RS-09-074, License Amendment Request to Modify Clinton Power Station Facility Operating License in Support of the Use of Isotope Test Assemblies

License Amendment Request to Modify Clinton Power Station Facility Operating License in Support of the Use of Isotope Test Assemblies

ML091801061
Person / Time
Site:	Clinton
Issue date:	06/26/2009
From:	Hansen J Exelon Generation Co, Exelon Nuclear
To:	Document Control Desk, Office of Nuclear Reactor Regulation
Shared Package
ML091801065	List: RS-09-074, License Amendment Request to Modify Clinton Power Station Facility Operating License in Support of the Use of Isotope Test Assemblies
References
DRF 0000-0100-5453, RS-09-074 NEDO-33505, Rev 0
Download: ML091801061 (103)

Text

Exelon .

Exelon Nuclear 4300 Winfield Road www.exeloncorp.com Nuclear Warrenville, IL 60555 10 CFR 50 .90 RS-09-074 June 26, 2009 U.S. Nuclear Regulatory Commission ATTN: Document Control Desk Washington, DC 20555-0001 Clinton Power Station, Unit 1 Facility Operating License No. NPF-62 NRC Docket No. 50-461

Subject:

License Amendment Request to Modify Clinton Power Station Facility Operating License in Support of the Use of Isotope Test Assemblies In accordance with 10 CFR 50.90, "Application for amendment of license, construction permit, or early site permit," Exelon Generation Company, LLC (EGC) requests an amendment to Facility Operating License No. NPF-62 for Clinton Power Station (CPS), Unit 1 . Specifically, the proposed change modifies CPS License Condition 2.B.(6) and creates new License Conditions 1 .J and 2.B.(7) as part of a pilot program to irradiate Cobalt (Co)-59 targets to produce Co-60. In addition to the proposed license condition changes, EGC also requests an amendment to Appendix A, Technical Specifications (TS), of the CPS Facility Operating License. This proposed change would modify TS 4.2.1, "Fuel Assemblies," to describe the Isotope Test Assemblies (ITAs) being used .

EGC is collaborating with Global Nuclear Fuel - Americas, LLC (GNF) and GE - Hitachi Nuclear Energy Americas, LLC (GEH) to develop and implement a pilot program for producing Co-60 in the CPS reactor during power operation. The Co-60 is intended for use in the medical industry for use in cancer treatments, blood and instrument sterilization, radiography and security industry for imaging, and in the food industry for cold pasteurization or irradiation sterilization . EGC plans to load 8 ITAs as part of the CPS Reload 12 Cycle 13 core reload, during the January 2010 refueling outage (i .e., C1 R12) .

Attachment 1 to this letter provides an evaluation supporting the proposed changes. The marked-up Operating License and TS pages, with the proposed changes indicated, are provided in Attachment 2 to this letter .

Attachment 3 to this letter provides GEH Report NEDC-33505P, "Safety Analysis to Support Introduction of GE14i Isotope Test Assemblies (ITAs) in Clinton Power Station," dated June 2009, which GEH considers to contain proprietary information. The proprietary information is identified by bracketed text . GEH requests that the proprietary information in Attachment 3 be withheld from public disclosure, in accordance with the requirements of 10 CFR 2.390, "Public inspections, exemptions, requests for withholding," paragraph (a)(4) . A signed affidavit supporting this request is provided in Attachment 3 to this letter . Attachment 4 to this letter provides a non-proprietary version of the GEH Report (i.e ., NEDC-33505).

June 26, 2009 U . S. Nuclear Regulatory Commission Page 2 Additionally, the analyses supporting the results documented in NEDC-33505P will be available for NRC review at the Washington D.C . offices of General Electric shortly after submittal of the attached license amendment request.

Attachment 5 summarizes the formal regulatory commitments pending NRC approval of the proposed amendment .

The proposed change has been reviewed by the CPS Plant Operations Review Committee and approved by the Nuclear Safety Review Board in accordance with the requirements of the EGC Quality Assurance Program . EGC requests approval of the proposed change by January 11, 2010, with the amendment being implemented within 30 days of issuance .

In accordance with 10 CFR 50 .91, "Notice for public comment; State consultation," EGC is notifying the State of Illinois of this application for a change to the TS by sending a copy of this letter and its attachments to the designated State Official .

Should you have any questions concerning this letter, please contact Mr. Timothy A. Byam at (630) 657-2804 .

I declare under penalty of perjury that the foregoing is true and correct. Executed on the 26th day of June 2009.

Respectfully, Jeffrey L-F-lansen Manager - Licensing Exelon Generation Company, LLC Attachments:  : Evaluation of Proposed Changes : Mark-up of Proposed Operating License and Technical Specification Pages : Safety Analysis Report to Support Introduction of GE14i Isotope Test Assemblies (ITAs) in Clinton Power Station (Proprietary) : Safety Analysis Report to Support Introduction of GE14i Isotope Test Assemblies (ITAs) in Clinton Power Station (Non-Proprietary)  : Summary of Regulatory Commitments

bcc: Project Manager, NRR - Clinton Power Station Illinois Emergency Management Agency - Division of Nuclear Safety Site Vice President - Clinton Power Station Plant Manager - Clinton Power Station Regulatory Assurance Manager - Clinton Power Station Director - Licensing and Regulatory Affairs (West)

Manager, Licensing - Clinton Power Station Licensing Engineer - Clinton Power Station Brenda Fore - Clinton Power Station (Electronic Copy)

Exelon Document Control Desk Licensing (Hard Copy)

Commitment Tracking Coordinator - West J . L. Peterson (CPS)

M. A. Vandermyde (CPS)

M. Reitmeyer (Kennett Square)

ATTACHMENT 1 Evaluation of Proposed Changes

Subject:

License Amendment Request to Modify Clinton Power Station Facility Operating License in Support of the Use of Isotope Test Assemblies 1 .0

SUMMARY

DESCRIPTION 2.0 DETAILED DESCRIPTION 3 .0 TECHNICAL EVALUATION

4.0 REGULATORY EVALUATION

4.1 Applicable Regulatory Requirements/Criteria 4.2 Precedent 4.3 No Significant Hazards Consideration 4.4 Conclusions

6.0 REFERENCES

Page 1 of 12

ATTACHMENT 1 Evaluation of Proposed Changes 1 .0

SUMMARY

DESCRIPTION In accordance with 10 CFR 50.90, "Application for amendment of license, construction permit, or early site permit," Exelon Generation Company, LLC (EGC) requests an amendment to the Facility Operating License No. NPF-62 for Clinton Power Station (CPS),

Unit 1 . Specifically, the proposed change modifies CPS License Condition 2.B .(6) and creates new License Conditions 1 .J and 2.B.(7) as part of a pilot program to irradiate Cobalt (Co)-59 targets to produce Co-60. The Co-60 would ultimately be sold to licensed users in the medical industry for use in cancer treatments, blood and instrument sterilization, radiography and security industry for imaging, and to the food industry for cold pasteurization or irradiation sterilization. In addition to the proposed license condition changes, EGC also requests an amendment to Appendix A, Technical Specifications (TS),

of the CPS Facility Operating License . This proposed change would modify TS 4.2.1, "Fuel Assemblies," to describe the Isotope Test Assemblies (ITAs) being used .

2.0 DETAILED DESCRIPTION EGC proposes to add a new License Condition 1 .J which states, "The receipt, production, possession, transfer, and use of Cobalt-60 as authorized by this license will be in accordance with the Commission's regulations in 10 CFR Part 30." This new license condition allows for the production and transfer of Co-60 under the CPS Facility Operating License in accordance with 10 CFR Part 30.

Condition 2 .B.(6) of the current Operating License for CPS is revised to state, "Exelon Generation Company, pursuant to the Act and 10 CFR Parts 30, 40, and 70, to possess, but not separate, such byproduct and special nuclear materials as may be produced by the operation of the facility . Mechanical disassembly of the GE14i isotope test assemblies containing Co-60 is not considered separation ." This change is intended to provide clarification of the term "separation" relative to the removal of enclosed rods containing Co-60 from the CPS reactor core, as distinguished from separation of byproduct material from the special nuclear material fuel in rods as intended by the original license restriction.

EGC proposes to add a new License Condition 2.B .(7) which states, "Exelon Generation Company, pursuant to the Act and 10 CFR Part 30, to intentionally produce, possess, receive, transfer, and use Co-60." This new License Condition supports the pilot bulk isotope generation project at CPS by allowing intentional production of byproduct material during operation of the CPS facility .

EGC also proposes to add the following paragraph to the end of TS 4.2.1 . This proposed change ensures that the Technical Specifications accurately describe all types of fuel assemblies used in the CPS reactor.

"A maximum of eight GE14i isotope test assemblies will be placed in non-limiting core regions, beginning with the Reload 12 Cycle 13 core reload, with the purpose of obtaining surveillance data to verify that the GE14i assemblies perform satisfactorily in service prior to use of these design features on a production basis . Each GE14i assembly contains a small number of zircaloy-2 clad isotope rods that contain Cobalt-Page 2 of 12

ATTACHMENT 1 Evaluation of Proposed Changes 59 targets . These Cobalt-59 targets will transition into Cobalt-60 isotope targets during the cycle irradiation of the assemblies . Details of the GE14i assemblies are contained in NEDC-33505P, "Safety Analysis Report to Support Introduction of GE14i Isotope Test Assemblies (ITAs) in Clinton Power Station," Revision 0, dated June 2009."

A copy of the affected CPS Operating License and TS pages marked-up to show the proposed changes identified above is provided in Attachment 2.

3.0 TECHNICAL EVALUATION

The radionuclide Co-60 is the most commonly used source of gamma radiation for radiation technology, both for medical and industrial purposes . There is a growing demand for Co-60 as applications for this material are expanding at a rapid rate, as is the demand within existing applications . Co-60 is currently used in applications such as neurosurgical devices, food irradiation, and gamma sterilization of medical devices.

In medicine, Co-60 has been used to effectively treat hundreds of thousands of patients over the last 30 years. It is widely used in applications for treating benign and malignant brain tumors, vascular malformations, and pain or other functional problems . It also plays a significant role in the food pasteurization industry . Food irradiation is the process of imparting ionizing energy to food to kill microorganisms . A third significant use of the Co-60 radionuclide is in the medical sterilization industry . More than 40% of U. S. made medical devices (i .e., syringes, bandages, etc.) are sterilized using medical isotopes .

EGC plans to load eight isotope test assemblies (ITAs) as part of CPS Reload 12 Cycle 13 core reload during the January 2010 refueling outage (i .e., C1 R12) . These assemblies, also referred to as GE14i ITAs, are planned to be in operation as part of a joint program between EGC, Global Nuclear Fuel - Americas, LLC (GNF) and GE - Hitachi Nuclear Energy Americas, LLC (GEH) . The purpose of this ITA program is to obtain surveillance data to verify that fuel assemblies with the design features of the GE14i ITAs perform satisfactorily in service, prior to use of those features on a production basis. This proposed program results in the introduction of a new type of fuel assembly to the CPS core.

While the GEl4i ITAs are not Lead Test Assemblies (LTA), EGC has decided to apply the LTA requirements from CPS TS 4.2.1 to the introduction of the GE14i new fuel design . CPS TS 4.2.1 states that a limited number of LTAs may be placed in non-limiting core locations .

EGC intends to introduce the GE14i assembly to the CPS reactor core in accordance with this TS requirement. A limited number of ITAs (i .e ., eight) will be loaded into the CPS core in C1 R12 . These ITAs will be located in non-limiting locations in the core with respect to thermal limit margins and shutdown margins.

GEH proprietary report NEDC-33505P, "Safety Analysis Report to Support Introduction of GE14i Isotope Test Assemblies (ITAs) in Clinton Power Station," dated June 2009, is provided in Attachment 3. This report contains a description of the GE14i fuel assembly, as well as, descriptions of the nuclear core design and applicability of nuclear and safety analysis methods . Attachment 3 also contains details relative to the licensing evaluations Page 3 of 12

ATTACHMENT 1 Evaluation of Proposed Changes that were performed by GNF and GEH in support of the introduction of these new fuel assemblies to the CPS core . A non-proprietary version of this report is provided in .

Production of radioactive cobalt starts with natural cobalt, which is an element composed of Co-59, a stable isotope. Each isotope rod in the GE14i ITA contains pellets made of cobalt .

As described in Attachment 3, all aspects of the GE14i ITAs are controlled under the GE Nuclear Energy Quality Assurance Program . The GE14i zirconium tubing and components are procured, fabricated, and handled under the same quality controls as standard production fuel rods at GNF. Target pellets are handled with similar quality controls as U02 pellets. The Cobalt targets were certified to meet the current drawing and specification requirements and are verified to meet the design requirements on a sampling basis.

The GE14i isotope test assemblies are placed in the nuclear reactor, where they stay for varying amounts of time that depend upon neutron flux and the desired specific activity.

While in the reactor, a Co-59 atom absorbs a neutron and is converted into a Co-60 atom .

The resulting irradiated isotope rods, now containing Co-60, are sent offsite for separation and further processing .

As documented in Attachment 3, the design of the GE14i assembly utilizes segmented Co target rods . Segmented fuel rods have operated successfully in a number of GNF fuel products and LUA programs for decades, most recently at Forsmark and Gundremmingen, with no evidence of cracking caused by vibration . The absence of failures in the segmented fuel rods and the similarity of the axial weight differential between segmented rods and isotope rods provides assurance that cracking from vibration due to non-uniform weight distribution will not be a significant issue for the isotope rods. Additionally, the heavier sections of the isotope rods are the Zircaloy connections, which are all supported laterally by a spacer at the same elevation .

GEH has completed the required non-cycle specific evaluations to support the loading of the GE14i ITAs in the CPS reactor. Cycle specific analyses will be performed for CPS Reload 12 Cycle 13 to establish fuel operating limits for the ITAs that assure compliance with regulatory limits . Results of the cycle specific analyses will be documented in the CPS Reload 12 Cycle 13 Supplemental Reload Licensing Report . The impact of the isotope rods on the anticipated operational occurrences (A00) and design basis accidents (DBA) has been analyzed and the results are documented in Attachment 3. This includes an evaluation of the impact on previously evaluated AOOs (e.g ., increase in reactor pressure events, decrease in reactor coolant system flow rate events, reactivity and power distribution anomalies, and increase in reactor coolant inventory events), other transients (e.g., Anticipated Transients without Scram), DBAs (e.g., Control Rod Drop Accident, Main Steam Line Break Accident, Fuel Handling Accident, and Loss of Coolant Accident), and radiological source terms and accident doses. These analyses have demonstrated that there is no impact to the current A00 and DBA results due to the introduction of this new fuel assembly. addresses the applicability of the current analysis methods and methodologies to the GE14i fuel design . It also addresses each NRC approved method that is used in the analyses, and provides qualification of methods in support of the GE14i geometry and Page 4 of 12

ATTACHMENT 1 Evaluation of Proposed Changes characteristics. In particular, the unique characteristics of GE14i that the methods must address are the impacts of non-power producing cobalt isotope rods and the impacts of the connector sections of the cobalt isotope rods . For example, the core and fuel modeling for CPS with the steady-state nuclear methods is consistent with prior application . The connector zones of the ITAs will not be directly modeled in the design, licensing, or core monitoring . Rather than modeling the connector sections it was decided that an additional margin adder be used . The choice to apply an additional adder instead of generation of an explicit model is consistent with current practices for other characteristics not specific to GE14i. The discussion in Section 3.2 .1 .3 of Attachment 3 describes the evaluation method of this assumption resulting in additional design requirements for power peaking control of neighboring rods . Methods changes are not required to add additional margin to accommodate characteristics of fuel designs that are not modeled explicitly .

The CPS containment analysis has also been evaluated to determine if there is an impact resulting from the introduction of the GE14i ITAs. As documented in Attachment 3, the key parameters determining containment response do not change for the GE14i ITA. Based on a GOTHIC analysis, EGC has verified that there is sufficient margin to the suppression pool temperature limit to support the introduction of the GE14i assemblies to the CPS core .

The addition of GE 14i ITAs in either the Reactor or spent fuel pool will not affect or alter the environmental qualification radiation dose requirements . Attachment 3 concludes that there is no significant increase in occupational radiation exposure and no significant radiological environmental impacts with the implementation of the GE14i ITAs . The radiation qualification dose for the affected zones is based on a 40 year normal integrated dose plus a post LOCA gamma dose. Since neither the normal nor post LOCA dose rates are affected by the GE 14i ITAs, the qualification dose for these areas does not change.

The thermal-mechanical design of the GE14i fuel assemblies has been completed and is documented in Attachment 3. This includes isotope rod failure mechanisms, design limits, drawings, and specifications . There has been no irradiated prototype testing completed on these fuel assemblies that EGC can take credit for. However, the analysis documented in shows that there will be no significant impact from operation in the CPS core.

The purpose of the introduction of these new fuel assemblies in an operating reactor core is to gain data on the performance of these assemblies under actual operational conditions .

Prior to loading the ITAs, cycle-specific analyses will be performed to establish fuel operating limits for the ITAs . This will ensure that the core loading has been designed such that the ITAs will not be the most limiting fuel assemblies at any time during the operating cycles, based on planned control rod patterns . Documentation of the results of the cycle-specific licensing analyses will be provided in the Supplemental Reload Licensing Report .

EGC commits to verify that all required licensing analyses as defined in the CPS Updated Safety Analysis Report (USAR) and NEDE-24011-P-A, "General Electric Standard Application for Reactor Fuel," latest approved revision, are completed prior to loading the ITAs in the CPS core in C1 R12 as stated in Attachment 5 .

Once the ITAs are introduced to the CPS reactor core they will remain in the core for a predetermined number of cycles . At subsequent refueling outages, a number of inspections will take place. These post-irradiation examinations are identified in Attachment 3. In Page 5 of 12

ATTACHMENT 1 Evaluation of Proposed Changes addition to these ITA inspections, Co-60 target rods will be periodically harvested intact from the ITAs using the fuel prep machine located in the CPS Fuel Building . EGC intends to remove one isotope rod from a GE14i assembly for inspection after one cycle in the core.

This rod will be replaced with a new cobalt target rod and the assembly will be returned to the reactor. After two cycles of operation, a number of GE -14i assemblies will be discharged from the CPS core . The isotope rods will be removed from these assemblies in accordance with the description provided in Attachment 3 and the assemblies will be stored in the spent fuel pool . After three cycles of operation the balance of the GE14i assemblies will be discharged and the isotope rods removed from these assemblies will be handled as described in Attachment 3. The intact Co-60 target rods will then be shipped in an approved shipping cask to the GEH Vallecitos facility in California for examination, separation and sale of the Co-60. No separation of Co-60 from the target rods will take place at CPS.

