LR-N09-0290, NEDO-33529, Rev. 0, Safety Analysis Report to Support Introduction of GE14i Isotope Test Assemblies (Itas) in Hope Creek Generating Station

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NEDO-33529, Rev. 0, Safety Analysis Report to Support Introduction of GE14i Isotope Test Assemblies (Itas) in Hope Creek Generating Station
ML093640199
Person / Time
Site: Hope Creek PSEG icon.png
Issue date: 12/31/2009
From:
GE-Hitachi Nuclear Energy Americas
To:
Office of Nuclear Reactor Regulation
References
LAR H09-01, LR-N09-0290 DRF 0000-0107-3743, NEDO-33529, Rev 0
Download: ML093640199 (126)
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HITACHI GE Hitachi Nuclear Energy NEDO-33529 Revision 0 Class I DRF 0000-0107-3743 December 2009 Non-ProprietaryInformation Safety Analysis Report to Support Introduction of GE14i Isotope Test Assemblies (ITAs) in Hope Creek Generating Station Copyright 2009 GE-HitachiNuclear Energy Americas LLC All Rights Resered

NEDO-33529 Revision 0 Non-Proprietary Information NON-PROPRIETARY NOTICE This is a non-proprietary version of the document NEDC-33529P, 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 design, engineering, and other information contained in this document is furnished for the purposes of obtaining NRC approval to support the introduction of GE14i Isotope Test Assemblies (ITAs) in Hope Creek Generating Station. The only undertakings of GEH with respect to information in this document are contained in the contract between PSEG Nuclear and GEH; nothing contained in this document shall be construed as changing the contract. The use of this information by anyone other than PSEG Nuclear 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 ii

NEDO-33529 Revision 0 Non-Proprietary Information Table of Contents Page Introd uction ................................................................................................................ 1

2. GE14i Fuel Product Description .......................................................................... 2 2.1 N ew D esign Features ............................................................................................. 2 2.2 Cobalt Isotope Rod Failure Mechanism Controls ............................................... 5 2.3 Online Failure Detection ..................................................................................... 10
3. Nuclear Design and Methods ............................................................................ 23 3.1 Nuclear Core Design ......................................................................................... 23 3 .2 M ethods .................................................................................................................... 23 3.2.1 N uclear M ethods .......................................................................................... 24 3.2.2 Thermal-Hydraulic Methodology ............................................................... 27 3.2.3 In-Core Instrumentation ............................................................................... 27 3.2.4 Safety Limit Methodology .......................................................................... 28 3.2.5 Transient Analysis Methodology ................................................................. 28 3.2.6 Stability Methodology ................................................................................. 28 3.2.7 Fuel Rod Thermal-Mechanical Methodology .............................................. 29 3.2.8 ECCS-SAFER/GESTR ............................................................................... 29 3.3 GEXL+ Correlation ............................................................................................ 29
4. Licensing Evaluations ......................................................................................... 45 4.1 Evaluation of Abnormal Operational Transients ................................................ 45 4.1.1 Decrease in Reactor Coolant Temperature .................................................. 46 4.1.2 Increase in Reactor Pressure ........................................................................ 47 4.1.3 Decrease in Reactor Coolant System Flow Rate .......................................... 47 4.1.4 Reactivity and Power Distribution Anomalies ............................................ 48 4.1.5 Increase in Reactor Coolant Inventory ......................................................... 48 4.1.6 Decrease in Reactor Coolant Inventory and Other Accidents ..................... 49 4.2 Evaluation of Other Transients ........................................................................... 49 4.2.1 Anticipated Transients Without Scram (ATWS) ......................................... 49 4.2.2 ASME Overpressure Protection .................................................................. 50 4.2.3 Single Loop Operation Pump Seizure Analysis ............................................ 50 4.2.4 Applicability of Off-Rated Limits to GE14i ITAs ....................................... 51 4.2.5 Flexibility and Equipment Out-of-Service (EOOS) Options ....................... 51 iii

NEDO-33529 Revision 0 Non-Proprietary Information 4.3 Evaluation of Design Basis Accidents ................................................................ 52 4.3.1 Control Rod Drop Accident ........................................................................ 52 4.3.2 M ain Steam Line Break Accident ............................................................... 53 4.3.3 Fuel Handling Accident ............................................................................... 53 4.3.4 Loss-of-Coolant Accident (LOCA) ............................................................. 53 4.4 Therm al-M echanical Evaluation ........................................................................ 54 4.5 Other Evaluations ................................................................................................ 55 4.5.1 Stability ............................................................................................................. 56 4.5.2 Decay Heat A ssessm ent .............................................................................. 57 4.5.3 Appendix R Safe Shutdown Fire ................................................................. 57 4.5.4 Station Blackout .......................................................................................... 58 4.5.5 Containm ent Response ................................................................................. 58 4.5.6 ECCS LOCA ............................................................................................... 58 4.5.7 Reactor Internal Pressure D ifference ........................................................... 59 4.5.8 Seism ic and D ynam ic Response ................................................................. 59 4.5.9 Reactor Internals Structural Evaluation ...................................................... 60 4.5.10 Recirculation System Evaluation ................................................................. 60 4.5.11 N eutron Fluence Im pact ............................................................................... 61 4.5.12 Hydrogen Injection ...................................................................................... 61 4.5.13 Post-LOCA Hydrogen Control ................................................................... 62 4.5.14 Fuel Storage Criticality Safety .................................................................... 62 4.5.15 Fresh Fuel Shipping ..................................................................................... 63 4.5.16 Fuel Channel D istortion ............................................................................... 64 4.5.17 Fuel Conditioning G uidelines ...................................................................... 64 4.5.18 Em ergency Operating Procedure Data ......................................................... 64 4.6 M anufacturing Quality A ssurance ...................................................................... 64 4.7 Post-Operational Evaluations ............................................................................. 65 4.7.1 Spent Fuel Pool Effects ............................................................................... 65 4.7.2 Environm ental Dose Considerations ........................................................... 67 4.7.3 Post-Irradiation Handling ............................................................................. 68 4.7.4 Post-Irradiation Exam ination ...................................................................... 72

5. Conclusion ................................................................................................................ 80
6. References ................................................................................................................ 81 Appendix A ............................................................................................................................ A -i iv

NEDO-33529 Revision 0 Non-Proprietary Information List of Tables Table Title Page Table 2-1 GE14i and GE14 Fuel Assembly Parameters ................................................... II Table 3-1 Summary of GNF Methods Applicability to GEI4i ......................................... 33 Table 3-2 Lattice RMS Fission Density Uncertainty Comparison at Zero Exposure ..... 34 Table 3-3 GE14i Control Blade Worth Comparison ......................................................... 34 Table 3-4 Internal PANACI I Parameters ........................................................................ 34 Table 3-5 GEXL14 Statistics for GE14 Critical Power Data with Zero-Power Rods ..... 35 Table 4-1 List of Analyzed Events for the Reload License with GE14i ITAs in the C ore ....................................................................................................................... 75 Table 4-2 GEI4i Data for Emergency Procedure Guidelines ........................................... 76 Table 4-3 Basket D ose Rate Values ................................................................................. 78 Table 4-4 Single Rod Dose Rate Values .......................................................................... 79 v

NEDO-33529 Revision 0 Non-Proprietary Information List of Figures Figure Title Page Figure 2-1 GE14i Bundle Cutaway View .......................................................................... 12 Figure 2-2 GE14i Lattice Arrangement ............................................................................ 13 Figure 2-3 Cobalt Target Isometric View .............. ....................... 14 Figure 2-4 Cobalt Target Orthographic View .................................................................... 14 Figure 2-5 )) Isometric View .................................. 15 Figure 2-6 (( )) Orthographic View ............................ 15 Figure 2-7 )) Isom etric View ...................................................................... 16 Figure 2-8 )) Orthographic View ............................................................... 16 Figure 2-9 )) Isom etric View ..................................................................... 17 Figure 2-10 (( 1] Orthographic View ............................................................... 17 Figure 2-11 (( )) Isom etric View ............................................................... 18 Figure 2-12 Er )) Orthographic View ........................................................ 18 Figure 2-13 Er Isom etric View ................................................ 19 Figure 2-14 Er )) Orthographic View ........................................ 19 Figure 2-15 Er )) Isometric View ............................................. 20 Figure 2-16 Er )) Orthographic View ....................................... 20 Figure 2-17 Er )) Isom etric View ................................................ 21 Figure 2-18 Er )) Orthographic View ........................................ 21 Figure 2-19 (( )) Isom etric View ................................................ 22 Figure 2-20 )) Orthographic View ........................................ 22 vi

NEDO-33529 Revision 0 Non-Proprietary Information Figure 3-1 TGBLA06 Cobalt 60 Inventory per Linear cm for GE14i for 00% In-C hannel V oid H istory ....................................................................................... 36 Figure 3-2 TGBLA06 Cobalt 60 Inventory per Linear cm for GEI4i for 40% In-C hannel V oid History ....................................................................................... 37 Figure 3-3 TGBLA06 Cobalt 60 Inventory per Linear cm for GE14i for 70% In-C hannel V oid H istory ....................................................................................... 38 Figure 3-4 Power Spike in Fuel Rods Adjacent to Cobalt Isotope Rod Connectors at 00% In-Channel Void Fraction ........................................................................ 39 Figure 3-5 Power Suppression Analysis for Rods (3,1) and (2,2) in Dominant Zone ........ 40 Figure 3-6 Rod-to-Rod Power Distributions with Zero-Power Rods ................................ 41 Figure 3-7 GEXL14 Test Conditions (P=1000 psia) ........................................................ 42 Figure 3-8 Typical Bundle Axial Power Shape Used for GEXL14 Testing ....................... 43 Figure 3-9 Calculated Versus Measured Critical Power for GEXLI4 .............................. 44 vii

NEDO-33529 Revision 0 Non-Proprietary Information ACRONYMS Term Definition AC Alternating Current ALARA As Low As Reasonably Achievable AOO Anticipated Operational Occurrence APF Axial Peaking Factor APLHGR Average Planar Linear Heat Generation Rate ASME American Society of Mechanical Engineers AST Alternate Source Term ATWS Anticipated Transients Without Scram ATWS-RPT Anticipated Transients Without Scram - Recirculation Pump Trip BOC Beginning of Cycle BPWS Banked Position Withdrawal Sequence BSP Backup Stability Protection BT Boiling Transition BWR Boiling Water Reactor COINS Combined Instrumentation Measurement System COLR Core Operating Limits Report CP Critical Power CPR Critical Power Ratio CR Control Room CRDA Control Rod Drop Accident CRGT Control Rod Guide Tube DBA Design Basis Accident DCF Dose Conversion Factor DIVOM Delta CPR over Initial MCPR Versus the Oscillation Magnitude EAB Exclusion Area Boundary ECCS Emergency Core Cooling System ECPR Experimental Critical Power Ratio EOC End-of-Cycle EOL End-of-Life EOOS Equipment Out-of-Service EPU Extended Power Uprate FES Fuel Examination Services FHA Fuel Handling Accident viii

NEDO-33529 Revision 0 Non-Proprietary Information ACRONYMS Term Definition FPCC Fuel Pool Cooling and Cleanup FPM Fuel Preparation Machine FWCF Feedwater Controller Failure - Maximum Demand GE General Electric GEH GE-Hitachi Nuclear Energy Americas, LLC GESTAR General Electric Standard Application for Reload GNF Global Nuclear Fuel - Americas, LLC HAC Hypothetical Accident Conditions HCGS Hope Creek Generating Station HPCI High Pressure Coolant Injection HWC Hydrogen Water Chemistry IMLTR Applicability of GE Methods to Expanded Operating Domains Licensing Topical Report INPO Institute of Nuclear Power Operations ITA Isotope Test Assembly LFWH Loss of Feedwater Heating LHGR Linear Heat Generation Rate LOCA Loss of Coolant Accident LOFW Loss of Feedwater Flow LPF Local Peaking Factor LPRM Local Power Range Monitor LPZ Low Population Zone LRNBP Generator Load Rejection with Bypass Failure LRWBP Generator Load Rejection with Bypass MAPLHGR Maximum Average Planar Linear Heat Generation Rate MLHGR Maximum Linear Heat Generation Rate MOC Middle of Cycle 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 Addition NRC Nuclear Regulatory Commission OLMCPR Operating Limit Minimum Critical Power Ratio ix

NEDO-33529 Revision 0 Non-Proprietary Information ACRONYMS Term Definition OPRM Oscillation Power Range Monitor PCT Peak Cladding Temperature PLR Part Length Rod PRFDS Pressure Regulator Failure - Closed PRFO Pressure Regulator Failure - Open PSEG PSEG Nuclear, LLC RCIC Reactor Core Isolation Cooling RG Regulatory Guide RHR Residual Heat Removal RIPD Reactor Internal Pressure Difference RMS Root Mean Square RPT Recirculation Pump Trip RSLB Recirculation Suction Line Break RWE Rod Withdrawal Error SACS Safety Auxiliary Cooling System SAR Safety Analysis Report SBO Station Blackout SCC Stress Corrosion Cracking SFP Spent Fuel Pool SLCS Standby Liquid Control System SLO Single Loop Operation SRLR Supplemental Reload Licensing Report SRV Safety Relief Valve TCV Turbine Control Valve TEDE Total Effective Dose Equivalent TIP Traversing In-Core Probe TPR Target Placement Rod TSV Turbine Stop Valve TTNBP Turbine Trip with Bypass Failure TTWBP Turbine Trip with Bypass UFSAR Updated Final Safety Analysis Report VNC Vallecitos Nuclear Center x

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1. Introduction PSEG Nuclear, LLC (PSEG) plans to load 12 GE14i Isotope Test Assemblies (ITAs) as part of Hope Creek Generating Station (HCGS) Reload 16 Cycle 17 during the 2010 refueling outage. These GEI4i bundles containing (( )) segmented cobalt isotope rods, also referred to as GE14i 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 PSEG.

The GEI4i 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 HCGS. The nuclear characteristics, mechanical characteristics, and thermal-hydraulic characteristics of these GE14i ITAs are compatible with those of the current GE1 4 fuel being loaded into HCGS.

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 PSEG in their evaluation of the effect of implementing GEI4i ITAs on the HCGS design and licensing bases.

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

(( )) fuel rods, (( )) cobalt isotope rods, and two large central water rods in a lOxlO 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 GEl4 fuel design. In addition to its similarities with the GEl4 design, GEI4i includes

(( )) cobalt isotope rods which convert cobalt-59 into cobalt-60 via neutron capture for use in various medical and food sterilization applications. Table 2-1 shows GE14i and GEl4 fuel assembly dimensions. Below is a list of new GE14i features.

  • GE14i Bundle Schematic
  • GEI4i Lattice Design
  • Er ]

A discussion of each of these new GEI4i features is provided in the remainder of this section.

Cobalt Target 2

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Defender Lower Tie Plate The GE14i fuel bundle will incorporate the Defender lower tie plate. The Defender lower tie plate on the GEI4i bundle, at a minimum, maintains the same resistance to foreign material debris as GE14 fuel with Defender or Debris Shield.

2.2 Cobalt Isotope Rod Failure Mechanism Controls The design discussed in Section 2.1 provides multiple features to prevent cobalt isotope rod failures. The main features that provide multiple levels of safety for the cobalt isotope rods are:

" Two layers of encapsulation before exposure of nickel-plated cobalt targets

" Solid Zircaloy connections at all spacer locations

" As discussed in Section 4.4, significantly lower heat generation rate compared to fuel rods This section provides the probable cobalt isotope rod failure mechanisms and controls in place to mitigate the failure and/or consequences of the failure during loading, operation, offloading and disassembly.

Fuel Handling Accident A summary of the FHA evaluation for the GEI4i ITAs is included as Section 4.3.3.

Errors in fuel assembly handling during loading and unloading could potentially result in failed cobalt isotope rods. The double containment design of the cobalt isotope rods provides additional protection against content release in comparison to normal fuel rods. The lack of gaseous fission products in the cobalt isotope rods ensures that the consequences in terms of radiological release of this accident are bounded by those of a standard fuel assembly.

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NEDO-33529 Revision 0 Non-Proprietary Information Administrative controls put in place to protect against fuel handling errors of standard fuel assemblies are similarly applied to GE14i ITAs.

Manufacturing Defects and Assembly Error As discussed in Section 4.6, the fabrication processes and materials for the GEI4i ITAs, including the cobalt isotope rod components, are controlled and handled under the same quality controls as fuel rods and other bundle components throughout fabrication to protect against manufacturing defects or assembly damage.

Pellet-Cladding Interaction No fuel pellets exist in the cobalt isotope rods so inherently there is no pellet cladding interaction.

Corrosion The target rod components use the same materials and are introduced into the same BWR environment as normal fuel rod cladding and components. Additionally, all composition elements, (cobalt, nickel, and zirconium) are known to not have problems in proximity to one another in a nuclear reactor environment.

Failure from cladding corrosion is much less likely for cobalt isotope rods than for standard production fuel rods. The outer cladding is the same material as standard GE14 production fuel rods, which has shown good in-reactor corrosion performance. The lack of fuel in the cobalt isotope rods results in lower cladding temperatures, and therefore significantly lower corrosion rates. The double containment design of the cobalt isotope rods provides additional protection against, content release in comparison to normal fuel rods. Both GNF manufacturing controls and PSEG reactor water chemistry controls put in place to protect against fuel rod cladding corrosion are similarly applied to assemblies containing cobalt isotope rods.

Primary Hydriding As discussed in Section 4.6, the isotope production rods are held to the same hydrogenous control standards as fuel rods. There are no additional sources of hydrogen in the cobalt isotope rod production and assembly process.

Cladding Creep Collapse As discussed in Section 4.4, even under conservative cobalt target heat generation rates, analyses indicate that thermal mechanical design criteria are met. Specifically, there is significant margin to interior melting and cladding integrity concerns. ((

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NEDO-33529 Revision 0 Non-Proprietary Information Rod Bow During rod bow, the deformation of the cladding results from creep of the Zircaloy material when subjected to operational temperatures, fluences, and axial loading from the rod weight and expansion spring. The cobalt isotope rods use standard fuel rod tubing for the outer structure. ((

)) The fluence seen by the outer tubing is no higher than that for a standard fuel rod. ((

)) Given these comparisons, it is seen that the creep of the Zircaloy cladding in the segmented rods is bounded by the creep in standard fuel rods.

