Waste Control Specialists LLC PO Box 1129

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RT-100 Transport Cask Contents Description

Rancho Seco Reactor Vessel Internals

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CONTENTS INTRODUCTION .......................................................................................................................................................... 3 1.0 RADIOLOGICAL CHARACTERIZATION ..... ..... ... ..... ... ..... ... ..... ... ..... ... ..... ... ..... ... ..... ... ..... ... ..... ... ..... ... .. ...... 6 2.0 DECAY HEAT (THERMAL LOAD) .............................................................................................................. 10 3.0 APPENDIX ........... ... ... ..... ... ..... ... ..... ... ..... ... ..... ... ..... ... ..... ... ..... ... ..... ... ..... ... ..... ... ..... ... ..... ... ..... ... ..... ... ..... ... .. .... 11 4.0 REFERENCES ................................................................................................................................................. 13

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INTRODUCTION

Rancho Seco was a 913-megawatt B&W des igned nuclear power plant ow ned by the Sacramento Municipal Utility District (SMUD) that began commercial operat ion in 1975. It was shut down in June of 1989 as the result of a voter referendum. Due to a minimal decommissioning fund balance, the decision was made to enter an extended period of SAFSTOR to allow the activity to decay and the fund to build to a level that would allow dismantlement.

The Sacramento Municipal Utility District (SMUD), operator of the Rancho Seco reactor site, has secured funding to ship a total of 23 liners of Class Band C waste in calendar year 2014. SMUD has contracted for disposal of all 23 B&C liners at WCS in 2014, with the only contract contingency being the availability of a suitable transportation cask. The current estimate of the number of liners that will need a Type B cask is a minimum of 8, and a maximum of 14. This estimate is based on calculation from surveys performed before the wastes were placed in shielded storage (circa 2006). Some of the liners will be suitable for shipment in a Type A cask; this determination depending on updated dose surveys that will be performed when the liners are removed from shielded storage in preparation for disposal.

SMUD has committed to the NRC and SMUD stakeholders to remove Class Band C wastes from the IOSB as soon as a disposal option becomes available. Authorization of a cask to transport the waste will support those commitments, and provide corresponding benefits in risk and dose reduction at the site. Removal of the Class B and C wastes stored there will remove a final barrier to completion of Part 50 license decommissioning activities that began in 1989 [l].

Rancho Seco 's Reactor Vessel and Internals are of B&W design and without the vessel head, weigh approximate ly 500 tons and are 18' - 6" in diameter and 38' in length. The major components of the Reactor Vessel Internals (RVI) consists of the follow ing:

Tablel - RVI Components - Volumes and Weights

Major Component Weight( Kg) Envelope Volume (m3) Baffle Plates (GTCC) 8172 1.02 Baffle Fo rmers (GTCC) 3464 0.43 Plenum Cover 24786 3.10 Plenum Cylinder 6650 .83 Control Rod Guide Tubes 6892 .86 Upper Grid 628 0.79 Co re Support Shield 28579 3.5 Core Barrel 18147 2.27 Thermal Shield 18799 2.35 Lower Internals- Top Section 12643 1.58 Lower Internals- M iddle 12697 1.59 Section Lower Internals- Lower 5154 0.64 Section

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Figure 1 - RVI Components

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Core Support Shield

ore Barrel

i----Thermal Shield

Lowe< Internals

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1.0 RADIOLOGICAL CHARACTERIZATION

From surveys conducted of the Core Shroud, Plenum Assembly and Lower Core Support Assembly, David James & Associates, in conjunction w ith Duratek, performed assessments on the majo r components within the Reactor Vessel; also know as the Reactor Vessel Internals (RVI) . More specifically, activity assessments were performed on the Baffle Plates, Baffle Formers, Core Ba rrel, Thermal Shield, Upper Grid, Lower Internals - Top Section, Lower Internals - Middle Section, Lower Internals - Bottom Section, Plenum Cover, Plenum Cylinder and Control Rod Gu ide Tubes (CRGTs). Taking into account weight, volume, radiation survey information, material composition, and operating history (flux distribution) the activated Co-60 concentration was determined for each of these major components. The other nuclides' activated activity was scaled to Co-60 [2].