Separation of the Co-60 sources from the target rods will only take place at the GEH facilities under the GEH license .

Attachment 3 describes the target rod removal process including dose rates, curie content, potential impact on the shielding design of the spent fuel pool, the design of the tools used to disassemble the isotope rods, effects of the introduction of the GE14i assemblies in the CPS spent fuel pool, and contingency plans if issues are encountered. As noted above, Attachment 3 concludes that there is no significant increase in occupational radiation exposure and no significant radiological environmental impacts with the implementation of the GE14i ITA's. Therefore, all measures in place to ensure there is no overexposure due to movement of an irradiated assembly in the Inclined Fuel transfer System (IFTS) will prevent overexposure when transferring a GE14i assembly . EGC and GEH do not intend to shuffle isotope rods between GE14i assemblies in subsequent power cycles . When the isotope rods are removed from the GE14i assembly, they are disassembled, placed in the shipping cask and transported to the GEH Vallecitos facility in California . The discharged isotope rods are not reinserted into the core as part of the same assembly or a different assembly .

The impact of the GE14i assembly on the results of the fuel handling accident has been evaluated in addition to the other design basis accidents . The results of the evaluation are documented in Attachment 3. Particulate entrainment in fuel pool coolant flow as a result of a fuel handling accident is not analyzed as part of the accident analysis . Prototype failure testing has shown that failure occurs at a benign point on the male threading of the connectors, not the cladding or welds. From this testing and because the GE14i segmented rod design utilizes double encapsulation of the cobalt targets, the targets are even less likely to reach the fuel pool cleanup system than pieces of fuel pellets.

If cobalt targets were released into the fuel pool, they are of similar size, shape and material properties to other metals in the fuel pool and therefore, have similar likelihood to be taken into the cooling system. Any target released would naturally fall to the bottom of the pool because their density is much greater than the density of water . The CPS spent fuel pool cooling system takes suction well above the top of the spent fuel bundles and does not have system flows great enough to remove metals from the floor of the spent fuel pool .

Therefore, it is highly unlikely that targets can be swept from the bottom of the spent fuel pool into the cooling system.

Page 6 of 12

ATTACHMENT 1 Evaluation of Proposed Changes CPS TS 4.2.1 provides a description of the fuel assemblies used at CPS. The TS states that each "assembly shall consist of a matrix of Zircaloy or ZIRLO clad fuel rods with an initial composition of natural or slightly enriched uranium dioxide (U02) as fuel material, and water rod(s)," and therefore, does not account for fuel assemblies containing other materials. As a result, EGC is proposing a change to TS 4.2.1 to provide additional detail on the GE-14i fuel assemblies to be introduced to the CPS core beginning in Cycle 13. This proposed change does not impact the description of the existing fuel assemblies or any assemblies other than the proposed ITAs .

The scope of the isotope production program is described in Attachment 3. This description includes the number of isotope rods in each of the eight GE14i isotope test assemblies as well as the location of these assemblies in the core . The criteria used for determining the locations of the isotope rods in the lattice design are provided in Attachment 3. All required measurements and examinations to be performed on the irradiated GE14i assemblies and isotope rods are also described in Attachment 3.

Activities requiring a byproduct material license are covered under the requirements specified in 10 CFR Part 30, "Rules of General Applicability to Domestic Licensing of Byproduct Material," section 30.3, "Activities requiring license ." This section states that except as provided for in 10 CFR 30.3, " . . .no person shall manufacture, produce, transfer, receive, acquire, own, possess, or use byproduct material except as authorized in a specific or general license issued in accordance with the regulations in this chapter." EGC has reviewed the requirements for a specific byproduct material license as defined in 10 CFR Parts 32, 33, 34, 35, 36, 39, and 40 to determine which of these requirements are applicable to the generation of Co-60 in the operating CPS reactor. It has been determined that none of these requirements are applicable to this situation . In addition, it has been determined that a general byproduct license, as defined in 10 CFR Part 31, is not required since Co-60 is the only byproduct material produced as part of this ITA program and it will be transferred to GEH for separation and handling. Therefore, it has been determined that a Part 30 byproduct material license will not be required to proceed with the ITA program described above.

EGC is proposing to revise several operating license conditions to provide clarification as to how the proposed ITA program continues to meet the requirements of 10 CFR Parts 30 and

50. This clarification is provided by acknowledging that CPS will be producing and transferring byproduct material (i .e., Co-60) in accordance with Parts 30 and 40. As described above, a Part 30 license is not required to support the introduction of the GE14i assembly at CPS. Generation of Co-60 in the GE14i assembly is not outside the authorizations of the current CPS Facility Operating License . EGC does not intend to handle or use the Co-60 material produced in the GE14i assembly in a manner that is not authorized in accordance with 10 CFR 30 and the Facility Operating License . In addition, the license condition indicating that EGC is licensed to possess but not separate byproduct material will be revised to clarify that the removal of the intact Co-60 target rods by disassembly of the fuel assembly does not constitute separation of the byproduct material as defined in 10 CFR Part 30. Finally, EGC is also proposing to add a new license condition that will acknowledge the intentional production of byproduct material in accordance with Part 30.

Page 7 of 12

ATTACHMENT 1 Evaluation of Proposed Changes Attachment 3 provides a discussion for the source tracking considerations associated with the Co-60 produced in the CPS ITAs. Details of how the individual rods are marked and tracked before and after removal from the reactor are provided in Attachment 3 as well . The target rods will be identified with unique tracking numbers to aid in assembly placement and post-processing core location identification . In accordance with the NRC requirements, as documented in the final rule for 10 CFR Parts 20 and 32, source tracking will begin after the irradiated cobalt target pellets are removed from the isotope rods and encapsulated into sealed sources for final product use.

In summary, EGC plans to use a new fuel design (i.e., GE14i ITA) at CPS that will result in the generation of byproduct material in the core during power operation. Use of these assemblies will not affect the operation of the plant as demonstrated in the safety analysis report provided in Attachment 3. EGC has determined that the proposed ITA program continues to comply with the requirements of 10 CFR Parts 30 and 40. EGC is proposing to revise the applicable license conditions, however, to provide clarification as to how CPS continues to meet the requirements for possessing and handling byproduct material .

4.0 REGULATORY ANALYSIS

4.1 Applicable Regulatory Requirements/Criteria The proposed changes have been evaluated to determine whether applicable regulations and requirements continue to be met. As described above, EGC has verified that CPS will continue to meet the requirements of 10 CFR 30 and 10 CFR 40 while implementing the ITA program .

10 CFR 50 .36, "Technical specifications," requires that the facility's TS will include a section addressing design features . In accordance with 10 CFR 50.36(d)(4), the design features to be included in this section 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" and are not addressed in other sections of the TS. While the proposed change adds clarification as to the ITA materials of construction, the proposed change has no significant effect on safety . As a result, this change is only editorial. Based on the above, the only required change to TS is the proposed clarification to TS 4.2.1 as a result of the intentional production of Co-60 and implementation of the ITA program. CPS will continue to meet the requirements of 10 CFR 50, "Domestic Licensing of Production and Utilization Facilities ."

EGC has determined that the proposed changes do not require any exemptions or relief from regulatory requirements, other than the facility operating license and technical specifications, and do not affect conformance with any General Design Criteria (GDC) differently than described in the USAR .

4.2 Precedent EGC has identified two instances where a commercial power reactor facility was specifically authorized by the NRC to be operated to produce byproduct material . The Tennessee Page 8 of 12

ATTACHMENT 1 Evaluation of Proposed Changes Valley Authority's (TVA) Watts Barr Nuclear Plant, Unit 1 and Sequoyah Nuclear Plant, Units 1 and 2 were authorized to produce tritium as documented in References 1 and 2, respectively . There was no need for an operating license change identified by TVA and the NRC.

4.3 No Significant Hazards Consideration In accordance with 10 CFR 50.90, "Application for amendment of license, construction permit, or early site permit," Exelon Generation Company, LLC (EGC) requests an amendment to Facility Operating License No . NPF-62 for Clinton Power Station (CPS), Unit 1 . Specifically, the proposed change modifies CPS License Condition 2.B.(6) and creates new License Conditions 1 .J and 2.B.(7) as part of a pilot program to irradiate Cobalt (Co)-59 targets to produce Co-60. The Co-60 would ultimately be sold to licensed users in the medical industry for use in cancer treatments, blood and instrument sterilization, radiography and security industry for imaging, and to the food industry for cold pasteurization or irradiation sterilization . In addition to the proposed license condition changes, EGC also requests an amendment to Appendix A, Technical Specifications (TS),

of the CPS Facility Operating License. This proposed change would modify TS 4.2.1, "Fuel Assemblies," to describe the Isotope Test Assemblies (ITAs) being used .

EGC has evaluated whether or not a significant hazards consideration is involved with the proposed amendment by focusing on the three standards set forth in 10 CFR 50.92, "Issuance of amendment," as discussed below :

1 . Does the proposed amendment involve a significant increase in the probability or consequences of an accident previously evaluated?

Response : No.

The proposed changes to the license conditions provide clarification and do not impact plant operation in any way. The handling of byproduct material (i.e., Co-60) will continue to be done in accordance with the requirements of 10 CFR 30 and the requirements of the CPS Facility Operating License. The proposed change to TS 4.2 .1 also provides clarification and additional description of the proposed ITAs to be used in the CPS core . These changes provide clarification and do not involve an increase in the probability or consequences of an accident previously evaluated .

The use of the GE14i ITAs, has been evaluated for impact on the previously evaluated transients and design basis accidents for CPS. GE-Hitachi report NEDC-33505P, "Safety Analysis Report to Support Introduction of GE14i Isotope Test Assemblies (ITAs) in Clinton Power Station," dated June 2009, documents the results of the analyses completed to demonstrate the impact on operation following introduction of the ITAs in the CPS core. The use of these ITAs does not adversely affect accident initiators or precursors nor alter the design assumptions, conditions, and configuration or the manner in which the plant is operated and maintained . The Cycle 13 (i.e., the first cycle of operation with the GE14i assembly) core, and subsequent cores, will be designed so that the ITAs will be placed in non-limiting locations with respect to thermal limit margins and shutdown margins. The ITAs do Page 9 of 12

ATTACHMENT 1 Evaluation of Proposed Changes not adversely affect the ability of any structures, systems or components (SSCs) to perform their intended safety function to mitigate the consequences of an initiating event within the assumed acceptance limits .

In addition to evaluation of the impact to operation with the introduction of the GE14i assemblies, EGC has also evaluated the effects of these assemblies on post-irradiation conditions. The effects on the spent fuel pool are minimal; post-irradiation handling of the assemblies and the isotope rods will be performed under approved procedures, by experienced personnel . Handling of the licensed transfer casks will be in accordance with the guidance in NUREG 0612, "Control of Heavy Loads at Nuclear Power Plants," using the Fuel Building Crane. These precautions will support safe movement of the casks within the Fuel Building .

The consequences of a previously analyzed event are dependent on the initial conditions assumed in the analysis, the availability and successful functioning of equipment assumed to operate in response to the analyzed event, and the setpoints at which these actions are initiated. The consequences of a previously evaluated accident are not significantly increased by the proposed change . As documented in NEDC-33505P, the proposed change does not affect the performance of any equipment credited to mitigate the radiological consequences of an accident .

Evaluation of operation with the GE14i assemblies in the CPS core, demonstrated that the licensing basis radiological analyses are not impacted by the introduction of eight GE14i assemblies at CPS. This includes the analyses done for transients and design basis accident events .

Therefore, the proposed change does not involve a significant increase in the probability or consequences of an accident previously evaluated .

2. Does the proposed amendment create the possibility of a new or different kind of accident from any accident previously evaluated?

Response : No.

The proposed revision to the CPS license conditions and TS 4.2.1 will not introduce any new or modified equipment since these changes are intended to provide clarification only. These clarifications will not result in operation of the facility in a different way than currently operated .

While the proposed ITA program does result in the introduction of several modified fuel assemblies (i .e., the GE14i assembly), these assemblies are essentially the same as the GE14 assemblies currently in use in the CPS core. The only difference being the use of a number of isotope rods in place of fuel rods . The GE14i assembly was designed for mechanical, nuclear, and thermal-hydraulic compatibility with the GE14 fuel design . The details of the design differences between the GE14 and GE14i are documented in NEDC-33505P. Use of the proposed ITAs does not involve the addition or modification of any plant equipment other than the assemblies modified to include the cobalt target rods . Also, use of the proposed ITAs will not alter the design configuration or method of operation of plant equipment beyond its Page 1 0 of 12

ATTACHMENT 1 Evaluation of Proposed Changes normal functional capabilities . The ITA program does not create any new credible failure mechanisms, malfunctions or accident initiators .

Therefore, the proposed change does not create the possibility of a new or different kind of accident from any previously evaluated .

3. Does the proposed amendment involve a significant reduction in a margin of safety?

Response : No.

The proposed change to the CPS operating license conditions are intended to provide clarification as to how the generation of byproduct material in the CPS reactor core meets the requirements of 10 CFR Part 30. The proposed change to TS 4.2.1 also provides clarification and additional description of the proposed ITAs to be used in the CPS core . These proposed changes would not affect the design or operation of any equipment important to safety . In addition, since the proposed changes to the license conditions and TS provide clarification only, these changes do not affect the results of any safety calculations .

Cycle specific analyses will be performed for CPS Reload 12 Cycle 13 to establish fuel operating limits for the ITAs that assure compliance with regulatory limits .

Results of these analyses will be documented in the CPS Reload 12 Cycle 13 Supplemental Reload Licensing Report. Furthermore, licensing analyses will be performed for the ITAs for each cycle of their operation, wherein the effect of the ITAs is considered for each of the appropriate licensing events and anticipated operational occurrences (AOOs) to establish the appropriate reactor thermal limits for operation .

The proposed introduction of the ITAs has no impact on equipment design or fundamental operation, other than the modifications made to the fuel assembly as part of the program. There are no changes being made to safety limits or safety system allowable values that would adversely affect plant safety as a result of the proposed ITAs . The performance of the systems important to safety is not significantly affected by the use of the proposed ITAs. The margin of safety can be affected by the thermal limits existing at the time of the postulated accident; however, the ITA design has been evaluated and demonstrated to have no significant effect on the calculated thermal limits as described above. The proposed change does not affect safety analysis assumptions or initial conditions and therefore, the margin of safety in the original safety analyses is maintained .

As documented above, the proposed change does not involve a significant reduction in a margin of safety.

Based on the above, EGC concludes that the proposed amendment does not involve a 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.

Page 1 1 of 12

ATTACHMENT 1 Evaluation of Proposed Changes 4.4 Conclusions 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 .

5.0 ENVIRONMENTAL CONSIDERATION

A review 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, or would change an inspection or surveillance requirement.

However, the proposed amendment does not involve (i) a significant hazards consideration, (ii) a significant change in the types or a 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(c)(9). Therefore, pursuant to 10 CFR 51 .22(b), no environmental impact statement or environmental assessment need be prepared in connection with the proposed amendment.

6 .0 REFERENCES 1 . Letter from Mr. R. E. Martin (NRC) to Mr. O. D. Kingsley, Jr . (TVA), "Issuance of Amendment on Tritium Producing Burnable Absorber Rod Lead Test Assemblies (TAC No. M98615)," dated September 15, 1997

2. Letter from Mr. R . W. Hernan (NRC) to Mr. J. A. Scalice (TVA), "Issuance of Amendments Regarding Technical Specification Change No. 00-06 (TAC No.

MB2972 and MB2973)," dated September 30, 2002 Page 1 2 of 12

ATTACHMENT 2 Mark-up of Proposed Operating License and Technical Specification Pages

G. The issuance of this license will not be inimical to the common defense and security or to the health and safety of the public; H. After weighing the environmental, economic, technical, and other benefits of the facility against environmental and other costs and considering available alternatives, the issuance of Facility Operating License No . NPF-62, subject to the conditions for protection of the environment set forth in the Environmental Protection Plan attached as Appendix B, is in accordance with 10 CFR Part 51 of the Commission's regulations and all applicable requirements have been satisfied; and The receipt, possession, and use of source, byproduct, and special nuclear material as authorized by this license will be in accordance with the Commission's regulations in 10 CFR Parts 30, 40, and 70 .

Based on the foregoing findings regarding this facility, and pursuant to approval by the Nuclear Regulatory Commission at a meeting on April 10, 1987, Facility Operating License No . NPF-62, which supersedes the license for fuel loading and low power testing, License No . NPF-55, issued on September 29, 1986, is hereby issued to Exelon Generation Company to read as follows :

A. This license applies to the Clinton Power Station, Unit No . 1, a boiling water nuclear reactor and associated equipment (the facility), owned by Exelon Generation Company. The facility is located in Harp Township, DeWitt County, approximately six miles east of the city of Clinton in east-central Illinois and is described in the licensee's Final Safety Analysis Report, as supplemented and amended, and in the licensee's Environmental Report-Operating License Stage, as supplemented and amended .

B. Subject to the condition and requirements incorporated herein, the Commission hereby licenses :

Exelon Generation Company, pursuant to section 103 of the Act and 10 CFR Part 50, to possess, use and operate the facility at the designated location in Harp Township, DeWitt County, Illinois, in accordance with the procedures and limitations set forth in this license; (2) Deleted (3) Exelon Generation Company, pursuant to the Act and 10 CFR Part 70, to receive, possess and to use at any time special nuclear material as reactor fuel, in accordance with the limitations for storage and amounts required for reactor operation, as described in the Final Safety Analysis Report, as supplemented and amended; Amendment No. 183

-3 (4) Exelon Generation Company, pursuant to the Act and to 10 CFR Parts 30, 40, and 70, to receive, possess, and use at any time any byproduct, source and special nuclear material as sealed neutron sources for reactor startup, sealed sources for reactor instrumentation and radiation monitoring equipment calibration, and as fission detectors in amounts as required ;

(5) Exelon Generation Company, pursuant to the Act and 10 CFR Parts 30, 40, and 70, to receive, possess, and use in amounts as required any byproduct, source or special nuclear material without restriction to chemical or physical form, for sample analysis or instrument calibration or associated with radioactive apparatus or components ; and (6) Exelon Generation Company, pursuant to the Act and 10 CFR Parts 30, 40, and 70, to possess, but not separate, such byproduct and special nuclear materials as may be produced by the operation of the facility . -o-e-C. This license shall be deemed to contain and is subject to the conditions specified in the Commission's regulations set forth in 10 CFR Chapter I and is subject to all applicable provisions of the Act and to the rules, regulations and orders of the Commission now or hereafter in effect ; and is subject to the additional conditions specified or incorporated below:

(1) Maximum Power Level Exelon Generation Company is authorized to operate the facility at reactor core power levels not in excess of 3473 megawatts thermal (100 percent rated power) in accordance with the conditions specified herein .