Unthreading of Segmented Rods Steps have been taken to specifically design this failure mode out of the cobalt isotope rod.

((I Additionally, there are no counteracting torsional loads on the rod segments to encourage unthreading. Similar threaded features have maintained engagement as designed using the same assembly torque values.

Stress The stress in the cladding during operation of the cobalt isotope rods was evaluated using the computer program that is used to calculate operational cladding stress in standard fuel rods.

Several layers of protection due to the nature of operation of an cobalt isotope rod are listed below and were used in the analysis:

0 [

  • Temperature increase at the spacer locations is negligible. As previously discussed, the spacers interface with the solid Zircaloy connectors, not with the cladding.

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NEDO-33529 Revision 0 Non-Proprietary Information S(()) As previously discussed, the spacers interface with the solid Zircaloy connectors, not with the cladding.

  • ((i I))

The nominal cladding outside surface temperature is assumed to be the same as the coolant, 550'F. The only power generated from cobalt isotope rods is from gamma heating.

((r )) The only power generated from cobalt isotope rods is from gamma heating.

Spacer spring force values are negligible. As previously discussed, the spacers interface with the solid Zircaloy connectors, not with the cladding.

Thus it is concluded that GE14i cobalt isotope rods are adequate with respect to cladding stress.

Seismic and Flow Induced Vibration

[r 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 fuel rods and cobalt isotope rods provides assurance that cracking from vibration due to non-uniform weight distribution will not be a significant issue for the cobalt isotope rods. Additionally, the heavier sections of the cobalt isotope rods are the Zircaloy connections, which are all supported laterally by a spacer at the same elevation.

Internal Fret from Inner Capsule Steps have been taken to specifically design this failure mode out of the cobalt isotope rod.

((

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NEDO-33529 Revision 0 Non-Proprietary Information Spacer Location Fretting

[r Mid-Span Fretting This potential failure mechanism of the cobalt isotope rods is mitigated by use of Defender debris filter. However, given the unlikely occurrence that outer cladding is perforated, allowing coolant to contact the inner cladding, the inner cladding provides an additional barrier preventing release of the cobalt targets to the reactor coolant. If the outer container is breached and the inner container is breached, then the two breach points would need to be aligned and of sufficient size to allow the nickel-plated cobalt targets to escape. Beyond this, the nickel coating on the cobalt targets provides a protective barrier against releasing cobalt from the targets to the reactor coolant.

Failures Occurring During Disassembly The failure scenarios for removal of a cobalt isotope rod from a bundle and subsequent handling of the rod are the same as those of a fuel rod. Cobalt isotope rods and rod segments can be dropped or manipulated in a manner that breaches the cladding just as with fuel rods.

Numerous administrative and procedural controls are in place to prevent these failures. The benefit with cobalt isotope rods is that the consequences of these failures are strongly mitigated because there is a second layer of sealed cladding to prevent content release and the escaping contents are not fuel pellets.

If isotope rod segments cannot be readily separated by unthreading the male-female connections as intended, they may be separated by a torque-induced failure of a necked point on the threaded extension of the male threaded connector. As discussed in Section 4.7.3.5, testing has shown that this failure consistently occurs at the intended location, preventing inner capsule and cobalt targets from escaping and locking the male extension into the female connector's threading.

The specific effects of a fuel or rod handling accident in the spent fuel pool have also been considered. Particulate entrainment in fuel pool coolant flow as a result of a fuel handling accident is not analyzed as part of the FHA analysis. Prototype failure testing has shown that failure occurs at a benign point on the male threading of the connectors, not the cladding or 9

NEDO-33529 Revision 0 Non-Proprietary Information welds. From this testing and because the GE14i cobalt isotope 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 cobalt target released would naturally fall to the bottom of the pool because its density is much greater than the density of water. The HCGS 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 cobalt targets can be swept from the bottom of the spent fuel pool into the cooling system.

2.3 Online Failure Detection Section 2.2 discusses multiple cobalt isotope rod design features to mitigate the failure and/or consequences of the failure during operation. Regardless of the failure mode, two layers of Zircaloy cladding and a layer of nickel plating must be breached before cobalt is exposed to reactor coolant. In order for an entire target to escape, the outer cladding and the inner cladding must be breached. Additionally, the two breach points would need to be aligned and of sufficient size to allow a cobalt target to escape.. Beyond this, the nickel coating on the cobalt targets provides a protective barrier against releasing cobalt from the targets to the reactor coolant. Online monitoring methods to indicate if cobalt isotope rod integrity has been significantly compromised (i.e., that cobalt may have escaped from the cobalt isotope rods) are explained below.

The plant chemistry sampling programs provide detection capability to measure significant increases in cobalt-60 activity and take appropriate response, which can include plant shutdown. The rate of activity increase is affected by the amount of cobalt exposed to the reactor coolant; therefore, a catastrophic failure of the cobalt assembly would be readily detectable.

Reactor water sampling procedures for HCGS describe the frequencies of analysis, chemistry control specifications, and corrective actions for reactor water chemistry control. These procedures also define the requirements for the Reactor Water Chemistry control program based on BWRVIP-190, "BWR Water Chemistry Guidelines -2008 Revision", TR-1016579.

These procedures include periodic sampling for cobalt-60 activity.

If an entire target were to become lodged where plant radiation monitors and radiological surveys provide detection capabilities, then appropriate response can be taken, which can include plant shutdown. If the target were to become lodged at a location remote to the plant radiation monitors, significant increases in radioactivity would be detected while performing radiological surveys during operation or shutdown.

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NEDO-33529 Revision 0 Non-Proprietary Information Table 2-1 GE14i and GE14 Fuel Assembly Parameters Fuel Assembly GE14i GE14 Total number of fuel rods R[ )) 92 Total number of cobalt isotope rods [U )) N/A Number of full length (( )) 78 Number of partial length 14 14

(( ](( )) N/A

(( ))[ )) N/A

(( )) F[ )) N/A

[F 1] F[ ]I] N/A Lattice array 1xO0 10xI10 Rod to rod pitch (in) 0.510 0.510 Number of water rods 2 2 Typical assembly fuel weight (IbU)a R[ )) FF ))

Total fuel assembly dry weight (lb) b R[ )) [F ))

Total fuel assembly submerged weight (lb)b FF )) Fr ))

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 Zircaloy-2 Zircaloy-2 Cladding tube diameter, outer (in) (( ))

Cladding tube wall thickness (in) (( Fr Pellet diameter, outer (in) (( )) (( ))

Fuel column stack density (g/cm 3 ) Fr )) (( ))

Fuel column stack density with burnable absorber (g/cm 3) (( )) F[ ]

Water Rod Tube material Zircaloy-2 Zircaloy-2 Maximum tube diameter, outer (in) [F )) (( ))

Tube wall thickness (in) (( )) ((

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

[F 11

NEDO-33529 Revision 0 Non-Proprietary Information Figure 2-1 GE14i Bundle Cutaway View 12

NEDO-33529 Revision 0 Non-Proprietary Information I[

1]

Figure 2-2 GE14i Lattice Arrangement 13

NEDO-33529 Revision 0 Non-Proprietary Information Er 1]

Figure 2-3 Cobalt Target Isometric View Er Figure 2-4 Cobalt Target Orthographic View 14

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[1 Figure 2-5 (( )) Isometric View 11 Figure 2-6 (( )) Orthographic View 15

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((

1]

Figure 2-7 (( )) Isometric View 1[

Figure 2-8 (( )) Orthographic View 16

NEDO-33529 Revision 0 Non-Proprietary Information 1r Figure 2-9 (( )) Isometric View Er Figure 2-10 Er )) Orthographic View 17

NEDO-33 529 Revision 0 Non-Proprietary Information Figure 2-11 11 )) Isometric View 11 Figure 2-12 (( )) Orthographic View 18

NEDO-33529 Revision 0 Non-Proprietary Information Er Figure 2-13 (( )) Isometric View Er Figure 2-14 (( )) Orthographic View 19

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[1

]1 Figure 2-15 (( )) Isometric View 1]

Figure 2-16 (( )) Orthographic View 20

NEDO-33529 Revision 0 Non-Proprietary Information 1[

Figure 2-17 [1 )) Isometric View Figure 2-18 (( )) Orthographic View 21

NEDO-33529 Revision 0 Non-Proprietary Information rr 1]

Figure 2-19 [1 )) Isometric View 1]

Figure 2-20 (( )) Orthographic View 22

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3. Nuclear Design and Methods The sections that follow discuss the core design approach and methods qualifications used to accommodate GE14i bundles in HCGS. Requirements for additional Critical Power Ratio (CPR) margin, limiting cobalt isotope rod adjacent fuel rod enrichment, and including additional shutdown margin are discussed in Section 3.1. Although many of the NRC approved methods used are unaffected by the characteristics of GE14i ITAs, Section 3.2 provides qualification of these methods for the specific application. Section 3.3 discusses the testing and evaluations completed for application of the GEXL14 correlation to GEI4i.

3.1 Nuclear Core Design HCGS will insert 12 GEI4i 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 GE14i fuel assemblies. The Cycle 17 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. Excluding the 12 GE14i bundles, the remainder of the Cycle 17 core will be comprised of all GE14 fuel design.

The applicability of the GEXLI4 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. ((

3.2 Methods This section addresses the applicability of the current methods and methodologies to the GE1 4i 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 GEI4i geometry and characteristics. In particular, the unique characteristics of GEI4i that the methods must address are the impacts of the non-power producing cobalt isotope rods and the impacts of the 23

NEDO-33529 Revision 0 Non-Proprietary Information 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.

The differences between GEI4i and GEl4 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 U0 2 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 3.2.1.3.

Table 3-1 shows the summary of the status of the applicability of codes andmethodologies to GE14i. A more detailed discussion follows.

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 GE14i ITAs. Input necessary to describe the GEI4i ITA lattice design is provided through the standard TGBLA06 input interface.

In GE14i ITAs, all U0 2 and Gd rod material attributes are identical to GEl4. No modifications to the methodology of TGBLA06 were required to model the GEI4i U0 2 and Gd rods. The material characteristics of the cobalt bearing regionsare 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-VII (Revision 0) 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 inReference 16 are provided in Table 3-2 and the control blade worth predictions are provided in Table 3-3. The current GE14 uncertainties for use in safety limit analysis will be used to model the GEI4i ITAs.

The introduction of GEI4i ITAs is accommodated by the TGBLA lattice physics methodology.

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NEDO-33529 Revision 0 Non-Proprietary Information 3.2.1.2 Steady-State Core Simulator PANACI I 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 PANACI I 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 GEI4i and GEI4. Only the number of heated rods is perturbed by the GE14i configuration. The following sections discuss unique features of the GEI4i ITAs and their impact on PANAC 11.

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 U0 2 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 PANACI I are required to model the thermal performance of the GE14i ITAs.

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 PANACI I are required to model the GEl4i ITAs.

3.2.1.2.c Pin Power Reconstruction The influence of zero-power rods on the PANACI I pin power reconstruction model was evaluated and no statistically meaningful differences were observed. Pin power reconstruction impacts in the GE14i ITAs and adjacent fuel assemblies have been reviewed

......... and the pin power reconstruction model was determined to be. adequate...............

No changes in PANACI 1 are required to model the GE14i ITAs.

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NEDO-33529 Revision 0 Non-Proprietary Information 3.2.1.3 ITA Margin Considerations For the HCGS ITA design, additional margins will be applied to the LHGR 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 HCGS 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 PANACI I and MCNP-05.

((

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NEDO-33529 Revision 0 Non-Proprietary Information 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 a GE14i ITA. An additional (( )) Ak shutdown margin in the control rod cell containing one GE14i ITA 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 GE14i ITA with PANAC II and evaluating the change in control rod worth of control blades adjacent to the GEI4i ITAs.

3.2.1.4 IMLTR Limitations and Conditions The Limitations and Conditions described in the Applicability of GE Methods to Expanded Operating Domains Licensing Topical Report (IMLTR) (Reference 18), are applied to the GE14i bundle in the same manner as they are applied to the GE14 bundle. See Appendix A for a description of the Limitations and Conditions.

3.2.2 Thermal-HydraulicMethodology 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 l0xl0 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 GEl4i, can be readily handled by ISCOR09 input quantities.

PANACI 1 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 I is capable of modeling the GEI4i ITA design for the HCGS core.

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

3.2.3 In-Core Instrumentation Given the fact that TGBLA06 correctly accounts for the change in the neutron or gamma flux distribution introduced by the Cobalt isotope rods, the adaption methodology in the core monitoring system is not impacted by the cobalt isotope rods for Local Power Range Monitors (LPRMs) and gamma TIP applications.

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NEDO-33529 Revision 0 Non-Proprietary Information Moreover, since the major contributors to the neutron or gamma detector response are the pins closest to the detector, the (( )) cobalt isotope rods located in the GEI4i bundle have an insignificant impact quantitatively on the instrument signal.

The J-factors are based upon the TGBLA06 calculations and are generated taking into account the quantity and location of the cobalt isotope rods. The gamma energy released from the neutron capture reactions in the cobalt isotope rods is included in the lattice heat generation summation. For LPRMs, thermal J-factors are computed from the TGBLA06 results of the multi-group flux in the instrument gap area. For gamma TIP applications, the gamma transport factors for the different axial GE14i fuel geometry variations "blend" together the specific characteristics of the GEI 4i fuel geometry (i.e., more or less PLRs, channel thickness variations, different thickness in fuel and water rods, etc.). The gamma source from the neutron capture reactions in the cobalt isotope rods is also included in the generation of the gamma TIP J-factors.

Therefore, the GE14i ITAs will not have a significant impact on the in-core instrumentation and core monitoring system.

3.2.4 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 GEI4i and was observed to be consistent with the RMS fission density uncertainty of GE14. The results of this evaluation at zero exposure can be found in Table 3-2. The GEl4 uncertainties will be used in the evaluation of the safety limit analysis for the HCGS safety limit evaluations. The capability of GESAM02 to model CPR related uncertainties is adequate for the HCGS GEI4i ITAs and is not impacted by the number of heated rods in GE14i.

3.2.5 TransientAnalysis Methodology The impact of the (( )) zero-power rods in the GE14i design will be reflected in the data that propagates through the PANACI I core model to the transient analysis methods. The GEI4i design characteristics will be used in the transient analysis methods ODYNMIO and TASC-03. The impact of the (( )) zero-power rods in each of the 12 HCGS GE14i ITAs will not challenge the adequacy of the transient analysis methods.

3.2.6 Stability Methodology ODYSY05 obtains the GEI4i geometry information from the ISCOR system and provides adequate results for the GEI4i ITAs. TRACG04 uses the PANACI I kinetics model and receives fuel neutronic information (nodal cross sections) through the PANACI I wrap-up 28

NEDO-33529 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 GEI4i bundle characteristics. The (( ))

zero-power rods in the HCGS GE14i ITAs will not challenge the adequacy of the stability methods.

3.2.7 Fuel Rod Thermal-MechanicalMethodology The design of the U0 2 and Gd rods in the GE14i ITAs is identical to the GEI4 and will therefore have no impact on the GSTRM07 methodology.

3.2.8 ECCS-SAFER/GESTR The Emergency Core Cooling System (ECCS) analysis methodology applicable to HCGS is SAFER/GESTR.

The zero-power rods can be described through 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 GEI 4 U0 2 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 r[ )) zero-power rods in the HCGS GEI4i ITAs 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 onthe specific GEXL14 correlation developed for the GE14 fuel and the R-factor methodology is given in*

Reference 6.

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NEDO-33529 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 GEXL14 correlation has been demonstrated to be an accurate predictor of the GEI4 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 the GEl4i ITAs. Due to the similarity between GE14i and GEl4, the currently approved GEXL14 correlation can be applied to the GEI4i ITAs, provided that the effects of the difference on the critical power performance are quantified and properly accounted for. The applicability of the GEXL14 correlation to the GEI4i ITAs is demonstrated by comparing the GEXL14 prediction to the critical power (CP) data with zero-power rods in the GE14 bundle.

Full-scale critical power and pressure drop testing for a simulated GE14 fuel bundle was performed in the Stern Laboratories test facility in Hamilton, Ontario. As a part of the Stern testing for the GE14 fuel, CP 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 a 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 3-6, where target rod(s), the highest R-factor rod(s), of each pattern were identified with a green background color. The peaking patterns JI/J2/J3 have (( )) zero-power rods and pattern DOxx has (( )) zero-power rods. Mass flux and inlet subcooling conditions are plotted in Figure 3-7. Typical bundle axial power shape is presented in Figure 3-8. The Stern Laboratories test assembly characteristics are provided in Reference 6.