Activation of major components

A radiological characterization of the was done using the Rancho Seco operating history to estimate the activity content of the Rancho Seco internals and distribute the activity by major sections. These calculations cover the Upper Internals (control rod guide tubes, and core support shield) [3], Core Region Internals (core baffle plates, baffle formers, core barrel, and thermal shield) [4], and Lower Internals (lower core grid rib assembly, in-core guide support plate, in-core guide tubes distributor late, flow distributor, and support ring) [5] . The plant was operated for approximately 7 cycles including and extended shut down period from December 1985 until April 1988. The material specification used for these calculations was a nominal specification of SS304, and the flux specification for the activation calculations were defined using one-dimensional Sn computer code (ANISN) radial and axial studies.

The ANISN models were developed using BUGLE-96, 47 group cross section libraries with a Legendre P3 order of scattering [6] [7] [8] . An ANISN model for the analysis of spent fuel hardware specific to a B&W reactor core with 165 inch fuel assemblies equivalent to those used at Rancho Seco. This model was extended to include the upper and lower internal reactor vessel internal components for Rancho Seco. The calculation was performed in two parts. The first extended upward from the core center line to the top of the control rod guide tubes, and the second extended downward to the inner surface of the reactor vessel bottom head. The radial model for Rancho Seco was developed using averaged core parameter from the axial model. It extended from the center of the core to 30 cm inside the rector biological shield [9].

Activation analysis of the major components was done using the average total flux that corresponded to the major component in a Simplified Neutron Activation Program (SNAP) [10]. Comparisons were run using ORIGEN 2.1 [11]. The calculations were run to a nominal date of 4/1/2003 corresponding to the time period when surveys were performed. The total decay from the end of the exposure was 5045 days.

Table 2 - Component Activated Co60 Activity Greater Than Class C {GTCC)

Major Component Co60 {Curies) Baffle Plates 22550 Baffle Formers 6030

Class B & C Waste Major Component Co60 {Curies) Upper Grid 199 Core Ba rrel 7440 Thermal Sh ield 3010 Lower Internals- Top Section 1910 Waste Control Specialists LLC Page 5of13 RT-100 Transport Cask Contents Description

Class A Waste Major Component Co60 {Curies) Plenum Cover Slight Activation Activity Pl enum Cylinder 0.142 Control Rod Guide Tubes (CRGTs) 2.06 Core Support Shield 27.5 Lower Internals- M iddle Section 32.8 Lower Internals- Lower Section 6.54

The table below provides the dose rate on contact of the container (liner) as modeled using Microshield®. The waste density was evaluated using the provided container weight, the tare weight and the volume of each container. The waste matrix was assumed to be carbon for resins, and aluminum was used to mimic the irradiated metal and filters. For activated metal, only Co-60 was considered a dose driver. The activity was decay corrected from 2005 to February 2014 and dose was also estimated and provided.

The dose rate obtained by modeling with Microshield® assumes a uniform distribution throughout the entire volume of the liner. This might explain the discrepancy between some of the measured and the estimated dose rate. This difference is larger for some activated metal and might be due to the geometrical form of the hardware and the way the waste was packed in the liner. Overall, a good agreement is found between the estimated and the measured dose rate values. The highest corrected dose rate is estimated to be 65 R/hr and 4 R/hr on contact and at 3 meters, respectively

Table 3 - Measured and estimated dose rate (mR/hr) using Microshield

Container Contact Measured Corrected Dose rate Corrected

ID Dose rate Dose rate Dose rate @3 meter Dose rate in 2005 in 2005 on contact in 2005 @3 meter

RVl-008 l.50E+05 7.00E+04 5.20E+04 9.10E+03 3.10E+03 RVl-016 2.00E+05 4.16E+05 6.50E+04 l.20E+04 4.00E+03 RVl-017 2.00E+05 6.20E+05 6.40E+04 l.10E+04 4.00E+03 RVl-018 2.00E+05 4.16E+05 6.40E+04 l.10E+04 4.00E+03 RVl-019 l.70E+05 l.OOE+06 4.80E+04 8.87E+03 3.45E+03 RVl-021 5.00E+04 7.20E+04 l.60E+04 3.00E+03 l.OOE+03 RVl-022 5.50E+04 5.00E+04 l.80E+04 3.70E+03 l.27E+03 RVl-026 6.20E+04 4.00E+04 l.50E+04 4.20E+03 l.50E+03 RVl-027 6.30E+04 3.50E+04 2.10E+04 4.25E+03 l.50E+03 RVl-029 6.50E+04 l.OOE+05 2.30E+04 4.50E+03 l.60E+03 RVl-033 6.70E+04 5.00E+04 2.30E+04 4.70E+03 l.60E+03 RVl-034 6.70E+04 4.50E+05 2.30E+04 4.70E+03 l.60E+03 RVl-035 6.50E+04 4.50E+05 2.20E+04 4.50E+03 l.55E+03 RVl-037 5.50E+04 5.00E+04 l.90E+04 3.80E+03 l.30E+03

