ML20236L350

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Rev 2 to Saxton Reactor Pressure Vessel & Internals Final Characterization. W/Two Oversize Drawings
ML20236L350
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Site: Saxton File:GPU Nuclear icon.png
Issue date: 04/30/1998
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ML20236L328 List:
References
1920-98-20343, 9801-7025, 9801-7025-R02, 9801-7025-R2, NUDOCS 9807100322
Download: ML20236L350 (66)


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{{#Wiki_filter:_ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ - - - _ _ - - l 1920-98 20343 ATTACINENT 5 1 TilRU CB l r f SAXTON REACTOR PRESSURE VESSEL AND INTERNALS FINAL CHARACTERIZATION l l l l Report WMG 9801-7025 1 April 1998 Prepared for: l Raytheon Engineers and Constructors l Prepared by: WMG, Inc. 16 Bank Street Peekskill, NY 10566 9807100322 900707 PDR ADOCK 0500 6

i l l l l l l i FOREWORD I This report details component characterization work performed by WMG, Inc. to l support the decommissioning of the Saxton Nuclear Power Plant by Raytheon. This work was performed by WMG, Inc. under Raytheon subcontract 77581803-49-1. i L l. l l i i Saxton Reactor Pressure Vessel and internals WMG Report 9801-7025 Final Characterization Rev. 2 4/98 i ]

TABLE OF CONTENTS

1.0 INTRODUCTION

1-1 2.0 EXECUTIVE

SUMMARY

.2-1

3.0 DESCRIPTION

OF CLASS 7 RADIOACTIVE MATERIAL. .3-1 3.1 Physical Description. .3-1 3.1.1 Core Baffle. .3-1 31.2 Thermal Shield. .3-2 3.1.3 Reactor Vessel Cladding. .3-2 3.1.4 Reactor Vessel.. .3-2 3.1.5 Reactor VesselInsulation. .3-2 3.1.6 Lower Core Plate. .3-2 3.1.7 Lower Guide Blocks. .3-2 3.1.8 Shroud. Tubes. .3-3 3.1.9 Lower Support Tie Rods. .3-3 3.1.10 Balance of the Lower Support Assembly. .3-3 3.1.11 Lower Core Barrel. .3-3 31.12 Upper Core Plate. .3-3 31.13 Upper Core Barrel.. .3-4 31.14 Reactor Vessel Insulation Can. .3-4 3.2 Radiological Characterization Results.. .3-4 3.2.1 Neutron Activation Products. .3-4 3.2.2 Internal Surface Contamination. .3-7 3.2.3 External Surface Contamination. .3-7 4.0 DOT CLASSIFICATION RESULTS.. .4-1 4.1 LSA Ill Determination.. .4-1 4.2 Estimated Dose Rates. .4-1 4.3 LSA Ill Test Requirements. .4-4 5.0 NRC CLASSIFICATION RESULTS., .5-1

6.0 REFERENCES

.6-1 APPENDICES APPENDIX A - REACTOR VESSEL INTERNAL CONTAMINATION DATA APPENDIX B - REACTOR VESSEL WETTED SURFACE AREA CALCULATION APPENDIX C - REACTOR VESSEL EXTERNAL CONTAMINATION DATA Saxton Reactor Pressure Vessel and internals WMG Report 9801-7025 l Final Characterization Rev. 2 4/98 ii - t

LIST OF FIGURES Figure Title Page 2-1 Reactor Vessel Assembly. .2-2 2-2 Reactor Vessel Transportation System.. .2-3 LIST OF TABLES Table Title Page 3-1 Radiological Characterization Results Summary. .3-5 3-2 Activation Analysis Results Summary. .3-6 4-1 DOT Classification Summary. .4-2 4-2 LSA Ill Determination Based on Lower Guide Block Data. .4-3 5-1 10 CFR 61 Classification Averaged Over All Components.. .5-3 5-2 BTP Concentration Averaging Summary. .5-4 l Saxton Reactor Pressure Vessel and Internals WMG Report 9801-7025 ) Final Characterization Rev. 2 4/98 iii L


J

1.0 INTRODUCTION

WMG was subcontracted by Raytheon Engineers and Constructors to provide waste characterization and classification services and regulatory liaison for the transportation and disposal of the GPU Saxton Nuclear Experimental Corporation (SNEC) facility reactor vessel with its intact internals. The waste charactenzation results quantify activities from both activation products and surface contaminants for the reactor vessel and internals. These results are based on extensive work performed by GPU NUCLEAR as documented in their Site Characterization Report (Ref.1), a previous study performed by TLG (Ref. 2) and recent sample analyses performed by GPU NUCLEAR. The final charactenzation results were used to classify the vessel and internals relative to the requirements of 10 CFR 61,49 CFR 173, Barnwell disposal site criteria and the NRC Branch Technical Position (BTP) on Concentration Averaging and Encapsulation (Ref. 3). i l l Saxton Reactor Pressure Vessel and internals WMG Report 9801-7025 Final Characterization Rev. 2 4/98 1-1