The database used in the comparison is adequate to confirm the applicability of the GEXL14 correlation and the R-factor methodology to the GE14 ITAs. It has been shown that the various fuel assembly and channel geometries such as cold rods (water rods) or vacant lattice positions due to the part length rods are well characterized by the R-factor methodology in the GEXL correlation. ((

The use of a single axial power shape data only to validate the use of GEXLI4 to GEI4i is justified.primarily based on the prior experience in GEXL correlations, which had. shown that the critical power data correlated well in the critical quality and boiling length plane independent of the axial power profiles. The axial power shape sensitivity has been well predicted by the GEXL correlation for a wide range of different designs such as the lattice design (9x9, IWxlO), the part length rod configuration (length, number and location), and spacer designs (type, number, pitch). The GEXL14 correlation was demonstrated to be an 30

NEDO-33529 Revision 0 Non-Proprietary Information accurate predictor for GE14 fuel for different axial power shapes as discussed in detail in Reference 6 A statistical analysis was performed for the GE14 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 GEXLI4 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 CPs. are compared to the measured CPs in Figure 3-9. It is shown from the mean ECPR that the GEXL14 correlation conservatively predicts the CP data with zero-power rods. ((

In summary, the applicability of the GEXL14 correlation to the GEl4i ITAs was demonstrated by comparing the GEXLI4 prediction to the CP data with zero-power rods in the GE14 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 evaluationof.the_ applicability. of _GEXLl4 to GE14i. As such, the R-factor methodology is confirmed applicable to GEI4i. The GEXLI4 correlation, on average, conservatively predicted the critical. powers, for the zero-power rod test data obtained at Stern Laboratories for. the GE 14 bundle with ((

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NEDO-33529 Revision 0 Non-Proprietary Information 32

NEDO-33529 Revision 0 Non-Proprietary Information Table 3-1 Summary of GNF Methods Applicability to GE14i Methodology Analysis Code and Revision Supported TGBLA06 X Nuclear PANACII X Thermal Hydraulic ISCOR09 X Safety Limit MCPR GESAM02 X ODYNMIO X Transient Analyses TASC-03 X ODYSY05 X Stability TRACG04 X TASC-03 X ATWS ODYNMIO X Thermal-Mechanical GSTRM07 X LAMB-08 X ECCS-LOCA TASC-03 X SAFER04 X 33

NEDO-33529 Revision 0 Non-Proprietary Information Table 3-2 Lattice RMS Fission Density Uncertainty Comparison at Zero Exposure In-Channel Void GEI4i GE14 Fraction 00%

40%

70%

Average ))

Table 3-3 GE14i Control Blade Worth Comparison MCNP TGBLA %Delta In-Channel Void Fraction (Kun,-Ko.)/K., (Kun,-Ko.)/Kun (WorthTcaLA-WorthMcNP)

/ WorthMCNP Cold 00%

40%

70% ]

Table 3-4 Internal PANAC1l Parameters Parameter Values GE14 GE14i Number of Fuel Rods (NBFURD) 92 (( ))

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

Hydraulic Diameter (in) 34

NEDO-33529 Revision 0 Non-Proprietary Information Table 3-5 GEXL14 Statistics for GE14 Critical Power Data with Zero-Power Rods Number of Data Points Mean ECPR Standard Deviation 35

NEDO-33529 Revision 0 Non-Proprietary Information 1[

Figure 3-1 TGBLA06 Cobalt 60 Inventory per Linear cm for GE14i for 00% In-Channel Void History 36

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[F Figure 3-2 TGBLA06 Cobalt 60 Inventory per Linear cm for GE14i for 40% In-Channel Void History 37

NEDO-33529 Revision 0 Non-Proprietary Information 11 Figure 3-3 TGBLA06 Cobalt 60 Inventory per Linear cm for GE14i for 70% In-Channel Void History 38

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((

Figure 3-4 Power Spike in Fuel Rods Adjacent to Cobalt Isotope Rod Connectors at 00% In-Channel Void Fraction 39

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[Fi

]

Figure 3-5 Power Suppression Analysis for Rods (3,1) and (2,2) in Dominant Zone 40

NEDO-33529 Revision 0 Non-Proprietary Information 1r Figure 3-6 Rod-to-Rod Power Distributions with Zero-Power Rods 41

NEDO-33529 Revision 0 Non-Proprietary Information 11 Figure 3-7 GEXL14 Test Conditions (P=1000 psia) 42

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]

Figure 3-8 Typical Bundle Axial Power Shape Used for GEXL14 Testing 43

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Figure 3-9 Calculated Versus Measured Critical Power for GEXL14 44

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4. Licensing Evaluations The ITAs will be monitored during operation via the process computer. 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 HCGS Reload 16 Cycle 17 to establish fuel operating limits for the ITAs that assure compliance with regulatory limits. Results of these analyses will be documented in the HCGS Reload 16 Cycle 17 Supplemental Reload Licensing Report (SRLR). Furthermore, normal reload licensing analyses will be performed for each cycle of operation with ITAs, 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.

PSEG intends to insert the GEI4i ITAs into HCGS and to operate them in Cycle 17 and subsequent cycles. Cycle specific analyses to establish fuel operating limits are not yet complete. When the cycle specific analyses are complete, GNF will document the results in the SRLR, and PSEG will update the HCGS 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 GE14i ITAs is discussed in Section 3.2. Because the analysis of the ITAs will meet the approved criteria, NRC approval of the cycle specific SRLR and COLR is not required prior to insertion.

The list of events supporting the HCGS Cycle 17 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 operational 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 HCGS Cycle 17 reload licensing analyses will include specific modeling of the GE14i ITAs in the determination of the OLMCPR. As discussed in Section 3.3, GEXL14 is.

demonstrated to conservatively apply to the GEI4i ITAs. The GE14i ITA U0 2 and Gd fuel rod mechanical designs are identical to the GE14 fuel rods (Section 3.2.7) and, therefore, the 45

NEDO-33529 Revision 0 Non-Proprietary Information normal GEI4 thermal and mechanical overpower LHGR limits ensure compliance with thermal-mechanical licensing requirements as specified in Reference 2.

The HCGS abnormal operational transients evaluated to support the introduction of GEI4i ITAs into HCGS are identified in the following subsections.

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

" Pressure Regulator Failure - Open (PRFO)

  • Inadvertent RHR Shutdown Cooling Operation The HCGS Cycle 17 reload licensing analyses will include specific modeling of the GE14i ITAs in determination of the OLMCPR. Plant characteristics, which include the steam line volume, steam line pressure losses, Turbine Control Valve (TCV) / Turbine Stop Valve (TSV) closure times, scram time, and the associated trip and delay times impact the core average response of the limiting events in this category. Such plant parameters are independent of fuel bundle design and are modeled by methods discussed in Section 3.2.5. The input to the ODYN model is updated on a cycle-specific basis, if needed, to incorporate the most recent HCGS plant characteristics. Therefore, in Cycle 17, as in HCGS Cycle 16 (Reference 7), the Inadvertent RHR Shutdown Cooling Operation, PRFO and Inadvertent Main Steam Relief Valve Opening events are bounded by the LFWH and FWCF events.

The transient response is affected by the core average reactivity characteristics. However, the introduction of GE14i ITAs has a negligible impact on the core average nuclear parameters affecting the transient response because the 12 GEI4i ITAs represent a small fraction of the total bundles in the core, and the hydraulic characteristics of the GE14i ITAs are similar to the GEl4 bundles (Section 3.2.2). Therefore, the GEl4 bundles dictate the core average nuclear parameters that affect the transient response. Consequently, for Cycle 17, as in Cycle 16 (Reference 7), only the FWCF and LFWH events will be analyzed as part of the cycle-specific reload licensing analyses, and it can be concluded that the reload licensing scope summarized in Table 4-1 bounds all other AOOs in this category.

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NEDO-33529 Revision 0 Non-Proprietary Information 4.1.2 Increasein Reactor Pressure The events in this category are:

" Pressure Regulator Failure - Closed (PRFDS)

  • Generator Load Rejection with Bypass (LRWBP)
  • Generator Load Rejection with Bypass Failure (LRNBP)

" Turbine Trip with Bypass Failure (TTNBP)

" Main Steam Isolation Valve Closures

  • Loss of Alternating Current (AC) Power

The HCGS Cycle 17 reload licensing analyses will include specific modeling of the GE14i ITAs. The GEI4i ITAs do not impact the core average response of the limiting events in this category because the core average nuclear characteristics are dictated by GEI4 bundles, as discussed in Section 4.1.1. The TTNBP and LRNBP events are analyzed as part of the cycle-specific reload licensing analyses.

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

  • Recirculation Flow Control Failure - Decreasing Flow The AOOs in this category are bounded by the events in the reload licensing analysis scope 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 GEI4i ITAs. Therefore, as in HCGS Cycle 16 (Reference 7), none of the AOOs in this category will be analyzed as part of the Cycle 17 specific reload licensing 47

NEDO-33529 Revision 0 Non-Proprietary Information analyses due to the introduction of GE14i ITAs. The Single Loop Operation (SLO) Pump Seizure accident is discussed in Section 4.2.3.

4.1.4 Reactivity and Power DistributionAnomalies The events in this category are:

" Rod Withdrawal Error (RWE)

" Control Rod Maloperation (System Malfunction or Operator Error)

  • Mislocated Fuel Assembly Accident

" Misoriented Fuel Assembly Accident

  • Abnormal Startup of Idle Recirculation Loop

" Recirculation Flow Control Failure with Increasing Flow The RWE and the Misoriented Fuel Assembly Accident are potentially limiting events at HCGS. The RWE bounds the Control Rod Maloperation event. The RWE and Misoriented Fuel Assembly Accident events have the potential to set an OLMCPR between the Beginning of Cycle (BOC) and Middle of Cycle (MOC) exposure range (Reference 7). The RWE and Misoriented Fuel Assembly Accident will be analyzed as part of the cycle-specific reload licensing analyses. As in Cycle 16 (Reference 7), the Mislocated Fuel Assembly Accident will be evaluated as part of the HCGS Cycle 17 reload licensing analyses. The Control Rod Maloperation event will not be analyzed as part of Cycle 17 reload transient analysis. The HCGS Cycle 17 reload licensing analyses will include specific modeling of the GE14i ITAs.

In Cycle 17, as in Cycle 16 (Reference 7), the off-rated limits (Section 4.2.4) will continue to bound the Abnormal Startup of Idle Recirculation Loop event and the Recirculation Flow Control Failure with Increasing Flow event which results in a fast recirculation flow runout, due to reasons specified in Section 4.1.1. 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 HCGS Cycle 17.

4.1.5 Increasein Reactor CoolantInventory The event in this category is:

  • Inadvertent High Pressure Coolant Injection (HPCI) Start-up The severity of the HPCI event is affected by plant parameters, which include- the steam line volume, steam line pressure losses, TCV/TSV closure times, scram time and HPCI system flow capacity. Such plant parameters are independent of fuel bundle design and are modeled by methods discussed in Section 3.2.5. The input to the ODYN model is updated on a cycle-specific basis, if needed, to incorporate the most recent HCGS plant characteristics.

Therefore, as in HCGS Cycle 16 (Reference 7), the Inadvertent HPCI Start-up event will 48

NEDO-33529 Revision 0 Non-Proprietary Information continue to be bounded by the cycle-specific reload licensing analyses scope tabulated in Table 4-1.

4.1.6 Decreasein Reactor CoolantInventory and Other Accidents The events in this category are:

  • 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 a LOCA, as a result of the introduction of GE14i 1TAs, are discussed in Section 4.5.6. 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 HCGS 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, which include the ATWS - Recirculation Pump Trip (RPT), Safety/Relief Valve, and Standby Liquid Control System (SLCS) operating parameters. The key core average nuclear parameters that affect the core response to an ATWS event are described in Reference 1. The GEI4i 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 HCGS Cycle 16 (Reference 7); a cycle-specific ATWS analysis is not required because of the introduction of GEI4i ITAs.

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NEDO-33529 Revision 0 Non-Proprietary Information The fuel and cycle-independent ATWS evaluation for HCGS is documented in Reference 8.

This evaluation demonstrates significant margin to the aforementioned ATWS acceptance criteria.

4.2.2 ASME OverpressureProtection 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. Scram is assumed to occur on high neutron flux in the reactor core. The assumption causes a delay in the scram. Consequently, the average system response for an MSIV closure with a flux scram is inherently worse than the events discussed in Section 4.1.2 that are terminated with a direct scram. As discussed in Section 4.1.1, the presence of 12 GEl4i-ITAs does not impact plant characteristics such as the scram delay time or the core average nuclear characteristics. Therefore, as in Cycle 16 (Reference 7), the MSIVF event will continue to bound all other pressurization events for overpressure protection and will be analyzed as part of the cycle-specific reload licensing analysis that will include specific modeling of the GEI4i ITAs.

4.2.3 Single Loop OperationPump Seizure Analysis This Single Loop Operation (SLO) Pump Seizure event was analyzed for GEl4 introduction into HCGS. ((

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NEDO-33529 Revision 0 Non-Proprietary Information 4.2.4 Applicability of Off-Rated Limits to GE14i 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, plant-specific off-rated limits. , 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 GE14i ITAs are similar to the GE14 bundles.

As such, the power and flow dependent limits are applicable to the GE14i ITAs. The off-rated limits are confirmed as part of the cycle-specific reload licensing analyses for HCGS Cycle 17.

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 GE14i ITAs is similar to those of GE14 fuel. In addition, the impact on the core average parameters that affect the transient response is negligible because the GE14i ITAs represent a small fraction of the total bundles in the core.

Consequently, the introduction of GE14i ITAs will not impact the flexibility and EOOS options supported for HCGS listed below:

" Maximum Extended Load Line Limit Analysis (MELLLA) (94.8% core flow at rated power)

  • Increased Core Flow (ICF, 105% at rated power),
  • One recirculation pump out-of-service or Single Loop Operation (SLO)
  • Recirculation Pump Trip Out-of-Service

NEDO-33529 Revision 0 Non-Proprietary Information The fast pressurization events ih combination with licensed flexibility and EOOS options will be evaluated as part of the reload transient analysis for HCGS Cycle 17. The GE14i transient analysis results will be summarized in the Supplemental Reload Licensing Report (SRLR) for HCGS Cycle 17.

4.3 Evaluation of Design Basis Accidents The HCGS Design Basis Accidents (DBAs) to be evaluated are identified in Chapter 15 of the HCGS Updated Final Safety Analysis Report (UFSAR). 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 HCGS licensing basis CRDA analyzed in Section 15.4.9 of the HCGS UFSAR assumes a failure of 850 rods (8x8 fuel). The mass fraction of fuel in the damaged rods that reaches or exceeds the initiation temperature of fuel melting is estimated to be 0.77%. Fuel reaching melt conditions is assumed to release 100% of the noble gas inventory and 50% of the iodine inventory. [

)) Therefore, the licensing basis CRDA radiological analysis is not impacted by the introduction of 12 GE14i assemblies at HCGS.

As described in Reference 9, 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 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 cal/gm fuel design limit. The reactivity characteristics of GE14i are similar to GEl4; therefore, the introduction of 12 GE14i assemblies at HCGS 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.

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NEDO-33529 Revision 0 Non-Proprietary Information 4.3.2 Main Steam Line Break Accident The HCGS licensing basis MSLB analyzed in Section 15.6.4 of the HCGS UFSAR 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 12 GE14i assemblies at HCGS.

4.3.3 Fuel HandlingAccident 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 HCGS licensing basis FHA is analyzed in Section 15.7.4 of the HCGS UFSAR. The licensing basis FHA postulates that an irradiated 8x8 fuel bundle is dropped 32.95 feet onto the reactor core and fails 124 rods. Of the failed rods, 8% of the 1-131, 10% of the Kr-85, 5%

of the other noble gases and halogen inventories, and 12% of the alkali metal inventory of the damaged rods are released from the rods. All other particulates are retained by the water.

Reference 1 documents that radiological consequences from a FHA involving the GE14 design are bounded by consequences from a FHA involving the 8x8 fuel design. ((

)) Therefore, the licensing basis FHA radiological analysis is not impacted by the introduction of 12 GEJ 4i assemblies at HCGS.

4.3.4 Loss-of-Coolant Accident (LOCA)

The HCGS LOCA source term was previously evaluated for an Extended Power Uprate (EPU). The impact of 12 GE14i assemblies on the EPU LOCA source term and radiological consequences was evaluated.

((I 53

NEDO-33529 Revision 0 Non-Proprietary Information The introduction of 12 GEI4i bundles at HCGS presents no significant impact on the AST LOCA source term.

4.4 Thermal-Mechanical Evaluation Thermal-mechanical characteristics of GE14i cobalt isotope rods were evaluated. For the GEI4i cobalt isotope rods, thermal-mechanical evaluations were performed ((

)) 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:

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NEDO-33529 Revision 0 Non-Proprietary Information 1.

2.

3.

4.

5.

GSTRM models (Reference 19) were used at multiple points during this evaluation ((

)) However, due to differences in geometry between the GE14i cobalt isotope rods and standard fuel rods, ((

were outside of the GSTRM application range.

)) has no impact on the results of the analyses and no additional margin is applied to the downstream safety analyses.

((

- ))

4.5 - Other Evaluations The results of other evaluations required to support the loading of the GEI4i ITAs are provided below.

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NEDO-33529 Revision 0 Non-Proprietary Information 4.5.1 Stability This section provides a qualitative assessment of the impact of 12 GE14i ITAs on thermal-hydraulic instability. In accordance with References 9 and 10, a review was done to ensure 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 I1(Reference 9).

A qualitative assessment was performed based on the GE14i bundles to evaluate the impact on decay ratio. With ((

differences between the GE14 and GEI4i channel decay ratio performance ((

Because a small number of GEl4i bundles are loaded, and since the hydraulic characteristics of the GEI4i bundles approximate the GE14 bundles, ((

HCGS is an Option III plant, and the plant will continue to use the Option II system for Cycle 17. For the Option III stability solution, two stability aspects must be considered. The first consideration is the Oscillation Power Range Monitor (OPRM) setpoint; the second is the size of the Backup Stability Protection (BSP) regions.

The OPRM setpoint with the GEl4i bundles included is expected to be comparable to that of GEl4 because ((

)) The Delta CPR over Initial MCPR Versus 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 two-recirculation pump 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. During the fuel design and licensing process, an appropriate OPRM setpoint is calculated. The OPRM setpoint is not expected to be affected by the introduction of 12 GEI4i ITAs in HCGS Cycle 17 beyond cycle-to-cycle variation.

An assessment indicates that the BSP regions are not expected to vary significantly with the substitution of the GEI4i design for the GEl4 design since ((

Moreover, the BSP regions are typically limited by characteristics of the full core, most of which remains GEI4 fuel. A stability evaluation'was performed and confirms this assessment. During the fuel design and licensing process, appropriate BSP regions are calculated addressing flexibility options. The size of these BSP regions is not expected to be affected by the introduction of 12 GEl4i ITAs in HCGS Cycle 17 beyond cycle-to-cycle variation.

