| contact person = 
| case reference number = RS-05-010
| document type = License-Application for Facility Operating License (Amend/Renewal) DKT 50, Technical Specifications
| page count = 34
| project = 
| stage = Request
}}

=Text=
{{#Wiki_filter:Exckin Generation                 wwwexeloncorp.cont                                    Nuclear 4300 Winfield Road Warrenville. IL 60555 10 CFR 50.90 RS-05-01 0 January 20, 2005 U. S. Nuclear Regulatory Commission ATTN: Document Control Desk Washington, DC 20555-0001 Dresden Nuclear Power Station, Units 2 and 3 Renewed Facility Operating License Nos. DPR-19 and DPR-25 NRC Docket Nos. 50-237 and 50-249 Quad Cities Nuclear Power Station, Units 1 and 2 Renewed Facility Operating License Nos. DPR-29 and DPR-30 NRC Docket Nos. 50-254 and 50-265

==Subject:==
Request for License Amendment Regarding Transition to Westinghouse Fuel In accordance with 10 CFR 50.90, "Application for amendment of license or construction permit," Exelon Generation Company, LLC (EGC) requests an amendment to Renewed Facility Operating License Nos. DPR-1 9 and DPR-25 for Dresden Nuclear Power Station (DNPS) Units 2 and 3, and Renewed Facility Operating License Nos. DPR-29 and DPR-30 for Quad Cities Nuclear Power Station (QCNPS) Units 1 and 2. The proposed change supports the transition to Westinghouse SVEA-96 Optima2 (Optima2) fuel at DNPS and QCNPS.
This request is subdivided as follows.
* Attachment 1 provides an evaluation supporting the proposed change.
* Attachments 2 and 3 contain the marked-up Technical Specifications (TS) pages for DNPS and QCNPS, respectively, with the proposed change indicated.
* Attachments 4 and 5 provide retyped TS pages for DNPS and QCNPS, respectively, with the proposed change incorporated.
The proposed change has been reviewed by the Plant Operations Review Committee and approved by the Nuclear Safety Review Board for the respective facilities in accordance with the requirements of the EGC Quality Assurance Program.

January 20, 2005 U. S. Nuclear Regulatory Commission Page 2 EGC plans to transition fuel vendors for DNPS and QCNPS beginning with the QCNPS Unit 2 refueling outage in March 2006. Therefore, EGC requests approval of the proposed change by January 20, 2006, since the core operating limits using the new analytical methods added to TS Section 5.6.5, "Core Operating Limits Report (COLR),"
will become effective upon startup following the QCNPS Unit 2 refueling outage. Once approved, the amendments shall be implemented within 60 days. This implementation period will provide adequate time for the affected station documents to be revised using the appropriate change control mechanisms.
The core reload analyses for the affected units may result in the need for additional TS changes to support the transition to Optima2 fuel, such as a change to the safety limit minimum critical power ratio. These changes, if any, will be submitted to the NRC in a separate license amendment request.
In accordance with 10 CFR 50.91(b), EGC is notifying the State of Illinois of this application for changes to the TS by transmitting a copy of this letter and its attachments to the designated State Official.
Should you have any questions related to this letter, please contact Mr. Kenneth M. Nicely at (630) 657-2803.
I declare under penalty of perjury that the foregoing is true and correct. Executed on the 20th day of January 2005.
Respectfully, Po R.
Patrick R. Simpson Manager - Licensing Attachments:
Attachment 1:   Evaluation of Proposed Change Attachment 2:   Markup of Proposed Technical Specifications Pages for DNPS Attachment 3:   Markup of Proposed Technical Specifications Pages for QCNPS Attachment 4:   Retyped Technical Specifications Pages for Proposed Change for DNPS Attachment 5: Retyped Technical Specifications Pages for Proposed Change for QCNPS cc:      Regional Administrator- NRC Region IlIl NRC Senior Resident Inspector - Dresden Nuclear Power Station NRC Senior Resident Inspector - Quad Cities Nuclear Power Station Illinois Emergency Management Agency - Division of Nuclear Safety

ATTACHMENT 1 Evaluation of Proposed Change

==1.0 DESCRIPTION==

==2.0 PROPOSED CHANGE==

==3.0 BACKGROUND==

==4.0 TECHNICAL ANALYSIS==

4.1  Proposed Change to TS 3.1.4 4.2  Proposed Change to TS Section 4.2.1 4.3  Proposed Change to TS Section 5.6.5

==5.0 REGULATORY ANALYSIS==

5.1  No Significant Hazards Consideration 5.2  Applicable Regulatory Requirements/Criteria

==6.0 ENVIRONMENTAL CONSIDERATION==

==7.0 REFERENCES==

Page 1 of 12

ATTACHMENT 1 Evaluation of Proposed Change

==1.0     DESCRIPTION==

In accordance with 10 CFR 50.90, "Application for amendment of license or construction permit,"
Exelon Generation Company, LLC (EGC) requests an amendment to Renewed Facility Operating License Nos. DPR-1 9 and DPR-25 for Dresden Nuclear Power Station (DNPS)
Units 2 and 3, and Renewed Facility Operating License Nos. DPR-29 and DPR-30 for Quad Cities Nuclear Power Station (QCNPS) Units 1 and 2. EGC will be transitioning to Westinghouse SVEA-96 Optima2 (Optima2) fuel at each DNPS and QCNPS unit beginning with QCNPS Unit 2 Cycle 19, which is currently scheduled to begin in late March 2006. Technical Specifications (TS) 3.1.4, "Control Rod Scram Times," TS Section 4.2.1, "Fuel Assemblies," and TS Section 5.6.5, "Core Operating Limits Report (COLR)," require revision to support this transition. The proposed change is described below.

==2.0      PROPOSED CHANGE==

TS 3.1.4 requires that each control rod scram time be within the limits specified in Table 3.1.4-1 and that no more than 12 control rods or two adjacent rods be "slow" in accordance with the Table. An editorial change is proposed to remove the phrase "for GE analyzed cores" from Table 3.1.4-1.;
TS Section 4.2.1 provides a description of fuel assemblies. The current description states, in part, "The assemblies may contain water rods or a water box." The proposed change revises this sentence to read, "The assemblies may contain water rods or other assembly bypass channels."
TS Section 5.6.5.b specifies the analytical methods used to determine the core operating limits.
Specifically, TS Section 5.6.5.b states:
         "b. The analytical methods used to determine the core operating limits shall be those previously reviewed and approved by the NRC, specifically those described in the following documents:"
The proposed change revises the list of analytical methods to add references to the Westinghouse analytical methods that will be used to determine the core operating limits to support cores containing Optima2 fuel. Specifically, the following references are added.
: 1. CENPD-300-P-A, "Reference Safety Report for Boiling Water Reactor Reload Fuel."
: 2. WCAP-16081-P-A, "10x10 SVEA Fuel Critical Power Experiments and CPR Correlation: SVEA-96 Optima2."
: 3. WCAP-15682-P-A, "Westinghouse BWR ECCS Evaluation Model: Supplement 2 to Code Description, Qualification and Application."
: 4. WCAP-16078-P-A, "Westinghouse BWR ECCS Evaluation Model: Supplement 3 to Code Description, Qualification and Application to SVEA-96 Optima2 Fuel."
: 5. WCAP-1 5836-P-A, "Fuel Rod Design Methods for Boiling Water Reactors -
Supplement 1."
Page 2 of 12