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Surface Contamination (Loose Contamination)

The activity due to surface contamination was determined by a Microshield evaluation based on a radiation survey of the Plenum cover (correcting for upper grid contribution) and the nuclide distribution was defined per SMUD sample # 9080-1. The result of the Microshield evaluation was that the Plenum Cover had 1.74 Curies of Co-60 and based on the surface area of the cover the contamination source term is equal to 1.83 eS Bq/cm2 total ~y, and l.92e3 Bq/cm2 total a. This is the assumed source term for all components.

Table 4- Evaluation Used to determine the activity from Surface Activity Microshield Results 201 mr/hr/Ci Avg Dose Rate 550.25 mr/hr Shine correction 200 mr/hr Eff Avg Dose Rate 350 mr/hr Total Co60 1.74 Ci

Table 5 - Sample results, Surface or "Contaminated" Activity, scaled to Co60 Sample # 9080-1 µCi/Sample Ratio To CoGO Co60 2.28e-1 l.OOe+O Cs137 3.83e-4 l.68e-3 C14 l.57e-2 6.90e-2 Fe 55 6.90e-2 3.03e-1 Ni63 l.03e+O 4.54e+O Sr90 l.25e-3 6.67e-3 Nb94 7.69e-4 3.38e-3 Tc99 6.52e-6 2.87e-5 Pu238 3.69e-4 l.62e-3 Pu239/Pu240 3.lSe-4 l.38e-3 Pu241 l.13e-2 4.98e-2 Am241 l.86e-3 8.17e-3 Cm242 l.32e-5 5.78e-5 Cm242/Cm243 2.59e-4 l.14e-3

MCNP was used to evaluate the dose rate contribution on the major components from contamination. Microshield was then used to estimate the contact dose rate based on 1 curie of contamination in the cover. The average dose rate from the survey was corrected to remove the influence of the activated component and a total cobalt activity was determined from the corrected dose using the Microshield conversion.

Class B & C Waste Segmentation

The governing procedure for the segmentation and tracking of the Reactor Vessel Internals was SMUD-OP-3. It contained various log sheets, detailing the segmenting, tracking, surveying and disposition of each component and each eventual package. The segmentation of the Reactor Vessel internals involved the generation of waste in three general categories: Greater Than Class C (GTCC) waste, Class B & C waste, and Class A waste. Much of the segmenting of the Class B & C components occurred concurrently with that of the Class A waste.

In preparing for the segmenting of the Class B & C waste, Duratek fabricated twenty-one, 8-120PL metal liners and overpacks [12]. Seventeen of these hold Class B or Class C waste whi le being kept in long-term IOSB, located at Rancho Seco. These liners were designed to be loaded, under water. This design incorporated vent and screened drain holes to accommodate the eventual draining of the water- after the liner was filled. The manufacturer's serial

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number, and a project-specific, alphanumeric package number identified each liner. These are the numbers logged on to the package accountability log sheets included in the Loaded Container Document Packages [13] [14] [15] [16] [17] [18] [19] [20] [21] [22] [23] [24] [25] [26]. The full package included an outside "overpack", designed to be automatically latched to the liner upon its loading into the transfer cask. The overpacks are marked with the matching manufacturer's serial number as its liner.

The major components that were classified as Class B or Class C Waste are: Core Barrel, Thermal Shield, Upper Grid, and Lower Internals-Top Section. The plan involved the separation of the Upper Grid assembly from the Control Rod Guide Tubes. The Core Barrel was separated from the Core Support Shield. The Core Barrel was segmented into 27 components. Each one was surveyed, weighed and loaded into a liner. The same technology and process was followed in segmenting the Thermal Shield; it also was segmented into 27 components. When separated, the Upper Grid segmentation was completed by making vertical cuts through horizontal surface. The remaining B & C waste consisted of components of the Lower internals - Top and Middle sections. The component loading process was the same for all Class B & C components; they were weighed, surveyed and loaded into liners. All Upper Grid, Core Barrel, and Thermal Shield, and Lower internals - Top and Middle components are B or C Waste, stored in 8-120 PL liners, and are in cells at the IOSB. After filling a liner, the lid was put on and bolted closed.