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NEDO-33529 Revision 0 Non-Proprietary Information 4.5.2 Decay Heat Assessment A comparative core decay heat assessment between GE14 and GE14i fuel types was performed ((

)) It was concluded that replacing 12 GEl4 bundles with 12 GE14i bundles presents no significant change in core decay heat at HCGS.

4.5.3 Appendix R Safe Shutdown Fire The limiting safety shutdown event for HCGS Appendix R fire protection in Reference 8 is mitigated with Reactor Core Isolation Cooling (RCIC), Safety Relief Valves (SRVs) and Residual Heat Removal (RHR) from a remote shutdown panel. As shown in Section 4.5.2,

)) 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 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 GEI4i ITA.

57

NEDO-33529 Revision 0 Non-Proprietary Information 4.5.4 Station Blackout As shown in Section 4.5.2, E[

)) provided other key parameters determining containment response such as, E[

)) do not change for GE14i ITAs. ((

)) for the licensing basis Station Blackout (SBO) analysis in Reference 8.

4.5.5 ContainmentResponse As shown in Section 4.5.2, ((

)) provided other key parameters determining containment response such as E[

)) do not change for GE14i ITAs. E[

)) including the Recirculation Suction Line Break (RSLB) and the NUREG-0783 evaluation documented in Reference 8. In addition, E[

4.5.6 ECCS LOCA As stated in Section 3.1, GE14i ITAs will be loaded into non-limiting locations with respect to thermal limit margins, including ECCS LOCA Maximum Average Planar Linear Heat Generation Rate (MAPLHGR) limits. The number of fueled rods will be (( )) less in the GEI 4i ITA bundle relative to the GE14 bundle, and MAPLHGR will not be averaged over the zero power rods. Therefore, the ECCS LOCA MAPLHGR limits in Reference 7 for GE 14 remain bounding for GE 14i 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 ITAs 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 GEI4i ITAs. The licensing basis PCT for GE14 reported in Reference 7 remains applicable for GEI4i ITAs.

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NEDO-33529 Revision 0 Non-Proprietary Information 4.5.7 Reactor InternalPressureDifference 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 GE14i bundle closely matches the overall pressure drop of previous designs. The pressure drop characteristics of the GE14i fuel are equivalent to those of the GE14 fuel and the method applied to the GE14 fuel pressure drop analysis is valid for the GE14i application. Therefore, the current pressure drops for reactor internals based on GE 14 (Reference 8) remain applicable for GEI 4i ITAs.

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 GEl4. 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 4.5.9.

The maximum CRGT lift force is determined by ((

)) These parameters do not change for GEI4i ITAs and thus the maximum CRGT lift force for GEI4 remains applicable for GE14i ITAs.

The introduction of 12 GE14i 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 GE14i ITAs.

4.5.8 'Seismic and Dynamic Response. .. .

Due to thenegligible full core weight variation ((( )) impact), the seismic/dynamic behavior of the core, the reactor internals, the balance of plant, and the primary structure will not be affected by the introduction of 12 GE14i ITAs. The dominant fuel type, GEl4 fuel, 59

NEDO-33529 Revision 0 Non-Proprietary Information dictates the seismic behavior of the core. The minor overall fuel bundle mass difference will not impact the seismic adequacy. The maximum fuel applicable accelerations increase to

(( )) G for horizontal and (( )) G for vertical, which are within the (( )) G horizontal and (( )) G vertical allowable peak seismic accelerations.

Dynamic fuel lift load analysis is not required for HCGS in accordance with the Mark I containment licensing basis.

4.5.9 Reactor InternalsStructuralEvaluation A qualitative structural assessment of the reactor internal components was performed with respect to the current design basis evaluation. The evaluation in Section 4.5.7 demonstrates..

that the current pressure drops for reactor internals remain applicable for GE 14i ITAs and that GE1 4i ITA fuel has no effect on the acoustic and flow-induced loads.

The evaluation in Section 4.5.8 demonstrates that operation with GEI4i ITAs will have no adverse effect on the structural integrity of the reactor internals relative to seismic loading.

All applicable Normal, Upset, Emergency, and Faulted condition loads for GEI4i ITAs such as seismic and dynamic loads, acoustic and flow induced loads, RIPDs, system flow loads, core flow loads, and thermal loads, as appropriate, were considered in the assessment. These 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 GE14i ITAs will have an insignificant effect on the structural integrity of the reactor internal components.

4.5.10 RecirculationSystem Evaluation An evaluation of the effects of introducing GE14i fuel on Reactor Recirculation System (RRS) performance for HCGS 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 I (Reference 11).

For the recirculation system, the primary impact of introducing GE14i fuel is a core pressure drop change. The evaluation results show that the core pressure drop does not change with the introduction of the GE14i fuel bundles. The values of the recirculation system pressures, temperatures, pump flow rate, and motor brake horsepower remain the same when there is no change in the core pressure drop.for a given core flow. There is no change to the recirculation pump required or available Net Positive Suction Head (NPSH) since the pump flow rates and 60

NEDO-33529 Revision 0 Non-Proprietary Information recirculation system pressures/temperatures are the same values as before GE14i fuel introduction.

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

4.5.11 Neutron FluenceImpact RPV fluence is negligibly impacted by the introduction of 12 GE14i ITA bundles. Given that the reactor power is unchanged and the core-wide void and relative power distribution remains approximately the same, the introduction of GEI4i fuel into the core will not significantly impact the magnitude of the RPVflux..

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 12 ITA bundles is only a small fraction of the total core loading of 764 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, GE14i fuel uses the same part length rod design as GE14; 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 HCGS.

4.5.12 Hydrogen Injection The evaluation performed with regard to Institute of Nuclear Power Operations (INPO)

Operating Experience report OE27774 (Reference 12) estimates the impact of inserting cobalt isotope rods in the core of the reactor on the downcomer gamma dose rates. OE27774 describes an incident where a core design change at Fermi 2, a moderate Hydrogen Water Chemistry (HWC) plant that does not use Noble Metal Chemical Addition (NMCA), 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 stress corrosion cracking (SCC) mitigation compared to the previous cy'cle of operation. Therefore, a key objective of the evaluation of OE27774 relative to GE14i was to determine whether there is any potential for a 61

NEDO-33529 Revision 0 Non-Proprietary Information decrease in gamma flux in the downcomer region of the reactor as a result of the cobalt isotope rod introduction.

BWRs that employ NMCA improve the efficiency of hydrogen injection, which allows the plant to operate at lower hydrogen injection rates. These plants do not rely on the downcomer dose rate to catalyze the hydrogen-oxygen recombination. Rather, the noble metals catalyze the reaction at the metal surface. Therefore, there is no negative impact on the hydrogen requirements for SCC mitigation with the introduction of 12 GEI4i ITAs into HCGS, a BWR with NMCA plus HWC.

4.5.13 Post-LOCA Hydrogen Control An evaluation was performed on the HCGS Post-LOCA Combustible Gas Control System with respect to the addition of 12 GE14i ITA bundles for Cycle 17. HCGS has incorporated the requirements of the revised 10 CFR 50.44 (68 FR 54123, dated September 16, 2003) which does not define a design basis LOCA hydrogen release and consequently eliminates the requirements for hydrogen control systems to mitigate such releases. These new requirements have been implemented at HCGS per Amendment No. 160 to Facility Operating License No.

NPF-57 (Reference 17). No further changes or evaluations are required for the HCGS Post-LOCA Combustible Gas Control System to support the addition of 12 GE14i ITAs.

4.5.14 Fuel Storage CriticalitySafety This evaluation addresses the introduction of GE14i ITA fuel at HCGS. Analyses have previously been performed to introduce GEl4 fuel to HCGS. These analyses address all fuel storage racks that are used at HCGS. The original analyses evaluated the peak reactive GEl4 lattice that meets the fuel storage rack reactivity safety limits at a maximum bounding uniform enrichment of no less than 4.9 wt% 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 (( )) regular fuel rods have been replaced with (( -. J] -

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 cobalt isotope rods. As allowed by fuel and core design constraints, the displaced enrichment may either be removed from the 62

NEDO-33529 Revision 0 Non-Proprietary Information assembly entirely or it may be placed in other locations within the same bundle or within bundles not utilizing cobalt isotope rods.

The maximum bounding uniform enrichment of no less than 4.9 wt% U-235 assumed in the original GEI4 models ensure that the models are insensitive to the spatial distribution of fissile material. In this way, the potential enrichment displacement proposed by the GEI4i ITA is already conservatively factored into the original GE 14 models. For these reasons, the GE14 fuel storage rack reactivity safety limits, including k. design limits, are appropriate for use with GE I4i ITAs.

4.5.15 Fresh Fuel Shipping In order to be shipped in the RAJ-11 container, the GE14i bundle must be shown to meet the technical shipping requirements specified inthe RAJ-I1 Certificate of Compliance (Reference 13). 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 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 U0 2 rods in Chapter 6 of the RAJ-II Safety Analysis Report (SAR) which bounds the cobalt isotope rods from a criticality safety standpoint. Therefore, although the GEI4i bundle with (( ))

total fuel rods is outside the range of 91-100 as specified in Table 3 of the Certificate of Compliance (Reference 13), this condition is bounded by what has been analyzed and demonstrated to be safe in the RAJ-I1 container under both Normal Conditions of Transport (NCT) and Hypothetical Accident Conditions (HAC).

Additionally, the GE14i bundle gadolinium rod requirements specifying the required minimum number of Gd rods at a minimum Gd rod wt% as a function of U-235 average lattice enrichment for shipment in the RAJ-II shipping container will be identical to those specified for the GE14 bundle. Since the five cobalt isotope rod locations are analyzed as containing enriched U0 2 rods in Chapter 6 of the RAJ-I1 SAR, this bounds the cobalt isotope rods "from a criticality safety standpoiht under both Normal Conditions Iof Transport (NCT).

and Hypothetical Accident Conditions (HAC).

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NEDO-33529 Revision 0 Non-Proprietary Information 4.5.16 Fuel ChannelDistortion Channel distortion that can cause channel interference is a function of 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 GEI4i bundles will be the same as of GEl 4 bundles.

4.5.17 Fuel Conditioning Guidelines There is no change to the Operating Practices to Reduce Risk of Duty-Related Fuel Failures 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.

4.5.18 Emergency OperatingProcedureData Table 4-2 provides the GE14i data for revising the applicable emergency operating procedures.

4.6 Manufacturing Quality Assurance All aspects of the GEl4i ITA program are controlled under the GE Nuclear Energy Quality Assurance Program Description (Reference 14).

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 U0 2 pellets. The controls include material traceability, material handling, hydrogenous material control, foreign material exclusion, and discrepant material control. During the fabrication process for the rod segments, the only production parts exposed to the nickel plated targets are the inside of the inner tubing and the target placement rod. Both the nickel plating of the targets and the lack of exposure to external cobalt isotope rod components (outer tubing, male threaded connectors, female threaded connectors, upper endplug extensions and lower endplug extensions) assure that no trace material transfer of cobalt is found on the portions of the rods in contact with reactor coolant.

Furthermore, the handling and loading of the cobalt targets occurs in a laboratory separate from the assembly of all other bundle components. This ensures that stray cobalt targets cannot find their way into other bundle components.

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 64

NEDO-33529 Revision 0 Non-Proprietary Information specific to the fabrication of the cobalt isotope rods, focusing on these critical to quality aspects of the fabrication process.

Cobalt targets are certified to meet the current drawing and specification requirements. The cobalt targets are verified to meet the design requirements on a sampling basis. Destructive sampling is performed per GNF quality assurance procedures. ((

Weld parameters for both internal and external rod segments are fully qualified prior to production and certified weld operators perform all production welding. Rod integrity is verified by helium leak check of both the inner and outer tubes following welding. Outer segment welds are also 100% verified by Level II 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 The subsequent sections discuss considerations for the GE14i ITAs following in-core operation. Section 4.7.1 deals with the effects of the ITAs on the physical components of the spent fuel pool and the fuel pool cooling and cleanup system. Section 4.7.2 discusses environmental dose considerations. Section 4.7.3 focuses on the disassembly processes, radiation shielding, and personnel dose associated with the ITAs. In addition, Section 4.7.3 addresses disassembly tooling, cask usage and source tracking considerations. Finally, Section 4.7.4 outlines the Post-Irradiation Examination plan for the ITAs. The detailed information and procedures in Sections 4.7.2, 4.7.3, and 4.7.4 are provided as a general description.

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 HCGS spent fuel pool (SFP), provided guidance for storage of GEI4i 65

NEDO-33529 Revision 0 Non-Proprietary Information bundles is followed to minimize the effect of gamma heating on the SFP concrete walls.

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

The introduction of GE14i fuel to the HCGS SFP 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 SFP. In both of these first two cases, the extra radiation from the cobalt isotope rods was conservatively added to the radiation in a "normal" GEI4 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 12 GEI4i bundles in an offload is (( )) after shutdown over that of an offload of all GE14 fuel. The current heat load calculated for refueling conditions from HCGS is 17.2 MBTU/hr, representing a margin of approximately 9.6% under the FPCC system heat removal capacity of 19 MBTU/hr. Adding ((

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

The gamma radiation effect on the SFP walls was evaluated for the case that the GE14i bundle is placed one, four, and six feet from the SFP wall. In the GEI4i 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/cm 2 /sec. The maximum energy deposition rate due to a GEI4i bundle placed one foot from the SFP walls is approximately 7.2E+10 MeV/cm 2/sec, so concrete heating due to gamma would be significant. At four feet, the energy deposition rate is 1.4E+8 MeV/cm 2/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 SFP, one foot from the wall, after three 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 GE14i bundle can be stored in the SFP, provided the bundle is stored at least one foot from the SFP wall to avoid integrated dose effects. Note that the four foot limit for gamma heating will be more limiting for storage locations.

The HCGS spent fuel storage procedures should be modified to specify that the GE14i bundles be stored at least four feet from the SFP walls. With the four foot distance 66

NEDO-33529 Revision 0 Non-Proprietary Information requirement in effect, there is no limitation on the amount of time a GEI4i bundle may remain in the SFP.

The GEI4i rods are clad with the same material as the GEI4 rods so that there will be no appreciable difference in the corrosion products from GEI4i versus GEl4. Therefore, there will be no adverse effect on the cleanup portion of the FPCC system.

4.7.2 EnvironmentalDose Considerations An evaluation was performed on the effects of dose from cobalt isotope rods on refueling equipment. For the refueling equipment, the dose rate from gamma radiation contained within each cobalt isotope rod is 0.02403E-3 R/hr. For all (( )) cobalt isotope rods in a single GE14i bundle, the dose rate would be less than 0.5 mR/hr. This is the dose rate calculated 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 (HCGS UFSAR 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 would be less than 0.1 R. 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 and electronic components. The radiation threshold is defined as the lowest radiation dose that 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, metals and electronic components. Of the above materials, Teflon TFE has the lowest dose threshold which is in the 2E+4 R range. All others are greater than 1E+5 R.

With regard to electronic components, some integrated circuits have damage thresholds in the 200 R range. As such, a radiation dose of less than 0.1 R is insignificant compared to the radiation threshold of all materials and components.

This total conservative dose for a 7-day exposure is well below the radiation threshold of the materials in the refueling equipment. This level will not affect the 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.

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NEDO-33529 Revision 0 Non-Proprietary Information 4.7.3 Post-IrradiationHandling

4. 7.3.1 Post IrradiationBundle 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 GEI4i fuel assembly and replaced 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 GEI4i 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 GE14i fuel assemblies between power cycles for additional irradiation time.

4.7.3.2 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. 7.3.3 Segmented Rod DisassemblyExperience 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 the equipment and tools required for the spent fuel pool.
  • Perform a receiving inspection on the fresh replacement rod segments (or stainless steel replacement rods at bundle 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 the ITA to the fuel preparation machine (FPM).
  • Remove the channel.
  • Visually inspect the ITA.
  • Remove the upper tie plate.
  • Grapple the cobalt isotope rod designated for removal with a standard collet tool and then withdraw the rod from the bundle by lowering FPM.

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  • Verify the segment locations and serial numbers.
  • Insert the rod into the segmented rod disassembly vise and secure.
  • Unscrew the top isotope segment by twisting the collet tool and the torque wrench.
  • Attach a second grapple tool with the proper collet for handling the full diameter of the isotope segments to the top of the second segment, release the vise, pull the rod up until the second junction is visible in the mirror, and the vise is again clamped shut.
  • Unscrew the second isotope segment and place it into a storage tube.

Repeat the 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 ITA.
  • Repeat the above steps until all cobalt isotope rods have been replaced.
  • Visually inspect the ITA and components.
  • Reinstall the upper tie plate.
  • Rechannel the ITA.
  • Remove the ITA from the FPM and place it into the designated location.

The current design includes a hex impression on each rod segment. This provides further gripping ease during the unscrewing steps.

Prior to segmented rod disassembly in the next refueling outage following HCGS Cycle 17, a similar isotope segmented rod procedure will be prepared and incorporated into the fuel inspection process. Any devices needed during the removal/replacement onto the FPM will adhere to all analysis requirements, including seismic.

4.7.3.4 GEH Tools Usedfor Rod Disassembly GEH has experience with rod removal and replacement. The specific tools that can be used to disassemble a GE14i cobalt isotope rod into segments are the Fuel Rod Collet Grapple, the Fuel Rod Side Grapple, and the Six Rod Universal Storage Rack. These same tools have been used for other 1Oxi0 disassembly efforts. They are designed to allow 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 male threaded connector 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 male threaded connector of the segment is unable to be held by the Fuel Rod Collet Grapple.

  • Six Rod Universal Storage Rack (side mount to the FPM) - This 12 foot rack attaches to the 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.

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NEDO-33529 Revision 0 Non-Proprietary Information 4.7.3.5 Segmented Rod Disassembly Contingency Plan 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 threaded connector 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 at the location of the cobalt targets. Furthermore, the broken male component is locked into the female receptor, preventing any debris inside the fuel pool.

4. 7.3.6 Shielding andDose 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 applied for the dose calculations during cobalt isotope rod disassembly.