ATTACHMENT I Evaluation of Proposed Change
: 6. WCAP-1 5942-P-A, "Fuel Assembly Mechanical Design Methodology for Boiling Water Reactors, Supplement 1 to CENPD-287-P-A."
: 7. CENPD-390-P-A, "The Advanced PHOENIX and POLCA Codes for Nuclear Design of Boiling Water Reactors."
Attachments 2 and 3 provide TS marked-up pages indicating the proposed change for DNPS and QCNPS, respectively. Attachments 4 and 5 provide the retyped TS pages incorporating the proposed change for DNPS and QCNPS, respectively.

==3.0     BACKGROUND==

EGC will be transitioning to Optima2 fuel at each DNPS and QCNPS unit beginning with QCNPS Unit 2 Cycle 19. Westinghouse, in conjunction with EGC, will be designing future core reloads for DNPS and QCNPS beginning with QCNPS Unit 2 Cycle 19, which is scheduled to begin in late March 2006. TS 3.1.4, TS Section 4.2.1, and TS Section 5.6.5 require revision to support this transition.

==4.0     TECHNICAL ANALYSIS==

4.1     Proposed Change to TS 3.1.4 TS 3.1.4 requires that each control rod scram time be within the limits specified in Table 3.1.4-1 and that no more than 12 control rods or two adjacent rods be "slow" in accordance with the Table. Currently, Table 3.1.4-1 lists scram time acceptance criteria, with an annotation that the times listed are for General Electric (GE) analyzed cores. Since Westinghouse will be designing future core reloads for DNPS and QCNPS, an editorial change is proposed to remove the phrase "for GE analyzed cores" from Table 3.1.4-1.
The need for including the phrase "for GE analyzed cores" in the existing TS is documented in Reference 1, which requested NRC approval of TS changes to support the change in fuel vendors from Siemens Power Corporation (SPC) to GE. Specifically, at the time of the change in fuel vendors, the methodology of modeling control rod insertion during a scram was different for SPC and GE. Therefore, Table 3.1.4-1 was modified to include scram times applicable to both the SPC and GE methodology. The NRC approved these changes in References 2 and 3 for DNPS and QCNPS, respectively. The associated NRC safety evaluation stated that this change is acceptable, based on consistency with the staff-approved GE methodology.
The proposed change does not alter the acceptance criteria for control rod scram times. Future core reloads will be analyzed using the NRC-approved methodology for modeling control rod insertion during a scram as described in Reference 1.
4.2      Proposed Change to TS Section 4.2.1 A complete description of the Optima2 fuel design is contained in Westinghouse topical report WCAP-15942-P. The Optima2 fuel channel consists of an inlet piece and a channel. This channel consists of an outer channel with a square cross section and an internal double-walled, cruciform structure, or "watercress," which forms channels for non-boiling water. The Page 3 of 12

ATTACHMENT 1 Evaluation of Proposed Change watercross structure is composed of a square central water channel and smaller water channels in each of the four wings. The watercress structure, along with the outer channel walls, form four sub-channels in which the sub-bundles are positioned.
TS Section 4.2.1 provides a description of fuel assemblies contained within the reactor. The description currently states, in part, that "The assemblies may contain water rods or a water box." Since the Optima2 fuel design contains a watercross structure, the description of fuel assemblies contained in TS Section 4.2.1 must be revised. Therefore, the proposed change revises the current description to state "The assemblies may contain water rods or other assembly bypass channels."
4.3      Proposed Change to TS Section 5.6.5 TS Section 5.6.5 requires, in part, the establishment of core operating limits prior to each reload cycle and that these limits be documented in the COLR. As stated in TS Section 5.6.5.b, the analytical methods used to determine the core operating limits shall be previously reviewed and approved by the NRC and documented in this section of the TS. The COLR contains a complete identification for each of the referenced topical reports used in the preparation of the COLR.
Future core reloads utilizing Optima2 fuel will use analytical methods described in topical reports that are currently not listed in TS Section 5.6.5.b. Since these methodologies will be used for the design and analysis of the DNPS and QCNPS core reloads, they must be added to the references included in TS Section 5.6.5.b.
The references being added have all been submitted to the NRC, and have either been approved or are currently under NRC review. Those methodologies that are currently under NRC review are scheduled to receive NRC approval prior to the first use of Optima2 fuel in a reload core at either DNPS or QCNPS. The following table lists the topical reports being added to TS Section 5.6.5.b, along with the current status of the NRC's review and approval.
Topical                                                   Submitted for           NRC Report                             Title                  NRC Review          Approval CENPD-300-P-A           Reference Safety Report for Boiling     12/8/94           5124/96 Water Reactor Reload Fuel WCAP-16081-P-A*         10xlO SVEA Fuel Critical Power          5/12/03           12/9/04 Experiments and CPR Correlation:
SVEA-96 Optima2 WCAP-15682-P-A          Westinghouse BWR ECCS                   2/8/02            3/10/03 Evaluation Model: Supplement 2 to Code Description, Qualification and Application WCAP-1 6078-P-A         Westinghouse BWR ECCS                   4/30/03           10/15/04 Evaluation Model: Supplement 3 to Code Description, Qualification and Application to SVEA-96 Optima2 Fuel Page 4 of 12