The liner maximum weight of 14,100 lbs was not exceeded. This maximum weight was based on the maximum weight allowed to be loaded in the 8-120B Transport Cask of 14,600 lbs including the 500 lb overpack. The maximum weight allowed to be loaded in the RT-100 Transport Cask is 15,000 lbs (6,804 kg).

In the initial cut plan, the segments of the Major Components were defined by size, weight, volume, and activity and given unique alphanumeric identifiers. Additionally, all of the planned components were assigned to specific packages defined in the characterization based on the components waste class and weight [27] .

In order to accurately account for activities from both sources (activated and contaminated) a sample spectrum was generated for each of the parent components that summed the activities, defined the major component's weight, volume and specific activity (µCi/gm). Using this input into the Integrated Shipping & Inventory Program (ISIP), sub-components of these major components were easily be generated and subsequently added to packages. This gave great flexibility in planning packages and when having to make changes to the package-loading plan as the project proceeded and the original cut plan changed. Using a "scale-up" method of component activity determination proved to be the best way to define a particular component's activity.

Table 6 - Example of Combination spectrum generated for Major Component, " Upper Grid "

Nuclide Upper Grid Specific Upper Grid Specific Upper Grid Specific Activity {µCi/gm)- Activity {µCi/gm)- Activity {µCi/gm)-Activated Nuclides Contaminated Combined Nuclides

Nuclides Co60 3.17e+l 7.9le-2 3.18e+l Cs137 2.77e-2 l.33e-4 l.33e-4 C14 5.46e-3 3.32e-2 Fe 55 8.12e+O 2.40e-2 8.14e+O Ni63 2.02e+l 3.58e-1 2.06e+l Sr90 5.27e-4 5.27e-4 Nb94 3.20e-4 2.67e-4 5.87e-4 Tc99 2.26e-6 2.26e-6 Pu238 l.28e-4 l.28e-4

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Pu239 l.09e-4 l.09e-4 Pu241 3.93e-3 3.93e-3 Am241 6.46e-4 6.46e-4 Cm242 4.57e-6 4.57e-6 Cm244 8.99e-5 8.99e-5 Mn54 2.37e-4 2.37e-4 Ni59 l.91e-1 l.91e-1 Zn65 4.25e-7 4.25e-7

2.0 DECAY HEAT (THERMAL LOAD)

The total energy emitted by the decay of activated nuclides contaminated nuclides is a sum of the average recoverable energies per disintegrations that gives the total average energy per disintegration that is available for conversion to heat. Average energies emitted by the nuclides is taken from values used in ORIG EN for recoverable decay energy values, q. The total energy is converted to power using the following relationships:

1 Mev = l.6e-13 J lW = 1 J s-1

1 Bq = 1 disintegration per second 1 Ci = 3.7e+10 Bq

Power (W U 1) = q MeV distegration-1 x l.6e-13 J/Mev x 1 W/ J s-1 x 1 dis s-1 x 3.7e+10 Bq U 1

= q x 5.92 mW u 1

Co-60, Fe-55, and Ni-63 comprises greater than 99 percent of the activity, and effectively 100 percent of the heat generated is generated by the Co-60. A maximum loading of 15,000 lbs (6,804 kg) activated metal with specific activity similar to the Upper Grid would generate approximately 3.34 W of energy available for conversion to heat within the RT-100 inner containment.

Table 6 - Decay Heat (Thermal Load) for Major Component, "Upper Grid "

Nuclide Recoverable Power Upper Grid Specific Upper Grid decay energy (mWU1

) Activity (µCi/gm)- Specific Power values Combined Nuclides (mW/kg)-q (MeV/ dis) Combined