(2) Technical Specifications and Environmental Protection Plan The Technical Specifications contained in Appendix A and the Environmental Protection Plan contained in Appendix B, as revised through Amendment No. 186, are hereby incorporated into this license.

Exelon Generation Company shall operate the facility in accordance with the Technical Specifications and the Environmental Protection Plan .

Amendment No. 186

Design Features 4 .0 4 .0 DESIGN FEATURES 4 .1 Site Location The site for the Clinton Power Station is located in Harp Township, DeWitt County, approximately six miles east of the city of Clinton in east-central Illinois . The exclusion area boundary shall have a radius of 975 meters from the Standby Gas Treatment System vent .

4 .2 Reactor Core 4 .2 .1 Fuel Assemblies The reactor shall contain 624 fuel assemblies . Each assembly shall consist of a matrix of Zircaloy or ZIRLO clad fuel rods with an initial composition of natural or slightly enriched uranium dioxide (U02 ) as fuel material, and water rod (s) . Limited substitutions of zirconium alloy 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 applicable NRC staff approved codes and methods and 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 .

3-4 .2 .2 Control Rod Assemblies The reactor core shall contain 145 cruciform shaped control rod assemblies . The control material shall be boron carbide or hafnium metal, or both .

(continued)

CLINTON 4 .0-1 Amendment No . -~s, 101

Insert #1 :

J. The receipt, production, possession, transfer, and use of Cobalt-60 as authorized by this license will be in accordance with the Commission's regulations in 10 CFR Part 30.

Insert #2:

Mechanical disassembly of the GEl4i isotope test assemblies containing Cobalt-60 is not considered separation .

Insert #3 :

(7) Exelon Generation Company, pursuant to the Act and 10 CFR Part 30, to intentionally produce, possess, receive, transfer, and use Cobalt-60 .

Insert #4:

"A maximum of eight GE14i isotope test assemblies will be placed in non-limiting core regions, beginning with the Reload 12 Cycle 13 core reload, with the purpose of obtaining surveillance data to verify that the GE14i assemblies perform satisfactorily in service prior to use of these design features on a production basis.

Each GE14i assembly contains a small number of zircaloy-2 clad isotope rods that contain Cobalt-59 targets. These Cobalt-59 targets will transition into Cobalt-60 isotope targets during the cycle irradiation of the assemblies. Details of the GE14i assemblies are contained in NEDC-33505P, "Safety Analysis Report to Support Introduction of GE14i Isotope Test Assemblies (ITAs) in Clinton Power Station,"

Revision 0, dated June 2009."

ATTACHMENT 4 Safety Analysis Report to Support Introduction of GE14i Isotope Test Assemblies (ITAs) in Clinton Power Station (Non-Proprietary)

HITACHI GE Hitachi Nuclear Energy NEDO-33505 Revision 0 Class I DRF 0000-0100-5453 June 2009 Non-Proprietary Information Safety Analysis Report to Support Introduction of GE14i Isotope Test Assemblies (ITAs) in Clinton Power Station Copyright 2009 GE-Hitachi Nuclear Energy Americas LLC All Rights Reserved

NEDO-33505 Revision 0 Non-Proprietary Information NON-PROPRIETARY NOTICE This is a non-proprietary version of the document NEDC-33505P, Revision 0, from which the proprietary information has been removed. Portions of the document that have been removed are identified by white space within double square brackets, as shown here (( )).

IMPORTANT NOTICE REGARDING CONTENTS OF THIS REPORT Please Read Carefully The only undertakings of GEH with respect to information in this document are contained in the contract between Exelon and GEH; nothing contained in this document shall be construed as changing the contract. The use of this information by anyone other than Exelon or for any purpose other than that for which it is intended is not authorized ; and with respect to any unauthorized use, GEH makes no representation or warranty, express or implied, and assumes no liability as to the completeness, accuracy, or usefulness of the information contained in this document, or that its use may not infringe privately owned rights .

Copyright 2009, GE-Hitachi Nuclear Energy Americas LLC, All Rights Reserved

NEDO-33505 Revision 0 Non-Proprietary Information Table of Contents Page

1. Introduction . . . .... .. . . . ..... . . . . . .... . . . . .... . . . . . ..... . . . . . .... . . . . .... . . . . . ... .. . . . ...... . . .... . . . . . ... . . . . . ...... . . . . ...1

2. GE 14i Fuel Product Description .. . . . . . ... . . . . . .... .. . . ...... . . . .. ... . . . ...... . . . ..... . . . ..... . . . ....... . . . . ....2 2.1 New Design Features . . . . ...... . . . . .... . . . . . ..... . . . ...... . . ...... . . . . . ... . . . ...... . . ...... . . . . .. . . . . . ...... . . . ..... . . .2

3. Nuclear Design and Methods .. . . . . . . ..... . . . ...... . . . . .... . . . . . ... . . . . . ..... . . . ..... . . . . ... . . . . . ...... . . . ..... .18 3 .1 Nuclear Core Design . . ..... . . . ...... . . . ..... . . . . . .... . . . . ...... . . . ... . . . . . ...... . . ...... . . . ... .. . . . ...... . . ...... . . .18 3 .2 Methods . . . .. ... . . . . ... .. . . . ...... . . ...... . . . . ..... . . . . .... . . . . . . . ... . . . ..... . . . . ..... . . ...... . . . ..... . . . . . .... . . . . ...... . . .18 3.2.1 Nuclear Methods ... . . . . ..... . . . . ..... . . . ...... . . .. .... . . . . . ... . . . . . .... . . . . ..... . . . . ... . . . . ..... . . . . .... .. . . . . .19 3 .2 .2 Thermal-Hydraulic Methodology .. . . . . .... . . . . . ... . . . . . ...... . . .... .. . . . ..... . . . ...... . . . ..... .. . . . ..22 3 .2 .3 Safety Limit Methodology .. . . . . . .... . . . . ...... . . . ..... . . . ...... . . ...... . . . ..... . . . ...... . . . ..... . . . . . ... .22 3 .2.4 Transient Analysis Methodology . . . ...... . . . ..... . . . ...... . . . . .... . . . . . ... . . . ...... . . . . .... . . . . . ..... .22 3 .2.5 Stability Methodology. . ..... . . . ...... . . ...... . . . . . ... . . . . . .... . . . . .... . . . . . ... . . . . . .... . . . . ...... . . . .. ... . . .22 3 .2.6 Fuel Rod Thermal-Mechanical Methodology .... . . . ...... . . . ..... . . . .... . . . . ...... . . . . . ... . . . . . 23 3.2.7 ECCS-SAFER/GESTR . . . . . . .... . . . . ..... . . . . ..... . . . ...... . . ...... . . . ..... . . . ...... . . ...... . . . . .... . . . . . ..23 3 .3 GEXL+ Correlation ...... . . . ..... . . . ...... . . ...... . . . ..... . . . . . .... . . . . .... . . . . . ... . . . . . .... . . . . .... . . . . . ... . . . . . . . ..23

4. Licensing Evaluations . . . ..... . . . ...... . . . . .... . . . .. ... . . . . . .... . . . . ..... . . . . ... . . . . . ...... . . . . ... . . . . ..... . . . . . ....37 4.1 Evaluation of Abnormal Operational Transients .. . . ..... . . . . ..... . . . . . .... . . ...... . . . . . .... . . . . .. ... . 37 4.1 .1 Decrease in Reactor Coolant Temperature . . . .... . . . . . ... . . . . . ...... . . ...... . . . . . .... 38 4.1 .2 Increase in Reactor Pressure .... . . . ..... . . . ...... . . ...... . . . . . ... . . . . . .... . . . . .... . . . . .... . . . . . ...... . . . .39 4 .1 .3 Decrease in Reactor Coolant System Flow Rate .. ... . . . . ..... . . . . ... ... . . . ..... . . . ...... . . . . .. 39 4.1 .4 Reactivity and Power Distribution Anomalies . ..... . . . .... .. . . . . .... . . . ..... . . . . . ...... . . . . .... 40 4.1 .5 Increase in Reactor Coolant Inventory. . ...... . . . ..... . . . . ..... . . . . ...... . . . . ,. . . . ., . . . . 40 4.1 .6 Decrease in Reactor Coolant Inventory and Other Accidents .... . . . . ...... . . . . ...... . .41 4.2 Evaluation of Other Transients . . . ..... . . . ..... . . . ...... . . . . . ... . . . . . .... . .  .... .. . . . . . . . _.  41 4.2.1 Anticipated Transients Without Scram (ATWS) .... . . . .... . . . . . .... . . . . ... . . . ...... . . . . . ...41 4 .2.2 ASME Overpressure Protection .... . . ...... . . . . . ... . . . ...... . . ...... . . . . . ... . . . .... . .42 4.2.3 Single Loop Operation Pump Seizure Analysis ... . . . . . ... . . . . . ... . . . . . .... . . . . .... .. . . . . . ... . . 42 4.2.4 Applicability of Off-Rated Limits to GE14i ITAs . .... . . . . . ... .. . . . .... . . . . ...... . . . ....... . . 42 4.2 .5 Flexibility and Equipment Out-of-Service (EOOS) Options ..... . . . . ..... . . . . . ... . . . . . .43 4 .3 Evaluation of Design Basis Accidents.. . . . . . .... . . . . .... . . . . ...... . . . ... . . . . . ......  .. ._ . ,. .43 4.3.1 Control Rod Drop Accident .. . . ...... . . . ..... . . . ... ... . . ...... . . . .... . . . . ...... . . ...... . . . . .... . . . . . .....43 4.3.2 Main Steam Line Break Accident . . ... .. . . . . . .... . . ...... . . . ..... . . . .... . . . . . .  ,_, .  ..44

NEDO-33505 Revision 0 Non-Proprietary Information 4.3 .3 Fuel Handling Accident . . . . . . .... .. . . .... .. . . . ... .. . . . ... . . . . . ..... . . . . . . ... .. . . . . . .... . . . . . ... . . . . . ..... . .44 4.3 .4 Loss-of-Coolant Accident (LOCA). . . . ... . . . . . .... . . . . .... . . . . . .. ... . . . . . .. .. . .  ..,_  . .44 4.4 Thermal-Mechanical Evaluation .. . . ... . . . . . . ... . . . . . ...... . . ...... . . . ..... . . . . . ...... . . . ..... . . . . . ..... . . . ...45 4.5 Other Evaluations .... . . . . . ... . . . . . ...... . . ...... . . . ..... . . . ... ... . . ...... . . . ..... . . . . . ...... . . . . .... . . . . . . . ... . . . . . ...46 4.5.1 Stability . . . . ...... . . . . . ... . . . . . ...... . . ...... . . . . .... . . . . ..... . . ...... . . . ..... . . . . . ...... . . . . .... . . . . . . . ... .. . . . .... .47 4 .5 2 , r ,. ,. ,.. . ~. .

.5 .3

.4 Station Blackout ... . . . ...... . . . . .... . . . . . .... . . . . ..... . . . . . .. . . . . . ... . . . . . ...... . . . . .... .. . . . ..... . . . . . .... . . ...48 4 .5 .4 Reactor Internal Pressure Difference ... . . . . ..... . . . . ... . . . . . ...... . . . . .... . .  ,. . . .  ,.  ,._48 4.5.5 Reactor Internals Structural Evaluation . ..... . . . . ... . . . . . .... .. . . .. ...... . . . . . . . . .  ,_ ,..49 4.5.6 Recirculation System Evaluation . . ..... . . ...... . . . .... . . . . ...... . . .... , . . . . ... , .  ... . . .50 4 .5.7 Seismic and Dynamic Response . . .... . . ..... . . . . ..... . . . .... . . . . .. .... . . . . . .... . . . . .. .... . . . . .... . . . . .50 4.5.8 Decay Heat Assessment . . . . . ... . . . . . .... . . . . .... . . . . . ... . . . ...... . . . ..... . . . . . ..... . . . ...... . . . . ..... . . . . . .50 4.5 .9 Neutron Fluence Impact . . . ... .. . . . ...... . . .... . . . . . ... . . . . .. ... . . ...... . . . . . ..... . . . . . .... . . . . ...... . . . ... .51 4.5 .10 Containment Response . . . ... . . . . . .... . . . . .... . . . . . ... . . . ...... . . ...... . . . . . ..... . . . . . . ... . . . . ...... . . . .... . .52 4 .5.11 ECCS LOCA ... . . . . .... .. . . . ..... . . . ...... . . . ... .. . . . .... . . . . .... . . . . .... . . . . . ... . . . . . . . .... . . . . .... .. . . . ..... . . .52 4 .5.12 Hydrogen Injection.. . . . ..... . . . . . .... . . . . .... . . . .. ... . . . ..... . . . .... . . . .. ... .. . . . .. .... . . . . ..... . . . . .... . . . . ..53 4.5.13 Post-LOCA Hydrogen Control . . ..... . . . .... . . . .. .... . . ... . . . . . . ..... . . . .. .... . . . . .... .. . . . . .... . . . ....53 4.5 .14 Environmental Dose Considerations . ... . . . . ..... . . .... .. . . . ... . . . . . .. ....  ,.. ,..,. ...54 4 .5 .15 Fuel Storage Criticality Safety ..... . . . ..... . . . . ... . . ...... . . . ..... . . . ...... . . . . ...... . . . ..... . . . . ..... . .55 4.5.16 Fresh Fuel Shipping . . . ...... . . . ... .... . . . ..... . . . .... . . ...... . . . ... . . . . . ...... . . . . ..... . . . . .... . . . . ..... . . . . .55 4.5.17 Fuel Channel Distortion .. . . . ...... . . . ..... . . . .... . . . . .... . . . .. ... . . . ...... . . .. .... . . . . . ..... . . . . . .... . . .. ..56 4.5.18 Fuel Conditioning Guidelines . . . ..... . . . ...... . . .... . . . ..... . . . .... .. . . ...... . . . . . ... . . . . . ...... . . .. ....56 4.6 Manufacturing Quality Assurance . . . .... . . . ...... . . .... . . . ..... . . . ...... . . ...... . . . . . ... . . . . . ...... . . .... . . . .56 4 .7 Post-Operational Evaluations .... . . . . . ... . . . .. . ... . . .... .. . . . ... . . . .... . . . . ...... . . . .. ... . . . . ...... . . . ..... . . . . .57 4.7.1 Spent Fuel Pool Effects .... . . . . . ... . . . . . .... . . ...... . . . ... . . . ..... . . . .. .... . . . . . ... . . . . . ...... . . ..... . . . . ...57 4.7.2 Post-Irradiation Handling . . . . ... . . . . . .... . . . ..... . . . ... . . . .... .. . . .... .. . . . ..... . . . . ..... . . . . .. ... . . . . ... . .58 4.7.3 Post-Irradiation Examination . . ...... . . ..... . . . . ... .. . . ..... . . ...... . . . ..... . . . . . .... . . .. .... . . . . . ... . . . .63

5. Conclusion . . .... . . . . . .... . . . . ...... . . . . . ... . . . . . .... . . . . .... . . . . .... . . . .... . . .... . . . . . ..... . . . . . . ... . . . ..... . . . . . .... . . . . .69

6. References .... . . . . . .... . . . . ...... . . . . . ... .. . . . .... .. . . .... . . . . .... . . . .... . . ...... . . . ..... . . . .. .... . . . ..... .. . . . ... . . . . . ...70

NEDO-33505 Revision 0 Non-Proprietary Information List of Tables Table Title Page Table 2-1 GE14i and GE14 Fuel Assembly Dimensions . . . ... ... . . . .... . . . . .... . . . . .... . . . . . ... .. . . . . . .... . . . . . 6 Table 3-1 Summary of GNF Methods Applicability to GE14i . ... . . . . . .... . . . ..... . . . . . ..... . . . ...... . . .  26 Table 3-2 Lattice RMS Fission Density Uncertainty Comparison ...... . . ...... . . . ..... . . . . . .... .. . . .. ... 27 Table 3-3 GE14i Control Blade Worth Comparison . . .. .... . . . . .... . . . ..... . . . ...... . . . ... . . . . . .. .... . . . . ...... . 27 Table 3-4 Internal PANACI I Parameters ... . . . . . .... . . . . ...... . . . ..... . . . ...... ,. . . , ...,_. , . . 27 Table 3-5 GEXL14 Statistics for GE14 CP Data with Zero-Power Rods ... . . . .. .... . . . . ...... . . . . . .. 28 Table 4-1 List of Analyzed Events for the Reload License with GE14i ITAs in the Core .. . 66 Table 4-2 Total Core Energy Comparison of GE14 and GE14i ..... . . . ... .. . . . ..... . . . ...... . . . . . .. , . 66 Table 4-3 Basket Dose Rate Values .. . . . . . .... . . . . ..... . . . . ..... . . . ...... . . ... . .. . . . ... . . . . . .... . . . . ...... . . . . . ... . . . . . .. 67 Table 4-4 Single Rod Dose Rate Values .. . . .. .... . . . . . ... . . . . . .... . . . . ....  , ... ..._ __ , . . . 68