ATTACHMENT 1 Evaluation of Proposed Change Topical                                                     Submitted for             NRC Report                             Title                    NRC Review            Approval WCAP-15836-P-A*          Fuel Rod Design Methods for              6/25/02             Currently Boiling Water Reactors -                                     under NRC Supplement 1                                                 review WCAP-15942-P-A*          Fuel Assembly Mechanical Design          10/31/04            Currently Methodology for Boiling Water                                under NRC Reactors, Supplement 1 to                                    review CENPD-287-P-A CENPD-390-P-A            The Advanced PHOENIX and                4/15/99              7/24/00 POLCA Codes for Nuclear Design of Boiling Water Reactors
* The proposed change includes the "-A" (i.e., designating accepted) following the topical report identification symbol due to the expectation that Westinghouse will reissue the topical report prior to NRC approval of the proposed change.
A brief description of each of these topical reports is provided below. EGC and Westinghouse have reviewed the associated NRC safety evaluations for those topical reports that have been previously approved by the NRC. The use of the topical reports for DNPS and QCNPS reload core designs will be limited to the extent specified and under the limitations delineated in the NRC safety evaluations.
4.3.1   CENPD-300-P-A, "Reference Safety Report for Boiling Water Reactor Reload Fuel" Topical report CENPD-300-P-A for BWR reload fuel describes the reload fuel design and safety analysis process used in specific plant applications. Specific topics related to the Westinghouse BWR reload fuel design and safety analysis methodology are contained in numerous licensing topical reports describing portions of the overall methodology. Topical report CENPD-300-P-A integrates all the separate reports into a single comprehensive reload fuel design and safety analysis methodology. Between the contents of the separate licensing topical reports and contents of CENPD-300-P-A, the code methods, code qualification, design bases, methodology, and sample applications are described for all fuel design and safety analyses performed in support of plant modifications requiring a safety evaluation of the fuel, core, reactor coolant pressure boundary, or containment systems, including BWR reload fuel applications. The NRC approved topical report CENPD-300-P-A in Reference 4.
One of the limitations identified in Reference 4 stated that the Westinghouse methodology for determining the operating limit minimum critical power ratio (OLMCPR) for non-Westinghouse fuel is acceptable only when each licensee application of the methodology identifies the value of the conservative adder to the OLMCPR. The correlation applied to the experimental data to determine the value of the adder must be shown to meet the 95/95 statistical criteria. In addition, the licensee's submittal must include the justification for the adder and reference the appropriate supporting documentation.
Page 5 of 12

ATTACHMENT I Evaluation of Proposed Change This adder will be calculated as part of the reload core design. Therefore, the information described above is currently not available. EGC will submit this information in a separate submittal during the fourth quarter of 2005.
4.3.2 WCAP-16081-P, "lOx10 SVEA Fuel Critical Power Experiments and CPR Correlation: SVEA-96 Optima2" Topical report WCAP-16081-P describes the critical power ratio correlation for SVEA-96 Optima2 BWR fuel and the experimental data supporting the correlation. The correlation for the SVEA-96 Optima2 assembly and the bases for its acceptance are presented in this report. The NRC approved topical report WCAP-1 6081-P in Reference 5.
4.3.3 WCAP-15682-P-A, "Westinghouse BWR ECCS Evaluation Model: Supplement 2 to Code Description, Qualification and Application" Topical report WCAP-15682-P-A describes changes to the Westinghouse emergency core cooling system evaluation model for BWRs. This version of the evaluation model is identified as USA4. The only difference between this version of the evaluation model and the previously approved evaluation model (i.e.,
USA2) is the methodology used to determine when the fuel rod cladding will rupture. This document provides the basis for improving the cladding rupture criteria such that rupture occurs when either there is contact between adjacent rods or the burst stress criterion has been exceeded.
The USA2 evaluation model, which predicts cladding rupture when the burst stress criterion is exceeded, is applied in a way that limits the maximum average planar linear heat generation rate (MAPLHGR) to prevent rod-to-rod contact.
The USA4 evaluation model predicts cladding rupture when either there is contact with a neighboring rod or the burst stress criterion is exceeded -
whichever comes first. The MAPLHGR is limited in the application of the USA4 evaluation model to ensure that the 10 CFR 50.46 criteria are met.
The NRC approved topical report WCAP-1 5682-P-A in Reference 6.
4.3.4 WCAP-16078-P-A, "Westinghouse BWR ECCS Evaluation Model: Supplement 3 to Code Description, Qualification and Application to SVEA-96 Optima2 Fuel" Topical report WCAP-1 6078-P-A describes changes to the Westinghouse emergency core cooling system evaluation model for BWRs. This version of the evaluation model is identified as USA5. The differences between this version and the previously approved version (i.e., USA4) are (1) a change to the counter-current flow limit correlation, (2) the addition of a fuel rod plenum model that is applicable to part-length fuel rods, and (3) incorporation of the applicable features of the improved STAV7.2 fuel performance model. In addition, this topical report also provides the basis for applying the USA5 evaluation model to the SVEA-96 Optima2 fuel design.
The NRC approved topical report WCAP-16078-P-A in Reference 7. The NRC safety evaluation states that Westinghouse has adequately demonstrated the Page 6 of 12

ATTACHMENT 1 Evaluation of Proposed Change reasonably conservative nature of its modified emergency core cooling system methodology, except for the implementation of STAV7.2 modeling features that are currently under review by the NRC.
4.3.5 WCAP-1 5836-P, "Fuel Rod Design Methods for Boiling Water Reactors -
Supplement 1" Topical Report WCAP-1 5836-P describes improvements to the Westinghouse BWR fuel rod performance codes in support of the program to extend the accepted application of these codes to a fuel rod average burnup of 62 MWdIkgU. This report describes the latest versions of the STAV, VIK, and COLLAPS codes. This document is a supplement to the approved topical report describing the STAV, VIK, and COLLAPS codes. This supplement provides a description of the revised models implemented in the latest code versions along with the qualification actions which demonstrate that these codes are qualified for fuel rod design and safety analyses to a rod average burnup of 62 MWd/kgU.
Westinghouse submitted topical report WCAP-1 5836-P to the NRC for review on June 25, 2002, and this topical report is currently under NRC review. It should be noted that both DNPS and QCNPS have license conditions that limit the maximum rod average burnup for any rod to 60 GWD/MTU until the completion of an NRC environmental assessment supporting an increased limit.
4.3.6 WCAP-15942-P, "Fuel Assembly Mechanical Design Methodology for Boiling Water Reactors, Supplement 1 to CENPD-287-P-A" WCAP-1 5942-P describes improvements to the methodology contained in CENPD-287-P-A associated with adoption of the latest versions of the STAV, VIK, and COLLAPS codes. Utilization of these improved code versions, in conjunction with the required revisions to the methodology associated with the use of these codes, justifies an extension of the burnup range for which the methodology can be applied to a rod-average burnup of 62 MWd/kgU.
This report also contains an application of the updated methodology to the latest version of the Westinghouse SVEA-96 fuel assembly referred to as SVEA-96 Optima2. In conjunction with an expanded fuel rod and assembly inspection data base and test basis, this sample application demonstrates that the SVEA-96 Optima2 assembly satisfies the Westinghouse design criteria to a rod-average burnup of 62 MWd/kgU for the sample plant application.
Westinghouse submitted topical report WCAP-1 5942-P to the NRC for review on October 31, 2004, and this topical report is currently under NRC review.
4.3.7 CENPD-390-P-A, "The Advanced PHOENIX and POLCA Codes for Nuclear Design of Boiling Water Reactors" The two principal computer programs for BWR steady-state nuclear design and analysis used by Westinghouse are PHOENIX and POLCA. The PHOENIX code is a two-dimensional multi-group transport theory code used to calculate the lattice physics constants of BWR fuel assemblies. The POLCA code is a two-Page 7 of 12

ATTACHMENT I Evaluation of Proposed Change group nodal code used for the three-dimensional simulation of the nuclear and thermal-hydraulic conditions in BWR cores. In addition, several auxiliary codes are also utilized in order to facilitate calculations and transfer of data between the aforementioned codes.
Topical report CENPD-390-P-A describes the changes made to the PHOENIX and POLCA codes since they were previously reviewed and approved by the NRC. It also provides assessment against operational data and measurements to demonstrate that the codes are capable of predicting power distributions, thermal limits, and critical conditions necessary to design a BWR. The report provides a detailed description of the verification that has been performed to qualify the computer codes and analysis methods which are used for the nuclear design and analysis of BWRs.
The NRC approved topical report CENPD-390-P-A in Reference 8.