Thirty years of GNF/GEH fuel examination experience has shown that placing fuel assemblies 6-7 feet underwater provides appropriate shielding for personnel regarding ALARA considerations. For cobalt isotope rod disassembly ALARA considerations, the GE14i fuel assemblies are planned to be handled approximately 9-101/2/2 feet underwater. The receiving basket is planned to 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 and 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. The doses in Table 4-4 can be multiplied by

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

4.7.3.7 Cask Movement Experience Currently, 'GEHowns 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 HCGS reactor building crane limit of 130 tons. NUREG 0612: Control of Heavy Loads at Nuclear Power Plants (Reference 15) applies to cask movement activities. Key steps for an approved site-specific procedure 70

NEDO-33529 Revision 0 Non-Proprietary Information include: removing the overpack, transferring the cask to the fuel floor, handling the cask lid, transferring the cask to and from the spent fuel pool cask pit, transferring the cask to the reactor building, and reinstalling the overpack. Typical procedure steps followed during GE Model 2000 cask movements include:

  • Ensure equipment and materials are available (e.g., fuel building crane, decontamination equipment, 11 OV 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 the licensee to accomplish the project tasks.
  • Review of the appropriate procedures by GEH personnel assigned to the project.
  • 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), has been load tested and inspections are current. Due to the planned water shielding, as discussed in Section 4.7.3.6, longer lifting slings may be used for handling the cask in the spent fuel pool and for loading the cobalt basket. Site-specific sling qualifications will be addressed during the pre-shipment evaluation for an approaching outage.
  • Ensure reactor building crane has been inspected in accordance with licensee procedures.
  • Ensure decontamination equipment and facilities are operational.
  • Conduct a pre-job brief, by the licensee supervisor, on expectations and personnel safety concerns.
  • Obtain permission from licensee management to start project activities.
  • Inform the control room supervisor as required by the licensee's procedures.
  • Establish a contamination control area for the pool hardware work.
  • Transfer the GE Model 2000 cask to the cask loading pool.
  • Load the GE Model 2000 cask.
  • Prepare the loaded GE Model 2000 cask.
  • Return to normal operations and demobilize cask rigging and equipment per licensee procedures. -
  • Notify the licensee's supervisor that the project on-site work is complete.

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NEDO-33529 Revision 0 Non-Proprietary Information 4.7.3.8 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 the 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. 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.

GEH has experience safely moving large amounts of high specific activity cobalt-60 in accordance with all state and federal regulations, and will continue to safely transport the isotope segments as part of the GE14i ITA program. The shipment of this quantity of material is governed by security regulations and orders issued by the NRC and will require that a similar level of detail 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.

4.7.4 Post-IrradiationExamination 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).

4.7.4.1 Poolside Visual Examination The GE14i ITA visual exam may include the following elements:

1. A full bundle periphery visual exam of all bundle mechanical elements
2. Assessment of rod-to-rod spacing of the cobalt isotope rods relative to nearby fuel rods
3. Assessment of rod growth of the cobalt isotope rods relative to nearby fuel rods 72

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4. Assessment of spacer cells with the cobalt isotope rods removed to verify no abnormal wear A GE14i ITA rod visual exam may include the following elements:
1. Visual exam of one or more cobalt isotope rods after brushing to remove the crud
2. Visual exam of one or more brushed fuel rods adjacent to cobalt isotope rods 4.7.4.2 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 an 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 GE14 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 GE14i ITAs may include the axial power distribution measurements of one or more fuel rods adjacent to the cobalt isotope rods.

4. 7.4.3 Poolside CombinedInstrumentationMeasurementSystem Inspection COINS is designed to measure corrosion and liftoff for a single fuel rod, which has been removed from a bundle. The poolside COINS inspection will be performed on cobalt isotope rods in order to non-destructively obtain information about outer surface corrosion and diameter.

COINS possesses two eddy-current probe lift-off instruments to measure the water side corrosion thickness of the cladding at two azimuthal locations 1800 apart and one probe lateral displacement instrument to measure the axial profile of the cladding.

. 4.7.4.4 Segmented Rod Hot Cell DestructiveExam

-One cobalt isotope rod will be sent to the GEH Vallecitos Nuclear Center (VNC) in Sunol, California for Hot Cell Examination.

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NEDO-33529 Revision 0 Non-Proprietary Information During hot cell examination, the outer rod will be inspected for corrosion. The outer rod will be cut at the VNC Hot Cell to inspect the inner capsule for the following:

  • Evidence of vibration damage and corrosion
  • Inner and outer oxide layer thickness The inner rod will also be cut to expose the TPR and cobalt targets. The cobalt targets may be inspected for the following:
  • Location specific activities along axis
  • Cobalt target conditions and status of nickel plating
  • Evidence of vibration damage and corrosion
  • Ease with which cobalt targets are released The results of these inspections are expected to confirm the successful performance of the GEl4i bundle design.

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NEDO-33529 Revision 0 Non-Proprietary Information Table 4-1 List of Analyzed Events for the Reload License with GE14i ITAs in the Core

'Evenit Method A..Fuel Th'ermal Ma~rgini Events

1. Generator Load Rejection with Bypass Failure ODYN
2. Turbine Trip with Bypass Failure ODYN
3. Feedwater Controller Failure-Maximum Demand ODYN
4. Loss of Feedwater Heating 3D-Simulator
5. Rod Withdrawal Error at Rated Power 3D-Simulator
6. Mislocated Fuel Assembly Accident 3D-Simulator
7. Misoriented Fuel Assembly Accident TGBLA B. Limiting Transient Overpressture Evnts
1. Main Steam Isolation Valve Closure with Flux ODYN Scram (MSIVF) (Failure of Direct Scram) 75

NEDO-33529 Revision 0 Non-Proprietary Information Table 4-2 GE14i Data for Emergency Procedure Guidelines Parameter Value Maximum subcritical banked control rod withdrawal position Cycle specific calculations suggested (MSBWP)

Water level at bottom of active fuel (WLrpv-baf)

Cold shutdown boron concentration requirement for naturally EE )) ppm' occurring boron (XB-cld-nat)

Hot shutdown boron concentration requirement for naturally (( )) ppm' occurring boron (XB-hot-nat) E[ )) ppm2 Decay heat fraction 10 minutes after shutdown (FQdh-1 0) ((_))

Decay heat as a function of time after reactor shutdown Time after Qdh-%0 3 Time after Qdh-%()3 (Qdh-%0) Shutdown (%) Shutdown (%)

(minutes) (minutes)

Minimum active fuel length fraction which must be covered to maintain PCT< 1500'F with injection (Fafl-15)

Minimum active fuel length fraction which must be covered to maintain PCT< 1800°F without injection (Fafl- 18)

Minimum bundle steam flow required to maintain PCT < 1500°F )) lbm/hr (wg- 1500)

Maximum core uncovery time before PCT exceeds 1500'F for Plant specific calculations recommended per SIL-peak linear heat generation rate of(( )) kW/ft (t-cu-15) 636, Revision I (issued June 6, 2001)

Specific heat of clad (c-clad) at (( ]O0F ((. )) Btu/lbm-°F Specific heat of fuel (c-fuel) at (( 0 O]F Btu/lbm-0 F BWR4-5 BWR4-6 Fuel Components Mass 7 (Mclad) (kg)

Fuel Mass (Mfuel) (( )) (kg) ))

Active Fuel Length (Lfuel) (inches) E[ ]

Notes:

I Value specified per the basis identified in EPG Revision 4. ((

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NEDO-33529 Revision 0 Non-Proprietary Information 2 Value specified per the basis identified in EPG/SAG Revision 2: control rods withdrawn to maximum rod block limit, core pressure 1100 psia saturated liquid, full power equilibrium xenon, no voids in core, no shutdown cooling, and initial reactor condition at Maximum Extended Operating Domain and most reactive exposure. ((

))

3 Values specified per a generic nominal decay heat curve based on ((

)) and the effect of SIL-636, Revision 1.

4 ((

5 [

6 This value is for a peak linear heat generation rate of (( )) kW/ft.

7 This includes all components except for the fuel pellets. ((

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NEDO-33529 Revision 0 Non-Proprietary Information Table 4-3 Basket Dose Rate Values Distance in Top Dose Rate (mr/hr) Side Dose Rate (mr/hr)

Water (ft) 0.5 4.727E+08 3.344E+08 1 1.393E+08 9.281E+07 1.5 4.434E+07 2.867E+07 2 1.457E+07 9.361 E+06 2.5 4.961E+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 1.811E+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 5.301E+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 1.461E+00 9.228E-01 10.5 5.521E-01 3.486E-01 11 2.089E-01 1.319E-01 78

NEDO-33529 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 7.931E+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 1.281E+05 4 8.281E+01 4.752E+04 4.5 2.900E+01 1.779E+04 5 1.028E+01 6.708E+03 5.5 3.681 E+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 79

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5. Conclusion The GEI4i fuel product has been described in Section 2 of this report. The core design approach and applicability of NRC approved methods to the GE14i ITAs were discussed in Section 3. Installation of 12 GE14i ITAs into HCGS Cycle 17 has been evaluated against the events addressed in Chapter 15 of the UFSAR, and against additional analyses typically performed during a fuel transition, as described in Section 4. These assessments confirm compliance to licensing requirements with the ITAs inserted.

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6. References
1. Global Nuclear Fuel, "GEl4 Compliance With Amendment 22 of NEDE-24011-P-A (GESTAR II)," NEDC-32868P, Revision 3, April 2009.
2. Global Nuclear Fuel, "General Electric Standard Application for Reactor Fuel (GESTAR 11)," 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 1I' - 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. Global Nuclear Fuel, "Supplemental Reload Licensing Report for Hope Creek Unit 1 Reload 15 Cycle 16," 0000-0088-3934-SRLR, Revision 0, March 2009.
8. GE Nuclear Energy, "Safety Analysis Report for Hope. Creek Constant Pressure Power Uprate," NEDC-33076P, Revision 2, August 2006.
9. GE Hitachi Nuclear Energy, "General Electric Standard Application for Reactor Fuel (Supplement for United States)," NEDE-2401 1-P-A-16-US, Revision 16, October 2007.
10. GE Nuclear Energy, "BWR Owners' Group Long-Term Stability Solutions Licensing Methodology," NEDO-31960-A, November 1995 (including Supplement 1).
11. GE Nuclear Services Information Letter (SIL) 465, Supplement 1, "Surface Observations on Jet Pump Mixers," April 30, 1993.
12. Institute of Nuclear Power Operations, OE27774, "Increased Hydrogen Injection Rates Required to Mitigate Intergranular Stress Corrosion Cracking in the Reactor Vessel,"

Event Date: October 16, 2008, Fermi 2 Reactor.

13. NRC Certificate of Compliance No. 9309, Revision 7, for Model RAJ-I1 Package, May 28, 2008.

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

NEDO-1 1209-04A, Revision 8, March 31, 1989.

15. US NRC, "Control of Heavy Loads at Nuclear Power Plants, Resolution of Generic Technical Activity A-36," NUREG-0612, 1980.
16. GE Nuclear Energy, "Methodology and Uncertainties for Safety Limit MCPR Evaluations," NEDC-32601 -P-A, August 1999.
17. Amendment No. 160 to Facility Operating License No. NPF-57, PSEG Nuclear LLC, Hope Creek Generating Station.
18. Thomas B. Blount (NRC) to Jerry G. Head (GEH), "Final Safety Evaluation for GE Hitachi Nuclear Energy Americas, LLC Licensing Topical Report NEDC-33173P,

'Applicability of GE Methods to Expanded Operating Domains' (TAC No. MD0277),"

July 21, 2009.

19. C.O. Thomas (NRC) to J. S. Charnley (GE), "Acceptance for Referencing of Licensing Topical Report NEDE-2401 1-P-A Amendment 7 to Revision 6, GE Standard Application for Reactor Fuel," March 1, 1985.

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NEDO-33529 Revision 0 Non-Proprietary Information Appendix A Limitations from Safety Evaluation for LTR NEDC-33173P 9.1 TGBLA/ The neutronic methods used to simulate the reactor core response Comply The GE 14i ITAs use TGBLA06/PANAC 1I PAN.AC and that feed into the downstream safety analyses supporting neutronic methods.

Version operation at EPU/MELLLA+ will apply TGBLA06/PANACI 1 or later NRC-approved version of neutronic method.

9.2 3D For EPU/MELLLA+ applications, relying on Not The GE14i ITAs do not use TGBLA04/PANAC10 Monicore TGBLA04/PANACIO methods, the bundle RMS difference Applicable as the neutronic method. See Limitation 9.1.

uncertainty will be established from plant-specific core-tracking data, based on TGBLA04/PANACIO. The use of plant-specific trend line based on the neutronic method employed will capture the actual bundle power uncertainty of the core monitoring system.

9.3 Power to Plant-specific EPU and expanded operating domain applications Comply This limitation is not dependent on fuel type.

Flow Ratio will confirm that the core thermal power to core flow ratio will not exceed 50 MWt/MlbrI/hr at any state point in the allowed operating domain. For plants that exceed the power-to-flow value of 50 MWt/Mlbm/hr, the application will provide power distribution assessment to establish that neutronic methods axial and nodal power distribution uncertainties have not increased.

A-I

NEDO-33529 Revision 0 Non-Proprietary Information 9.4 SLMCPR1 For EPU operation, a 0.02 value shall be added to the cycle- Comply Based upon the recommendation within NEDC-specific SLMCPR value. This adder is applicable to SLO, which is 33173P, the 0.02 adder for SLMCPR is a bounding derived from the dual loop SLMCPR value. value for EPU operation. Section 3 demonstrates both that the SLMCPR process remains valid and that pin power uncertainty basis for GE14i remains consistent with GE 14 such that the margin basis remains valid. The additional margin in the

)) so it remains applicable to GE14i.

9.5 SLMCPR2 For operation at MELLLA+, including operation at the EPU power Not This limitation is Not Applicable, because HCGS is levels at the achievable core flow state point, a 0.03 value shall be Applicable not licensed for the MELLLA+ domain.

added to the cycle-specific SLMCPR value.

9.6 R-Factor The plant specific R-factor calculation at a bundle level will be Comply This limitation is implemented for all fuel types, consistent with lattice axial void conditions expected for the hot including GEI4i.

channel operating state. The plant-specific EPU/MELLLA+ As noted in Section 3.1, the GE14i bundles are application will confirm that the R-factor calculation is consistent placed in non-limiting locations.

with the hot channel axial void conditions.

9.7 ECCS- For applications requesting implementation of EPU or expanded Not This limitation was addressed for the LOCA 1 operating domains, including MELLLA+, the small and large Applicable implementation of EPU at HCGS with GE14.

break ECCS-LOCA analyses will include top-peaked and mid- The ECCS-LOCA evaluation of the GEI4i bundles peaked power shape in establishing the MAPLHGR and documented in Section 4.5.6 demonstrates that the determining the PCT. This limitation is applicable to both the Licensing Basis PCT is not affected by the licensing bases PCT and the upper bound PCT. The plant-specific ictin of T i ntas b applications will report the limiting small and large break licensing basis and upper bound PCTs.

A-2

NEDO-33529 Revision 0 Non-Proprietary Information 9.8 ECCS- The ECCS-LOCA will be performed for all state points in the Not This limitation is Not Applicable, because HCGS is LOCA 2 upper boundary of the expanded operating domain, including the Applicable not licensed for the MELLLA+ domain.

minimum core flow state points, the transition state point as defined in Reference A-I and the 55 percent core flow state point.

The plant-specific application will report the limiting ECCS-LOCA results as well as the rated power and flow results. The SRLR will include both the limiting state point ECCS-LOCA results and the rated conditions ECCS-LOCA results.

9.9 Transient Plant-specific EPU and MELLLA+ applications will demonstrate Comply This limitation is implemented for all fuel types, LHGR 1 and document that during normal operation and core-wide AGOs, including the GEI4i ITAs. The GEI4i ITA fuel the T-M acceptance criteria as specified in Amendment 22 to rods are identical to the GE 14 rods which have GESTAR II will be met. Specifically, during anl AOO, the been demonstrated to meet the limitation.

licensing application will demonstrate that tile: (I) loss of fuel rod mechanical integrity will not occur due to fuel melting and (2) loss of fuel rod mechanical integrity will not occur due to pellet-cladding mechanical interaction. The plant-specific application will demonstrate that the T-M acceptance criteria are met for the both the U0 2 and the limiting GdO 2 rods.

9.10 Transient Each EPU and MELLLA+ fuel reload will document the Comply This limitation is implemented for all fuel types, LHGR 2 calculation results of the analyses demonstrating compliance to including GE14i ITAs.

transient T-M acceptance criteria. The plant T-M response will be domente in th SRLR. e provided with the SRLR or COLR, or it will be reported directly to COLR.

the NRC as an attachment to the SRLR or A-3

NEDO-33529 Revision 0 Non-Proprietary Information 9.11 Transient To account for the impact of the void history bias, plant-specific Comply The basis of the void history bias is Monte Carlo to LHGR 3 EPU and MELLLA+ applications using either TRACG or ODYN TGBLA response comparisons. As demonstrated will demonstrate an equivalent to 10 percent margin to the fuel in Section 3, Monte Carlo evaluations indicate good centerline melt and the 1 percent cladding circumferential plastic comparisons between GE14i and GE14 throughout strain acceptance criteria due to pellet-cladding mechanical the range of void conditions. Additionally, interaction for all of limiting AOO transient events, including considering the direct modeling of GE14i in equipment out-of-service. Limiting transients in this case, refers to nuclear, thermal-hydraulic, and transient methods, transients where the void reactivity coefficientplays a significant this limitation is implemented for all fuel types, role (such as pressurization events). If the void history bias is including GEI4i ITAs.

incorporated into the transient model within the code, then the additional 10 percent margin to the fuel centerline melt and the 1 percent cladding circumferential plastic strain is no longer required.