NEDO-33505 Revision 0 Non-Proprietary Information List of Figures Figure Title Page Figure 2-1 GE14i Bundle Cutaway View . . .... . . . . .... . . . . .... . . . . . .. ... . . . . . .... . . . . .... . . . . . ..... . . . . . . . .... . . . . .... . . . 7 Figure 2-2 GE 14i Lattice Arrangement . . . ..... . . . .... . . . . .... . . . . . ..... . . . . . .... . . . . . . .... . . . . . ... .. . . . . ..... . . . . .... . . . . . 8 Figure 2-3 Cobalt Target Isometric View ... . . . .... . . . . ... . . . . . . ... . . . . . ...... . . . . ...... . . . ..... . . . . . .,__ ._  ,_._, 9 Figure 2-4 Cobalt Target Orthographic View.... . . ..... . . . . ..... . . . ...... . . . . ...... . . . ... .. . . . . . .. .... . . ...... . . . . .... . 9 Figure 2-5 Isometric View.. . . . . .. .... . . ...... . . . . . .. . . . . . . ...... . . . . ...... . . . ..... . 10 Figure 2-6 (( )) Orthographic View.... . . . ..... . . . . . .... . . . . ..... . . . . . .... .. . . . ..... . . . 10 Figure 2-7 (( )~ Cladding Isometric View . . . ...... . . .. .... . . . . . ... . . . . . .... . . . . . ..... . . . . . ... .. . . . .. 11 Figure 2-8 (( )) Cladding Orthographic View . . . . ..... . . . .. ... . . . . . .... . . . . .. ....  . . . ...  , .... 11 Figure 2-9 Isometric View .... . . . ..... . . . ...... . . ...... . . . .. ... . . . . . .... . . . . .... .. . . . . . ... . . . . . .... . . 12 Figure 2-10 Orthographic View.... . . . ..... . . . ...... . . . ..... . . . . ..... . . ...... . . . . . ..... . . . .. .... . . .. 12 Figure 2-11 Isometric View . .... . . . ..... . . . ...... . . ..... . . . . ..... . . . ...... . . ...... . . . . . ... .. . . . ...... . . .... 13 Figure 2-12 Orthographic View ..... . . . .... . . . . .... . . . . . ... . . . . . ..... . . . .... .. . . . ..... . . . . . .... . . . . .... . . 13 Figure 2-13 (( )) Isometric View. . ... . . . . . .... . . . . .... . . . . . ... . . . . . ...... . . . . .... . . . . 14 Figure 2-14 (( )) Orthographic View . .... . . . . .... . . . . .... . . . . . ...... . . ...... . . . .. . 14 Figure 2-15 (( )) Isometric View ...... . . ...... . . . ..... . . . ...... . . . . .... . . . . . .... . 15 Figure 2-16 (( )) Orthographic View ...... . . . ..... . . . .... . . . . ...... . . . . . ... . . . . 15 Figure 2-17 (( ]) Isometric View.... . . . ...... . . .... . . . . . ... .. . . . .... . . . . ...... . . . ..... . . . ... 16 Figure 2-18 (( )) Orthographic View .. . . . . .... . . . . . ... . . . . ..... . . ...... . . . . . .... . . . . .... . 16 Figure 2-19 Isometric View. . ..... . . . ..... . . . ..... . . . ...... . . .... . . . . . .. ... . . . . . .... . . . 17 Figure 2-20 Orthographic View ...... . . . ... . . . . . ..... . . . .... . . . . . ... .. . . . ...... . . ... 17 vi

NEDO-33505 Revision 0 Non-Proprietary Information Figure 3-1 TGBLA06 Cobalt 60 Inventory per Linear cm for GE14i for 00% In-Channel Void History. . . .... .. . . . ... . . . . . ...... . . . . .... . . . . . ..... . . . . . .... .. . . ...... . . . .... . . . . .. .. . . . . .... .. . . . ..... . . . . . .... . 29 Figure 3-2 TGBLA06 Cobalt 60 Inventory per Linear em for GE 14i for 40% In-Channel Void History . .... . . . .. ... . . . . . .... . . . . .... . . . . . ..... . . . . . .... .. . . . . .... . . . . . ... . . . ...... . . .... . . . . . . .. .. . . . . . .... .. . . . 30 Figure 3-3 TGBLA06 Cobalt 60 Inventory per Linear cm for GE 14i for 70% In-Channel Void History ... . . . .... . . . . .... .. . . ...... . . . . . ... . . . . . ...... . . . . .... . . . . . .... . . . . .... . . . . .... . . . . . ... . . . . . ...... . . . . ... 31 Figure 3-4 Power Spike in Face Adjacent Fuel Rods in the Wide-Wide Corner . . . . .... .. . . .. .... . 32 Figure 3-5 Rod-to-Rod Power Distributions with Zero-Power Rods . . . .... . . . .. ... . . . . . .... . . . . ...... . . . 33 Figure 3-6 GEXL14 Test Conditions (P=1000 psia) ..... . . . ..... . . . . . .... . . . . .... . . . ..... . . . . ..... . . . . ...... . . . . . 34 Figure 3-7 Typical Bundle Axial Power Shape Used for GEXL14 Testing . . . . .. .... . . . . ...... . . . . . .. 35 Figure 3-8 Calculated Versus Measured Critical Power for GEXL14 . . ..... . . . ...... . . . . .... . . .  . . ... . 36

NEDO-33505 Revision 0 Non-Proprietary Information ACRONYMS Term Definition ALARA As Low As Reasonably Achievable AOO Anticipated Operational Occurrence APLHGR Average Planar Linear Heat Generation Rate ASME American Society of Mechanical Engineers ATWS Anticipated Transients Without Scram ATWS-RPT Anticipated Transients Without Scram - Recirculation Pump Trip BHL Beginning of Heated Length BSP Backup Stability Protection BT Boiling Transition BWR Boiling Water Reactor CGCS Combustible Gas Control System COINS Combined Instrumentation Measurement System COLR Core Operating Limits Report CPR Critical Power Ratio CPS Clinton Power Station CRDA Control Rod Drop Accident CRGT Control Rod Guide Tube DAS Data Acquisition System DCF Dose Conversion Factor DIVOM Delta CPR over Initial MCPR _Versus the _Oscillation Magnitude EAB Exclusion Area Boundary ECCS Emergency Core Cooling System(s)

ECPR Experimental Critical Power Ratio EOC End-of-Cycle EOL End-of-Life EOOS Equipment Out-of-Service FES Fuel Examination Services FHA Fuel Handling Accident FPCC Fuel Pool Cooling and Cleanup FPM Fuel Prep Machine FWCF Feedwater Controller Failure GE General Electric GEH GE-Hitachi Nuclear Energy Americas, LLC viii

NEDO-33505 Revision 0 Non-Proprietary Information ACRONYMS Term Definition GESTAR General Electric Standard Application for Reload GNF Global Nuclear Fuel - Americas, LLC HAC Hypothetical Accident Conditions HPCS Inadvertent High Pressure Core Spray (HPCS) Pump Start-Up HPGe High Purity Germanium HWC Hydrogen Water Chemistry ITA Isotope Test Assembly LFWH Loss of Feedwater Heating LHGR Linear Heat Generation Rate LOCA Loss of Coolant Accident LOFW Loss of Feedwater Flow LPRM Local Power Range Monitor LPZ Low Population Zone LRNBP Load Rejection (Turbine Control Valve Fast Closure) with Bypass Failure LRWBP Load Rejection (Turbine Control Valve Fast Closure) with Bypass MAPLHGR Maximum Average Planar Linear Heat Generation Rate MCPR Minimum Critical Power Ratio MSIV Main Steam Isolation Valve MSIVF Main Steam Isolation Valve Closure with Flux Scram MSLB Main Steam Line Break NCT Normal Conditions of Transport NMCA Noble Metal Chemical Application NRC Nuclear Regulatory Commission OLMCPR Operating Limit Minimum Critical Power Ratio OPRM Oscillation Power Range Monitor PCT Peak Cladding Temperature PLR Part Length Rod(s)

PRFDS Pressure Regulator Downscale Failure PRFO Pressure Regulator Failure Open RCIC Reactor Core Isolation Cooling RG Regulatory Guide RHR Residual Heat Removal RIPD Reactor Internal Pressure Difference RMS Root Mean Square RV Relief Valve

NEDO-33505 Revision 0 Non-Proprietary Information ACRONYMS Term Definition RWE Rod Withdrawal Error SAR Safety Analysis Report SBO Station Blackout SCC Stress Corrosion Crack SFP Spent Fuel Pool SLCS Standby Liquid Control System SLO Single Loop Operation SRLR Supplemental Reload Licensing Report SRV Safe Relief Valve SV Safety Valve TEDE Total Effective Dose Equivalent TIP Traversing In-Core Probe TPR Target Placement Rod TTNBP Turbine Trip with Bypass Failure TTWBP Turbine Trip with Bypass USAR Updated Safety Analysis Report VNC Vallecitos Nuclear Center

~WSGS Water Submersible Gamma Scanner

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1. Introduction Exelon Generation Company, LLC (Exelon) plans to load eight (8) GE14i Isotope Test Assemblies (ITAs) as part of the Clinton Power Station (CPS) Reload 12 Cycle 13 during the 2010 refueling outage. These GE14i bundles containing (( )) segmented cobalt isotope rods, also referred to as GE 14i ITAs, are planned to be in operation as part of a joint program with Global Nuclear Fuel - Americas, LLC (GNF), GE-Hitachi Nuclear Energy Americas, LLC (GEH), and Exelon Generation Company, LLC.

The GE141 fuel design is described in Section 2 . GE14i is designed to be compatible with other GE fuel designs . The external envelope of the fuel assembly is comparable to the GE14 fuel assembly currently supplied to CPS . The nuclear characteristics of these GE14i ITAs are compatible with those of the current GE14 fuel being loaded into CPS.

Section 3 addresses the nuclear core design and applicability of nuclear and safety analysis methods. Section 4 provides information relative to the evaluations that were performed by GNF and GEH to assist Exelon in their evaluations according to their configuration change process .

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2. GE14i Fuel Product Description The GE14i bundle is identical to the GE14 bundle described in Reference 1 with the exception of the (( )) cobalt isotope rods as illustrated in Figure 2-1 . GE14i consists of

(( )) fuel rods, (( )) cobalt isotope rods, and two large central water rods in a 10 x 10 array. The two water rods encompass eight fuel rod positions . For the GE14i product, the cobalt isotope rods will be limited to the (( )) locations identified in the lattice design shown in Figure 2-2. With this specific bundle design, there will be no ((

)) In addition, rods are located towards the outside of the bundle where enrichment is typically lower relative to internal locations . Consequently, a ((

)) is displaced when a GE14i bundle is used. ((

)) as allowed by fuel and core design constraints .

2.1 New Design Features GE14i was designed for mechanical, nuclear, and thermal-hydraulic compatibility with the GE14 fuel designs . In addition to its similarities with the GE14 design, GE14i includes ((

)) cobalt isotope rods which converts cobalt-59 into cobalt-60 via neutron capture to be used in various medical and food sterilization applications . Table 2-1 shows GE14i and GE14 fuel assembly dimensions. Below is a list of new GE 14i features .

" GE 14i Bundle Schematic

" GE14i Lattice Design

" Cobalt Target

" ((

" (( JJ

" (( JJ A discussion of each of these new GE14i features is provided in the remainder of this section .

Cobalt Target 11

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NEDO-33505 Revision 0 Non-Proprietary Information 11 Defender Lower Tie Plate The GE14i fuel bundle will incorporate the Defender lower tie plate. The Defender lower tie plate maintains the same resistance to foreign material debris as the balance of GE14 fuel assemblies in the reload .

NEDO-33505 Revision 0 Non-Proprietary Information Table 2-1 GE14i and GE14 Fuel Assembly Dimensions Fuel Assembly GE14i GE14 Total number of fuel rods 92 Total number of cobalt isotope rods N/A Number of full length 78 Number of partial length 14 14 N/A N/A N/A N/A N/A Lattice array I Ox 10 10x 10 Rod to rod itch (in) 0.510 0.510 Number of water rods 2 2 a

Typical assembly fuel weight (IbU)

Total fuel assembly d weight (lb) b Total fuel assembly submerged weight (lb) b Typical assembly active fuel full length (in) 150.00 150.00 Typical assembly active fuel partial length (in) 84.00 84 .00 Fuel Rod Cladding material with zirconium inner liner Zr-2 Zr-2 Cladding tube diameter, outer (in)

Cladding tube wall thickness (in)

Pellet diameter, outer (in)

Fuel column stack density ( cm3)

Fuel column stack density with burnable absorber (( )) (( ))

( cm3)

Water Rod Tube material Zirc-2 Zirc -2 Maximum tube diameter, outer (in)

Tube wall thickness (in)

Spacer Number of spacers 8 8 Axial locations See Reference 1 See Reference 1 Material Zr-2 ferrule and bands with Zr-2 ferrule and bands with Inconel X-750 springs Inconel X-750 springs

NEDO-33505 Revision 0 Non-Proprietary Information Figure 2-1 GE14i Bundle Cutaway View

NEDO-33505 Revision 0 Non-Proprietary Information Figure 2-2 GE14i Lattice Arrangement

NEDO-33505 Revision 0 Non-Proprietary Information Figure 2-3 Cobalt Target Isometric View Figure 24 Cobalt Target Orthographic View

NEDO-33505 Revision 0 Non-Proprietary Information Figure 2-5 (( )) Isometric View Figure 2-6 11 )) Orthographic View

NEDO-33505 Revision 0 Non-Proprietary Information Figure 2-7 (( )) Cladding Isometric View

))

Figure 2-8 (( )) Cladding Orthographic View

NEDO-33505 Revision 0 Non-Proprietary Information Figure 2-9 (( )) Isometric View Figure 2- 1 0 (( )) Orthographic View

NEDO-33505 Revision 0 Non-Proprietary Information Figure 2-11 (( )) Isometric View Figure 2-12 (( )) Orthographic View

NEDO-33505 Revision 0 Non-Proprietary Information Figure 2-13 (( )) Isometric View Figure 2-1 4 (( )) Orthographic View

NEDO-33505 Revision 0 Non-Proprietary Information Figure 2-15 (( )) Isometric View Figure 2-1 6 (( )) Orthographic View

NEDO-33505 Revision 0 Non-Proprietary Information Figure 2-17 (( )) Isometric View

))

Figure 2-18 (( )) Orthographic View

NEDO-33505 Revision 0 Non-Proprietary Information Figure 2-19 (( )) Isometric View Figure 2-20 (( )) Orthographic View

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3. Nuclear Design and Methods 3.1 Nuclear Core Design CPS will insert 8 GE14i bundles in this ITA program. The purpose of these bundles will be to confirm ITA performance and provide confidence in overall design, prior to inserting large numbers of GE 14i fuel assemblies . The Cycle 13 core will be designed so that the ITAs will be placed in non-limiting locations with respect to thermal limit margins and shut down margins .

The applicability of the GEXL14 correlation to the GE14i ITAs is demonstrated by comparing the GEXL14 prediction to the critical power data with zero-power rods in the GE14 bundle. It is shown that the GEXL14 correlation conservatively predicts the critical power data with zero-power rods. f [

)) Also, the ITA assemblies will be designed into non-limiting locations with respect to the other fuel assemblies .

3 .2 Methods This section addresses the applicability of the current methods and methodologies to the GE14i fuel design . It also addresses each NRC approved method (References 2, 3, and 4) that is used in the analyses, and provides qualification of methods in support of GE14i geometry and characteristics . In particular, the unique characteristics of GE14i that the methods must address are the impacts of the non-power producing cobalt isotope rods and the impacts of the connector sections of the cobalt isotope rods. Many of the methods are unaffected by either of these characteristics, but a few require explanation as to how they are qualified for this application.

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NEDO-33505 Revision 0 Non-Proprietary Information The differences between GE14i and GE14 fuel products are minimal . All geometry, fuel rod, and water rod characteristics are identical. Differences are limited to the (( )) locations where cobalt isotope rods replace UOZ fuel rods. The cobalt isotope rods are segmented for disassembly at time of discharge . The impact of the segmented rod connector sections is discussed in Section 3 .2.1 .2 .

Table 3-1 shows the summary of the status of the applicability of codes and methodologies to GE14i. A more detailed discussion follows to explain those items for which changes are planned or underway .

3.2.1 Nuclear Methods 3.2.1.1 Lattice Physics TGBLA06 is the two-dimensional transport corrected diffusion theory model used to model the details of nuclear transport at the lattice level. The fundamental methodology for TGBLA06 will not be changed to model GE 14i. Input necessary to describe the GE 14i lattice design is provided through the standard TGBLA06 input interface.

In GE14i, all U and Gd rod material attributes are identical to GE14. No modifications to the methodology of TGBLA06 were required to model the GE14i U and Gd rods. The material characteristics of the cobalt bearing regions are provided through the standard TGBLA06 input parameters .

The qualification of TGBLA06 was performed by comparisons with a Monte Carlo simulation (Reference 5) of the depletion of the (( )) rod GE14i Co-59 isotope target design. The Monte Carlo method used was MCNP-05 (Reference 5) with ENDFB-VI (Revision 8) cross sections. The comparison of the Co-60 inventory as a function of lattice exposure and in-channel void history is shown in Figure 3-1, Figure 3-2, and Figure 3-3 .

Infinite lattice reactivity, pin fission density distributions, pin power distributions, gamma source distributions, and nuclear instrumentation responses as functions of lattice exposure and void history are generated by TGBLA06 for use in downstream applications such as PANAC 11 . Representative uncertainties for the fission density consistent with the methodology described in Reference 19 are provided in Table 3-2 and the control blade worth predictions are provided in Table 3-3 . The current GE 14 uncertainties for use in safety limit analysis will be used to model the GE14i fuel product.

The introduction of GE 14i is accommodated by the TGBLA lattice physics methodology .

NEDO-33505 Revision 0 Non-Proprietary Information 3.2.1.2 Steady-State Core Simulator PANAC 11 is the three-dimensional core simulator utilized for design, licensing, and core monitoring of BWR cores . PANACI I correctly handles varying axial geometry in nuclear and thermal-hydraulic modeling through use of its lattice dependent geometry, nodal thermal-hydraulic properties, and axial meshing routines . This allows PANAC 11 to handle multiple Part Length Rods (PLRs), varying water rod diameter, and other axially varying features when modeled at the bundle/lattice library level. All fuel and spacer geometries are consistent between GE 14i and GE 14 . Only the number of heated rods is perturbed by the GE 14i configuration . The following sections discuss unique features of the GE 14i and their impact on PANAC11 .

3 .2.1 .2 .a Zero-Power Rods The introduction of zero-power rods impacts the calculation of the heated perimeter, average fuel rod temperature, average planar linear heat generation rate (APLHGR), and the fuel pin linear heat generation rate (LHGR) for the isotope bearing bundles . The heat deposition from gamma effects in the cobalt isotope rods is expected to be less than 1 kW/ft under maximum nodal power conditions . For purposes of critical power, average fuel rod temperature, average planar power and peak UOZ rod power, all gamma energy is assumed to be deposited in the fuel rods. The correct count of heated rods, zero-power rods, and total rods is provided to PANACI I as input quantities . The quantities shown in Table 3-4 are significant to the proper processing of thermal limits in PANAC 11 .