==5.0 REGULATORY ANALYSIS==

5.1 No Significant Hazards Consideration Exelon Generation Company, LLC (EGC) will be transitioning to Westinghouse SVEA-96 Optima2 (Optima2) fuel at Dresden Nuclear Power Station (DNPS), Units 2 and 3, and Quad Cities Nuclear Power Station (QCNPS), Units 1 and 2. Westinghouse, in conjunction with EGC, will be designing future core reloads for DNPS and QCNPS beginning with QCNPS Unit 2 Cycle 19, which is scheduled to begin in late March 2006.
Technical Specifications (TS) 3.1.4, "Control Rod Scram Times," TS Section 4.2.1, "Fuel Assemblies," and TS Section 5.6.5, "Core Operating Limits Report (COLR)," require revision to support this transition. Specifically, the proposed change: (1) incorporates an editorial change to TS 3.1.4 to clarify that the control rod scram times specified in Table 3.1.4-1 will continue to apply independent of whether the core is analyzed by General Electric; (2) revises TS Section 4.2.1 to modify the description of fuel assemblies to be more generic yet envelope key fuel assembly characteristics; and (3) revises TS Section 5.6.5 to add analytical methods that support design of core reloads utilizing Optima2 fuel.
According to 10 CFR 50.92, "Issuance of amendment," paragraph (c), a proposed amendment to an operating license involves 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 different kind of accident from any accident previously evaluated; or (3)     Involve a significant reduction in a margin of safety.
EGC has evaluated the proposed change to the TS for DNPS, Units 2 and 3, and QCNPS, Units 1 and 2, using the criteria in 10 CFR 50.92, and has determined that the Page 8 of 12

ATTACHMENT 1 Evaluation of Proposed Change proposed change does not involve a significant hazards consideration. The following information is provided to support a finding of no significant hazards consideration.
: 1.      Does the proposed change involve a significant increase in the probability or consequences of an accident previously evaluated?
Response: No The proposed change has no effect on any accident initiator or precursor previously evaluated and does not change the manner in which the core is operated. The type of fuel is not a precursor to any accident. The new methodologies for determining core operating limits have been validated to ensure that the output accurately models predicted core behavior, and use of the methodologies will be within the ranges previously approved. The new methodologies being referenced have all been submitted to the NRC, and have either been approved or are currently under NRC review. Those methodologies that are currently under NRC review are scheduled to receive NRC approval prior to the first use of Optima2 fuel in a reload core at either DNPS or QCNPS.
There is no change in the consequences of an accident previously evaluated.
The proposed change in the administratively controlled analytical methods does not affect the ability to successfully respond to previously evaluated accidents and does not affect radiological assumptions used in the evaluations. Source term from Optima2 fuel will be bounded by the source term assumed in the accident analyses. There is no affect on the type or amount of radiation released, and there is no affect on predicted offsite doses in the event of an accident.
Therefore, the proposed change does not involve a significant increase in the probability or consequences of an accident previously evaluated.
: 2. Does the proposed change create the possibility of a new or different kind of accident from any-accident previously evaluated?
Response: No The proposed change does not affect the performance of any DNPS or QCNPS structure, system, or component credited with mitigating any accident previously evaluated. The use of new analytical methods, which have either been reviewed and approved by the NRC or are currently being reviewed by the NRC, for the design of a core reload will not affect the control parameters governing unit operation or the response of plant equipment to transient conditions. The proposed change does not introduce any new modes of system operation or failure mechanisms.
Therefore, the proposed change does not create the possibility of a new or different kind of accident from any previously evaluated.
Page 9 of 12

ATTACHMENT I Evaluation of Proposed Change
: 3.      Does the proposed change involve a significant reduction in a margin of safety?
Response: No The proposed change to TS 3.1.4 clarifies that analyses for design basis accidents and transients will continue to support the scram times listed in TS Table 3.1.4-1, independent of whether General Electric analyzes the core. The proposed change does not alter the acceptance criteria for control rod scram times. Future core reloads will be analyzed using the NRC-approved methodology for modeling control rod insertion during a scram. The proposed change to TS Section 4.2.1 revises the description of fuel assemblies to envelope the Optima2 fuel characteristics. The proposed change to TS Section 5.6.5 adds new analytical methods for design and analysis of core reloads to the list of methods currently used to determine the core operating limits. The NRC has either previously approved the analytical methods being added, or is currently reviewing the methods.
The proposed change does not modify the safety limits or setpoints at which protective actions are initiated, and does not change the requirements governing operation or availability of safety equipment assumed to operate to preserve the margin of safety.
Therefore, the proposed change does not involve a significant reduction in a margin of safety.
Based upon the above, EGC concludes that the proposed amendment presents no significant hazards consideration under the standards set forth in 10 CFR 50.92(c), and, accordingly, a finding of no significant hazards consideration is justified.
5.2 Applicable Regulatory Requirements/Criteria 10 CFR 50.36, "Technical specifications," provides the regulatory requirements for the content required in a licensee's TS. 10 CFR 50.36, paragraph (c)(5) states that TS will include administrative controls that address the provisions relating to organization and management, procedures, recordkeeping, review and audit, and reporting necessary to assure operation of the facility in a safe manner. The COLR is required as part of the reporting requirements specified in the DNPS and QCNPS TS administrative controls.
In addition, it is required that the analytical methods used to determine the core operating limits be approved and described in the administrative controls section of the TS. The proposed change ensures that these requirements are met.
10 CFR 50.36, paragraph (c)(4) states that design features to be included are those features of the facility such as materials of construction and geometric arrangements, which, if altered or modified, would have a significant effect on safety. The proposed change revises TS Section 4.2.1 to modify the description of fuel assemblies to be more generic yet envelope key fuel assembly characteristics. The revised description meets the requirements of 10 CFR 50.36, paragraph (c)(4).
Page 10 of 12

ATTACHMENT 1 Evaluation of Proposed Change In conclusion, based on the considerations discussed above, (1) there is reasonable assurance that the health and safety of the public will not be endangered by operation in the proposed manner, (2) such activities will be conducted in compliance with the Commission's regulations, and (3) the issuance of the amendment will not be inimical to the common defense and security or to the health and safety of the public.