i I i 2.0 EXECUTIVE

SUMMARY

i The SNEC reactor vessel and internals consist of the activated internal components i and the activated reactor vessel that surrounds the intact immovable internals 1 l components and are shown in Figure 2-1. The vessel and intact internals comprise the l Class 7 (radioactive) material proposed for transportation to the Barnwell disposal site. These materials will be packaged within an outer container, which together with its contents and grouting, form a transportation system providing equivalent safety to an Industrial Package Type 2 (IP-2). A sketch showing the configuration of the transportation system is shown in Figure 2-2. i Final radiological characterization of the SNEC reactor vessel transportation system yielded a total activity of 1,285 curies as of 8/1/98 with 1,282 curies attnbuted to neutron activation and the remaining 3 curies from surface contamination. The SNEC reactor vessel transportation system meets the DOT specific activity requirements for Low Specific Activity (LSA) lli material. Each component within the vessel and the vessel itself have specific activities below the LSA-lli limit of 2 x 10-8 A2/g. On average, the vessel and the internals, excluding the grout, has a specific 4 activity of 1.7 x 10 A2/g ( < 0.1 % of the LSA Ill limit) and the worst case component 4 has a specific activity of 7.1 x 10 A2/g (< 4% of the limit). i The radioactive material meets the requirements of 10 CFR 61.55 for disposal as Class C waste. While some of the internal components have concentrations which exceed the Class C limits, they can be blended within the guidance of the NRC Branch Technical Position on Concentration Averaging. The reactor vessel and intact internal components also meet the Barnwell site criteria for disposal as low level waste. l-Saxton Reactor Pressure Vessel and Internals WMG Report 9801-7025 Final Characterization Rev. 2 4/98 l 2-1 l 2

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3.0 DESCRIPTION

OF CLASS 7 RADIOACTIVE MATERIAL The SNEC reactor vessel is a reactor system, excluding the nuclear fuel, which will be transported and disposed as a single, intact component. For analytical purposes, it has been broken down into a collection of components which were quantified separately based on historic decommissioning projects where internal components were segmented and packaged in shielded shipping casks for transportation. The SNEC reactor vessel and internals are irradiated metals which have become i activated from neutron exposure during plant operations. These activated metal l components are also surface contaminated from contact with primary coolant. In order to characterize the reactor vessel and internal components a detailed physical description was required which included component dimensions, weights, metal volumes (displacement volumes), surface areas and proximity to the reactor core. Estimated activities by radionuclides from both neutron activation and surface contamination decay corrected to August 1,1998 are provided herein. The surface l contaminants are broken down into contamination levels on the interior surfaces in direct contact with primary coolant and the exterior surfaces of the vessel which are quantified separately. 3.1 Physical Description The SNEC reactor is a 23.5 MWt (nominal) expenmental pressurized water test reactor located in Saxton, PA. The reactor vessel is 18 feet high, has a diameter of about 5 feet and is shown in the sketch in Figure 2-1 above. The SNEC reactor vessel and internals amount to nearly 1,285 curies in 112,000 pounds of activated metal. Descriptions of the components of interest taken from the TLG study (Ref. 2) are presented below with the weights rounded for simplicity. Understandably, the components in regular contact with primary coolant were fabricated from corrosion resistant materials, primarily stainless steel. The reactor vessel is clad with stainless steel while the remainder of the vessel was fabricated of carbon steel. 3.1.1 Core Baffle The core baffle is a '/s inch thick box-like structure immediately surrounding the reactor core. The baffle assembly weighs approximately 2 1,210 pounds and has a surface area of 1.21E+05 cm I Saxton Reactor Pressure Vessel and internals WMG Report 9801-7025 Final Characterization Rev. 2 4/98 3-1 l

i 3.1.2 - Thermal Shield The thermal shield is attached to the inner wall of the reactor vessel. It is l a 94 inch tall hollow right circular cylinder with an inner radius of 24.5 l inches and a thickness of 3 inches. The thermal shield has an estimated I weight of 10,500 pounds and an estimated surface area of 1.68 E+05 cm* 3.1.3 Reactor Vessel Claddina The inner surfaces of the reactor vessel are lined with stainless steel to a' nominal thickness of '/s inch. The cladding has an inner radius of 28 /. 7 inches, an estimated weight of 2.040 pounds, and a surface area of 2.47 2 E+05 cm 3.1.4 Reactor Vessel l l The reactor vessel includes a removable closure head and a 143 inch tall i-laminated plate cylinder flanged at one end and fused to a hemi-spherical l lower head at the other. All components are fabricated from carbon steel to a minimum thickness of 4.5 inches and nominal inner radius of 29 inches. The unit weighs a combined 87,300 pounds. 3.1.5 Reactor Vessel Insulation i The outer wall of the reactor vessel is insulated with fiberglass which is not activated to any appreciable extent. This insulation is contained within the Reactor Vessel Insulation Can. 3.1.6 Lower Core Plate The lower core plate is 1.5 inches thick,43.5 inches in diameter, and has l penetrations for the control rods and coolant flow. The lower core plate l has a weight of 500 pounds and an estimated surface area of 1.40 E+04 2 cm. 3.1.7 Lower Guido Blocks There are 32 guide blocks attached to the lower core plate which were used to align and locate fuel assemblies. The guide blocks are % inch thick square plates,5 '/,. inches on a side with a 3'/,. inch diameter hole in the center. Two adjacent sides of the square plate have a 3 inch tall '/. inch thick sheet attached to help align fuel assemblies. The guide blocks weigh a total of 150 pounds and have a total estimated surface area of 2.3 E+04 cm'. The lower guide blocks have radionuclides concentrations in excess of Part 61 Class C limits if considered separately. Saxton Reactor Pressure Vessel and Internals WMG Report 9801-7025 Final Characterization Rev. 2 4/98 3-2 i