No changes in PANAC 11 are required to model the thermal performance of the GE 14i fuel design .

3 .2 .1 .2.b Nodal Quantities The impacts on nodal reactivity, nodal pin power distributions, and nodal instrument response functions are explicitly provided by lattice physics evaluations with TGBLA06.

No changes in PANAC 11 are required to model the GE 14i fuel design .

3 .2 .1 .2.c Pin Power Reconstruction The influence of zero-power rods on the PANAC 11 pin power reconstruction model was evaluated and no statistically meaningful differences were observed. Pin power reconstruction impacts in the GE14i and adjacent fuel assemblies have been reviewed and the pin power reconstruction model was determined to be adequate.

No changes in PANAC 11 are required to model the GE 14i fuel design .

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NEDO-33505 Revision 0 Non-Proprietary Information 3.2.1.3 ITA Margin Considerations For the CPS ITA design, additional margins will be applied to the Linear Heat Generation Rate limit and the cell Shutdown Margin Limit. The need for these additional margins stems from the use of Zircaloy-2 connector sections in the cobalt isotope rods. The neutron absorption cross section of the connector section is lower than the neutron absorption cross section of the cobalt bearing section . This is typical of segmented rod applications .

The connector/spacer zones will not be modeled in the 3-dimensional simulator PANACI I in the CPS ITA program . However, 2 and 3 dimensional modeling of the connector/spacer zones was performed as part of design studies to determine the appropriate assumptions to accommodate cobalt isotope rod geometric modeling assumptions . The 2-dimensional models were evaluated with TGBLA06 and the 3-dimensional models were evaluated with PANAC 11 and MCNP-05 .

The lower absorption cross section of the connector zone increases the reactivity of this section relative to the cobalt isotope bearing zone. This will potentially reduce the shutdown margin in the control rod cell that contains the GE14i bundle . An additional (( )) Ak shutdown margin in the control rod cell containing the GE 14i bundle will provide the necessary margin to accommodate this geometric modeling assumption . This additional margin was determined by explicitly modeling all axial zones (connector and cobalt) in a GE 14i bundle with PANAC 11 and evaluating the change in control rod worth of control blades adjacent to the GE14i bundles. This evaluation was performed over the complete CPS Cycle 12 operation .

NEDO-33505 Revision 0 Non-Proprietary Information 3.2.2 Thermal-Hydraulic Methodology GNF uses the "New Dix" void-quality correlation in its thermal-hydraulics treatment in GNF thermal-hydraulic methodology. This void correlation has previously been shown to be applicable for all current GE BWR fuel designs, including 10x10 lattices with part length rods.

ISCOR09 is a thermal-hydraulic core analysis program wherein different fuel types can be designated to represent various types of bundles within a core. The introduction of bundles with zero-power rods, such as in GE14i can be readily handled by ISCOR09 input quantities.

PANACI I uses the "New Dix" void-quality correlation in its thermal-hydraulics treatment and accounts for bundle leakage and water rod flow by parameterized input from ISCOR simulations. As discussed in Section 3.2.1 .2, PANACI is capable of modeling the GE 14i bundle design for the CPS core.

All other thermal-hydraulic characteristics of GE14i are identical to GE14.

3.2.3 Safety Limit Methodology GESAM02 utilizes the PANACI I physics models to calculate CPR distribution . The lattice Root Mean Square (RMS) fission density uncertainty has been evaluated for GE14i and was observed to be consistent with the RMS fission density uncertainty of GE 14 . The results of this evaluation can be found in Table 3-2. The GE 14 uncertainties will be used in the evaluation of the safety limit analysis for the CPS safety limit evaluations. The capability of GESAM02 to model CPR related uncertainties is adequate for the CPS GE14i bundles and is not impacted by the number of heated rods in GE14i .

3.2.4 Transient Analysis Methodology The impact of the (( )) zero-power rods in the GE14i design will be reflected in the data that propagates through the PANAC 11 core model to the transient analysis methods . The GE14i design characteristics will be used in the transient analysis methods ODYNV09 and TASC-03 . The impact of the (( )) zero-power rods in GE14i in the eight CPS GE14i bundles will not impact the adequacy of the transient analysis methods.

3.2.5 Stability Methodology ODYSY05 obtains the GE14i geometry information from the ISCOR system and provides adequate results for the GE 14i bundle design . TRACG04 uses the PANAC 11 kinetics model and receives fuel neutronic information (nodal cross sections) through the PANAC 11 wrap-up 22

NEDO-33505 Revision 0 Non-Proprietary Information information. The fuel geometry information is provided through the TRACG04 user input data and is sufficiently flexible to model the GE14i bundle characteristics . The (( ))

zero-power rods in the CPS GE14i bundles will not impact the adequacy of the stability methods.

3.2.6 Fuel Rod Thermal-Mechanical Methodology The design of the U and Gd rods in the GE14i bundle is identical to the GE14 and will therefore have no impact on the GSTRM07 methodology.

3.2.7 ECCS-SAFERIGESTR The Emergency Core Cooling System (ECCS) analysis methodology applicable to CPS is SAFER/GESTR.

The zero-power rods can be described though the SAFER04 input. The GESTR fuel characteristics data is based on GE14 fuel rod evaluations. No changes to the GESTR fuel characteristics are required as a result of the use of GE 14 U02 fuel rod design characteristics .

The gamma energy generated in the cobalt isotope rods is assumed to be deposited in the uranium fuel rods. The total gamma energy generated in the cobalt isotope rods varies from 2% to 3% of the total gamma energy released in the lattice as a function of lattice exposure and void history. This assumption will provide a small conservatism in the SAFER/GESTR analysis .

The impact of the (( )) zero-power rods in the CPS GE14i bundles will not impact the adequacy of SAFER/GESTR analysis methodology.

3.3 GEXL+ Correlation The critical quality - boiling length correlation (GEXL+) was developed to accurately predict the onset of boiling transition in boiling water reactor (BWR) fuel assemblies during both steady-state and reactor transient conditions. In the GEXL+ correlation, critical quality is expressed as a function of boiling length, thermal diameter, mass flux, pressure, R-factor, and annular flow length. The R-factor is an input to the GEXL+ correlations and it accounts for the effects of the pin power distributions and the geometry of the assembly/channel/spacer on the assembly critical power . Its formulation for a given rod location depends on the power of that rod, as well as the power of the surrounding rods. A detailed discussion on the specific GEXL14 correlation developed for the GE14 fuel and the R-factor methodology is given in Reference 6.

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NEDO-33505 Revision 0 Non-Proprietary Information The measure of the capability of a boiling transition prediction correlation is its ability to predict the test data. The GEXL 14 correlation has been demonstrated to be an accurate predictor of the GE14 fuel for wide ranges of fluid conditions, a number of different rod-to-rod power distributions, and different axial power shapes as provided in Reference 6.

The GE14i ITAs are identical to the GE14 fuel bundles except for the cobalt isotope rods in GE14i. Due to the similarity between GE14i and GE14, the GEXL14 correlation can be applied to the GE14i ITAs . The applicability of the GEXL14 correlation to the GE14i ITAs is demonstrated by comparing the GEXL 14 prediction to the critical power data with zero-power rods in the GE 14 bundle .

Full-scale critical power and pressure drop testing for a simulated GE 14 fuel bundle was performed in the Stern Laboratories test facility in Hamilton, Ontario. As a part of the Stern testing for the GE 14 fuel, critical power data was collected with zero-power rods and ((

)) Four different rod-to-rod power distributions were tested for a wide range of inlet flow and inlet subcooling conditions at pressure of 1000 psia. The rod-to-rod power distributions or local peaking patterns tested with zero-power rods at Stern Laboratories are presented in Figure 3-5, where cobalt isotope rod(s) or the highest R-factor rod(s) of each pattern were identified with a green background color. The peaking patterns J1/J2/J3 have (( )) zero-power rods and pattern DOxx has (( )) zero-power rods. Mass flux and inlet subcooling conditions are plotted in Figure 3-6. Typical bundle axial power shape is presented in Figure 3-7. The Stern Laboratories test assembly characteristics are provided in Reference 6.

A statistical analysis was performed for the GE 14 database with zero-power rods consisting of

(( )) data points obtained from the Stern test assembly . To facilitate the statistical evaluation of the predictive capability of the GEXL14 correlation, the concept of an experimental critical power ratio (ECPR) is used.

The ECPR is determined from the following relationship:

ECPR = (Predicted Critical Power)/(Measured Critical Power)

A summary of the ECPR statistics is provided in Table 3-5 and the predicted critical powers are compared to the measured critical power in Figure 3-8. It is shown from the mean ECPR that the GEXL 14 correlation conservatively predicts the critical power data with zero-power rods . ((

NEDO-33505 Revision 0 Non-Proprietary Information In summary, the applicability of the GEXL14 correlation to the GE14i ITAs was demonstrated by comparing the GEXL 14 prediction to the critical power data with zero-power rods in the GE 14 bundle. The R-factor methodology as described in Reference 6 was applied in generating the R-factors for the test assembly containing zero-power rods as part of the overall evaluation of the applicability of GEXL14 to GE14i. As such, the R-factor methodology is confirmed applicable to GE 14i. The GEXL 14 correlation, on average, conservatively predicted the critical powers for the zero-power rod test data obtained at Stern Laboratories for the GE14 bundle with ((

NEDO-33505 Revision 0 Non-Proprietary Information Table 3-1 Summary of GNF Methods Applicability to GE14i Methodology Analysis Code and Revision Supported TGBLA06 X Nuclear PANAC I I X Thermal Hydraulic ISCOR09 X Safety Limit MCPR GESAM02 X ODYNV09 X Transient Analyses TASC-03 X ODYSY05 X Stability TRACG04 X TASC-03 X ATWS ODYNV09 X Thermal-Mechanical GSTRM07 X LAMB-08 X ECCS-LOCA TASC-03 X SAFER04 X

NEDO-33505 Revision 0 Non-Proprietary Information Table 3-2 Lattice RMS Fission Density Uncertainty Comparison In-Channel Void GE14i GE14 Fraction CPS (Typical) 00%

40%

70%

Average Table 3-3 GE14i Control Blade Worth Comparison

%Delta In-Channel Void MCNP TGBLA Fraction (Kunc- K,)/K (K, nc-~on )/Kunc (WorthTCSCA-WorthMCNP) on nc

/ WorthMCNP Cold 00%

40%

70%

Table 3-4 Internal PANACII Parameters Parameter Values GE14 GE14i Number of Fuel Rods (NBFURD) 92 (( ))

Quantity of Like Rods (QNTYPR) 92 92 Thermal Diameter (GEYL Definition) (in) ((

Hydraulic Diameter (in)

NEDO-33505 Revision 0 Non-Proprietary Information Table 3-5 GEXL14 Statistics for GE14 CP Data with Zero-Power Rods Number of Data Points Mean ECPR Standard Deviation

NEDO-33505 Revision 0 Non-Proprietary Information Figure 3-1 TGBLA06 Cobalt 60 Inventory per Linear cm for GE14i for 00% In-Channel Void History

NEDO-33505 Revision 0 Non-Proprietary Information Figure 3-2 TGBLA06 Cobalt 60 Inventory per Linear cm for GE14i for 40% In-Channel Void History

NEDO-33505 Revision 0 Non-Proprietary Information Figure 3-3 TGBLA06 Cobalt 60 Inventory per Linear cm for GE14i for 70% In-Channel Void History

NEDO-33505 Revision 0 Non-Proprietary Information Figure 34 Power Spike in Face Adjacent Fuel Rods in the Wide-Wide Corner

NEDO-33505 Revision 0 Non-Proprietary Information Figure 3-5 Rod-to-Rod Power Distributions with Zero-Power Rods 33

NEDO-33505 Revision 0 Non-Proprietary Information Figure 3-6 GEXL14 Test Conditions (P=1000 psia)

NEDO-33505 Revision 0 Non-Proprietary Information Figure 3-7 Typical Bundle Axial Power Shape Used for GEXL14 Testing

NEDO-33505 Revision 0 Non-Proprietary Information Figure 3-8 Calculated Versus Measured Critical Power for GEXL14

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4. Licensing Evaluations The ITAs will be monitored during operation via the process computer and 3DM. Cycle-specific operating limits are established to assure compliance with licensing limits.

Furthermore, ITAs will be inserted into core locations projected to be non-limiting with planned, steady state control rod patterns.

Cycle-specific analyses will be performed for CPS Reload 12 Cycle 13 to establish fuel operating limits for the ITAs that assure compliance with regulatory limits. Results of these analyses will be documented in the CPS Reload 12 Cycle 13 Supplemental Reload Licensing Report (SRLR). Furthermore, licensing analyses will be performed for the ITAs for each cycle of their operation, wherein the effect of the ITAs is considered for each of the appropriate licensing events and anticipated operational occurrences (AOOs) to establish appropriate reactor core thermal limits for operation .

Exelon intends to insert the GE14i ITAs into CPS and to operate them in Cycle 13 . Cycle specific analyses to establish fuel operating limits are not yet complete. When the cycle specific analyses are complete, GNF-A will document the results in the SRLR and Exelon will update the CPS Core Operating Limits Report (COLR) accordingly .

The application of approved methods to analyze events and accidents whose results could be affected by the inclusion of GE 14i ITAs is discussed in Section 3.2. Because the analysis of the ITAs will meet the approved criteria, it is not anticipated that NRC approval of the cycle specific SRLR and COLR is required prior to insertion.

The list of events deemed adequate to support the CPS Cycle 13 reload transient analysis is summarized in Table 4-1 .

4.1 Evaluation of Abnormal Operational Transients Current approved methods described in Reference 2 are appropriate to determine the impact of abnormal operational transients on the GE14i ITAs . As described in Reference 2, cycle-specific analyses of the limiting transient events are performed to establish the plant Operating Limit Minimum Critical Power Ratio (OLMCPR), demonstrate thermal/mechanical compliance and demonstrate compliance with the ASME overpressure protection criteria .

The CPS Cycle 13 reload licensing analyses will include specific modeling of the GE14i ITAs in the determination of the OLMCPR. As discussed in Section 3.3, GEXL 14 is demonstrated to conservatively apply to the GE 14i ITAs . The GE 14i ITA U and Gd fuel mechanical designs are identical to the GE14 fuel rods (Section 3.2.7) and, therefore, the normal GE14 thermal and mechanical overpower LHGR limits ensure compliance with thermal-mechanical licensing requirements as specified in Reference 2.

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NEDO-33505 Revision 0 Non-Proprietary Information The CPS abnormal operational transients evaluated to support the introduction of GE 14i ITAs into CPS are identified in the following subsections .

4.1.1 Decrease in Reactor Coolant Temperature The events in this category are:

" Maximum Demand Feedwater Controller Failure (FWCF)

" Loss of Feedwater Heating (LFWH)

" Inadvertent HPCS Pump Start-up (HPCS)

" Pressure Regulator Failure Open (PRFO)

" One RV/SV Opening The CPS Cycle 13 reload licensing analyses will include specific modeling of the GE 14i ITAs in determination of the OLMCPR. Plant characteristics, which include the ((

)) Such plant parameters are independent of fuel bundle design and are modeled by methods discussed in Section 3.2.5 . ((

The transient response is affected by the core average reactivity characteristics . However, the introduction of GE14i ITAs have a negligible impact on the core average nuclear parameters affecting the transient response because the eight GE14i ITAs represent a small fraction of the total bundles in the core, and the hydraulic characteristics of the GE14i ITAs are similar to the GE14 bundles (Section 3 .2.2). Therefore, the GE14 bundles dictate the core average nuclear parameters that affect the transient response. ((

NEDO-33505 Revision 0 Non-Proprietary Information 4.1.2 Increase in Reactor Pressure The events in this category are:

" Load Rejection (Turbine Control Valve Fast Closure) with Bypass Failure (LRNBP)

" Load Rejection (Turbine Control Valve Fast Closure) with Bypass (LRWBP)

" Turbine Trip with Bypass Failure (TTNBP)

" Turbine Trip with Bypass (TTWBP)

" Main Steam Isolation Valve Closure with Flux Scram (MSIVF)

" Pressure Regulator Downscale Failure (PRFDS)

" Loss of Auxiliary Power / Loss of Condenser Vacuum

" Loss of Feedwater Flow (LOFW)

" Loss of Instrument Air System

(( )) will continue to bound the TTWBP, Loss of Auxiliary Power/Loss of Condenser Vacuum, LOFW and Loss of Instrument Air System events due to reasons specified in Section 4.1 .1 . ((

))

The FWCF event includes a system pressure increase due to the turbine trip from reactor high water level. However, the FWCF event is categorized as a reactor coolant temperature decrease event and is discussed in Section 4.1 .1 . The MSIVF event is analyzed for overpressure protection and is discussed in Section 4.2 .2.

)) The GE14i ITAs do not impact the core average response of the limiting events in this category because the core average nuclear characteristics are dictated by GE14 bundles, as discussed in Section 4.1 .1 . The TTNBP, LRNBP and PRFDS events are analyzed as part of the cycle-specific reload licensing analyses.

4.1.3 Decrease in Reactor Coolant System Flow Rate The events in this category are :

" Trip of One Recirculation Pump

" Trip of Two Recirculation Pumps

" Recirculation Flow Control Failure - Decreasing Flow The AOOs in this category are bounded by the events listed in Table 4-1 . The decrease in core flow causes a decrease in reactor power, and consequently, the thermal limits are not challenged . The core-wide decrease in reactor power instigated by decreasing core flow is a BWR characteristic that remains unchanged with the introduction of the GE14i ITAs . ((

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)) The Single Loop Operation (SLO) Pump Seizure accident is discussed in Section 4 .2.3 .

4.1.4 Reactivity and Power Distribution Anomalies The events in this category are:

" Rod Withdrawal Error (RWE) at Power

" Mislocated Fuel Assembly Accident

" Misoriented Fuel Assembly Accident

" Startup of an Inactive Recirculation Loop

" Recirculation Manual Controller Failure Increasing Flow The Rod Withdrawal Error, Mislocated Fuel Assembly Accident and the Misoriented Fuel Assembly Accident are potentially limiting events at CPS. ((

)) The slow recirculation flow runout event has been previously analyzed to develop the flow dependent MCPR and LHGR limits (Reference 7).