==6.0      ENVIRONMENTAL CONSIDERATION==

EGC has determined that the proposed amendment would change a requirement with respect to installation or use of a facility component located within the restricted area, as defined in 10 CFR 20, "Standards for Protection Against Radiation." However, the proposed amendment does not involve: (i) a significant hazards consideration, (ii) a significant change in the types or significant increase in the amounts of any effluent that may be released offsite, or (iii) a significant increase in individual or cumulative occupational radiation exposure. Accordingly, the proposed amendment meets the eligibility criterion for categorical exclusion set forth in 10 CFR 51.22, "Criterion for categorical exclusion; identification of licensing and regulatory actions eligible for categorical exclusion or otherwise not requiring environmental review," Paragraph (c)(9). Therefore, pursuant to 10 CFR 51.22, Paragraph (b), no environmental impact statement or environmental assessment need be prepared in connection with the proposed amendment.

==7.0      REFERENCES==
: 1.       Letter from R. M. Krich (Commonwealth Edison Company) to U. S. NRC, "Request for Technical Specifications Change, Transition to General Electric Fuel," dated September 29, 2000
: 2.       Letter from S. N. Bailey (NRC) to 0. D. Kingsley (Exelon Generation Company, LLC),
        "Issuance of Amendments (TAC Nos. MB0170, MB0171, MB1337, MB1338, MB2715, and MB2716)," dated November 2, 2001
: 3.      Letter from S. N. Bailey (NRC) to 0. D. Kingsley (Exelon Generation Company, LLC),
        "Issuance of Amendments (TAC Nos. MB0168, MB0169, MB13327, and MB1328)," dated December 20, 2001
: 4.      Letter from R. C. Jones (NRC) to D. Ebeling-Koning (ABB CENO Fuel Operations),
        "CENPD-300-P, 'Reference Safety Report for Boiling Water Reactor Reload Fuel,' (TAC No. M91197)," dated May 24,1996
: 5.      Letter from H. N. Berkow (NRC) to J. A. Gresham (Westinghouse Electric Company),
        "Final Safety Evaluation for Topical Report (TR) WCAP-16081-P, '10xlO SVEA Fuel Critical Power Experiments and CPR Correlation: SVEA-96 Optima2' (TAC No.
MB901 1)," dated December 9, 2004
: 6.      Letter from H. N. Berkow (NRC) to P. W. Richardson (Westinghouse Electric Company),
        "Acceptance for Referencing Topical Report WCAP-15682-P, Westinghouse BWR Page 11 of 12

ATTACHMENT 1 Evaluation of Proposed Change ECCS Evaluation Model: Supplement 2 to Code Description, Qualification and Application' (TAC No. MB4276)," dated March 10, 2003
: 7. Letter from H. N. Berkow (NRC) to J. A. Gresham (Westinghouse Electric Company),
   "Final Safety Evaluation for Topical Report WCAP-1 6078-P, Westinghouse BWR ECCS Evaluation Model: Supplement 3 to Code Description, Qualification and Application to SVEA-96 Optima2 Fuel' (TAC No. MB8908)," dated October 15, 2004
: 8. Letter from S. A. Richards (NRC) to 1.C. Rickard (Combustion Engineering Nuclear Power), "Acceptance for Referencing of CENPD-390-P, 'The Advanced Phoenix and Polca Codes for Nuclear Design of Boiling Water Reactors' (TAC No. MA5659)," dated July 24, 2000 Page 12 of 12

ATTACHMENT 2 Markup of Proposed Technical Specifications Pages DRESDEN NUCLEAR POWER STATION, UNITS 2 AND 3 RENEWED FACILITY OPERATING LICENSE NOS. DPR-19 AND DPR-25 REVISED TECHNICAL SPECIFICATIONS PAGES 3.1.4-3 4.0-1 5.6-4

Control Rod Scram Times 3.1.4 Table 3.1.4-1. (page 1 of 1)
Control Rod Scram Times
 ----        ----------------             NOTES --     ----------------
: 1. OPERABLE control rods with scram times not within the limits of this Table are considered "slow."
: 2. Enter applicable Conditions and Required Actions of LCO 3.1.3, "Control Rod OPERABILITY," for control rods with scram times > 7 seconds to 90%
insertion. These control rods are inoperable, in accordance with SR 3.1.3.4, and are not considered "slow."
SCRAM TIMES(a)(b)(seconds) when REACTOR STEAM DOME PSSURE 2 800 psia PERCENT INSERTION                       tFo_            zd      reg) 5                                          0.48 20                                          0.89 50                                          1.98 90                                          3.44 (a) Maximum scram time from fully withdrawn position based on de-energization of scram pilot valve solenoids at time zero.
(b) Scram times as a function of reactor steam dome pressure when ( 800 psig are within established limits.
Dresden 2 and 3                         3.1.4-3                      Amendment No.

Design Features 4.0 4.0  DESIGN FEATURES 4.1  Site Location 4.1.1   Site and Exclusion Area Boundaries The site area boundary follows the Illinois River to the north, the Kankakee River to the east, a country road from Divine extended eastward to the Kankakee River on the south, and the Elgin, Joliet, and Eastern Railway right-of-way on the west. The exclusion area boundary shall be an 800 meter radius from the centerline of the reactor vessels.
4.1.2   Low PoDulation Zone The low population zone shall be a five mile radius from the centerline of the reactor vessels.
4.2  Reactor Core 4.2.1   Fuel Assemblies The reactor shall contain 724 fuel assemblies. Each assembly shall consist of a matrix of Zircaloy clad fuel rods with an initial composition of natural or slightly enriched uranium dioxide (UOD) as fuel material. The assemblies may contain water rods o a ter ox.       Limited substitutions of Zircaloy, ZIRLO, or t stainless steel filler rods for fuel rods, in accordance with approved applications of fuel rod configurations, may be used.
Fuel assemblies shall be limited to those fuel designs that have been analyzed with NRC staff approved codes and methods and have been shown by tests or analyses to comply with all safety design bases. A limited number of lead test assemblies that have not completed representative testing may be placed in nonlimiting core regions of'Aer-    dssA/y Ay~ss e                      ez/
4 2.2   Control Rod Ass The reactor core shall contain 177 cruciform shaped control rod assemblies. The control material shall be boron carbide and hafnium metal as approved by the NRC.
(continued)
Dresden 2 and 3                       4.0-1                Amendment No.