9.12 LHGR and In MFN 06-481, GE committed to submit plenum fission gas and Not This limitation pertains to future license Exposure fuel exposure gamma scans as part of the revision to the T-M Applicable applications for EPU and MELLLA+ referencing Qualification licensing process. The conclusions of the plenum fission gas and LTR NEDC-33173P.

fuel exposure gamma scans of GE IOx 10 fuel designs as operated will be submitted for NRC staff review and approval. This revision will be accomplished through Amendment to GESTAR II or in a T-M licensing LTR. PRIME (a newly developed T-M code) has been submitted to the NRC staff for review (Reference A-3).

Once the PRIME LTR and its application are approved, future license applications for EPU and MELLLA+ referencing LTR NEDC-33173P must utilize the PRIME T-M methods.

A-4

NEDO-33529 Revision 0 Non-Proprietary Information 9.13 Application Before applying 10 weight percent Gd to licensing applications, Not This limitation is Not Applicable to HCGS as the of 10 Weight including EPU and expanded operating domain, the NRC staff Applicable Gd concentration is less than 8%.

Percent Gd needs to review and approve the T-M LTR demonstrating that the T-M acceptance criteria specified in GESTAR II and Amendment 22 to GESTAR 11can be met for steady-state and transient conditions. Specifically, the T-M application must demonstrate that the T-M acceptance criteria can be met for TOP and MOP conditions that bounds the response of plants operating at EPU and expanded operating domains at the most limiting state points, considering the operating flexibilities (e.g., equipment out-of-service).

Before the use of 10 weight percent Gd for modern fuel designs, NRC must review and approve TGBLA06 qualification submittal.

Where a fuel design refers to a design with Gd-bearing rods adjacent to vanished or water rods, the submittal should include specific information regarding acceptance criteria for the qualification and address any downstream impacts in terms of the safety analysis. The 10 weight percent Gd qualifications submittal can supplement this report.

9.14 Part 21 Any conclusions drawn from the NRC staff evaluation of the GE's Comply The HCGS GE14 T-M basis includes the prescribed Evaluation Part 21 report will be applicable to the GESTR-M T-M assessment 350 psi penalty from Appendix F of the Methods of GESTR- of this. SE forfuture license application. GE submitted the T-M LTR SE (Reference A-2). The GE14i fuel rods M Fuel Part 21 evaluation, which is currently under NRC staff review. have the same T-M basis as the GE14 fuel rods.

Temperature Upon completion of its review, NRC staff will inform GE of its Further, the GE141LTAs are placed in non-limiting Calculation conclusions. Frhr h ~4 Tsaepae nnnlmtn locations and additional margin is provided for fuel rods in the GE14i ITAs that are adjacent to the cobalt bearing rods.

9.15 Void The void reactivity coefficient bias and uncertainties in TRACG Not This limitation is Not Applicable to HCGS because Reactivity 1 for EPU and MELLLA+ must be representative of the lattice Applicable TRACG is not used for AOOs.

designs of the fuel loaded in the core A-5

NEDO-33529 Revision 0 Non-Proprietary Information

.. , m itatlI ion .t 9.16 Void A supplement to TRACG /PANACI 1 for AOO is under NRC staff Not This limitation is Not Applicable to HCGS because Reactivity 2 review (Reference A-4). TRACG internally models the response Applicable TRACG is not used for AGOs.

surface for the void coefficient biases and uncertainties for known dependencies due to the relative moderator density and exposure on nodal basis. Therefore, the void history bias determined through the methods review can be incorporated into the response surface "known" bias or through changes in lattice physics/core simulator methods for establishing the instantaneous cross-sections. Including the bias in the calculations negates thle need for ensuring that plant-specific applications show sufficient margin.

For application of TRACG to EPU and MELLLA+ applications, the TRACG methodology must incorporate the void history bias.

The manner in which this void history bias is accounted for will be established by the NRC staff SE approving NEDE-32906P, Supplement 3, "Migration to TRACG04/PANAC 1I from TRACG02/PANAC 10," May 2006 (Reference A-4). This limitation applies until the new TRACG/PANAC methodology is approved by the NRC staff.

9.17 Steady-State The instrumentation specification design bases limit the presence Comply The basis of the local power distribution is Monte 5 Percent of bypass voiding to 5 percent (LRPM levels). Limiting the bypass Carlo to TGBLA response comparisons. As Bypass voiding to less than 5 percent for long-term steady operation demonstrated in Section 3, Monte Carlo evaluations Voiding ensures that instrumentation is operated within the specification. indicate good comparisons between GEI4i and For EPU and MELLLA+ operation, the bypass voiding will be GE14 throughout the range of void conditions.

evaluated on a cycle-specific basis to confirm that the void fraction Additionally, considering the direct modeling of remains below 5 percent at all LPRM levels when operating at GEl4i in nuclear and thermal-hydraulic methods, steady-state conditions within the MELLLA+ upper boundary. this limitation is implemented for all fuel types, The highest calculated bypass voiding at any LPRM level will be including GE 14i ITAs.

provided with the plant-specific SRLR.

A-6

NEDO-33529 Revision 0 Non-Proprietary Information 9.18 Stability Set The NRC staff concludes that the presence bypass voiding at the Comply This limitation is not dependent on fuel type. The points low-flow conditions where instabilities are likely can result in application of the specified calibration errors is Adjustment calibration errors of less than 5 percent for OPRM cells and less directed from the NRC staff based on conservative than 2 percent for APRM signals. These calibration errors must be estimates.

accounted for while determining the set points for any detect and The hydraulic characteristics of GE14i are suppress long-term methodology. The calibration values for the essentially the same as those of GE14. Therefore, different long-term solutions are specified in the associated there is no basis to change these penalties with the sections of this SE, discussing the stability methodology.

introduction of the GE 14i ITAs.

9.19 void- For applications involving PANACEA/ODYN/ISCOR/TASC for Comply This limitation is implemented for all fuel types, Quality operation at EPU and MELLLA+, an additional 0.01 will be added including the GE14i ITAs. Section 3 establishes Correlation 1 to the OLMCPR, until such time that GE expands the experimental that the Findlay-Dix void correlation has previously database supporting the Findlay-Dix void-quality correlation to been shown to be applicable for all current GE demonstrate the accuracy and performance of the void-quality BWR fuel designs, including lOx 10 lattices with correlation based on experimental data representative of the current part length rods.

fuel designs and operating conditions during steady-state, transient, and accident conditions.

9.20 Void- The NRC staff is currently reviewing Supplement 3 to NEDE- Not This limitation is Not Applicable to HCGS because Quality 32906P, "Migration to TRACG04/PANAC1 1 from Applicable TRACG is not used.

Correlation 2 TRACG02/PANAC 10," dated May 2006. The adequacy of the TRACG interfacial shear model qualification for application to EPU and MELLLA+ will be addressed under this review. Any conclusions specified in the NRC staff SE approving Supplement 3 to LTR NEDC-32906P will be applicable as approved.

9.21 Mixed Core Plants implementing EPU or MELLLA+ with mixed fuel vendor Not This limitation addresses mixed fuel vendor cores.

Method I cores will provide plant-specific justification for extension of GE's Applicable Therefore, this limitation is not applicable to analytical methods or codes. The content of the plant-specific HCGS.

application will cover the topics addressed in this SE as well as subjects relevant to application of GE's methods to legacy fuel.

Alternatively, GE may supplement or revise LTR NEDC-33173P for mixed core application.

A-7

NEDO-33529 Revision 0 Non-Proprietary Information 9.22 Mixed Core For any plant-specific applications of TGBLA06 with fuel type Comply GEI4i does not deviate from the approved GEl4 Method 2 characteristics not covered in this review, GE needs to provide product line when considering geometry and assessment data similar to that provided for the GE fuels. The materials. The introduction of cobalt bearing rods Interim Methods review is applicable to all GE lattices up to GE14. with GEI4i is the subject of Section 3 which Fuel lattice designs, other than GE lattices up to GE14, with the supports the use of the TGBLA06/PANACI I following characteristics are not covered by this review: methodology for the GEI4i ITAs.

Square internal water channels water crosses Gd rods simultaneously adjacent to water and vanished rods Ilx IxIlattices MOX fuel The acceptability of the modified epithermal slowing down models in TGBLA06 has not been demonstrated for application to these or other geometries for expanded operating domains.

Significant changes in tile Gd rod optical thickness will require an evaluation of the TGBLA06 radial flux and Gd depletion modeling before being applied. Increases in the lattice Gd loading that result in nodal reactivity biases beyond those previously established will require review before the GE methods may be applied.

A-8

NEDO-33529 Revision 0 Non-Proprietary Information Limlitationl Text 9.23 MELLLA+ In the first plant-specific implementation of MELLLA+, the cycle- Not This limitation is Not Applicable, because HCGS is Eigenvalue specific eigenvalue tracking data will be evaluated and submitted Applicable not licensed for the MELLLA+ domain.

Tracking to NRC to establish the performance of nuclear methods under the operation in the new operating domain. The following data will be analyzed:

Hot critical eigenvalue, Cold critical eigenvalue, Nodal power distribution (measured and calculated TIP comparison),

Bundle power distribution (measured and calculated TIP comparison),

Thermal margin, Core flow and pressure drop uncertainties, and The MIP Criterion (e.g., determine if core and fuel design selected is expected to produce a plant response outside the prior experience base).

Provision of evaluation of the core-tracking data will provide the NRC staff with bases to establish if operation at the expanded operating domain indicates: (1) changes in the performance of nuclear methods outside the EPU experience base; (2) changes in the available thermal margins; (3) need for changes in the uncertainties and NRC-approved criterion used in the SLMCPR methodology; or (4) any anomaly that may require corrective actions.

A-9

NEDO-33529 Revision 0 Non-Proprietary Information 9.24 Plant The plant-specific applications will provide prediction of key Not The limited number of GEI4i ITAs are placed in Specific parameters for cycle exposures for operation at EPU (and Applicable non-limiting locations in the HCGS core.

Application MELLLA+ for MELLLA+ applications). The plant-specific Therefore, the inclusion of GE l4i ITAs in the prediction of these key parameters will be plotted against the EPU HCGS core does not affect the key parameter maps Reference Plant experience base and MELLLA+ operating which were provided when the EPU was approved experience, if available. For evaluation of the margins available in by the NRC.

the fuel design limits, plant-specific applications will also provide quarter core map (assuming core symmetry) showing bundle power, bundle operating LHGR, and MCPR for BOC, MOC, and EOC. Since the minimum margins to specific limits may occur at exposures other than the traditional BOC, MOC, and EOC, the data will be provided at these exposures.

References A-i GE Nuclear Energy, "General Electric Boiling Water Reactor Maximum Extended Load Line Limit Analysis Plus," NEDC-33006P-A, Revision 3, June 2009.

A-2 Thomas B. Blount (NRC) to Jerry G. Head (GEH), "Final Safety Evaluation for GE Hitachi Nuclear Energy Americas, LLC Licensing Topical Report NEDC-33173P, 'Applicability of GE Methods to Expanded Operating Domains' (TAC No.

MD0277)," July 21, 2009.

A-3 GNF Letter (FLN-2007-001), A. A. Lingenfelter to NRC, "The PRIME Model for Analysis of Fuel Rod Thermal-Mechanical Performance," January 19, 2007. (ADAMS Package Accession No. ML070250414).

A-4 GE Nuclear Energy, "Migration to TRACG04/PANAC1 1 from TRACG02/PANAC 10," NEDE-32906P, Supplement 3, May 2006.-

A-10

GE-Hitachi Nuclear Energy Americas LLC AFFIDAVIT I, James F. Harrison, state as follows:

(1) I am Vice President, Fuel Licensing, Regulatory Affairs, GE-Hitachi Nuclear Energy Americas LLC ("GEH"). I have been delegated the function of reviewing the information described in paragraph (2) which is sought to be withheld, and have been authorized to apply for its withholding.

(2) The information sought to be withheld is contained in 0000-0110-2515, Additional Information Providing Hope Creek GeneratingStation (HCGS) Responses to Requests for Additional Information (RAI) from the Clinton Power Station (CPS) GE14i License Amendment Request (LAR), Revision 0, dated December 2009. The proprietary information in 0000-0110-2515, Additional Information Providing Hope Creek Generating Station (HCGS) Responses to Requests for Additional Information (RAI) from the Clinton.Power Station (CPS) GE14i License Amendment Request (LAR), Revision 0, dated December 2009, is identified by a [doKtted underline insjde double square brackets!.1)). In each case, the superscript notation kTlrefers to Paragraph (3) of this affidavit, which provides the basis for the proprietary determination.

(3) In making this application for withholding of proprietary information of which it is the owner or licensee, GEH relies upon the exemption from disclosure set forth in the Freedom of Information Act ("FOIA"), 5 USC Sec. 552(b)(4), and the Trade Secrets Act, 18 USC Sec. 1905, and NRC regulations 10 CFR 9.17(a)(4), and 2.390(a)(4) for "trade secrets" (Exemption 4). The material for which exemption from disclosure is here sought also qualify under the narrower definition of "trade secret", within the meanings assigned to those terms for purposes of FOIA Exemption 4 in, respectively, Critical Mass Energy Project v. Nuclear Regulatory Commission, 975F2d871 (DC Cir. 1992), and Public Citizen Health Research Group v. FDA, 704F2d1280 (DC Cir. 1983).

(4) Some examples of categories of information which fit into the definition of proprietary information are:

a. Information that discloses a process, method, or apparatus, including supporting data and analyses, where prevention of its use by GEH's competitors without license from GEH constitutes a competitive economic advantage over other companies;
b. Information which, if used by a competitor, would reduce his expenditure of resources or improve his competitive position in the design, manufacture, shipment, installation, assurance of quality, or licensing of a similar product;
c. Information which reveals aspects of past, present, or future GEH customer-funded development plans and programs, resulting in potential products to GEH;
d. Information which discloses patentable subject matter for which it may be desirable to obtain patent protection.

Additional Information Providing HCGS Responses to RAIs from the CPS GE14i LAR Affidavit Page I of 3

GE-Hitachi Nuclear Energy Americas LLC The information sought to be withheld is considered to be proprietary for the reasons set forth in paragraphs (4)a. and (4)b. above.

(5) To address 10 CFR 2.390(b)(4), the information sought to be withheld is being submitted to NRC in confidence. The information is of a sort customarily held in confidence by GEH, and is in fact so held. The information sought to be withheld has, to the best of my knowledge and belief, consistently been held in confidence by GEH, no public disclosure has been made, and it is not available in public sources. All disclosures to third parties, including any required transmittals to NRC, have been made, or must be made, pursuant to regulatory provisions or proprietary agreements which provide for maintenance of the information in confidence. Its initial designation as proprietary information, and the subsequent steps taken to prevent its unauthorized disclosure, are as set forth in paragraphs (6) and (7) following.

(6) Initial approval of proprietary treatment of a document is made by the manager of the originating component, the person most likely to be acquainted with the value and sensitivity of the information in relation to industry knowledge, or subject to the terms under which it was licensed to GEH. Access to such documents within GEH is limited on a "need to know" basis.

(7) The procedure for approval of external release of such a document typically requires review by the staff manager, project manager, principal scientist, or other equivalent authority for technical content, competitive effect, and determination of the accuracy of the proprietary designation. Disclosures outside GEH are limited to regulatory bodies, customers, and potential customers, and their agents, suppliers, and licensees, and others with a legitimate need for the information, and then only in accordance with appropriate regulatory provisions or proprietary agreements.

(8) The information identified in paragraph (2), above, is classified as proprietary because it contains detailed results including the process and methodology for the design and analysis of the GE14i Isotope Test Assembly. The GE14i Isotope Test Assembly has been developed at a significant cost to GEH.

The development of the GE14i Isotope Test Assembly is derived from the extensive experience database that constitutes a major GEH asset.

(9) Public disclosure of the information sought to be withheld is likely to cause substantial harm to GEH's competitive position and foreclose or reduce the availability of profit-making opportunities. The information is part of GEH's comprehensive BWR safety and technology base, and its commercial value extends beyond the original development cost.

The value of the technology base goes beyond the extensive physical database and analytical methodology and includes development of the expertise to determine and apply the appropriate evaluation process. In addition, the technology base includes the value derived from providing analyses done with NRC-approved methods.

Additional Information Providing HCGS Responses to RAIs from the CPS GEI4i LAR Affidavit Page 2 of 3

GE-Hitachi Nuclear Energy Americas LLC The research, development, engineering, analytical and NRC review costs comprise a substantial investment of time and money by GEH.

The precise value of the expertise to devise an evaluation process and apply the correct analytical methodology is difficult to quantify, but it clearly is substantial.

GEH's competitive advantage will be lost if its competitors are able to use the results of the GEH experience to normalize or verify their own process or if they are able to claim an equivalent understanding by demonstrating that they can arrive at the same or similar conclusions.

The value of this information to GEH would be lost if the information were disclosed to the public. Making such information available to competitors without their having been required to undertake a similar expenditure of resources would unfairly provide competitors with a windfall, and deprive GEH of the opportunity to exercise its competitive advantage to seek an adequate return on its large investment in developing and obtaining these very valuable analytical tools.

I declare under penalty of perjury that the foregoing affidavit and the matters stated therein are true and correct to the best of my knowledge, information, and belief.

Executed on this 17th day of December 2009.

r James F. Harrison Vice President, Fuel Licensing, Regulatory Affairs GE-Hitachi Nuclear Energy Americas LLC Additional Information Providing HCGS Responses to RAIs from the CPS GE14i LAR Affidavit Page 3 of 3

0000-0110-2515 Revision 0 Non-Proprietary Information Additional Information Providing Hope Creek Generating Station (HCGS)

Responses to Requests for Additional Information (RAI) from the Clinton Power Station (CPS) GE14i License Amendment Request (LAR)1 December 2009 NOTE 1: All references and references to Attachments in the RAI questions below refer to the CPS LAR submittal dated June 26, 2009 (Agencywide Documents Access and Management System Accession No. ML091801061); all references and references to Attachments in the RAI Responses below refer to this HCGS LAR submittal, unless otherwise noted.

NOTE 2: The equivalent Safety Analysis Report (SAR) for HCGS is GEH Report NEDC-33529P, "Safety Analysis Report to Support Introduction of GE14i Isotope Test Assemblies (ITAs) in Hope Creek Generating Station," dated December 2009.