The off-rated limits documented in Reference 7 are validated as part of reload licensing analyses for application to Cycle 13.

4.1.5 Increase in Reactor Coolant Inventory The event in this category is :

" Inadvertent HPCS Pump Start-up In Cycle 13, as in CPS Cycle 12 (Reference 7), the Inadvertent HPCS Pump Start-up event will continue to be bounded ((

)) due to reasons specified in Section 4.1 .1 .

NEDO-33505 Revision 0 Non-Proprietary Information 4.1.6 Decrease in Reactor Coolant Inventory and Other Accidents The events in this category are:

" Control Rod Drop Accident

" Main Steam Line Break Accident

" Fuel Handling Accident

" Loss of Coolant Accident (LOCA)

All events in this category are classified as limiting faults or design basis accidents . The Main Steam Line Break and LOCA events result in a decrease in reactor coolant inventory . The consequences of the LOCA as a result of the introduction of GE14i ITAs are discussed in Section 4.5 .11 . The radiological consequences of all accidents are discussed in Section 4.3 .

4.2 Evaluation of Other Transients 4.2.1 Anticipated Transients Without Scram (ATWS)

The evaluation of ATWS events is not a design basis requirement . However, specific requirements for ATWS are provided in 10 CFR 50.62 . In particular, BWRs are required to have an alternate rod insertion system, automatic recirculation pump trip, and an 86 gpm equivalent boron injection system . These features are included in the CPS plant. The current licensing basis ATWS analyses demonstrate reactor integrity, containment integrity and fuel integrity. Reactor integrity is demonstrated by ensuring that peak reactor vessel pressure is less than the ASME Service Level C limit. Containment integrity is demonstrated by ensuring that the peak suppression pool temperature is below the maximum allowed bulk suppression pool temperature and containment pressure is less than the containment design pressure limit. Fuel integrity is demonstrated by ensuring that the Peak Cladding Temperature (PCT) and fuel cladding oxidation is below the 10 CFR 50.46 limit.

The ATWS response is primarily affected by the key plant characteristics, ((

)) The GE14i ITAs represent a small fraction of the total bundles in the core. As a result, their impact on the core average nuclear parameters is negligible . Furthermore, the hydraulic characteristics of the GE14i ITAs are similar to the GE14 bundles. Therefore, as in CPS Cycle 12 (Reference 7), a cycle-specific ATWS analysis is not required because of the introduction of GE 14i ITAs .

The fuel and cycle-independent ATWS evaluation for CPS is documented in Reference 8 .

This evaluation demonstrates significant margin to the aforementioned ATWS acceptance criteria . The flexibility and EOOS options supported in Reference 7 are unaffected by the introduction of the GE 14i ITAs .

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NEDO-33505 Revision 0 Non-Proprietary Information 4.2.2 ASME Overpressure Protection ASME overpressure protection is demonstrated by the analysis of an assumed closure of all Main Steam Isolation Valves (MSIVs) with no credit for the direct scram signal on MSIV closure (MSIVF). Scram is assumed to occur on high neutron flux in the reactor core. ((

)) The flexibility and EGOS options supported in Reference 7 are unaffected by the introduction of the GE 14i ITAs .

4.2.3 Single Loop Operation Pump Seizure Analysis This Single Loop Operation (SLO) Pump Seizure event was analyzed for GE14 introduction into CPS (Reference 9). ((

)) The loading of GE14i ITAs will not affect the results of this analysis because a conservative multiplier is applied to the core average void coefficient that results in a void coefficient corresponding to the lower range of values documented in Reference 1 . The GEXL14 correlation is conservatively applied to the GE14i ITAs ; therefore, ((

4.2.4 Applicability of Off-Rated Limits to GEl4i ITAs The off-rated limits are constructed to assure that thermal limits are not violated when a transient event (AOO) is initiated while the reactor is operating at an off-rated power/flow condition . The off-rated limits (or multipliers) are confirmed applicable for new fuel designs as part of the Amendment 22 process outlined in GESTAR II (Reference 2), cycle-independent analyses for a New Fuel Introduction reload application, or as in the case of CPS, plant-specific off-rated limits (Reference 7). The important bundle characteristic that influences the transient response and operating thermal limits is the critical power performance of the new fuel . As discussed in Section 3.3, GEXL14 is conservatively applied to the GE14i ITAs . In addition, the impact on the core average nuclear parameters that affect the transient response is negligible because the GE14i ITAs represent a small fraction of the total bundles in the core, and the hydraulic characteristics and fuel mechanical design of the GE 14i ITAs are similar to the GE 14 bundles . As such, the power and flow dependent limits are applicable to the GE14i ITAs . ((

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NEDO-33505 Revision 0 Non-Proprietary Information 4.2.5 Flexibility and Equipment Out-of-Service (EOOS) Options As discussed in Section 3.3, the thermal-hydraulic characteristics, fuel mechanical design, and critical power performance of the GE 14i ITAs is similar to GE 14 fuel. In addition, the impact on the core average parameters that affect the transient response is negligible because the GE 14i ITAs represent a small fraction of the total bundles in the core. Therefore, the Flexibility and EOOS options supported in Reference 7 remain unchanged and continue to be supported with the introduction of the GE 14i ITAs . The cold-water events, fast pressurization and ASME overpressurization events in combination with the licensed flexibility options for CPS will be analyzed as part of the cycle-specific reload licensing analyses .

4.3 Evaluation of Design Basis Accidents The CPS Design Basis Accidents (DBAs) to be evaluated are identified in Chapter 15 .0 of the CPS Updated Safety Analysis Report (USAR). The Control Rod Drop Accident (CRDA),

Main Steam Line Break (MSLB) accident outside containment, Fuel Handling Accident (FHA), and Loss-of-Coolant Accident (LOCA) are licensed under 10 CFR 50.67 utilizing Alternate Source Term (AST) methodology per Regulatory Guide (RG) 1 .183 .

4.3.1 Control Rod Drop Accident The CPS licensing basis CRDA analyzed in Section 15 .4.9 of the CPS USAR assumes a failure of 1200 rods (for 10 x 10 fuel). An estimated mass fraction of 0.77% of the fuel in the damaged rods is assumed to reach or exceed the initiation temperature of fuel melting. Fuel reaching melt conditions is assumed to release 100% of the noble gas inventory and 50% of the iodine inventory. The remaining fuel in the damaged rods is assumed to release 10% of both the noble gas and iodine inventories. ((

11 Therefore, the licensing basis CRDA radiological analysis is not impacted by the introduction of eight GE14i assemblies at CPS.

As described in Reference 10, compliance with licensing limits governing CRDA is assured through adherence to the Banked Position Withdrawal Sequence (BPWS). The associated analyses have generically demonstrated large margin to licensing limits governing acceptable 43

NEDO-33505 Revision 0 Non-Proprietary Information enthalpy insertions . The BPWS analyses demonstrated that the characteristic control rod worth associated with limiting rods in a BPWS sequence are low as compared to that required to challenge the 280 caUgm fuel design limit. The reactivity characteristics of GE14i are similar to GE14; therefore, the introduction of eight GE14i assemblies at CPS will have negligible effects on the existing CRDA margin . In addition to similar fuel reactivity characteristics the impact on the rod worths is constrained by other design factors such as shutdown margin and in-sequence rod worths.

4.3.2 Main Steam Line Break Accident The CPS licensing basis MSLB analyzed in Section 15 .6.4 of the CPS USAR assumes no fuel damage occurs as a result of the event. ((

)) Therefore, the licensing basis MSLB radiological analysis is not impacted by the introduction of eight GE 14i assemblies at CPS .

4.3.3 Fuel Handling Accident The existing GE 14 fuel handling accident analysis takes the available potential energy from a dropped fuel assembly and calculates the number of failed fuel rods, assuming the rods fail by 1% strain in compression using a number of conservative assumptions . Given the reduced weight of the GE14i fuel assembly, the potential energy from a dropped fuel assembly is reduced and the resulting number of failed rods is also reduced.

The CPS licensing basis FHA is analyzed in Section 15.7.4 of the CPS USAR. The licensing basis FHA postulates that an irradiated GE 14 fuel assembly is dropped 34 feet onto the reactor core and fails 172 total rods . Of the failed rods, the entire noble gas fission product inventory and a fraction of the iodine fission product inventory are assumed to be released to the air above the water. All particulates are retained by the water.

)) Therefore, the licensing basis FHA radiological analysis is not impacted by the introduction of eight GE14i assemblies at CPS.

4.3.4 Loss-of-Coolant Accident (LOCH)

The CPS licensing basis LOCA is analyzed in Section 15.6.5 of the CPS USAR. The impact of eight GE14i ITAs on LOCA radiological consequences was evaluated .

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NEDO-33505 Revision 0 Non-Proprietary Information The introduction of eight GE14i bundles at CPS presents no significant impact on the AST LOCA source term.

4.4 Thermal-Mechanical Evaluation Thermal-mechanical characteristics of GE 14i cobalt isotope rods were evaluated . For the GE14i cobalt isotope rods, thermal-mechanical evaluations were performed f f 45

NEDO-33505 Revision 0 Non-Proprietary Information

)) The failure modes considered are the same as for a fuel rod (Reference 2) : internal melting and loss of cladding integrity . These evaluations demonstrate that the internal geometry (no melting) and cladding integrity is maintained for the cobalt isotope rods during steady-state operation and anticipated operational occurrences (AOOs). In particular, the following conclusions have been made:

. cc 2.

3.

4.

5.

4.5 Other Evaluations The results of other evaluations required to support the loading of the GE 14i ITAs are provided below.

NEDO-33505 Revision 0 Non-Proprietary Information 4.5.1 Stability This section provides a qualitative assessment of the impact of GE14i ITAs on thermal-hydraulic instability. In accordance with Reference 11, a review was done on the GE14i ITAs to demonstrate that an ITA is very unlikely to result in single-channel instability .

The licensed Option III solution uses NRC-approved methodologies as outlined in GESTAR II (Reference 10).

A qualitative assessment was performed based on the GE 14i bundles to evaluate the impact on decay ratio . With((

))

differences between the GE14 and GE14i channel decay ratio performance ((

))

Because a small number of GE 14i bundles are loaded, and since the hydraulic characteristics of the GE14i bundles approximate the GE14 bundles, the impact on the core decay ratio is negligible .

CPS is an Option III plant, and the plant will continue to use the Option III system for Cycle 13 . For the Option III stability solution, two stability aspects must be considered. The first consideration is the Oscillation Power Range Monitor (OPRM) system setpoint; the second is the size of the Backup Stability Protection (BSP) regions .

The OPRM setpoint with the GE14i bundles included is expected to be comparable to that of GE14 because ((

)) The Delta CPR over Initial MCPR Versus the Oscillation Magnitude (DIVOM) curve is evaluated on a cycle-specific basis and is expected to show cycle-to-cycle variation. The change in MCPR due to a 2-pump recirculation trip is expected to be similar since it relates only to the most limiting bundle in the core, and is evaluated on a cycle-specific basis. The fuel design and licensing process calculates an appropriate OPRM setpoint. This value is not expected to be affected by the introduction of 8 GE14i ITAs in CPS Cycle 13 beyond cycle-to-cycle variation.

An assessment indicates that the BSP regions are not expected to vary significantly with the substitution of the GE 14i design for the GE 14 design since ((

)) Moreover, the BSP regions are typically limited by characteristics of the full core, most of which remains GE14 fuel . A stability evaluation was performed and confirms this assessment. The fuel design and licensing process calculates appropriate BSP regions, addressing flexibility options like reduced feedwater temperature.

The size of these BSP regions is not expected to be affected by the introduction of 8 GE14i ITAs in CPS Cycle 13 beyond cycle-to-cycle variation.

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NEDO-33505 Revision 0 Non-Proprietary Information 4.5.2 Appendix R Safe Shutdown Fire The limiting safety shutdown method for CPS Appendix R fire protection in Reference 8 is Method R, which uses Reactor Core Isolation Cooling (RCIC), Safety Relief Valves (SRVs) and Residual Heat Removal (RHR) from a remote shutdown panel. As shown in Section 4.5 .8, ((

)) provided the other parameters determining Appendix R fire protection containment response remain unchanged, such as ((

)) Also, RCIC is used from a remote shutdown panel to maintain the water level above the top of active fuel, and the peak cladding temperature (PCT) for GE14i ITA is the initial steady state fuel temperature, which is well below the Appendix R PCT limit of 1500°F . The fire event evaluation results and acceptance criteria in Reference 8 remain applicable for GE 14i ITA.

4.5.3 Station Blackout As shown in Section 4 .5.8, ((

provided other key parameters determining containment response such as ((

)) do not change for GE141 ITAs . ((

)) for the licensing basis SBO analysis .

4.5.4 Reactor Internal Pressure Difference The reactor internal pressure difference (RIPD) evaluates the maximum pressure drop for reactor internals, the minimum fuel bundle lift margin, and the maximum control rod guide tube (CRGT) lift force, as well as acoustic and flow-induced loads on jet pump, core shroud, and shroud support .

The thermal hydraulic design for the GE 14i bundle closely matches the overall pressure drop of previous designs . The pressure drop characteristics of the GE14i ITA fuel are equivalent to those of the GE14 fuel and the method applied to the GE14 fuel pressure drop analysis is valid for the GE 14i application. Therefore, the current pressure drops for reactor internals based on GE 14 (Reference 8) remain applicable for GE 14i ITA.

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NEDO-33505 Revision 0 Non-Proprietary Information The minimum fuel bundle lift margin is determined by ((

)) The bundle weight for GE14i bundles is ((

)) GE14. Other key parameters determining the fuel lift margin remain unchanged due to similar thermal-hydraulic design. Therefore, the GE14i bundle results in ((

)) than the minimum fuel bundle lift margins for GE14. The limiting faulted condition fuel lift margin for GE14i is (( )) . The impact on fuel lift load and other reactor internal loads due to this decreased fuel lift margin is assessed by structural analysis in Section 4.5 .7.

The maximum CRGT lift force is determined by ((

)) These parameters do not change for GE14i ITAs and thus the maximum CRGT lift force for GE 14 remains applicable for GE 14i ITA.

The introduction of 8 GE 14i ITAs has no effect on the acoustic and flow-induced loads on jet pump, core shroud and shroud support, which are caused by pressure waves as a result of a recirculation suction line break. ((

which remain unchanged for the GE 14i ITA.

4.5.5 Reactor Internals Structural Evaluation A qualitative structural assessment of the reactor internal components was performed with respect to the current design basis evaluation. The evaluation in Section 4.5 .4 demonstrates that the current pressure drops for reactor internals and maximum CRGT lift force for GE 14 remain applicable for GE 14i ITA and that GE 14i ITA fuel has no effect on the acoustic and flow-induced loads The evaluation in Section 4.5.7 demonstrates that operation with GE14i ITAs will have no adverse effect on the structural integrity of the reactor internals relative to seismic loading .

The weight variation of the full core as reported in Section 4.5 .7 is negligible relative to structural integrity . The symmetrical location of the GE 14i bundles will have no eccentricity impact on the mass center of the full core.

All applicable Normal, Upset, Emergency, and Faulted condition loads for GE14i ITAs such as seismic loads, acoustic and flow induced loads, fuel lift loads, RIPDs, system flow loads, core flow loads, and thermal loads, as appropriate, were considered in the assessment. These 49

NEDO-33505 Revision 0 Non-Proprietary Information loads are either bounded by, remain unaffected, or have an insignificant effect on the structural integrity of the reactor internals with respect to the current design basis evaluation.

The introduction of GE 14i ITAs will have an insignificant effect on the structural integrity of the reactor internal components .

4.5.6 Recirculation System Evaluation An evaluation of the effects of introducing GE14i fuel on Reactor Recirculation System (RRS) performance for CPS was performed. The evaluation is based on clean equipment conditions rather than current plant operating conditions. This evaluation does not consider the potential effects of crud deposition on jet pumps, which lowers their efficiency, as discussed in SIL 465 Supplement 1 (Reference 12).

For the recirculation system evaluation, the primary impact of introducing a different fuel assembly would be a core pressure drop change . The evaluation results show the core pressure drop does not change with the introduction of the GE14i fuel bundles . The recirculation system pressures, temperatures, pump flow rate, and motor brake horsepower remain the same values when the core pressure drop does not change . There is no change to the recirculation pump required or available Net Positive Suction Head (NPSH) since the pump flow rates and recirculation system pressure/temperature is the same value as before GE14i fuel introduction .

Consequently, it is concluded that no modifications to RRS equipment or setpoints are required with the introduction of GE14i ITAs at CPS .

4.5.7 Seismic and Dynamic Response Due to the negligible full core weight variation (less than (( )) impact), the seismic/dynamic behavior of the core, the reactor internals, and the balance of plant will not be affected by the introduction of eight GE14i ITAs. The dominant fuel type, GE14 fuel, dictates the seismic behavior of the core.

There is a minor reduction in fuel lift loads based on the Reactor Internal Pressure Difference assessment in Section 4.5 .4. The minor reduction of mass of GE 14i may slightly increase its dynamic fuel lift height to (( )) inches; however, this is still less than the (( ))

inches allowable (Reference 13).

4.5.8 Decay Heat Assessment A comparative core decay heat assessment between GE 14 and GE 14i fuel types was performed ((

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)) It was concluded that replacing eight GE 14 bundles with eight GE 14i bundles presents no significant change in core decay heat at CPS .

4.5.9 Neutron Fluence Impact RPV fluence is negligibly impacted by the introduction of 8 GE14i ITAs. Given that the reactor power is unchanged and the core-wide void and relative power distribution remains approximately the same, the introduction of GE 14i fuel into the core will not significantly impact the magnitude of the RPV flux.

RPV fluence is highly dependent on the core peripheral bundle power distribution, which is affected by the cycle operating plan and the core loading pattern. The loading pattern constraints and limitations are applicable to each reload fuel cycle, regardless of the fuel type.

The substitution of neutron absorber material for fuel in (( )) of 92 fuel rods in the GE14i design will contribute to insignificant changes in power density for this bundle design.

Furthermore, the loading of eight ITAs is only a small fraction of the total core loading of 624 bundles, and therefore, is not expected to significantly impact the core-wide power distribution and peripheral bundle power.

Shifting from one fuel type to another with different part length rod designs may cause slight variation in the axial flux distribution . However, GE 14i fuel uses the same part length rod design as GE 14, thus no variation in the axial flux distribution is expected.