Reporting Requirements 5.6 5.6   Reporting Requirements 5.6.5        CORE OPERATING LIMITS REPORT (COLR)   (continued)
: 9. ANF-89-98(P)(A). Generic Mechanical Design Criteria for BWR Fuel Designs.
: 10. ANF-91-048(P)(A). Advanced Nuclear Fuels Corporation Methodology for Boiling Water Reactors EXEM BWR Evaluation Model.
: 11. Commonwealth Edison Company Topical Report NFSR-0091.
                        "Benchmark of CASMO/MICROBURN BWR Nuclear Design Methods."
: 12. EMF-85-74(P), RODEX2A (BWR) Fuel Rod Thermal Mechanical Evaluation Model.
: 13. NEDE-24011-P-A. "General Electric Standard Application for Reactor Fuel (GESTAR)."
: 14. NEDC-32981P, "GEXL96 Correlation for ATRIUM 9B Fuel,"
September 2000.
The COLR will contain the complete identification for each of the TS referenced topical reports used to prepare the COLR (i.e.. report number, title, revision, date, and any supplements).
: c. The core operating limits shall be determined such that all applicable limits (e.g.. fuel thermal mechanical limits.
core thermal hydraulic limits, Emergency Core Cooling Systems (ECCS) limits, nuclear limits such as SDM, transient analysis limits, and accident analysis limits) of the safety analysis are met.
: d. The COLR, including any midcycle revisions or supplements.
shall be provided upon issuance for each reload cycle to the NRC.
5.6.6        Post Accident Monitorina (PAM) Instrumentation ReDort When a report is required by Condition B or F of LCO 3.3.3.1, "Post Accident Monitoring (PAM) Instrumentation," a report shall be submitted within the following 14 days. The report shall outline the preplanned alternate method of monitoring, the cause of the inoperability. and the plans and schedule for restoring the instrumentation channels of the Function to OPERABLE status.
Dresden 2 and 3                        5.6-4              Amendment No.

Insert 5.6.5
: 15. CENPD-300-P-A, "Reference Safety Report for Boiling Water Reactor Reload Fuel."
: 16. WCAP-16081-P-A, "10x1 0 SVEA Fuel Critical Power Experiments and CPR Correlation: SVEA-96 Optima2."
: 17. WCAP-15682-P-A, "Westinghouse BWR ECCS Evaluation Model: Supplement 2 to Code Description, Qualification and Application."
: 18. WCAP-1 6078-P-A, "Westinghouse BWR ECCS Evaluation Model: Supplement 3 to Code Description, Qualification and Application to SVEA-96 Optima2 Fuel."
: 19. WCAP-1 5836-P-A, "Fuel Rod Design Methods for Boiling Water Reactors -
Supplement 1."
: 20. WCAP-1 5942-P-A, "Fuel Assembly Mechanical Design Methodology for Boiling Water Reactors, Supplement 1 to CENPD-287-P-A."
: 21. CENPD-390-P-A, 'The Advanced PHOENIX and POLCA Codes for Nuclear Design of Boiling Water Reactors."

ATTACHMENT 3 Markup of Proposed Technical Specifications Pages QUAD CITIES NUCLEAR POWER STATION, UNITS 1 AND 2 RENEWED FACILITY OPERATING LICENSE NOS. DPR-29 AND DPR-30 REVISED TECHNICAL SPECIFICATIONS PAGES 3.1.4-3 4.0-1 5.6-4

Control Rod Scram Times 3.1.4 Table 3.1.4-1 (page 1 of 1)
Control Rod Scram Times
 ------------------------------------ NOTES------------------------------------
: 1. OPERABLE control rods with scram times not within the limits of this Table are considered "slow."
: 2. Enter applicable Conditions and Required Actions of LCO 3.1.3, "Control Rod OPERABILITY," for control rods with scram times > 7 seconds to 90X insertion. These control rods are inoperable, in accordance with SR 3.1.3.4, and are not considered "slow."
SCRAM TIMES(a)(b) (seconds) when REACTOR STEAM DOME PERCENT INSERTION          PRESSURE 2 800 PSIG2for E a yzed       ores 5                                     0.48 20                                    0.89 50                                     1.98 90                                     3.44 (a) Maximum scram time from fully withdrawn position based on de-energization of scram pilot valve solenoids at time zero.
(b)   Scram times as a function of reactor steam dome pressure when
       < 800 psig are within established limits.
Quad Cities 1 and 2                   3.1.4-3                Amendment No.

Design Features 4.0 4.0  DESIGN FEATURES 4.1  Site Location 4.1.1  Site and Exclusion Area The site consists of approximately 784 acres on the east bank of the Mississippi River opposite the mouth of the Wapsipinicon River, approximately three miles north of the village of Cordova.
Rock Island County, Illinois. The exclusion area shall not be less than 380 meters from the centerline of the chimney.
4.1.2  Low Population Zone The low population zone shall be a three mile radius from the centerline of the chimney.
4.2  Reactor Core 4.2.1  Fuel Assemblies The reactor shall contain 724 fuel assemblies. Each assembly shall consist of a matrix of Zircaloy clad fuel rods with an initial composition of natural or slightly enriched uranium dioxide (UO ) as fuel material. The assemblies may contain water rods o a     ter     . Limited substitutions of Zircaloy or ZIRLO filler rods for fuel rods, in accordance with approved applications of fuel rod configurations, may be used. Fuel assemblies shall be limited to those fuel designs that have been analyzed with NRC staff approved codes and methods and have been shown by tests or analyses to comply with all safety design bases.
A limited number of lead test assemblies that have not completed rtative          testing may be placed in nonlimiting core regions.
4.2.2  Control Rod Assembl The reactor core shall contain 177 cruciform shaped control rod assemblies. The control material shall be boron carbide and hafnium metal as approved by the NRC.
(continued)
Quad Cities 1 and 2                   4.0-1                Amendment No.

Reporting Requirements Reporting Requirements 5.6 5.6   Reporting Requirements 5.6.5        CORE OPERATING LIMITS REPORT (COLR)   (continued)
: 10. Advanced Nuclear Fuels Corporation Critical Power Methodology for Boiling Water Reactors/Advanced Nuclear Fuels Corporation Critical Power Methodology for Boiling Water Reactors: Methodology for Analysis of Assembly Channel Bowing Effects/NRC Correspondence, ANF-524(P)(A).
: 11. COTRANSA 2: A Computer Program for Boiling Water Reactor Transient Analyses, ANF-913(P)(A).
: 12. Advanced Nuclear Fuels Corporation Methodology for Boiling Water Reactors EXEM BWR Evaluation Model, ANF-91-048(P)(A).
: 13. Commonwealth Edison Topical Report NFSR-0091, "Benchmark of CASMO/MICROBURN BWR Nuclear Design Methods."
: 14. ANFB Critical Power Correlation Application for Coresident Fuel, EMF-1125(P)(A).
: 15. EMF-85-74(P), RODEX2A(BWR) Fuel Rod Thermal Mechanical Evaluation Model, Supplement 1(P)(A) and Supplement 2 22v:ASE^     S.S.5          (P)(A), Siemens Power Corporation, February 1998.
: 16. NEDC-32981P. "GEXL96 Correction for ATRIUM 9B Fuel."
The COLR will contain the complete identification for each of the TS referenced topical reports used to prepare the COLR (i.e., report number, title, revision, date, and any supplements).
: c. The core operating limits shall be determined such that all applicable limits (e.g., fuel thermal mechanical limits, core thermal hydraulic limits, Emergency Core Cooling Systems (ECCS) limits, nuclear limits such as SDM, transient analysis limits, and accident analysis limits) of the safety analysis are met.
: d. The COLR. including any midcycle revisions or supplements, shall be provided upon issuance for each reload cycle to the NRC.
(continued)
Ouad Cities I and 2                    5.6-4               Amendment No. Amen