e 1 3.1.8 Shroud Tubes l These tubes support the lower core plate and align fuel followers. The upper portion of the tube is a cylinder of 3.5 inch radius that is tapered to approximate the envelope of a cruciform 11 inches long. The lower l portion of the tube is a 38 inch long right circular cyhnder of 6.5 inches in diameter. The portion of the tubes down to 3 feet below the core was included in the activation calculations, weighs 340 pounds and has an estimated surface area of 7.61 E+04 cm 2 3.1.9 Lower Support Tie Rods The periphery of the lower core plate is connected to a spider assembly by eight, 43.5 inch long 1 inch diameter lower support tie rods. The portion of the tie rods down to 3 feet below the core was included in the activation calculations, weighs 54 pounds and has an estimated surface area of 5.84 E+03 cm 2 3.1.10 Balance of the Lower Support Assembly The portions of the lower support tie rods, shroud tubes and other items more than 3 ft below the active core region weigh a combined 700 pounds. 3.1.11 Lower Core Barrel The lower core barrel separates the core intet and outlet plenums and is attached to the upper end of the core baffle. For the purposes of this report, the lower core barrel refers to the entire lower core barrel (Lower Core Barrel and the Balance of Lower Core Barrel Assembly, Ref.1 Table 4-24). This component has a (combined) estimated weight of 2.900 pounds and a surface area of 4.82 E+04 cm2 3.1.12 Upper Core Plate The upper core plate is another core alignment structure that is 1 inch thick and 40.8 inches in diameter. It has an estimated weight of 260 pounds and a surface area of 5.70 E+03 cm2 l L Saxton Reactor Pressure Vessel and Internals WMG Report 9801-7025 Final Characterization Rev. 2 4/98 3-3 I w

r y 3.1.13 Upper Core Barrel The upper core barrel supports the upper core plate and is a 63 inch tall cylindrical structure with a diameter of 39.5 inches and a thickness of % inch. For the purposes of this report, the upper core barrel refers to the entire upper core barrel (Upper Core Barrel and the Balance of Upper Core Barrel Assembly, Ref.1 Table 4-24). This component has a (combined) estimated weight of 1,300 pounds and a surface area of 7.71 2 E+04 cm 3.1.14 Reactor Vessel Insulation Can This % inch thick,100 inch tall stainless steel nght circular cylinder of 77 '/e anch diameter was used to prevent wetting of the insulation. It weighs 5,000 pounds, including a massive support flange, and has a surface area of 1.57 E+05 cm' 3.2 Radiological Characterization Results The radiological characterization of the SNEC reactor vessel and internals is based on an activation analysis study prepared by TLG (Ref. 2) and extensive representative sampling performed by GPUN as documer'ed in the Saxton Site Characterization Report (Ref.1). Additional swipe samples obtained by GPU NUCLEAR as recently as April 1998 were also incorporated into the final charactenzation results. The results from neutron activation are combined with the internal surface contaminants and external contamination estimates provided herein to determine the total activity by radionuclides which is summarized in Table 3-1. The characterization results from these three sources of activity are discussed separately below. 3.2.1 Neutron Activation Products The activation product radionuclides inventory was based on a study performed by TLG (Ref. 2). The methodology used by TLG utilized one dimensional neutron transport calculations with the ANISN (Ref. 4) program and activation analysis with the ORIGEN2 (Ref. 5) program This ANISN/ORIGE N2 methodology has been widely used in the industry for both decommissioning studies and waste characterization by both WMG and TLG. Activation analysis results for the irradiated components are summarized in Table 3-2. Note that some components such as the lower core barrel were broken down into segments to facilitate use of this methodology. As shown in the table, the total activity of the components is 1,282 curies as of 8/1/98. These activation analysis results have been benchmarked extensively relative to empirical sample data as documented in the Saxton Site Characterization Report (Ref.1) and are therefore considered a reasonable estimate for final waste characterization.. Saxton Reactor Pressure Vessel and internals WMG Report 9801-7025 Final Characterization Rev. 2 4/98 3-4