Therefore, the GE14i fuel introduction will not have any significant impact on the current overall fluence values for CPS .

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NEDO-33505 Revision 0 Non-Proprietary Information 4.5.10 Containment Response As shown in Section 4.5 .8, ((

provided other key parameters determining containment response ((

)) do not change for GE14i ITA. ((

)) including the Recirculation Suction Line Break (RSLB), the Main Steam Line Break (MSLB), the alternate shutdown cooling and the Humphrey concerns . In addition, ((

)) as discussed in Section 4.5 .14.

4.5.11 ECCS LOCA As stated in Section 3 .1, GE14i ITAs will be loaded into a non-limiting location (not hot channel) with respect to ECCS LOCA Maximum Average Planar Linear Heat Generation Rate (MAPLHGR) limits . The number of fueled rods will be (( )) less in the GE14i ITA bundle relative to the GE 14 bundle, and MAPLHGR will not be averaged over the zero power rods . Therefore, the ECCS LOCA MAPLHGR limits in Reference 7 for GE14 remain bounding for GE 14i ITAs .

Incorporating errors reported under 10 CFR 50.46 through 2008-01, ((

)) per Reference 8. The licensing PCT for GE14 reported in Reference 8 remain applicable for GE14i ITAs .

)) Furthermore, because the PCT and the maximum local oxidation values remain within licensing limits, a coolable geometry is assured. Finally, introduction of the GE 14i ITA does not affect the reflooding capability of the ECCS or the operation of the core spray systems, thus assuring long-term cooling. Therefore, the five acceptance criteria established by 10 CFR 50.46 remain satisfied with the introduction of the GE14i ITAs .

In addition, the EOOS such as ADSOOS and the flexibility options such as MELLLA, SLO, and FFWTR for the current analyses remain applicable for GE 14i.

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NEDO-33505 Revision 0 Non-Proprietary Information 4.5.12 Hydrogen Injection The evaluation performed with regard to OE27774 estimates the impact of inserting cobalt isotope rods in the core of the CPS reactor on the downcomer gamma dose rates. OE27774 (Reference 14) describes an incident where a core design change at Fermi 2, a moderate Hydrogen Water Chemistry (HWC) plant that does not use NobleChern , resulted in a lower gamma flux in the downcomer region of the reactor causing a reduction in the hydrogen-oxygen recombination reaction. The decreased gamma flux necessitated an increase in the feedwater hydrogen injection rate to maintain SCC mitigation compared to the previous cycle of operation. Therefore, a key objective of the evaluation of this OE relative to GE14i was to determine whether there is any potential for a decrease in gamma flux in the downcomer region of the CPS reactor as a result of the cobalt containing rod insertion.

Regardless of the location of the cobalt isotope rods, there is no possibility for gamma flux in the downcomer region to decrease for the inclusion of 8 GE14i ITAs . The CPS plant, being a Noble Metal Chemical Addition (NMCA) plant, the recombination of hydrogen and oxygen is expected to be more efficient in the downcomer region; hence no negative impact is expected on the hydrogen requirement for SCC mitigation.

4.5.13 Post-LOCH Hydrogen Control A qualitative analysis of the metal-water reaction has been performed with respect to the addition of 8 GE 14i ITA bundles inserted in the Cycle 13 core with the goal of comparing any possible addition of hydrogen produced to the hydrogen produced as a result of the previous extended power uprate analysis with a core of GE 14 fuel . The section of Regulatory Guide 1 .7 Revision 2 (Reference 15) which addresses the metal-water reaction is based on the fact that the decay heat generated in the irradiated fuel rods in the post-LOCA core is the primary vehicle for the metal-water reaction. During the first two minutes following the LOCA when the metal-water reaction is assumed to occur in accordance with Regulatory Guide 1 .7 (Reference 15), the cobalt isotope rods will not produce any significant gamma heating greater than the core ambient conditions . This leaves the cobalt isotope rod's temperature well below the threshold to cause the metal-water reaction, such that with respect to the metal-water reaction there is no increase in the quantity of hydrogen produced in a post-LOCA situation. As such, there are no adverse impacts from this perspective with the utilization of the GE14i ITA.

With respect to radiolysis, a comparison of the gamma energy of these two types of fuel bundles was performed. This was done to determine the significance of any additional hydrogen produced from radiolysis as a result of the gamma heating caused by the energy profile of the irradiated cobalt isotope rods within the GE14i bundle . The cobalt isotope rods will have a different energy profile than that of a normal fuel rod within a GE 14 bundle due to 53

NEDO-33505 Revision 0 Non-Proprietary Information the relatively long half-life of the Co-60 isotope . This comparison is shown in Table 4-2 and includes the percentage difference in energy between the two types of fuel bundles.

Hydrogen production as a result of corrosion was not considered for this analysis since this is typically the result of changes in containment coolant chemistry including pH and containment metal and coolant temperatures . The use of the GE 14i bundles will not produce changes in these parameters .

Based on the comparison shown in Table 4-2, the percentage difference in total core energy throughout the 30 day period is significantly less than 1 percent such that the relative impact on any additional hydrogen produced is considered insignificant. Therefore, there is no appreciable impact on the ability of the CPS Combustible Gas Control System (CGCS) to mitigate the effects of a combustible gas mixture in a post-LOCA situation with the addition of 8 GE 14i ITA bundles.

4.5.14 Environmental Dose Considerations An evaluation was performed on the effects of dose from cobalt isotope rods on refueling equipment . For the refueling equipment, the gamma radiation contained within each cobalt isotope rod is 0 .02403E-3 R/hr. For all (( )) cobalt isotope rods, the dose rate would be 0.12E-3 R/hr. This is the dose rate measured at the water surface with the top of the fuel submerged 8 feet below, whereas the refueling bridge is approximately 10 feet above the water surface . Eight feet is as close to the water surface as allowed by the fuel handling equipment (CPS USAR Chapter 9). Consequently, the dose rate is even lower on the bridge due to the additional air gap.

Using the above dose rate as the worst case, the dose accumulation on the refueling equipment during a refueling outage of approximately 7-day duration could be in the 0.02 R range . The 7-day value is a conservative estimate for transporting fuel that is normally stored at the bottom of the fuel pool. This radiation dose is well below the radiation threshold of all materials . The radiation threshold is defined as the lowest radiation dose which induces permanent change in a measured property of a material and the first detectable change in a property of a material due to the effect of radiation.

In general, the refueling equipment may contain synthetic organic materials, inorganic materials, and metals. Of the above materials, the synthetic organic materials are the most susceptible to radiation dose effects. Teflon TFE has the lowest dose threshold which is in the 2E4 R range. All others are greater than IE5 R. As such, a radiation dose of 0.02 R is insignificant compared to the radiation threshold of these materials .

A total conservative exposure dose of 0.02 R for a seven-day exposure is well below the radiation threshold of the materials in the refueling equipment . This level will not affect the 54

NEDO-33505 Revision 0 Non-Proprietary Information functionality of the materials or the components in the refueling equipment . Therefore, the GE14i fuel introduction will not have any impact on the refueling equipment .

4.5.15 Fuel Storage Criticality Safety This evaluation addresses the introduction of GE14i ITA fuel at CPS . Analyses have previously been performed to introduce GE14 fuel to CPS. These analyses address all fuel storage racks that are used at CPS. The original analyses evaluated the peak reactive GE 14 lattice that meets the fuel storage rack reactivity safety limits at a maximum bounding uniform enrichment of no less than 4.8wt% U-235 .

The scope of this analysis assumes that mechanically equivalent stainless steel rods will be used to replace any cobalt isotope rods that are removed from the bundle in order to maintain mechanical integrity of the stored bundle. Use of the mechanically equivalent stainless steel rods lends greater stability to the system and displaces the interstitial water in order to conserve the relative moderator effects of the previous analyses.

For the purposes of criticality safety, the only difference between the GE14i and a standard GE14 fuel assembly is that up to (( )) regular fuel rods have been replaced with ((

)) cobalt isotope rods in the GE14i ITA. This replacement introduces neutron absorbers to the system . In this case, the absorber is in the form of cobalt targets, but the neutronics involved apply to all neutron absorbers that may be introduced by the isotope rods . The displaced enrichment may be either removed from the assembly entirely or it may be placed within other locations within the same bundle or bundles not utilizing cobalt isotope rods as allowed by fuel and core design constraints .

The maximum bounding uniform enrichment of no less than 4.8wt% U-235 assumed in the original GE14 models ensure that the models are insensitive to the spatial distribution of fissile material. In this way, the potential enrichment displacement proposed by the GE14i ITA is already conservatively factored into the original GE14 models . For these reasons, the GE14 fuel storage rack reactivity safety limits, including k.. design limits, are appropriate for use with GE 14i ITAs .

4.5.16 Fresh Fuel Shipping In order to be shipped in the RAJ-II container, the GE 14i bundle must be shown to meet the technical shipping requirements specified in the RAJ-11 Certificate of Compliance (Reference 16). These requirements specify, in part, the "Type and Form of Material" contained in the fuel bundles and are documented in Section 5(b)(1) of the certificate . Since the technical requirements specified in Section 5(b)(1) pertain specifically to "enriched commercial grade uranium or enriched reprocessed uranium, uranium oxide or uranium carbide fuel rods enriched to no more than 5.0 weight percent in U-235", these technical requirements do not 55

NEDO-33505 Revision 0 Non-Proprietary Information apply to the cobalt isotope rods since these rods do not contain uranium. Furthermore, Section 5(b)(1) contains no material type and form restrictions on non-uranium bearing components contained in fuel rod locations .

There is no licensing impact on the fresh fuel shipping container criticality analysis since these (( )) cobalt isotope rod locations are analyzed as containing 5% enriched UOZ rods in Chapter 6 of the RAJ-11 Safety Analysis Report (SAR) which bounds the cobalt isotope rods from a criticality safety standpoint. Therefore, although the GE14i bundle with (( ))

total fuel rods is outside the range of 91-100 as specified in Table 3 of the Certificate of Compliance (Reference 16), this condition is bounded by what has been analyzed and demonstrated to be safe in the RAJ-1I container under both Normal Conditions of Transport (NCT) and Hypothetical Accident Conditions (HAC).

4.5.17 Fuel Channel Distortion Channel distortion that can cause channel interference is a function the fluence gradient (fluence bow), early life control (shadow bow) and the pressure gradient across the channel (channel bulge). The presence of non-fueled rods will not significantly affect these parameters, and therefore, the channel performance of GE14i bundles will be the same as of GE14 bundles.

4.5.18 Fuel Conditioning Guidelines There is no change to the barrier fuel operating guidelines for the use of the isotope bundles.

The fuel conditioning guidelines are based on the peak nodal powers in the bundle; thresholds are exposure dependent . The presence of cobalt isotope rods does not modify these guidelines.

4.6 Manufacturing Quality Assurance All aspects of the GE14i ITA program are controlled under the GE Nuclear Energy Quality Assurance Program Description (Reference 17).

GE 14i zirconium tubing and components are procured, fabricated, and handled under the same quality controls as standard production fuel rods at GNF. Target pellets are handled with similar quality controls as UOZ pellets. The controls include material traceability, material handling, hydrogenous material control, foreign material exclusion, and discrepant material control. Independent quality oversight is performed during production to assure that these quality controls are being met, as in standard fuel rod production. All operators receive training specific to the fabrication of the cobalt isotope rods, focusing on these critical to quality aspects of the fabrication process.

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NEDO-33505 Revision 0 Non-Proprietary Information Cobalt targets were certified to meet the current drawing and specification requirements . The cobalt targets are verified to meet the design requirements on a sampling basis.

Weld parameters for both internal and external rod segments are being fully qualified prior to production and certified weld operators perform all production welding. Rod integrity is being verified by helium leak check of both the inner and outer tubes following welding .

Outer segment welds are also 100% verified by Level 11 Ultrasonic Evaluators . These rod integrity checks help ensure that there is no opportunity for cobalt pellets to escape. The outer surface of assembled rods is visually inspected before bundle assembly and during final bundle inspection to further ensure rod integrity and to ensure there is no foreign material on the surface of the rods .

Isotope bundle assembly is performed using robust automated controls to ensure the physical and nuclear configuration of the final assembly . Rod locations and makeup are identified, verified and downloaded directly to the manufacturing equipment, eliminating the potential for manual transposition errors.

All product records are reviewed by the GNF Quality organization prior to product release for shipment .

4.7 Post-Operational Evaluations 4.7.1 Spent Fuel Pool Effects Through analysis, it has been shown that there are no adverse effects from the introduction of GE14i fuel in the CPS spent fuel pool, provided guidance for storage of GE14i bundles is followed to minimize the effect of gamma heating on the spent fuel pool concrete walls.

Irradiated fuel storage procedures should be modified to specify that the GE14i bundles be stored at least 4 feet from the pool walls. With the 4 foot distance requirement in effect, there is no limitation on the amount of time a GE14i bundle may remain in the pool.

The introduction of GE14i fuel to the CPS spent fuel pool was evaluated for three effects.

First was the effect of the additional heat from the Co-60 decay. Second was the effect of increased gamma radiation on the concrete walls of the spent fuel pool. In both of these first two cases, the extra radiation from the cobalt isotope rods was conservatively added to the radiation in a "normal" GE14 bundle. No credit was taken for the removal of (( )) fuel rods in each bundle. The third evaluation was the effect of GE14i bundles on the cleanup portion of the Fuel Pool Cooling and Cleanup (FPCC) system.

The additional heat from the Co-60 decay is insignificant when compared to the total heat from a normal refueling discharge . The additional heat added by 8 GE14i bundles in an offload is (( )) after shutdown over that of an offload of 57

NEDO-33505 Revision 0 Non-Proprietary Information all GE 14 fuel. The current heat load calculated for refueling conditions from CPS calculation OIFC25 is 41 .2 MBTU/hr, representing a margin of approximately 3% under the worst case FPCC system heat removal capacity of 42.54 MBTU/hr . Adding ((

)) The small amount of extra heat added by the cobalt isotope rods poses no additional risk of spent fuel pool local boiling over that previously analyzed.

The gamma radiation effect on the spent fuel pool walls was evaluated for the case that the GE14i bundle is placed 1, 4, and 6 feet from the pool wall. In the GE14i analysis, no credit was taken for shielding provided by the spent fuel and racks in the outer rows, however, water and self-shielding were credited.

Significant concrete heating due to gamma radiation begins at 1E+10 MeV/cm2/sec. The maximum incident radiation due to a GE 14i bundle placed one foot from the spent fuel pool walls is approximately 7.2E+10 MeV/cm2/sec, so concrete heating due to gamma would be significant. At 4 feet, the energy deposition rate is 1 .4E+8 MeV/cm2/sec, well below that required to cause significant concrete heating .

Long-term concrete degradation begins with a total integrated gamma dose of approximately 1E+10 R. The total integrated dose from a GE14i assembly left in the spent fuel pool, one foot from the side, after 3 years is less than 3.65E+9 R, without taking into account any decay of the Co-60 or fission products. Therefore, there is no restriction on the amount of time a GE 14i bundle can be stored in the Spent Fuel Pool (SFP), provided the bundle is stored at least one foot from the pool wall to avoid gamma heating effects. Note that the four foot limit for gamma heating will be more limiting for storage locations .

Per CPS Reactor Engineering, procedures exist to guide placement of irradiated fuel bundles in the SFP to avoid gamma heating of the wall concrete. These procedures should be modified to specify that the GE14i bundles be stored at least 4 feet from the pool walls. With the 4 foot distance requirement in effect, there is no limitation on the amount of time a GE14i bundle may remain in the pool.

The GE 14i rods are clad with the same material as the GE 14 rods so that there will be no appreciable difference in the corrosion products from GE 14i versus GE 14. Therefore, there will be no adverse effect on the cleanup portion of the FPCC system.

4.7.2 Post-Irradiation Handling 4.7.2.1 Post Irradiation Bundle Disassembly Timing To reduce impact on plant outage planning and operations, normal cobalt isotope rod extraction will occur as post-outage activities after a fuel assembly's end of life (EOL) . The cobalt isotope rods will be removed from the discharged GE14i fuel assembly and replaced 58

NEDO-33505 Revision 0 Non-Proprietary Information with mechanically equivalent stainless steel rods to maintain integrity of the stored bundle.

However, following the first cycle of operation it will be necessary to conduct an ITA fuel inspection during the outage . As part of this examination process, a single cobalt isotope rod from one of the GE14i bundles will be extracted and sent to the Vallecitos Nuclear Center (VNC) in Sunol, CA. To maintain nuclear and mechanical equivalence for this bundle, a single cobalt isotope rod will be replaced with an identical cobalt isotope rod. There are no plans to shuffle rod segments out of spent GE 14i fuel assemblies between power cycles for additional irradiation time.

4.7.2.2 Rod Removal and Replacement Fuel rod removal and replacement is part of the standard scope of work for GEH's Fuel Examination Services (FES) team. FES routinely disassembles, inspects, and where required, replaces individual rods . The processes, tooling, and methods needed for isotope rod removal and replacement are the same as those currently in place. Segmented rods have also been retrieved using FES procedures . Therefore, the retrieval of segmented cobalt isotope rods is within GNF/GEH experience .

4.7.2.3 Segmented Rod Disassembly Experience GEH has experience removing and replacing segmented rods within the spent fuel pool, via procedure 246-GP-20 Revision 1 . This procedure also includes disassembly of the segments and subsequent placement into a segment storage rack. The segmented rod procedure steps include :

" Assemble, set-up, and install equipment and tools required for the spent fuel pool.

" Perform receiving inspection on fresh replacement rod segments (or stainless steel replacement rods at bundle End of Life (EOL)).

" Load new full length cobalt isotope rods (or stainless steel replacement rods at bundle EOL) into the full length fuel rod storage can.

" Move ITA to the fuel preparation machine .

" Remove channel.

" Visually inspect ITA.

" Remove upper tie plate.

" Grapple cobalt isotope rod designated for removal with a standard collet tool and then withdraw rod from bundle by lowering fuel preparation machine .

" Verify segment locations and serial numbers .

" Insert rod into the segmented rod disassembly vise and secure.

" Unscrew the top isotope segment by twisting the collet tool and the torque wrench.

" A second grapple tool with the proper collet for handling the full diameter of the isotope segments is attached to the top of the second segment, the vise is released, the rod is pulled up until the second junction is visible in the mirror, and the vise is again clamped shut.