In Reference 11, the NRC requested that Exelon Generating Company (EGC) submit the following documents to support completion of their review of "License Amendment Request to Modify Clinton Power Station Facility Operating License in Support of the Use of Isotope Test Assemblies," dated June 26, 2009.

1. "GE14 Fuel Design Cycle-Independent Analyses for Clinton Power Station,"

GE-NE-L12-00879-00-O1P, April 2001, or latest version of the same type document.

2. "Evaluation of Hydraulic Characteristics of GE14i Fuel," No. 0000-0100-9379 RO, 04/22/2009
3. "GE I4i Thermal Mechanical Evaluation," 0000-0097-1904, 04/24/2009
4. Revision 0 of DRF Section 0000-0100-9366, 04/26/2009 Reports prepared in response to items 2, 3 and 4 above have been previously docketed by CPS as Attachments 4, 5, 6, 7, 8, and 9 of the letter dated November 4, 2009 (ADAMS Accession Number ML093100313 and Reference 13). These reports (References 1, 6 and 7) are fuel type specific but plant independent and therefore, equally applicable to HCGS. Response to item I is addressed in RAI 7 of this document.

'This document addresses RAIs received by Clinton Power Station in two letters dated October 16, 2009 (References 11 and 12).

Page 1 of 20

0000-0110-2515 Revision 0 Non-Proprietary Information NRC RAI:

In reviewing the Exelon Generation Company's submittal dated June 26, 2009 (Agencywide Documents Access and Management System Accession No. ML091801061)

(Reference 1), related to modifying License Condition 2.B.(6) and create new License Conditions 1.J and 2.B. (7) as part of a pilot program to irradiateCobalt (Co)-59 targets to produce Co-60, for the Clinton Power Station, Unit No. 1 (CPS), the Nuclear Regulatory Commission (NRC) staff has determined that the following information is needed in order to complete its review:

RAI 1: , Evaluation of Proposed Changes, Section 3.0, Reference 1 Page4 of Attachment 1 states that: "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 desiredspecific activity."

Clarify the terms "varying amounts of time, " "desired specific activity," by specifying examples of amounts of time and specific activities. (See relatedRAI 7)

Response 1:

Varying amounts of time The term "varying amounts of time" refers to the operating time of the fuel rods. As with GEM4 bundles, the GE14i Isotope Test Assemblies (ITAs) are limited by the accumulated time period in the reactor core spent at operating temperature and by the peak pellet exposure. The peak pellet exposure limit of the GE14i ITAs is identical to the GE14 limit. This allows the GE14 and GE14i bundles to remain in the core for one to five 18-month cycles or less as long as they do not exceed the exposure limit.

For HCGS, the ITA bundles are planned to remain in the core for three to four 18-month cycles depending on subsequent core designs. This is from four-and-a-half to six years at operating temperature in the reactor core.

Desiredspecific activity The term "desired specific activity" refers to the specific activities that are sought after in the radioactive cobalt industries. The radiotherapy industries consider High Specific Activity (HSA) cobalt greater than 200 Ci/gram. Low Specific Activity (LSA) cobalt is also desired in the sterilization industries. LSA cobalt is considered to be less than 200 Ci/gram but usually greater than 100 Ci/gram. This program intends to produce HSA cobalt at HCGS. See response to RAI 9.c for examples of cobalt specific activity vs.

exposure.

Page 2 of 20

0000-0110-2515 Revision 0 Non-Proprietary Information RAI 2: , Section 4.3, Reference 1 In the section for "No Significant Hazards Consideration" of Attachment 1, it is stated that: "the effects on the spent fuel pool are minimal; post-irradiationhandling of the assemblies and the isotope rods will be performed under approved procedures, by experienced personnel." of the application (NEDC-33505P, Section 4.7.1) indicates that the maximum incident (gamma) radiation due to a GE14i bundle placed one foot from the spent fuel pool (SFP) wall is approximately 7.2E+10 MeV/cm 2 -sec, so concrete heating due to gamma would be significant. At 4 feet, it has been shown that the energy deposition rate of 1.4E+8 Me V/cm 2-sec is well below that required to cause significant concrete heating. Therefore, in order to minimize the effect of gamma heating on the SFP concrete walls, the irradiatedfuel storage procedures are modified to specify that the GE14ifuel bundles be stored at least 4 feet from the pool walls with no limitation on the amount of time a GE14i bundle may remain in the pool.

The applicantshould modify the section on "No Significant Hazards Consideration," the significant heating effect of gamma on the SFP and the procedures adopted to mitigate the heating effect.

Response 2:

The No Significant Hazards Consideration of the HCGS submittal can be found in the HCGS LAR.

RAI3:

Evaluation of Hydraulic Characteristicsof the GE14ifuel.

Section 5 of DRF Section 0000-0100-9379, "Scope of Verification" lists 4 items to be verified.

1. Verify that the analysis in "GE14i DP Evaluation.pdf'(neFile 0000-0100-9380) performed for the GEl4 pressure drop data with and without cold rods are accurate and adequate.
2. Verify that flow area calculation for the Hex region in "GE14i DP Evaluation.pdf'is performed accurately.
3. Verify that the conclusions in this design study are adequate.
4. Verify that Tables and Figures in this design study summary file are prepared correctly.

Page 3 of 20

0000-0110-2515 Revision 0 Non-Proprietary Information Please confirm that these verifications are completed.

Response 3:

All information developed by GEH to support licensing actions is based on verified design records per procedures that meet the requirements of IOCFR50 Appendix B. This is specifically true for the information developed to support the HCGS License Amendment Request.

RAI 4

Qualitative Assessment of Impact on Updated Safety Analysis Report Chapter 15 Transient & Anticipated Transient Without Scram (ATWS) Events (0000-0100-9699-Supplement-RO)

Section 3.3.4, Flexibility Options, indicate that: "The thermal-hydraulic characteristics and criticalpower performance of the GE14i ITAs [Isotope Test Assemblies] are not expected to be different than the GEl4 fuel (Reference 2) as explained in previous sections. In addition, 8 GE14i assemblies represent a very small fraction of the bundles in the Clinton Cycle 13 core to affect average core performance. Therefore, the flexibility options available to CPS will remain unaffected by the implementation of the GE14i assemblies. The fast pressurization events in combination with the licensed flexibility options for CPS will be analyzed as part of the reload transient analysis in Cycle 13. The thermal limits of the GE14i ITAs will be computed explicitly."

Providea summary of this analysis.

Response 4:

The response to RAI 4 is incorporated into Section 4.2, Evaluation of Other Transients, of NEDC-33529P - Revision 0, "Safety Analysis Report to Support Introduction of GE14i Isotope Test Assemblies (ITAs) in Hope Creek Generating Station".

RAI 5

Thermal-MechanicalEvaluation (Section 4.4 of NEDC-33505P), and GE14i Thermal-MechanicalEvaluation (DesignStudy summary, 4/29/2009)

1. Provide detailedproof to support the conclusion made in the Thermal-Mechanical evaluation regarding cladding failure expected due to mechanical interaction between the inner and outer cladding,applied over the lifetime of ITAs in the core.

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0000-0110-2515 Revision 0 Non-Proprietary Information

2. Regarding the thermal-mechanical evaluation, there is considerable pressure difference between the system and the internalcavity even though there is no fission product gas generation in the ITA fuel rods.

Explain in detail how cladding collapse due to this pressure difference and cladding creep is prevented from occurring during the lifetime of the ITA in the core.

3. Section 6 of the Design Study summary indicates that some of the inputs were outside of the range of acceptable inputs in to the GE Thermal-Mechanical Methodology, GSTRM, and these inputs were modified as necessary and were confirmed to have no impact on the final results. List the input parameters that were outside of the range of acceptable inputs to GSTRM and discuss if the extensions of these specific parameters are acceptable and explain whether additional margin is applied to the affected downstream safety analyses (as specified in the safety evaluationfor NEDC-33173P)

Response 5:

The generic report on thermal mechanical performance that was submitted on the Clinton Power Station (CPS) docket is applicable to HCGS as noted below:

1. The detailed response to RAI 5.1, the "GE14i Thermal-Mechanical Evaluation" (Reference 1), was previously docketed by CPS as Attachments 5 and 8 of the letter dated November 4, 2009 (ADAMS Accession Number ML093100313).

Specific to RAI 5.1, the analysis documented in Section 7.3 and 7.3.1 of Reference 1 concludes that the ((

)). Thus cladding failure due to mechanical interaction is not plausible. These results meet the 1% strain limit.

2. The detailed response to RAI 5.1, the "GE14i Thermal-Mechanical Evaluation" (Reference 1), was previously docketed by CPS as Attachments 5 and 8 of the letter dated November 4, 2009 (ADAMS Accession Number ML093100313).

Specific to RAI 5.2, the analysis documented in Section 7.4 of Reference 1 was performed to ensure that the cladding does not collapse ((

)) of the ITA cobalt isotope rods. This analysis was performed per the approved methodology in Reference 8.

This analysis explicitly accounts for the pressure difference between the system and internal cavity with no pressure credit for fission gas release. The results showed ovality to be unchanged before and after the end-of-life applied overpressure; thus the ITA cobalt isotope rods meet the (no) creep collapse requirement.

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0000-0110-2515 Revision 0 Non-Proprietary Information

3. The response to RAI 5.3 is incorporated into Section 4.4, Thermal-Mechanical Evaluation, of NEDC-33529P - Revision 0, "Safety Analysis Report to Support Introduction of GE 14i Isotope Test Assemblies (ITAs) in Hope Creek Generating Station".

RAI 6

Stability (Section 4.5.1 ofNEDC-33505P)

Section 4.5.1 states that "the section provides a qualitative assessment of the impact of GEJ4i ITAs on thermal hydraulic instability and demonstrated that an ITA is very unlikely to result in single-channelinstability."

DRF 0000-0100-6431 states that: "When implementing the long term solution, a procedure to review the thermal hydraulicstability of the lead use assemblies (LUA) in a core reload should be established. The review should ensure that inclusion of the LUA as proposed in the core reload is very unlikely to result in single-channel instability."

(a) Provide reasonable assurance with supporting analyses that, the thermal hydraulic stability as prescribed by the Option III with respect to the two stability aspects, namely, (i) the OPRM system setpoint, and (ii) the size of the backup stability protection (BSP) regions is maintained during the lifetime of the ITAs in the CPS core.

(b) Provide typical calculationssupporting(a) using core and channel decay ratios for CPSfacility.

Response 6:

(a)(i) The Option III stability long-term solution OPRM system setpoint has three component calculations: Hot Channel Oscillation Magnitude (HCOM), Delta CPR over Initial MCPR Versus Oscillation Magnitude curve (DIVOM), and change in Minimum Critical Power Ratio (MCPR) due to a 2-recirculation pump trip (DIRPT).

In accordance with Reference 2, the HCOM is independent of fuel type and is therefore unchanged by the implementation of ITAs in a core.

The DIVOM curve is calculated on a plant- and cycle-specific basis in accordance with Reference 3. The DIVOM calculation includes selection of limiting channels in the core, including the ITAs. For a mixed GE14/ITA core, limiting channels may be either GE14 or an ITA. Therefore, the DIVOM calculation incorporates the ITAs.

In accordance with Reference 2, the DIRPT is calculated on a plant- and cycle-specific basis. The MCPR, at both the rated power and off-rated power Page 6 of 20

0000-0110-2515 Revision 0 Non-Proprietary Information conditions, is the most limiting CPR in the core. For a mixed GE14/ITA core, the MCPR may occur in either GE14 or an ITA. The DIRPT calculation accounts for the ITAs explicitly.

A table of OPRM amplitude setpoint versus stability-based Operating Limit MCPR (OLMCPR) is provided for each reload in the SRLR. This provides assurance that the OPRM system setpoint is acceptable for each cycle that includes an ITA.

(a)(ii) The Option III stability BSP regions provide protection in the case that the OPRM system is inoperable. In accordance with Reference 4, the BSP regions are calculated on a plant- and cycle-specific basis. The BSP region is expanded or contracted each cycle in accordance with the specific ODYSY computer code acceptance criteria for core and channel decay ratio in accordance with Reference 5.

The core decay ratio is calculated for each cycle-specific core design, for example a mixed GE14/ITA core, in accordance with Reference 4. A hot channel decay ratio is also calculated for each reload in accordance with Reference 4. For a mixed GE14/ITA core, the hot channel may be either GE14 or an ITA.

Therefore, the core and channel decay ratio calculations incorporate the ITAs.

The plant- and cycle-specific BSP regions are provided for each reload in the SRLR. This provides assurance that the BSP regions are acceptable for each cycle that includes an ITA.

(b) Commitments for providing typical calculations supporting use of core and channel decay ratios for HCGS facility are discussed in the HCGS LAR.

RAI .7:

Reference 9 of Safety Analysis Report, GEJ4 Fuel Design Cycle-Independent Analyses for CPS This document summarizes the cycle-independent analyses that are performed for the GEJ4 reloads and reported in the plant and cycle unique Supplemental Reload Licensing Reports. This report primarilyprovides the result of cycle-independent GEM4 analyses and evaluations. These analyses include (1) Loss-of-Coolant Accident, (2) Reactor Internal Pressure Differences, (3) Reactor Pressure Vessel Internals Structural Evaluation/Assessment, and (4) Recirculation Pump Seizure: Single-Loop Operation.

This report contains other technical areas,such as, Decay Heat, Fuel HandlingAccident, Neutron Fluence Impact, ATWS, and Special Transient - Pressure Regulator System Single Failure (Cycle 8 specific analysis including GEM4 and current operating fuel types).

Page 7 of 20

0000-0110-2515 Revision 0 Non-Proprietary Information Because of the importance of the analyses describedin this report to the evaluation of the license amendment application, the staff requests that this document be submitted to the NRC document desk.

Response 7:

NEDC-33529P - Revision 0, "Safety Analysis Report to Support Introduction of GE14i Isotope Test Assemblies (ITAs) in Hope Creek Generating Station" includes sufficient information such that the HCGS Fuel Transition Report did not need to be referenced.

Applicable portions of NEDC-33158P, Revision 4, "Fuel Transition Report for Hope Creek Generating Station," March 2005, pertaining to reactor recirculation pump seizure are included in Section 4.2, Evaluation of Other Transients, ofNEDC-33529P - Revision 0, "Safety Analysis Report to Support Introduction of GE14i Isotope Test Assemblies (ITAs) in Hope Creek Generating Station".

RAI 8

Evaluation of Hydraulic Characteristics of the GE14i fuel (DRF Section 0000-0100-9379, April 22, 2009), and the application of GEXL + correlation.

(a) If full scale pressure drop data from GE14i fuel is not available, explain the procedure how pressure drop data were collected from the full-scale GEJ4 criticalpower testing with zero-power rods.

(b) With the number and location of cold rods in the GE14ifuel are different from the stern test configuration with zero-power rods, and the use of as few as four different rod-to-rodpower distributions with zero power rods, how reasonably accurate is the statement "the cold rod impact on the pressure drop characteristics of the GE14i fuel is negligible, or within the uncertainty establishedfor the GE14 fuel."

(c) Explain the validity of the derived GEXL correlation and the R-factor with the statisticallylimited number of trials and data points?

(d) Explain why only outlet-peaked axial power shape was used in the GEXL correlation development for the GE14i fuel assembly, and not bottom/cosine peaked shapes.

Response 8:

(a) Pressure drop measuring instrumentation was installed along the heated length of the GE14 full-scale test section. During the full-scale GE14 critical power testing with zero-power rods, the pressure drop data were also collected. ((

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0000-0110-2515 Revision 0 Non-Proprietary Information

)) the pressure drop data are adequate to evaluate the pressure drop characteristics of the GE I4i fuel as further discussed in item (b) below.

(b) The detailed response to RAI 8(b), "Evaluation of Hydraulic Characteristics of the GE14i Fuel" (Reference 6), was previously docketed by CPS as Attachments 4 and 7 of the letter dated November 4, 2009 (ADAMS Accession Number ML093100313) and is applicable to HCGS.

Specific to RAI 8(b), ((

Although the rod-to-rod power distributions are not expected to have a more significant impact on the pressure drop characteristics than the cold rod configuration, four different rod-to-rod power distributions tested are considered as representative patterns for the evaluation of the cold rod impact as further explained in item (c) below.

In summary, it is concluded from the justifications provided above that data collected during the GEI4 testing with zero-power rods are adequate to evaluate the cold rod impact on the pressure drop characteristics of the GE14i fuel. The evaluation of the hydraulic characteristic of the GE14i fuel is provided in Reference 6.

(c) The detailed response to RAI 8(c), "Additional GEXL Information on the GE14i ITAs" (Reference 7), was previously docketed by CPS as Attachments 6 and 9 of the letter dated November 4, 2009 (ADAMS Accession Number ML093100313) and is applicable to HCGS.

Specific to RAI 8(c), it is noted that the GEXL critical power correlation for the GE14 fuel (GEXL14) was developed using data obtained from the ATLAS critical power test facility. The GE14 critical power database used in the GEXL14 development covers wide ranges of fluid conditions and a number of rod-to-rod power distributions (( )). The GEXL14 correlation was further validated against additional ((

)) data generated in the ATLAS facility and ((

)) data generated in the Stern facility as discussed in Reference 10.

Requirements on GEXL correlation are provided in GESTAR II Sections 1.1.7 and 1.2.7 (Reference 9). Relevant texts in GESTAR II are excerpted below:

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0000-0110-2515 Revision 0 Non-Proprietary Information 1.1.7 CriticalPower Correlation A. The currently approved criticalpower correlationswill be confirmed or a new correlation will be established when there is a change in wetted parametersof the flow geometry; this specifically includesfuel and water rod diameter, channelsizing and spacer design.

1.2.7 CriticalPower Correlation A. The coefficients for the criticalpower correlation of a fuel design will be determined generically based on the criteria documented in Subsection 1.1.7 (GESTAR-JI). The fuel design parameters given in these criteria are those which have the primary effect on determining the need for a new critical power correlation when there is a change in the fuel design. New coefficients for the critical power correlation will be provided in the fuel design information report.