i Table 3-1 Radiological Characterization Results Summary Intemal Extemal Total l Activation Contamination Contamination Activity Nuclide Curies Curies Curies Curies Fe-55 2.17E+01 2.40E-02 3.22E-06 2.17E+01 Nb-94 1.27E-02 4.22E-04 1.31E-02 Co-60 4.52E+02 4.39E-01 9.25E-04 4.52E+02 Ni-59 6.55E+00 2.06E-02 6.57E+00 Ni-63 7.99E+02 1.64E+00 1.68E-05 8.01E+02 Eu-154 2.59E-03 2.59E-03 Eu-155 5.88E-04 5.88E-04 H-3 1.49E+00 LLD 2.58E-07 1.49E+00 C-14 1.16E+00 LLD 1.42E-09 1.16E+00 Tc-99 2.23E-03 LLD 7.99E-08 2.23E-03 1-129 LLD LLD LLD Cs-137 8.05E-02 1.16E-03 8.16E-02 Sr-90 7.76E-02 2.96E-06 7.76E-02 Pu-238 3.03E-02 3.99E-07 3.03E-02 Pu-239/40 6.91E-02 1.06E-06 6.91E-02 Pu-241 8.20E-01 3.61E-06 8.20E-01 Pu-242 3.09E-09 3.09E 09 Am-241 1.44E-01 1.68E-06 1.44E-01 Cm-242 5.85E-06 1.29E-08 5.86E-06 Cm-243/244 1.27E-03 3.87E-08 1.27E-03 U-234 2.29E-05 2.29E-05 Totals 1.282E+03 3.350E+00 2.119E-03 1.285E+03 Saxton Reactor Pressure Vessel and intemals WMG Report 9801-7025 Final Characterization Rev. 2 4/98 3-5

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i 3.2.2 Internal Surface Contamination The surface contamination on the internal surfaces exposed to primary coolant was determined by direct sample analysis performed by GPUN as presented in the Saxton Site Characterization Report and recent samples obtained in Apnl 1998. Representative samples of the surface contaminants were obtained for areas of high coolant flow and for areas where flow was stagnant and contaminants are expected to collect in crud traps. As expected, the contamination levels in the high flow areas were considerably lower than those in the areas of low flow. During site characterization, GPU NUCLEAR obtained representative samples at two locations within the vessel; at the safety injection piping and in the bottom head. These samples were sent off-site to B&W for independent laboratory analysis and the results are enclosed as Appendix A along with a copy of Section 4.3.6 from the Site Characterization Reoort. Comparison of the isotopic distr bution data between the two samples is in excellent agreement and the average isotopic distribution of the two was used. The sample of the safety injection piping was taken by acid etching a known surface area of piping and was assumed representative of the quantitative contamination levels over the wetted surface of the areas of low flow within the reactor vessel. This yields an estimated contamination level of 9.2 uCi/cm (as of 8/1/98). 2 The areas within the reactor vessel considered to be low flow areas consist of the bottom of the vessel where insoluble particulate can accumulate over time and the region between the vessel wall and thermal shield where flow is relatively stagnant. Therefore, entire interior surface of the vessel and the external surface of the thermal shield were conservatively assumed to be low flow areas. The surface area of the exterior of the thermal shield is estimated at 8.42E+04 cm and the 2 intarnal surface of the reactor vessel is estimated at 2.47E+05 cm as 2 documented in GPUN calculation 5340 95-011 enclosed in Appendix B. This surface area in conjunction with the 9.2 uCi/cm y elds a total activity 2 of about 3 curies from internal surface contamination in the areas of low flow. Use of the quantitative data from the safety injection piping (Ref. 9) was considered to be overly conservative for the remainder of the internal components since piping represents a crud trap where the coolant flow is stagnant. Therefore, additional swipe samples were obtained by GPU NUCLEAR in April 1998 in areas considered to be representative of the contamination levels on the remaining internals. The samples were analyzed in house and the results are enclosed in Appendix A. The in house results were used in conjunction with Co-60 based scaling factors derived from the average nuclide distribution described above. Saxton Reactor Pressure Vessel and Internals WMG Report 9801-7025 Final Characterization Rev. 2 4/98 3-7

Three swipe samples were obtained at different locations on the internals with sample areas varying from 100 to 450 cm' The Co-60 sample analysis results per unit area (cm') were in very good agreement between the three swipes and varied from 4 42E-04 uCi/ cm to 4.98E-04 uCil em' 2 with an average of 4.7 E-04 uCi/ cm2 This average Co-60 contamination level was used to estimate total contamination levels assuming 1% of the total activity was removed (10% swipe removal efficiency per NRC (E 85-46 and a fixed to removable ratio of 10). The Co-60 value of 4.7 uCi/cm' was used in conjunction with Co-60 based scaling factors derived from the average nuclide distribution desenbed above to estimate surface contamination levels in the areas of high coolant flow. Based on an estimated surface area of the remaining components of 7.99E+0S cm' [1.13E+06 cm' (total) - 8 42E+04 cm (thermal shield exterior)- 2. 2 cm* (vessel interior)). This yielded a total surface activity in these areas of about 0.3 curies. 3.2.3 External Surface Contamination Surface contamination levels and corresponding activitios by radionuclides were estimated based on historic swipe surveys taken from June 1995 to October 1997. The swipe survey results are enclosed in Appendix C. The worst case contamination survey taken on the external surface of the 1 vessel showed 120,000 DPM/100 cm' This worst case contamination level was used in conjunction with representative isotopic sample data from Area 6 in the " Site Characterization Report" (Rei.1 Table 4-6), an estimated surface ~ area of 392,000 cm and an assumed swipe removal 2 efficiency of 10% (NRC IE 85-46) to determine the activity by radionuclides. This data yielded a total activity of 2.1 millicuries of removable surface contamination. The total contamination was estimated based on an assumed fixed to removable contamination ratio of 10. Thi results in a total exterior contamination activity of level of 21.2 millicuries. The supporting data for this estimate is enclosed in Appendix C Saxton Reactor Pressure Vessel and Internals WMG Report 9801-7025 L Final Characterization Rev. 2 4/98 3-8 l l