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" The second isotope segment is unscrewed and placed into a storage tube.

" Repeat above process for all remaining segments .

" Grapple the correct replacement rod in the full-length fuel rod storage can.

" Insert the rod into the proper location in the isotope fuel assembly .

" Repeat above steps until all cobalt isotope rods have been replaced.

" Visually inspect the ITA and components .

" Reinstall the upper tie plate.

" Rechannel the ITA

" Removing the ITA from the fuel preparation machine and place into the designated location.

While previous segmented rod campaigns demonstrated disassembly success, current designs also include a hex impression on each rod segment . This provides improvement on visual inspection, as well as, further gripping ease during unscrewing steps .

Prior to segmented rod disassembly during the 2012 CPS outage, a similar isotope segmented rod procedure will be prepared and incorporated into the fuel inspection process . Per standard GEH and Exelon analysis process, any devices needed during the removal/replacement process onto the fuel preparation machine will adhere to all analysis requirements, including seismic .

4.7.2.4 GEH Tools Usedfor Rod Disassembly GEH has experience with rod removal and replacement . The specific tools that will be used to disassemble a GE 14i cobalt isotope rod into segments are the Fuel Rod Collet Grapple, the Fuel Rod Side Grapple, and the 6 Rod Universal Storage Rack . These same tools have been used for other 10x10 disassembly efforts. They are designed for operators to work underwater and have lengths required for As Low as Reasonably Achievable (ALARA) considerations . Below is a brief description of these existing GEH tools.

" Fuel Rod Collet Grapple - This tool grabs the upper end plug of each segment.

" Fuel Rod Side Grapple - This tool grabs the tubing of the rod where segments are threaded together. GEH will use two of these tools if the upper end plug of the segment is unable to be held by the Fuel Rod Collet Grapple.

" 6 Rod Universal Storage Rack (side mount to fuel prep machine) - This 12 foot rack attaches to the fuel prep machine (FPM) so that the top of the rod is the same height as the top of the rods in the bundle. Appropriate length pedestals will be in each tube to facilitate segment disassembly from the top.

NEDO-33505 Revision 0 Non-Proprietary Information 4.7.2.5 Contingency Plans IfIssues Encountered Design features have been included to protect cobalt encapsulation integrity if segment disassembly problems occur. Specifically, the male/female connection has a thread size that will allow for disassembly after years of irradiation . However, if disassembly under normal conditions is not possible, this will not be a problem since the male end plug of a cobalt isotope rod has been designed with a strategic break zone so that large amounts of torque will force a fracture at this known breaking point, not the location of cobalt targets. Furthermore, the broken male component is locked into the female receptor preventing any debris inside the fuel pool. Prototype tests have shown the failure torque to be high enough to prevent failure during normal operation, but low enough for contingency plans with existing tooling.

4.7.2.6 Shielding and Dose Considerations The activity of the cobalt-60 targets will be less than (( )) However, most analyses in this document assume a maximum activity of (( )) to ensure conservative compliance. This strategy was also applied during the dose calculations during cobalt isotope rod disassembly.

Thirty years of GNF/GEH fuel examination experience has shown that fuel assemblies placed 6-7 feet underwater provides appropriate shielding regarding ALARA considerations to personnel . For cobalt isotope rod disassembly ALARA considerations, the GE 14i fuel assemblies will be placed approximately 9-10 /2 ' feet underwater . The receiving basket will be hung from a depth (to top of basket) of approximately 10 '/2 feet underwater . This depth allows for the top of the longest segment to be 9 feet underwater when manually lowering the segment into the basket before releasing the segment from the collet (or side) grapple .

If the receiving basket is filled with one bundle of segmented rods, this would correspond to a conservative estimate of (( )) of Co-60. Therefore, the result of a completely filled receiving basket with ((

)) which is the limiting dose consideration . Table 4-3 shows the basket dose rates at incremental distances for both the top and side views. Table 4-3 can be used to determine a multitude of dose conditions. Similarly, Table 4-4 shows dose calculations for individual cobalt isotope rods from the top and side views. These dose calculations can be multiplied by

(( )) to represent a (( )) cobalt isotope rod configuration.

4.7.2.7 Cask Movement Experience Currently, GEH owns and operates two GE Model 2000 casks capable of moving cobalt isotope rods. The Model 2000 cask has a gross assembly weight of 33,550 lbs, with a cask weight of 23,750 lbs. This cask weight is well below the CPS fuel building crane limit of 60 tons. This cask has been used for movement of radioactive materials from inside Exelon spent fuel pools. NUREG 0612 : Control of Heavy Loads at Nuclear Power Plants (Reference 61

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18) applies to cask movement activities . Key steps for an approved site specific procedure include removing the overpack, transferring the cask to the fuel floor, cask lid handling, transferring the cask to and from the spent fuel pool cask pit, transferring the cask to the reactor building, and reinstalling the overpack . Specific procedure steps during GE Model 2000 movements at one of the Exelon plants include the following :

" Ensure equipment and materials are available (e.g. . fuel building crane, decontamination equipment, 1 IOV single phase 60 Hz power, 80 psig air, underwater lighting, fuel handling platform, appropriate hooks, HEPA unit, face shields, pressure washer, stainless steel cable, approved cask lifting equipment, strap wrench, leader rope, chain hoist, all purpose pool tooling, torque wrenches, various sockets, "Never Seeze" (or equivalent), shackles, ratchet, tamper-proof security seals, demineralized water supply, pliers, nylon slings, binoculars, plastic sheeting, underwater survey instrument, bottle of certified helium, bottle of compressed nitrogen and liquid nitrogen) .

" Establish work orders as required by Exelon to accomplish the project tasks.

" Review of the appropriate procedures by GEH personnel assigned to the project.

" Ensure fuel building ventilation system is in operation .

" Train personnel assigned to the project on the GE Model 2000 cask handling operations/process .

" Ensure all lifting equipment, if used (e.g. slings, shackles, tools), have been load tested and inspections are current. Note : due to the increased water shielding, required above for ALARA you would need approximately 18 to 19 ft length lifting slings for handling the cask in the spent fuel pool and loading of the cobalt basket. Currently GEH has standard 11 ft and 15 ft lift slings approved .

" Ensure reactor building crane has been inspected in accordance with Exelon procedures .

" Ensure decontamination equipment and facilities are operational.

" Prejob briefing by Exelon supervisor on expectations and personnel safety concerns .

" Obtain permission to start project activities from Exelon management .

" Inform control room supervisor as required by Exelon procedures.

" Establish a foreign material exclusion area.

" Establish a contamination control area for the pool hardware work.

" Transfer GE Model 2000 cask-to-cask loading pool.

" Load GE Model 2000 cask.

" Prepare loaded GE Model 2000 cask.

" Return to normal operations and demobilize cask rigging and equipment per Exelon procedures .

" Notify the Exelon supervisor that the project on-site work is complete .

NEDO-33505 Revision 0 Non-Proprietary Information 4.7.2.8 Source Tracking Considerations As described in Section 2 of this document, the cobalt targets are double encapsulated and backfilled with helium to enable leak testing during initial weld process and to ensure weld integrity . Furthermore, each segment will be provided a unique tracking number to aid in bundle placement and post processing core location identification . For discussion purposes, these encapsulated devices are defined as `targets' . While some of these characteristics appear similar to the description of sealed sources, other characteristics indicate that they are not .

The final rule for 10 CFR Parts 20 and 32: National Source Tracking of Sealed Sources, November 8, 2006, clearly states that "For the purpose of this rulemaking, the term nationally tracked source does not include material encapsulated solely for disposal, or nuclear material contained in any fuel assembly, subassembly, fuel rod, or fuel pellet." The targets are manufactured with non-radioactive materials . In the irradiation process, both the encapsulating and target material become irradiated . The encapsulating material also becomes contaminated from contact with reactor coolant . Additionally, no verification steps of encapsulation integrity are intended .

GEH has experience safely moving large amounts of high specific activity cobalt-60 within all state and federal regulations, and will continue to safely transport the isotope segments of the GE 14i ITA program . The shipment of this quantity of material is governed by security regulations and orders issued by the NRC and will require a similar level of detail to be communicated to the NRC, preserving the control envisioned by 10 CFR Part 32 . Per NRC requirements, source tracking will begin after the irradiated cobalt pellets are removed from the cobalt isotope rods and encapsulated into sealed sources for final product use.

4.7.3 Post-Irradiation Examination Post-Irradiation Examination (PIE) of a GE14i ITA bundle and rods may include all or part of the following four inspections : Poolside Visual, Poolside Gamma Scan Measurements, Poolside Combined Instrumentation Measurement System (COINS) and Segmented Rod Hot Cell Destructive Exam. This PIE plan applies to the end of the first cycle of operation . These tests may also be performed after subsequent fuel cycles and at the bundle's end of life (EOL) .

4.7.3.1 Poolside Visual Examination The GE 14i ITA visual exam may include the following elements :

1. A full bundle periphery visual exam of all bundle mechanical elements

2. Assess rod-to-rod spacing of the cobalt isotope rods relative to nearby fuel rods

3. Assess rod growth of the cobalt isotope rods relative to nearby fuel rods 63

NEDO-33505 Revision 0 Non-Proprietary Information 4 . Assess spacer cells with the cobalt isotope rods removed to verify no abnormal wear A GE 14i ITA rod visual exam may include the following elements :

1 . Visual exam of 1 or more cobalt isotope rods after brushing to remove the crud

2. Visual exam of 1 or more brushed fuel rods adjacent to cobalt isotope rods 4.7.3.2 Poolside Gamma Scan Measurements Gamma scanning is a non-destructive method to determine the gamma emission from a radioactive source and can be used to measure the relative fission product inventory in irradiated nuclear fuel rod or the gamma activity of a cobalt isotope rod. A multi-channel analyzer is used to capture gamma scan counts at discrete energy levels in order to determine the activity for all isotopes of interest for a given decay chain. An axial Gamma scan can determine the distribution of activity over the component's active fuel length . Measuring the relative ratios of certain fission fragments and decay products present at a specific time provides a power measurement . The exposure of the fuel rods or cobalt isotope rods can also be determined by measurement of a specific isotopic distribution of activity over a component's length .

To date, pin-by-pin gamma scans on GE 14 fuel assemblies have provided data on relative fuel burn up and power profiles of reactor fuels, fission gas disposition in the fuel rods, position and dimension of internal structures within fuel assemblies and relative distribution of various isotopes in fuel . The applications of interest for the GE 14i ITA may include the axial power distribution measurements of one or more fuel rods adjacent to the cobalt isotope rods.

4.7.3.3 Poolside Combined Instrumentation Measurement System Inspection The COINS system is for measuring corrosion and liftoff for a single fuel rod, which has been removed from a bundle. The poolside COINS will be performed on cobalt isotope rods in order to non-destructively obtain information about outer surface corrosion and diameter.

The COINS system possesses two eddy-current probe lift-off instruments to measure the waterside corrosion thickness of the cladding at two azimuthal locations 180° apart and one probe lateral displacement instrument to measure the axial profile of the cladding.

4.7.3.4 Segmented Rod Hot Cell Destructive Exam One cobalt isotope rod will be sent to the GEH Vallecitos Nuclear Center (VNC) in Sunol, CA for Hot Cell Examination .

During hot cell examination, the outer rod will be inspected for corrosion . The outer rod will be cut at the Vallecitos Hot Cell to inspect the inner capsule for the following :

" Vibration and corrosion

" Inner and outer oxide layer thickness 64

NEDO-33505 Revision 0 Non-Proprietary Information The inner rod will also be cut to expose the TPR and cobalt targets . The cobalt targets can be inspected for the following:

" Location specific activities along axis

" Cobalt target conditions and status of Ni plating

" Vibration and corrosion

" Ease with which cobalt targets are released The results of these inspections are expected to confirm the successful performance of the GE 14i bundle design.

NEDO-33505 Revision 0 Non-Proprietary Information Table 4-1 List of Analyzed Events for the Reload License with GE14i ITAs in the Core Table 4-2 Total Core Energy Comparison of GE14 and GE14i

NEDO-33505 Revision 0 Non-Proprietary Information Table 4-3 Basket Dose Rate Values Distance in Top P Dose Rate ( mr/hr) Side Dose Rate ( mr/hr)

Water (ft) 0.5 4.727E+08 3.344E+08 1 1 .393E+08 9.281 E+07 1 .5 4 .434E+07 2.867E+07 2 1 .457E+07 9.361 E+06 2 .5 4.961 E+06 3.166E+06 3 1 .732E+06 1 .098E+06 3 .5 6.114E+05 3.869E+05 4 2.176E+05 1 .379E+05 4 .5 7 .852E+04 4.975E+04 5 2 .862E+04 1 .811 E+04 5.5 1 .049E+04 6.627E+03 6 3.866E+03 2.436E+03 6.5 1 .429E+03 8.993E+02 7 5.301 E+02 3.332E+02 7.5 1 .968E+02 1 .239E+02 8 7.336E+01 4.624E+01 8.5 2.745E+01 1 .732E+01 9 1 .030E+01 6.502E+00 9.5 3.876E+00 2.447E+00 10 1 .461 E+00 9 .228E-01 10 .5 5.521 E-01 3.486E-01 11 2.089E-01 1 .319E-01 l

NEDO-33505 Revision 0 Non-Proprietary Information Table 4-4 Single Rod Dose Rate Values Distance in Top Dose Rate (mr/hr) Side Dose Rate (mr/hr)

Water (ft) 0.5 3.269E+05 1 .005E+08 1 7.869E+04 2.521 E+07 1 .5 2.205E+04 7.931 E+06 2 6.692E+03 2.702E+06 2 .5 2.140E+03 9.567E+05 3 7.080E+02 3.474E+05 3.5 2.399E+02 1 .281 E+05 4 8 .281 E+01 4.752E+04 4.5 2 .900E+01 1 .779E+04 5 1 .028E+01 6.708E+03 5.5 3.681E+01 2.538E+03 6 1 .329E+00 9.627E+02 6.5 4.836E-01 3.653E+02 7 1 .769E-01 1 .390E+02 7.5 6.507E-02 5.296E+01 8 2.403E-02 2.023E+01 8.5 NA 7.735E+00 9 NA 2.964E+00 9.5 NA 1 .134E+00 10 NA 4.384E-01 10.5 NA 1 .674E-01 11 NA 6.418E-02

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5. Conclusion Installation of 8 GE 14i ITAs into CPS Cycle 13 has been evaluated against the events addressed in Chapter 15 of the USAR as well as additional analyses typically performed during a fuel transition in Section 4 of this report . These assessments confirm compliance to licensing requirements with the ITAs inserted.

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6. References

1. GE Hitachi Nuclear Energy, "GE14 Compliance With Amendment 22 of NEDE-24011-P-A (GESTAR II)," NEDC-32868P, Revision 3, April 2009.

2. GE Hitachi Nuclear Energy, "General Electric Standard Application for Reactor Fuel (GESTAR II)," NEDE-2401 I -P-A-16, October 2007.

3. Letter from USNRC to G. A. Watford (GE), "Amendment 26 to GE Licensing Topical Report NEDE-24011-P-A, 'GESTAR II' - Implementing Improved GE Steady-State Methods," November 10, 1999.

4. Letter, "Implementation of Improved GE Steady-State Nuclear Methods," MFN 098-96, July 2, 1996.

5. "MCNP - A General Monte Carlo N-Particle Transport Code," Version 5, Los Alamos National Laboratory, LA-UR-03-1987, April 2003 .

6. GE Hitachi Nuclear Energy, "GEXL14 Correlation for GE14 Fuel," NEDC-32851P-A, Revision 4, September 2007.

7. GE Hitachi Nuclear Energy, "Supplemental Reload Licensing Report for Clinton Power Station Unit 1 Reload 11 Cycle 12," 0000-0067-2201-SRLR, Revision 0, December 2007.

8. GE Nuclear Energy, "Safety Analysis Report for Clinton Power Station Extended Power Uprate," NEDC-32989P, June 2001.

9. GE Nuclear Energy, "GE14 Fuel Design Cycle-Independent Analyses for Clinton Power Station," GE-NE-L12-00879-00-OIP, April 2001 .

10. GE Hitachi Nuclear Energy, "General Electric Standard Application for Reactor Fuel (Supplement for United States)," NEDE-24011-P-A-16-US, Revision 16, October 2007.

11 . GE Nuclear Energy, "BWR Owners' Group Long-Term Stability Solutions Licensing Methodology," NEDO-31960-A, November 1995 (including Supplement 1).

12 . GE Nuclear Services Information Letter (SIL) 465, Supplement 1, "Surface Observations on Jet Pump Mixers," April 30, 1993 .

13 . GE Nuclear Energy, "BWR Fuel Assembly Evaluation of Combined Safe Shutdown (SSE) and Loss-of-Coolant Accident (LOCA) Loadings (Amendment No. 3)," NEDE-21175-3-P-A, October 1984.

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14. OE27774, "Increased hydrogen injection rates required to mitigate intergranular stress corrosion cracking in the reactor vessel," Event Date: 10/16/08, Fermi 2 Reactor.

15. USNRC Regulatory Guide 1 .7, Revision 2, "Control of Combustible Gas Concentrations in Containment Following a Loss-of-Coolant Accident".

16. NRC Certificate of Compliance No. 9309, Revision 7, for Model RAJ-II Package, May 28, 2008.

17. GE Nuclear Energy, "GE Nuclear Energy Quality Assurance Program Description,"

NEDO-11209-04A, Revision 8, March 31, 1989.

18. US NRC, "Control of Heavy Loads at Nuclear Power Plants, Resolution of Generic Technical Activity A-36," NUREG-0612, 1980.

19. GE Nuclear Energy, "Methodology and Uncertainties for Safety Limit MCPR Evaluations", NEDC-32601-P-A, August 1999.

ATTACHMENT 5

SUMMARY

OF REGULATORY COMMITMENTS The following table identifies those actions committed to by Exelon Generation Company, LLC (EGC) in this document. Any other statements in the submittal are provided for information purposes and are not considered to be regulatory commitments.

COMMITTED COMMITMENT TYPE COMMITMENT DATE OR ONE-TIME A Programmatic "OUTAGE" Yes/No _es/No Verify that all required licensing analyses as Prior to loading Yes No defined in the CPS USAR and NEDE-24011-P- the ITAs in the A, "General Electric Standard Application for CPS core in Reactor Fuel," latest approved revision, are C1 R12 completed prior to loading the ITAs in the CPS core .