The GE14i ITA is identical to the GE14 bundle with the exception of the cobalt isotope rods in GEI4i ITAs. Due to the similarity between GE14 and GE14i, the currently approved GEXL14 correlation can be applied to the GE14i ITAs, provided that the effects of the difference on the critical power performance are quantified and properly accounted for.

It was shown in the Section 3.3 of the subject SAR that the GEXL14 correlation conservatively predicted the critical power data with zero-power rods. The database used in the comparison is adequate to confirm the applicability of the GEXL14 to the GE14i ITAs. It has been shown that the various fuel assembly and channel geometries such as cold rods (water rods) or vacant lattice positions due to the part length rods are well characterized by the R-factor methodology in the GEXL correlation. ((

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0000-0110-2515 Revision 0 Non-Proprietary Information In summary, the GE14 critical power data with zero-power rods areadequate to confirm the applicability of the GEXL14 and the R-factor methodology to the GE14i ITAs and evaluate the effects of the difference between GE14 and GE14i ITAs on the critical power performance. As a result of the evaluation, additional CPR margin is applied to the GE14i ITAs. Supporting GEXL analysis to the additional CPR margin is provided in Reference 7.

(d) The response to RAI 8(d) is incorporated into Section 3.3, GEXL+ Correlation, of NEDC-33529P - Revision 0, "Safety Analysis Report to Support Introduction of GE 14i Isotope Test Assemblies (ITAs) in Hope Creek Generating Station".

RAI 9

Letter of May 19, 2009, from NRC (PeterBamford) to Exelon Nuclear (Charles Pardee),

Clinton Power Station, Unit No. 1 - Withdrawal of License Amendment Request RegardingBulk Isotope Generation Project(TAC No. ME0657)

"InformationNeeded" section of this letter requiredthat "this specific request is beyond the scope of both the GESTAR-If and the LTA programs," as well as the NRC staffs approval of these processes. The letter further stated that: "In order to review this application,it must be structuredso that it can be evaluated without relianceon the LTA program and GESTAR-JI." The NRC staff requests the licensee provides the following information that was listed in the letter.

(a) Provide, in detail, all probable isotope production rod failures. List and explain all probable means by which the target rods can fail during loading, operation and offloading from the core. Explain also, the administrative and other controls which will be in place to mitigate consequences of suchfailures.

(b) Provide isotope production rod design limits, including, but not limited to, the expected and design maximum Co-60 activities per rod model. Also, describe the prototype testing associated with the conditions of use (high neutron and gamma fields for years, exposure to corrosive materials, temperature,pressure, puncture, dropped source, torque, and build up of expected radioactivity including activation of contaminants).

(c) Using the estimated neutron flux at the location of target rod, provide a mathematical analysis to show the time to reach equilibrium activity between production and decay of Co-60 isotope.

(d) Provide,for agency records, detailed engineering drawings and specifications of the assembly and target rods. Specificallyprovide the items listed below:

DrawingNumber Title

  • 147C1233 Canister
  • 147C1236 Inner Tube Page 11 of 20

0000-0110-2515 Revision 0 Non-Proprietary Information

  • 147C1237 Inner Tube Cap 0 147C1238 Outer Tube
  • 147C1239 Female Threaded Canister 0 147C1240 Male Threaded Connector a 147C1241 Lower End Plug Extension
  • 147C1242 Upper End Plug Extension
  • 147C3356 Inner Capsule (Co-59 Bearing) 0 147C3357 Rod Segment a 147C3358 Rod (segmented) 0 147C3359 Isotope Target (e) Providedetails of the impact of Co-60 rods on predictedcore power.

(f) Provide details of the impact of Co-60 rods on instrumentation and measured core power.

Response 9:

(a) The response to RAI 9(a) is incorporated into Section 2.2, Cobalt Isotope Rod Failure Mechanism Controls, of NEDC-33529P - Revision 0, "Safety Analysis Report to Support Introduction of GE14i Isotope Test Assemblies (ITAs) in Hope Creek Generating Station".

(b) All mechanical design limits applicable to standard GE14 fuel rods have also been applied to isotope production rods.

The activity of the cobalt rods is not a design limit for the GE14i ITAs. The activity of the cobalt isotope rods is a function of the bundle exposure and bundle residence time. The expected activity of the cobalt isotope rods is given in Figures 3-1 through 3-3 of NEDC-33529P.

There has been no irradiated prototype testing completed on these fuel assemblies that HCGS can take credit for. However, the analysis documented in NEDC-33529P - Revision 0, "Safety Analysis Report to Support Introduction of GE14i Isotope Test Assemblies (ITAs) in Hope Creek Generating Station" shows that there will be no significant impact from operation in the HCGS 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.

Although there is no direct prototype experience for the GE14i ITAs, the materials and bundle configuration were purposely selected to be the same as GE14 - the design for which GNF has now deployed in approximately 26,000 bundles with over 10 years of successful operating experience. The selection of this well-established bundle design reduces risk and performance uncertainty such that a specific prototyping campaign was not necessary.

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0000-0110-2515 Revision 0 Non-Proprietary Information The GE14i ITAs are being introduced under similar constraints as any GNF new fuel design. The GE14i ITAs are being introduced with a number of conservatisms:

1) Introduced in limited quantities of approximately 2% of the core size
2) Introduced in non-limiting locations with respect to thermal limit margins and shutdown margins
3) Receive additional inspections after first outage and after final discharge This conservative introductory approach and associated inspection during and after operation are performed to gather information and confirm expected performance. This data will provide additional data in terms of bundle performance in high neutron and gamma fields for years, exposure to corrosive materials, temperature, pressure, puncture, dropped source, torque and build up of expected radioactivity including activation of contaminants.

(c) The response to (c) is generic information that is of general interest to cobalt production.

The number density of the cobalt-60 can be calculated as:

N 60 =((N N59

  • 0-59
  • 0)/) e,*t where N - number density of isotope of interest a Co 59microscopic thermal (2200 m/s) absorption cross section 0- thermal flux in region of interest 2A- Co 6 ° decay constant t - Irradiation time For t -> oo N 60 - (N 59
  • o-59 ,)/

/ /2 1

This calculation assumes that the number density of cobalt-59 is constant, the thermal flux is constant, and the absorption cross section is a constant as a function of time. The equilibrium results (i.e., t=oo) of an evaluation for a node operating at 50 kw/l and with a 40% average in-channel void fraction are provided in Table 1.

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0000-0110-2515 Revision 0 Non-Proprietary Information Table 1: Equilibrium Cobalt-60 Inventory Rod Co-59 Co-59 Half- Flux XS COW Co-60 Co-60 Activity Activity/

Location Density Atom Life (15Gwd/st, (0.025ev) Atom Density Total /Linear- Segment (smeared Density 40% VH) Density Mass/ cm (36.6cm) over rod Linear-volume) cm (gm/cc) Atom/ (sec) N/cm^2-sec barns Atom/ (gm/cc) grams Curies Curies barn-cm barn-cm (1,1) a 1.66e8 5.96e13 37.0 ((

(2,1) 1.66e8 4.84e13 37.0 (10,1) 1.66e8 4.92e13 37.0 ))

In the actual application, the peak cobalt-60 inventory is reached at approximately

)) Gwd/st lattice average exposure. This exposure level is significantly greater than the exposure at which the U0 2 pellet exposure limit is reached. A typical GEl4 bundle is limited to approximately (( )) Gwd/st lattice average exposure by the U0 2 (( )) Gwd/mt peak pellet exposure limit. The end of irradiation for the GE14i bundles is expected to be less than (( )) Gwd/st lattice average exposure. The TGBLA predicted cobalt-60 inventory is provided in Figure 1.

Figure 1: TGBLA Predicted Co60 Inventory (d) As a general policy, GEH does not provide detailed engineering drawing and specifications for the NRC's records. The NRC Staff may examine these drawings at the GEH facilities in an audit forum at any time. Further, if the NRC Page 14 of 20

0000-0110-2515 Revision 0 Non-Proprietary Information desires to perform confirmatory calculations, GEH will respond to specific requests for design information.

(e) Total Core power or reactor power is determined by a heat balance based on measured reactor parameters such as feedwater flow, reactor pressure, control rod drive flow, feedwater temperature and other parameters. The GE14i bundles will not impact the determination of total core power.

The core power (nodal) distribution is predicted by the three-dimensional reactor simulator PANAC 11. The characteristics of the GE14i bundle are provided by the lattice physics method TGBLA06. The impact of the GE14i bundle and the cobalt isotope rods is explicitly modeled in the TGBLA06 methodology and all contributions to the core power and local fuel rod power from the cobalt isotope rods are accounted for. Heat generation from the n,y reaction is accounted for in the explicit model for the lattice effective Mev/fission. This effective Mev/fission model accounts for the heat generation from neutron capture events in all actinides, fission products, structural materials, and cobalt material.

A comparison of the uncertainty of the pin-wise fission density was provided in Table 3-2 of NEDC-33529P. This comparison indicates that pin-wise fission density or power is consistent with the uncertainty of the standard GE 14 product.

(f) The response to RAI 9(f) is incorporated into Section 3.2.3, In-Core Instrumentation, of NEDC-33529P - Revision 0, "Safety Analysis Report to Support Introduction of GE14i Isotope Test Assemblies (ITAs) in Hope Creek Generating Station".

RAI 10

Letter referenced in RAI 9 (a) Provide the necessary and sufficient technical information to show that the target pellets are of high purity to minimize the production of unwanted/unanalyzedisotopes.

(b) Provide assurance that there will be no cobalt contamination (i.e., no encapsulatedCo-59) loaded in to the reactor.

Response 10:

(a) The response to RAI 10(a) is incorporated into Section 2.1, New Design Features, and Section 4.6, Manufacturing Quality Assurance, of NEDC-33529P - Revision 0, "Safety Analysis Report to Support Introduction of GE14i Isotope Test Assemblies (ITAs) in Hope Creek Generating Station".

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0000-0110-2515 Revision 0 Non-Proprietary Information (b) The response to RAI 10(b) is incorporated into Section 4.6, Manufacturing Quality Assurance, of NEDC-33529P - Revision 0, "Safety Analysis Report to Support Introduction of GE14i Isotope Test Assemblies (ITAs) in Hope Creek Generating Station".

RAI 11

Letter referenced in RAI 9 The applicationdoes not indicate reintroductionor any restrictionson the reintroduction of cobalt rod segments. The applicationshould describe if there are any plans to shuffle rod segments (i.e., disassemble rods and swap rod segments within a rod, or with a rod in a different location in the same or different Co-60 isotope production assembly),

between power cycles (during refueling outages).

Response 11:

The response to RAI 11 is incorporated into Section 4.7.3, Post-Irradiation Handling, of NEDC-33529P - Revision 0, "Safety Analysis Report to Support Introduction of GE14i Isotope Test Assemblies (ITAs) in Hope Creek Generating Station".

To further clarify, there are no plans to shuffle rod segments (i.e. disassemble rods and swap rod segments within a rod, or with a rod in a different location in the same or different cobalt-60 isotope production assembly), between power cycles (during refueling outages).

RAI 12

Section 4.6, ManufacturingQuality Assurance, of NEDC-33505PRev 0 (a) Provide detailedprocedure of how the rod integrity is verified by helium leak check of both the inner and outer tubes following welding of the tubes during the manufacturingprocess.

(b) Describe details of methods andprocedures that are in place to ensure the rod integrity during the lifetime of the target rods while in the CPS core.

Response 12:

The following response is applicable to Section 4.6, Manufacturing Quality Assurance, of NEDC-33529P Revision 0, "Safety Analysis Report to Support Introduction of GE14i Isotope Test Assemblies (ITAs) in Hope Creek Generating Station".

(a) ((

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0000-0110-2515 Revision 0 Non-Proprietary Information (b) The response to RAI 12(b) is incorporated into Section 2.3, Online Failure Detection, of NEDC-33529P - Revision 0, "Safety Analysis Report to Support Introduction of GE14i Isotope Test Assemblies (ITAs) in Hope Creek Generating Station."

RAI 13

Non-limiting Core Locations.

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0000-0110-2515 Revision 0 Non-Proprietary Information (a) The applicant intends to place a limited number of ITAs in the CPS non-limiting core locations with respect to thermal limit margins and shutdown margin.

Provide the details of analyses and methodologies used to identify the locations that are non-limiting locations. Provide the criteria and the key parameters used to determine the non-limiting locations.

(b) Since ITAs are to be used in the subsequent cycles, provide justification and assurancefor the assumption/prediction that the ITAs will remain in non-limiting core regions during the subsequent CPS cycles.

Response 13:

(a) The analysis and methodologies used to identify the non-limiting locations are identical to the analysis and methodologies used to design current reloads. Three-dimensional analysis is performed to simulate the fuel performance based on the chosen locations for the ITAs and assure that these bundles are not limiting in regards to both thermal margins and reactivity margins.

The definition of a non-limiting location for thermal margins is a bundle location that does not result in the highest core MFLCPR, MFLPD, and MAPRAT values at any exposure throughout the cycle.

Definitions:

MFLCPR: Maximum Fraction of Limiting Critical Power Ratio MFLPD: Maximum Fraction of Linear Power Density MAPRAT: Maximum Average (nodal) Power RATio The OLMCPR for the ITA bundles will include an additional (( )) adder compared to the other GE14 bundles, such that all bundles are monitored to the same MFLCPR margin value. The thermal-mechanical limits applied to the ITA bundles will include additional power suppression at high lattice exposures such that all bundles are monitored to the same MFLPD margin value.

The definition of a non-limiting location for reactivity margins is any four-bundle cell containing a single ITA that does not result in the minimum core Shutdown Margin (SDM) value at any exposure statepoint throughout the cycle. The ITA cells will include an additional (( )) SDM with respect to other limiting SDM cells at the same cycle exposure statepoint.

(b) During the cycle of introduction, three-dimensional analyses are performed for subsequent cycles to assure that the ITAs will remain in non-limiting locations.

During the detail design of subsequent cycles, the ITA locations will be chosen in order to assure they are in non-limiting locations as was done for the cycle of introduction.

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0000-0110-2515 Revision 0 Non-Proprietary Information

RAI 14

NEDC-33505PRev 0 Section 3.2.1.3 ITA Margin Consideration.

Regarding the reference section, explain which of the power suppression options is used to accommodate the power peak and provide the technical basesfor the selection.

Response 14:

The response to RAI 14 is incorporated into Section 3.2.1.3, ITA Margin Considerations, of NEDC-33529P - Revision 0, "Safety Analysis Report to Support Introduction of GE 14i Isotope Test Assemblies (ITAs) in Hope Creek Generating Station".

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0000-0110-2515 Revision 0 Non-Proprietary Information

References:

1. Global Nuclear Fuel, "GE14i Thermal-Mechanical Evaluation," GNF-0000-0108-6874-RO, October 2009.
2. GE Nuclear Energy, "Reactor Stability Detect and Suppress Solutions Licensing Basis Methodology for Reload Applications," NEDO-32465-A, August 1996.
3. GE Nuclear Energy, "Plant-Specific Regional Mode DIVOM Procedure Guideline,"

GE-NE-0000-0028-9714-Ri, June 2005.

4. GE Nuclear Energy, "Backup Stability Protection (BSP) for Inoperable Option III Solution," OG 02-0119-260, July 2002.
5. GE Nuclear Energy, "ODYSY Application for Stability Licensing Calculations,"

NEDC-32992P-A, July 2001.

6. Global Nuclear Fuel, "Evaluation of Hydraulic Characteristics of the GE14i Fuel,"

GNF-0000-0 108-9509-RO-P, October 2009.

7. Global Nuclear Fuel, "Additional GEXL Information on the GE14i ITAs," GNF-0000-0108-9523-RO-P, October 2009.
8. Global Nuclear Fuel, "Cladding Creep Collapse," NEDC-33139P-A, July 2005.
9. GE Hitachi Nuclear Energy, "GE14 Compliance With Amendment 22 of NEDE-24011-P-A (GESTAR II)," NEDC-32868P, Revision 3, April 2009.
10. GE Hitachi Nuclear Energy, "GEXL14 Correlation for GE14 Fuel," NEDC-32851 P-A, Revision 4, September 2007.
11. Letter from NRC (Cameron S. Goodwin) to Exelon (Charles G. Pardee), "Clinton Power Station, Unit No. 1 - Request for Additional Information Related to License Amendment Request to Modify Clinton Power Station Facility Operating License in Support of the Use of Isotope Test Assemblies (TAC NO. ME1643)," October 16, 2009 (ADAMS Accession Number ML092800374).
12. Letter from NRC (Cameron S. Goodwin) to Exelon (Charles G. Pardee), "Clinton Power Station, Unit No. 1 - Request for Additional Information Related to License Amendment Request to Modify Clinton Power Station Facility Operating License in Support of the Use of Isotope Test Assemblies (TAC NO. ME1643)," October 16, 2009 (ADAMS Accession Number ML092800426).
13. Letter from Jeffrey L. Hansen (Exelon Generation Company, LLC) to U. S. NRC, "Additional Information Supporting the Request for a License Amendment to Modify Clinton Power Station Facility Operating License in Support of the Use of Isotope Test Assemblies (TAC NO. ME1643)," November 4, 2009 (ADAMS Accession Number ML093100313).

Page 20 of 20 LAR H09-01 LR-N09-0290 LIST OF REGULATORY COMMITMENTS The following table identifies those actions committed to by PSEG in this document. Any other statements in this submittal are provided for information only purposes and are not considered to be regulatory commitments.

Commitment Type One-Time Programmatic Regulatory Commitment Committed Date AcTion (Yes/No)

Action (Yes/No)

(Yes/No)

Revise applicable Spent Fuel Pool Implementation of the Yes No Storage procedures to require approved amendment storage of irradiated GE14i fuel bundles at least four feet from the wall of the SFP Typical calculations supporting use July 8, 2010 Yes No of core and channel decay ratios for HCGS facility will be provided.

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