4 I 4.0 DOT CLASSIFICATION RESULTS The SNEC vessel and internals characterization results were classified according to the following DOT criterian Type A quantity determination 49 CFR 173.433 Reportable quantity limits of 49 CFR 172.101 LSA Ill concentration limits of 49 CFR 173.40 3 Fissile Excepted as per 49 CFR 173.453 Unshielded dose rate limits of 49 CFR 173.427 LSA lit leach testing requirements of 49 CFR 173.468 i The DOT classification results are summanzed in Table 4-1. As shown in Table 4-1, the matenal contains a greater than Type A quantity of radioactive material with a Reportable Quantity of radionuclides and is Fissile Excepted. The basis for the LSA 111 determination, unshielded dose rates and LSA 111 leach test are discussed separately below. 4.1 LSA Ill Determination As discussed in Section 3 above the activities of individual components vary considerably due to their location relative to the active core region during operation. Therefore, the radioactivity can not be considered to be "distnbuted throughout" the material. The material must be considered as a " collection of solid objects" for classification as LSA Ill. The worst case specific activities from the core region were compared to the LSA Ill limit of 2x10 A2/g and these results are shown in Table 4-2. As shown in the table, even the worst case component is well within the concentration limits of 49 CFR 173.403 at < 4% of the limit. 4.2 Estimated Dose Rates 49 CFR 173.427(a) 1 requires that the dose rate at 3 meters from the unshielded LSA material be less than 1 Rem /hr (10 mSv/hr). If the intact vessel and internals are the LSA material to which this limit is applied, the limit is satisfied. The worst case dose rate at 3 meters from the vessel exterior is 660 mrem /hr l (6.6 mSv/hr). However, when the irradiated reactor vessel and internal components are considered as a collection of solid objects, some components, if considered separately will exceed 1 Rem /hr at 3 meters. The worst case component from an exposure standpoint is the lower core plate and guide blocks, due to the high specific activity of this component and its physical configuration. The estimated unshielded dose rate at 3 meters from the lower core plate is 11 R/hr. However, this component is an integral part of the reactor vessel and will also be grouted in place. Therefore, there is no credible scenario under normal transport conditions where direct exposure to the lower core plate is possible. Saxton Reactor Pressure Vessel and Internals WMG Report 9801-7025 Final Characterization Rev. 2 4/98 4-1

l Table 41 j DOT Classification Summary ) I Total A2 A2 (1) RQ RQ(2) Fissile Nuclide Curies Curies Fraction Curies Fraction Grams (3) i Fe 55 2.17E+01 1.08E+03 2.01E-02 1.00E+02 2.17E-01 Nb-94 1.31 E-02 1.62E+01 8.07E-04 1.00E+01 1.31E-03 I Co-60 4.52E+02 1.08E+01 4.19E+01 1.00E+01 4.52E+01 Ni-59 6.57E+00 1.08E+03 6.08E-03 1.00E+02 6.57E-02 Ni-63 8.01E+02 8.11E+02 9.88E-01 1.00E+02 8.01E+00 Eu-154 2.59E-03 1.35E+01 1.92E-04 1.00E+01 2.59E-04 Eu-155 5.88E-04 5.41E+01 1.09E-05 1.00E+01 5.88E-05 H-3 1.49E+00 1.08E+03 1.38E-03 1.00E+02 1.49E-02 C-14 1.16E+00 5.41 E+01 2.13E-02 1.00E+01 1.16E-01 Tc-99 2.23E-03 2.43E+01 9.18E-05 1.00E+01 2.23E-04 l-129 LLD ' Cs-137 8.16E-02 1.35E+01 6.05E 03 1.00E+00 8.16E-02 Sr-90 7.76E-02 2.70E+00 2.87E 02 1.00E-01 7.76E-01 Pu-238 3.03E-02 5.41E-03 5.60E+00 1.00E-02 3.03E+00 1.78E-03 Pu-239/40 6.91E-02 5.41E-03 1.28E+01 1.00E-02 6.91E+00 1.11E+00 Pu 241 8.20E-01 2.70E-01 3.04E+00 1.00E+00 8.20E-01 8.20E-03 Pu-242 3.09E-09 5.41 E-03 5.72E-07 1.00E-02 3.09E-07 7.93E-07 Am-241 1.44E-01 5.41 E-03 2.66E+01 1.00E-02 1.44E+01 Cm-242 5.86E-06 2.70E-01 2.17E-05 1.00E+00 5.86E-06 Cm-243/24 1.27E-03 8.11 E-03 1.56E-01 1.00E-02 1.27E-01 U-234 2.29E-05 2.70E-02 8.49E-04 1.00E-01 2.29E-04 1.29E+03 9.11 E+01 7.98E+01 1.12E+00 Notes: (1) The A2 fraction is greater than 1. Therefore it is greater than a Type A quantity. (2) The RC fraction is greater than 1 rnaking it a Reportable Quantity of radionuclides (3) The grarns of fissile material is less than 15 making it fissile excepted. Saxton Reactor Pressure Vessel and internals WMG Report 9801-7025 Final Characterization - Rev. 2 4/98 4-2