Nuclides Co60 2.606 15.43 3.18e+l 4.91E-01 Cs137 0.1798 1.06 1.33e-4 l.42E-10 C14 0.0495 0.29 3.32e-2 9.73E-09 Fe 55 0.0056 0.03 8.14e+O 2.70E-07 Ni63 0.0174 0.10 2.06e+l 2.12E-06 Sr90 0.1958 1.16 5.27e-4 6.llE-10 Nb94 1.7034 10.08 5.87e-4 5.92E-09 Tc99 0.0552 0.33 2.26e-6 7.39E-13 Pu238 5.5886 33.08 l.28e-4 4.23E-09 Pu239 5.2413 31.03 l.09e-4 3.38E-09 Pu241 5.2510 31.09 3.93e-3 l.22E-07 Am241 5.5960 33.13 6.46e-4 2.14E-08 Waste Control Specialists LLC Page 9of13 RT-100 Transport Cask Contents Description

Cm242 6.6240 39.21 4 .57e-6 l.79E-10 Cm244 5.8989 34.92 8.99e-5 3.14E-09 Mn54 0.8400 4.97 2.37e-4 l.18E-09 Ni59 0.0066 0.04 l.91e-1 7.46E-09 Zn65 0.5888 3.49 4.25e-7 l.48E-12

3.0 APPENDIX [13]

Liner Contents Loading Summary for Major Components

Liner ID Segment ID Date loaded in liner Document Package RVI008 UG 2-2 11/22/2005 RS-RVl-008 [13]

UG 2-3 11/22/2005

RVI016 CBl-1 10/06/05 RS-RVl-016 [14] CBl-2 10/06/05 CBl-3 10/06/05 CB2-1 10/12/05 CB2-2 10/12/05 CB2-3 10/12/05 CB3-1 10/15/05 CB3-2 10/15/05

RVI017 CBl-5 10/10/05 RS-RVl-017 [15] CBl-6 10/10/05 CB2-4 10/12/05 CB2-5 10/13/05 CB2-6 10/13/05 CB3-3 10/18/05 CB3-4 10/18/05 CB3-5 10/18/05

RVI018 CBl-7 10/11/05 RS-RVl-018 [16] CBl-8 10/11/05 CBl-9 10/11/05 CB2-7 10/14/05 CB2-8 10/14/05 CB2-9 10/14/05

RVI019 CB3-7 10/19/05 RS-RVl-019 [17] CB3-8 10/19/05 CB3-9 10/19/05 CBl-4 10/26/05 CB3-6 10/28/05 TS 1-1 10/26/05 TSl-2 10/27/05 TSl-3 10/27/05

RVI021 LIGS2-2 3/29/06 RS-RVl-021 [18] LIGS5-2 4/6/06 LIGS3-2 4/25/06

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LIGS4-2 4/21/06 RVI022 LIGS6-1 3/31/06 RS-RVl-022 [19]

LIGS2-1 4/03/06 LIGS6-2 3/31/06 LIGS6-3 4/4/06 331NST tubes 4/5/06

RVI026 LIGS4-3 4/27/06 RS-RVl-026 [20] LIGS5-1 4/7 /06 LIGS5-3 4/7 /06

RVI 027 LIGS7-2 4/12/06 RS-RVl-027 [21] LIGSl-2 4/18/06

RVI029 LIGS2-3 4/19/06 RS-RVl-029 [22] LIGS7-3 4/13/06 LIGS7-1 4/13/06 LIGSl-3 4/18/06 LIGSl-1 4/18/06

RVI033 TSl-5 10/31/05 RS-RVl-030 [23] TSl-4 10/31/05 TSl-6 10/31/05 TSl-7 11/01/05 TSl-8 11/01/05 TSl-9 11/01/05 TS2-1 11/03/05 TS2-2 11/03/05

RVI034 TS2-3 11/03/05 RS-RVl-034 [24] TS2-4 11/03/05 TS2-5 11/04/05 TS2-6 11/04/05 TS2-7 11/04/05 TS2-8 11/07/05 TS2-9 11/07/05 TS3-1 11/07/05

RVI035 TS3-2 11/08/05 RS-RVl-035 [25] TS3-3 11/08/05 TS3-4 11/08/05 TS3-5 11/08/05 TS3-6 11/08/05 TS3-7 11/09/05 TS3-8 11/09/05 TS3-9 11/09/05

RVI037 LIGS3-3 4/25/06 RS-RVl-037 [26] LIGS3-1 4/26/05 LIGS4-1

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REFERENCES [l] Rancho Seco Post Shutdown Decommissioning Activities Report, Amendment 4, dated July 31 2003. [2] Karl Johnson, "Rancho Seco Reactor Vessel Internals Segmetation Project," , Albuquerque, NM, 2006. [3] D.W. James, "Radiological Characterization of Rancho Seco Upper Internals," DURATEK ENGINEERING