Insert 5.6.5
: 17. CENPD-300-P-A, "Reference Safety Report for Boiling Water Reactor Reload Fuel."
: 18. WCAP-16081-P-A, "10x10 SVEA Critical Power Experiments and CPR Correlation:
SVEA-96 Optima2."
: 19. WCAP-1 5682-P-A, "Westinghouse BWR ECCS Evaluation Model: Supplement 2 to Code Description, Qualification and Application."
: 20. WCAP-16078-P-A, "Westinghouse BWR ECCS Evaluation Model: Supplement 3 to Code Description, Qualification and Application to SVEA-96 Optima2 Fuel."
: 21. WCAP-1 5836-P-A, "Fuel Rod Design Methods for Boiling Water Reactors -
Supplement 1."
: 22. WCAP-1 5942-P-A, "Fuel Assembly Mechanical Design Methodology for Boiling Water Reactors, Supplement 1 to CENPD-287-P-A."
: 23. CENPD-390-P-A, "The Advanced PHOENIX and POLCA Codes for Nuclear Design of Boiling Water Reactors."

ATTACHMENT 4 Retyped Technical Specifications Pages for Proposed Change DRESDEN NUCLEAR POWER STATION, UNITS 2 AND 3 RENEWED FACILITY OPERATING LICENSE NOS. DPR-19 AND DPR-25 REVISED TECHNICAL SPECIFICATIONS PAGES 3.1.4-3 4.0-1 5.6-4 5.6-5

Control Rod Scram Times 3.1.4 Table 3.1.4-1 (page 1 of 1)
Control Rod Scram Times
------------------------------------- NOTES ----------------
: 1. OPERABLE control rods with scram times not within the limits of this Table are considered "slow."
: 2. Enter applicable Conditions and Required Actions of LCO 3.1.3, "Control Rod OPERABILITY," for control rods with scram times > 7 seconds to 90%
insertion. These control rods are inoperable, in accordance with SR 3.1.3.4, and are not considered "slow."
SCRAM TIMES(a)(b)(seconds) when REACTOR STEAM DOME PERCENT INSERTION                       PRESSURE 2 800 psig                 I 5                                      0.48 20                                      0.89 50                                      1.98 90                                      3.44 (a)   Maximum scram time from fully withdrawn position based on de-energization of scram pilot valve solenoids at time zero.
(b) Scram times as a function of reactor steam dome pressure when < 800 psig are within established limits.
Dresden 2 and 3                      3.1.4- 3                        Amendment No.

Design Features 4.0 4.0   DESIGN FEATURES 4.1   Site Location 4.1.1  Site and Exclusion Area Boundaries The site area boundary follows the Illinois River to the north, the Kankakee River to the east, a country road from Divine extended eastward to the Kankakee River on the south, and the Elgin, Joliet, and Eastern Railway right-of-way on the-west. The exclusion area boundary shall be an 800 meter radius from the centerline of the reactor vessels.
4.1.2  Low Population Zone The low population zone shall be a five mile radius from the centerline of the reactor vessels.
4.2  Reactor Core 4.2.1  Fuel Assemblies The reactor shall contain 724 fuel assemblies. Each assembly shall consist of a matrix of Zircaloy clad fuel rods with an initial composition of natural or slightly enriched uranium dioxide (U02) as fuel material. The assemblies may contain water rods or other assembly bypass channels. Limited substitutions of l Zircaloy, ZIRLO, or stainless steel filler rods for fuel rods, in accordance with approved applications of fuel rod configurations, may be used. Fuel assemblies shall be limited to those fuel designs that have been analyzed with NRC staff approved codes and methods and have been shown by tests or analyses to comply with all safety design bases. A limited number of lead test assemblies that have not completed representative testing may be placed in nonlimiting core regions.
4.2.2   Control Rod Assemblies The reactor core shall contain 177 cruciform shaped control rod assemblies. The control material shall be boron carbide and hafnium metal as approved by the NRC.
(continued)
Dresden 2 and 3                       4.0- 1                      Amendment No.

Reporting Requirements 5.6 5.6   Reporting Requirements 5.6.5        CORE OPERATING LIMITS REPORT (COLR)   Continued)
: 9. ANF-89-98(P)(A), Generic Mechanical Design Criteria for BWR Fuel Designs.
: 10. ANF-91-048(P)(A), Advanced Nuclear Fuels Corporation Methodology for Boiling Water Reactors EXEM BWR Evaluation Model.
: 11. Commonwealth Edison Company Topical Report NFSR-0091, "Benchmark of CASMO/MICROBURN BWR Nuclear Design Methods."
: 12. EMF-85-74(P), RODEX2A (BWR) Fuel Rod Thermal Mechanical Evaluation Model.
: 13. NEDE-224011-P-A, "General Electric Standard Application for Reactor Fuel (GESTAR)."
: 14. NEDC-32981P, "GEXL96 Correlation for ATRIUM 9B Fuel,"
September 2000.
: 15. CENPD-300-P-A, "Reference Safety Report for Boiling Water Reactor Reload Fuel."
: 16. WCAP-16081-P-A, "lOxlO SVEA Fuel Critical Power Experiments and CPR Correlation: SVEA-96 Optima2."
: 17. WCAP-15682-P-A, "Westinghouse BWR ECCS Evaluation Model: Supplement 2 to Code Description, Qualification and Application."
: 18. WCAP-16078-P-A, "Westinghouse BWR ECCS Evaluation Model: Supplement 3 to Code Description, Qualification and Application to SVEA-96 Optima2 Fuel."
: 19. WCAP-15836-P-A, "Fuel Rod Design Methods for Boiling Water Reactors - Supplement 1."
: 20. WCAP-15942-P-A, "Fuel Assembly Mechanical Design Methodology for Boiling Water Reactors, Supplement 1 to CENPD-287-P-A."
: 21. CENPD-390-P-A, "The Advanced PHOENIX and POLCA Codes for Nuclear Design of Boiling Water Reactors."
(continued)
Dresden 2 and 3                       5.6- 4                       Amendment No.

Reporting Requirements 5.6 5.6   Reporting Requirements 5.6.5        CORE OPERATING LIMITS REPORT (COLR)   (continued)
The COLR will contain the complete identification for each of the TS referenced topical reports used to prepare the COLR (i.e., report number, title, revision, date, and any supplements).
: c. The core operating limits shall be determined such that all applicable limits (e.g., fuel thermal mechanical limits, core thermal hydraulic limits, Emergency Core Cooling Systems (ECCS) limits, nuclear limits such as SDM, transient analysis limits, and accident analysis limits) of the safety analysis are met.
: d. The COLR, including any midcycle revisions or supplements, shall be provided upon issuance for each reload cycle to the NRC.
5.6.6        Post Accident Monitorina (PAM) Instrumentation Report When a report is required by Condition B or F of LCO 3.3.3.1, "Post Accident Monitoring (PAM) Instrumentation," a report shall be submitted within the following 14 days. The report shall outline the preplanned alternate method of monitoring, the cause of the inoperability, and the plans and schedule for restoring the instrumentation channels of the Function to OPERABLE status.
Dresden 2 and 3                       5.6- 5                       Amendment No.