i l Table 4-2 l l L LSA lli Determination l Based on Lower Guide Block Data i i Total A2 A2 A2/ gram LSA Ill Nuclide Curies Curies Fraction Fraction Fe-55 2.89E+00 1.08E+03 2.68E-03 3.88E-08 1.94E-05 l l l Nb-94 1.22E-03 1.62E+01 7.56E 05 1.09E-09 5.47E-07 l Co-60 5.06E+01 1.08E+01 4.68E+00 6.77E-05 3.39E-02 l Ni-59 8.90E-01 1.08E+03 8.24E-04 1.19E-08 5.96E-06 Ni-63 1.09E+02 8.11E+02 1.34E-01 1.94E-06 9.69E-04 Eu-154 5.50E-06 1.35E+01 4.08E-07 5.90E-12 2.95E-09 Eu-155 1.25E-06 5.41E+01 2.31E-08 3.34E-13 1.67E-10 l H-3 1.68E-01 1.08E+03 1.56E-04 2.25E-09 1.13E-06 1 C-14 1.56E-01 5.41E+01 2.88E-03 4.16E-08 2.08E-05 Tc-99 1.45E-04 2.43E+01 5.95E.06 8.61E-11 4.31E-08 l-129 0.00E+00 Cs-137 1.70E-04 1.35E+01 1.26E-05 1.82E-10 9.08E-08 Sr-90 1.64E-04 2.70E+00 6.06E-05 8.77E-10 4.39E-07 Pu-238 6.61E-05 5.41E-03 1.22E-02 1.77E-07 8.83E-05 Pu-239/40 1.50E-04 5.41 E-03 2.77E-02 4.01E-07 2.00E-04 Pu-241 1.78E-03 2.70E-01 6.58E-03 9.52E-08 4.76E-05 Pu-242 3.18E-04 5.41E-03 5.88E-02 8.50E-07 4.25E-04 Am-241 3.04E-08 5.41E-03 5.62E-06 8.13E-11 4.07E-08 Cm-242 4.31 E-06 2.70E-01 1.59E-05 2.31E-10 1.15E-07 Cm-243/24 4.85E-07 8.11E-03 5.98E-05 8.65E-10 4.32E-07 l U-234 2.29E-05 2.70E-02 8.49E 04 1.23E-08 6.14E-06 Totals 1.63E+02 4.93E+00 7.13E-05 3.57E-02 i l Weight Grams = 6.91 E+04 l l 1 Saxton Reactor Pressure Vessel and Intemals WMG Report 9801-7025 l Final Characternaton Rev.2 4/98 4-3

l 4.3 LSA W Test Requirements Sinc.e the reactor vessel and internals are comprised of activated metal, only the surface contaminants could potentially be released under the leach testing requirements for LSA Ill material in 49 CFR 173.468. The internals components and the reactor vessel interior have contaminants on their surfaces which are conservatively estimated to consist of 3.3 curies as discussed in Section 3.2.2 above. Uniform distnbution of this activity results in contamination levels of 3 2 uCi/cm These contaminants will be grouted onto their surfaces and enclosed within the reactor vessel. This contamination is further contained by the grout between the vessel and the outer container and by the steel outer container itself. This double containment system ensures that the internal surface contaminants could not be released under normal transport conditions Therefore, it is assumed that only the surface contamination on the exterior surfaces of the vessel and insulation can would be applicable in regards to the leach testing requirements. As shown in Section 3.2.3 above, the exterior surfaces of the reactor vessel contain relatively low levels of surface contaminants. It is conservatively assumed that both the removable and fixed contamination would be available for j release under the leach testing requirements of 49 CFR 173.468. Based on swipe sample data, the level of exterior surface contaminants is estimated at 21.2 mci. Based on the radionuclides distribution of these contaminants, the total l amount corresponds to 7.7x10 of an A2 quantity [@ 49 CFR 173.468 (b) 4), which is well within the limit of 0.1 A2. l l i i Saxton Reactor Pressure Vessel and internals WMG Report 9801-7025 Final Characterization Rev. 2 4/98 4-4