REPORT, ER-05-005, Revision 0 (DWJ-RS-002,Revision 0) 2005. [4] D. W. James, "Radiological Characterization of Rancho Seco Core Region Internals," DURATEK

ENGINEERING REPORT, ER-05-006, Revision 0 (DWJ-RS-003, Revison 0) 2005. [5] D. W. James, "Radiological Characterization of Rancho Seco Lower Internal Assembly," DURATEK

ENGINEERING REPORT, ER-05-007, Revision 0 (DWJ-RS-004, Revison 0) 2005. [6] A. Luksic, "PNL-9606 Vol. 1, Spent Fuel Asembly Hardware: Characterization and 10CFR61 Classification for

Waste Disposal," Pacific Northwest Laboratory, 1989. [7] A. Luksic, "PNL-9606 Vol. 2, Spent Fuel Asembly Hardware: Characterization and 10CFR61 Classification for

Waste Disposal," 1989. [8] A. Luksic, "PNL-9606 Vol. 3, Spent Fuel Asembly Hardware: Characterization and 10CFR61 Classification for

Waste Disposal," 1989. [9] D.W. James, "Rancho Seco Flux Profiles," DURATEK ENGINEERING REPORT, ER-05-004 (DWJ-RS-001,

Revison 0) 2005. [ 1 O] SNAP, "Simplified Neutron Activation Program", D. W James & Associates, 2001, St. Paul, Minnesota].. [11] ORJGEN 2.1, Isotope Generation and Depletion code, Matrix Exponential Method, RSCICC Computer Code

Collection, CC-3 71. [12] DURATEK, "Specification for Fabrication of Rancho Seco Fuel Pool Liner and Overpack Assemblies,"

ENGINEERING SPECIFICATIN, ES-G-002, Rev. 0 2004. [13] TRANSNUCLEAR, "Reactor Vessel Internal Segmentation, Loaded Container Documentation Package, RS-

RVI-008, Upper Grid and Chips," 2006. [ 14] TRANSNUCLEAR, "Reactor Vessel Internal Segmentation, Loaded Container Documentation Package, RS-

RVI-016, Core Barerel," 2006. [15] TRANSNUCLEAR, "Reactor Vessel Internal Segmentation, Loaded Container Documentation Package,RS-

RVI-017, Core Barrel Liner," 2006. [ 16] TRANSNUCLEAR, "Reactor Vessel Internal Segmentation, Loaded Container Documentation Package, RS-

RVI-018, Upper Grid," 2006. [17] TRANSNUCLEAR, "Reactor Vessel Internal Segmentation, Loaded Container Documentation Package, RS-

RVI-019, Lower Internals," 2006. [ 18] TRANSNUCLEAR, "Reactor Vessel Internal Segmentation, Loaded Container Documentation Package,RS-

RVI-021, Lower Internals upper Segment," 2006. [19] TRANSNUCLEAR, "Reactor Vessel Internal Segmentation, Loaded Container Documentation Package, RS-

RVI-022, Lower Internals Upper," 2006. [20] TRANSNUCLEAR, "Reactor Vessel Internal Segmentation, Loaded Container Documentation Package, RS-

RVI-026, Lower Internals," 2006. [21] TRANSNUCLEAR, "Reactor Vessel Internal Segmentation, Loaded Container Documentation Package, RS-

RVI-027, Lower Internals Upper Segment," 2006. [22] TRANSNUCLEAR, "Reactor Vessel Internal Segmentation, Loaded Container Documentation Package, RS-

RVI-029, Lower Internals Upper Segment," 2006. [23] TRANSNUCLEAR, "Reactor Vessel Internal Segmentation, Loaded Container Documentation Package, RS-

RVI-033, Thermal Shield," 2006. [24] TRANSNUCLEAR, "Reactor Vessel Internal Segmentation, Loaded Container Documentation Package, RS-

RVI-034, Thermal Shield," 2006. [25] TRANSNUCLEAR, "Reactor Vessel Internal Segmentation, Loaded Container Documentation Package, RS-

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[26] TRANSNUCLEAR, "Reactor Vessel Internal Segmentation, Loaded Container Documentation Package, RS-RVI-037, Lower Internals," 2006.

[27] TRANSNUCLEAR, Segmentation Strategy and Sequence.

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