ATTACHMENT 5 Retyped Technical Specifications Pages for Proposed Change QUAD CITIES NUCLEAR POWER STATION, UNITS 1 AND 2 RENEWED FACILITY OPERATING LICENSE NOS. DPR-29 AND DPR-30 REVISED TECHNICAL SPECIFICATIONS PAGES 3.1.4-3 4.0-1 5.6-4 5.6-5

Control Rod Scram Times 3.1.4 Table 3.1.4-1 (page 1 of 1)
Control Rod Scram Times
          ---------------              NOTES------------------------------------
: 1. OPERABLE control rods with scram times not within the limits of this Table are considered "slow."
: 2. Enter applicable Conditions and Required Actions of LCO 3.1.3, "Control Rod OPERABILITY," for control rods with scram times > 7 seconds to 90%
insertion. These control rods are inoperable, in accordance with SR 3.1.3.4, and are not considered "slow."
SCRAM TIMES(a)(b) (seconds) when REACTOR STEAM DOME PERCENT INSERTION                        PRESSURE 2 800 psig                 I 5                                       0.48 20                                       0.89 50                                       1.98 90                                       3.44 (a) Maximum scram time from fully withdrawn position based on de-energization of scram pilot valve solenoids at time zero.
(b)  Scram times as a function of reactor steam dome pressure when
     < 800 psig are within established limits.
Quad Cities 1 and 2                  3.1.4- 3                        Amendment No.

Design Features 4.0 4.0   DESIGN FEATURES 4.1   Site Location 4.1.1   Site and Exclusion Area The site consists of approximately 784 acres on the east bank of the Mississippi River opposite the mouth of the Wapsipinicon River, approximately three miles north of the village of Cordova, Rock Island County, Illinois. The exclusion area shall not be less than 380 meters from the centerline of the chimney.
4.1.2   Low Population Zone The low population zone shall be a three mile radius from the centerline of the chimney.
4.2  Reactor Core 4.2.1   Fuel Assemblies The reactor shall contain 724 fuel assemblies. Each assembly shall consist of a matrix of Zircaloy clad fuel rods with an initial composition of natural or slightly enriched uranium dioxide (U02) as fuel material. The assemblies may contain water rods or other assembly bypass channels. Limited substitutions of l Zircaloy or ZIRLO filler rods for fuel rods, in accordance with approved applications of fuel rod configurations, may be used.
Fuel assemblies shall be limited to those fuel designs that have been analyzed with NRC staff approved codes and methods and have been shown by tests or analyses to comply with all safety design bases. A limited number of lead test assemblies that have not completed representative testing may be placed in nonlimiting core regions.
4.2.2  Control Rod Assemblies The reactor core shall contain 177 cruciform shaped control rod assemblies. The control material shall be boron carbide and hafnium metal as approved by the NRC.
(continued)
Quad Cities 1 and 2                   4.0-1I                      Amendment No.

Reporting Requirements 5.6 5.6   Reporting Requirements 5.6.5        CORE OPERATING LIMITS REPORT (COLR)   (continued)
: 10. Advanced Nuclear Fuels Corporation Critical Power Methodology for Boiling Water Reactors/Advanced Nuclear Fuels Corporation Critical Power Methodology for Boiling Water Reactors: Methodology for Analysis of Assembly Channel Bowing Effects/NRC Correspondence, ANF-524(P)(A).
: 11. COTRANSA 2: A Computer Program for Boiling Water Reactor Transient Analyses, ANF-913(P)(A).
: 12. Advanced Nuclear Fuels Corporation Methodology for Boiling Water Reactors EXEM BWR Evaluation Model, ANF-91-048(P)(A).
: 13. Commonwealth Edison Topical Report NFSR-0091, "Benchmark of CASMO/MICROBURN BWR Nuclear Design Methods."
: 14. ANFB Critical Power Correlation Application for Coresident Fuel, EMF-1125(P)(A).
: 15. EMF-85-74(P), RODEX2A(BWR) Fuel Rod Thermal Mechanical Evaluation Model, Supplement 1(P)(A) and Supplement 2 (P)(A), Siemens Power Corporation, February 1998.
: 16. NEDC-32981P, "GEXL96 Correction for ATRIUM 9B Fuel."
: 17. CENPD-300-P-A, "Reference Safety Report for Boiling Water Reactor Reload Fuel."
: 18. WCAP-16081-P-A, "lOxlO SVEA Critical Power Experiments and CPR Correlation: SVEA-96 Optima2."
: 19. WCAP-15682-P-A, "Westinghouse BWR ECCS Evaluation Model: Supplement 2 to Code Description, Qualification and Application."
: 20. WCAP-16078-P-A, "Westinghouse BWR ECCS Evaluation Model: Supplement 3 to Code Description, Qualification and Application to SVEA-96 Optima2 Fuel."
: 21. WCAP-15836-P-A, "Fuel Rod Design Methods for Boiling Water Reactors - Supplement 1."
(continued)
Quad Cities 1 and 2                    5.6- 4                       Amendment No.

Reporting Requirements Reporting Requirements 5.6 5.6   Reporting, Requirements 5.6.5        CORE OPERATING LIMITS REPORT (COLR)    (continued)
: 22. WCAP-15942-P-A, "Fuel Assembly Mechanical Design Methodology for Boiling Water Reactors, Supplement 1 to CENPD-287-P-A."
: 23. CENPD-390-P-A, "The Advanced PHOENIX and POLCA Codes for Nuclear Design of Boiling Water Reactors."
The COLR will contain the complete identification for each of the TS referenced topical reports used to prepare the COLR (i.e., report number, title, revision, date, and any supplements).
: c. The core operating limits shall be determined such that all applicable limits (e.g., fuel thermal mechanical limits, core thermal hydraulic limits, Emergency Core Cooling Systems (ECCS) limits, nuclear limits such as SDM, transient analysis limits, and accident analysis limits) of the safety analysis are met.
: d. The COLR, including any midcycle revisions or supplements, shall be provided upon issuance for each reload cycle to the NRC.
5.6.6        Post Accident Monitoring (PAM) Instrumentation Report When a report is required by Condition B or F of LCO 3.3.3.1, "Post Accident Monitoring (PAM) Instrumentation," a report shall be submitted within the following 14 days. The report shall outline the preplanned alternate method of monitoring, the cause of the inoperability, and the plans and schedule for restoring the instrumentation channels of the Function to OPERABLE status.
Quad Cities I and 2                    5.6-5                        Amendment No.}}