i I l l 5.0 NRC CLASSIFICATION RESULTS i Component specific activities were compared to the waste classification limits of 10 CFR 6155 and the concentration averaging criteria from the 1995 BTP. A summary of the results for the vessel and internals if considered as a single entity is shown in Table 5-1 As shown in the table, the average concentrations are well within the Class C limits of 10 CFR 61.55. In fact, if the concentrations are averaged over the entire metal weight and volume the vessel and internals would meet the Class B limits. However, i there is one component, the lower guide blocks, which exceeds the Class C limits if i considered separately. Although the lower core guide blocks are an integral part of the I lower core plate, they were conservatively considered as separate components for the purpose of concentration averaging since they were characterized separately in the TLG report (Ref. 2). The concentration averaging BTP does not specifically address large components when shipped intact within a reactor vessel. This could be addressed under the alternative provisions in Section 3.9 of the BTP. However, the concentration averaging criteria for activated metals in Section 3.3 was applied to the reactor vessel and internals as if they were separate components in a shipping container. Application of I the BTP cnteria is summarized in Table 5-2 and shows that the concentration averaging cntena are satisfied. The concentration averaging PTP stipulates that components can be averaged for the purpose of waste classification provided that the maximum and minimum concentrations for classification controlling nuclides are within a factor of 10 to the average concentration. The classification controlling nuclides are defined as those with concentrations greater than 1 percent of the 10 CFR 61.55 limit for that nuclide (Class C limits in this case). On this basis, the classification controlling nuclides in this case are C-14, Ni-59, Ni-63, Nb-94 and Am-241 since these nuclides all have either Table 1 or Table 2 average Class C fractions greater than 0.01 as shown in Table 5-2. The BTP also includes more stringent requirements for averaging within a factor of 1.5 for primary gamma emitters (Co-60, Cs-137 and Nb-94) if these nuclides dictate the waste classification. Based on the data in Table 5-2, the primary gamma emitters yield a Table 1 Class C fraction of 0.71 for Nb-94 in the lower guide blocks. Whereas, the i Table 2 Class C fraction of for Ni-63 in the lower guide blocks 1.80. Therefore, the gamma emitters do not dictate the waste classification and the factor of 1.5 is not applicable. The maximum, minimum and volume averaged concentrations for each controlling nuclide are tabulated in Table 5-2 for the components of interest. As shown in the table, the ratio of maximum to the average concentrations are all within a factor of 10 with the worst case at a factor of 2.5. The ratio of the average to minimum l concentrations are also within a factor of 10 with the worst case at a factor of 1.9. Thus the concentration averaging criteria from the 1995 BTP are satisfied. Saxton Reactor Pressure Vessel and internals WMG Report 9801-7025 Final Characterization Rev. 2 4/98 5-1

i i I 1. l l l l By averaging the component that exceed Class C limits with the lower core, the Class C limits are satisfied within the concentration averaging guidance of the BTP. Thereforec the material is Class C waste and is acceptable for disposal as low level waste. It should also be noted that the materials satisfy the Barnwell Rule of 10 since the maximum and minimum Part 61 fractions are within a factor of 10. Therefore, the vessel and intact internal components meet the current Barnwell site enteria for concentration averaging and are acceptable for disposal as low level waste. l l l 1 i Saxton Reactor Pressure Vessel and internals WMG Report 9801-7025 l Final Characterization Rev. 2 4/98 i 5-2

f Table 5-1 10 CFR 61 Classification Averaged Over All Components i i Pan 61 Pan 61 Pan 61 Pan 61 Table 1 Table 1 l Table 2 Table 2 Activity Curies / nCi/ , Class C Class C i Class C Class C Nuclide Curies m3 gm Limit Fraction ' Limit Fraction Fe-55 2.17E+01 3.51 E+00 Nb-94 ' 1.31E-02 2.11 E-03 0.2 0.011 l Co-60 l 4.52E+02 7.30E+01 Ni-59 , 6.57E+00 1.06E+00 i 220 0.005 i Ni-63 . 8.01E+02 1.29E+02 i j 7000 0.018 1 Eu-154 2.59E-03 4.18E-04 j Eu-155 5.88E-04 9.50E-05 l l H-3 1.49E+00 2.'41 E-01 l C-14 1.16E+00 1.87E-01 80 0.002 Tc-99 2.23E-03 3.60E-04 { 3 0.000 1-129 LLD LLD l Cs-137 ' 8.16E-02 1.32E-02 4600 0.000 . Sr-90

7.76E-02 1.25E-02 3

7000 0.000 Pu-238 ! 3.03E-02 6.18E-01 100 0.006 Pu-239/40 6.91 E-02 1.41E+00 i 100 0.014 Pu-241 8.20E-01 1.67E+01 ; 3500 0.005 Pu-242 3.09E-09 6.31E-08 l 100 0.000 Am-241

1.44E-01 2.94E+00 100 0.029 Cm-242 5.86E-06 1.20E-04 i 20000 0.000 t

Cm-243/244' 1.27E-03 2.58E-02 l 100 0.000 U-234 2.29E-05 O.018 Totals 1.29E+03 2.07E+02 2.17E+01 0.073 L Total Metal Volume (m3)= 6.19E+00 l Total Weight (gm) = 4.90E+07 l l l l 1 Saxton Reactor Pressure Vessel and Internals WMG Report 9801-7025 Final Characterization Rev. 2 4/98 5-3

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6.0 REFERENCES

1. Saxton Nuclear Experimental Corporation (SNEC) Site Characterization Repon, GPU Nuclear.1996. 2. The Saxton Facility Reactor Vessel, Internals, Ex-Vessel Lead, Structural Steel and Reactor Compartment Concrete Shield Wall Radlonuclide Inventory, G01-1192-003, Revision 0, Levin, A. H., et al, TLG Services Inc., Bridgewater, Connecticut, December 1995. 3. Final Branch Technical Position on Concentration Averaging and Encapsulation, U.S. Nuclear Regulatory Commission, Knapp, M.R. et al., January 1995. 4. A Users Manual for ANISN, A One Dimensional Discrete Ordinates Transport. Code, ORNL CCC-254, Engle, W.W., Jr., et al. Oak Ridge National Laboratory, April 1991. 5. Isotope Generation and Depletion Code - Matrix Exponential Method, CCC-371, Oak Ridge National Laboratory, May 1991. 6. MegaShield, A Windows" Based Point-Kernel Shielding Program, WMG, Inc., December 1996. 7. (Drawing) Saxton Reactor Plant Lower Core Guide Block Assy.. 881D181, Westinghouse Electric Corporation, February 1961. 8. (Drawing) Saxton Reactor Plant Lower Core Support Plate, 539F370, Westinghouse Electric Corporation, September 1960. 9. Curie Estimate-Saxton Piping Systems, Calculation 5830-95-023, Rev. l 0, GPU Nuclear, l Saxton Reactor Pressure Vessel and Internals WMG Report 9801-7025 I Final Characterization Rev. 2 4I98 6-1 J

i 1 l APPENDIX A { REACTOR VESSEL I INTERNAL CONTAMINATION DATA I ) l Saxton Reactor Pressure Vessel and Internals WMG Report 9801-7025 Final Characterization Rev. 2 4/98 I

APPENDIX A REACTOR VESSEL INTERNAL CONTAMINATION DATA l l l l l l Saxton Reactor Pressure Vessel and Intemals WMG Report 9801-7025 Final Characterization Rev. 2 4/98 t

santon Nuclear Experimental corporation (sNEC) site charactertration Repo.rt Table 4-24, Reactor Vessel Curie Estimate-Calculated Values Components Weight is lbs. Curies 24 Year After Shotdown Cae BafDe 1,210 642.6 Thermal Shield 10,491 132.7 Reactor Vessel Clad 2,043 4J85 Reactor Vessel Wall s7,313 4.737 RVIa==i=aan Can(1) 4,994 1.073 RY Support Can (1) 5.442 3.034 lawer Core Guide Blocks 152 1s2.9 Lawer Core Plane 500 226.7 Lower Support Shroud Tubes (2) 33s s.323 Lower Support Tie Rods (2) 54 2.113 Balance aflawer Support Assamhly (3) 106 0.0744 _lawer Core Barmi(4) 1,067 Ss.46 Upper Core Plass 264 165 Upper Core Banel(4) 454 19.71 Balance oflawer Core Barrel Assembly (5) 1,s23 0.2593 Balance of Upper Core Barrel Assembly (6) s32 0.1183 ,?j(, Notes- (1) includes weight of component between 3' above and 3* below the aceve core. (2) includes weeght of component down to 3' below tie aseve core. t (3) Weight of lower support shroud tees, lower support tie rods and associated components snore that 3' below tie active core. l (4) includes weigit of component up to 3' above tie andve core. (5) Weight of lower core barrel and associated components more tien 3' above the active oore. (6) Weight of upper oore barrel and assomsted components more then 3' above the aseve m. 4.3.8 Reactor Vessel Intemal Surface Materials Sampilng Results A sample of the internal surface deposited materials was coEscted on a swab. The swab sample was taken of surface materials present in the RV intertor, at the bottom of the vessel. The swab was attached to a long segmented rod and inserted into the vessel from instrument access port (N-3), located at the top of the vessel head, wwdTiately 9" from the center of the vessel %L, Mi (see Figure 4 8). An additional sample was taken from a section of the safety injection piping system between the vessel and installed check valve. Both of these matenals are of sammar us B-I l

Saxton Nuclear Experimental Corporation (SNEC) Site Characterization Report composition. The safety injection materials are similar in relative radionudide composition, but have a higher specific activity, making their inclusion a more conservative choice. The combined materials from bcth samples (as % of total) were used to aid in determining the RV waste classification. The complete sample results for these two samples are presented in Appendix D. The following table contains the positive values from both B&W analysis results for these two sample materials, except where marked with an asterisk ("less than values *). The intamal surface area of the reactor vessel was determined from examining SNEC RV drawings. As stated earlier, an estimate of the curie content for intemal surface contamination in the reactor vessel is approximated from S1 & RV sample results and calculated RV intemal surface areas. Table 4-25, Radionuclides Composition Of Si And RV Sample Materials MtV SA MPLE SI SAMPLE RADIONUCLIDES % OF TOTAL % oF TOTAL H3 10853 0.0040' Fe-55 E1328 2.5315 A*, Ni-59 0.5477 0.5477 Ni43 410s49 43.5136 sr 90 4.1387 0.2839 Nb-94 0.0145' O.0079 Tc-99 E0033 0.0003* Pu 238 0.5548 1.0905 l Pu.239/240 L2723 2.3973 Pu-241 22.5424 27.4796 Am-241 3.9449 3.7232 Cm 242 10236 0.0390* Cm-243/244 E0682 0.0462* U 234 a.0002 0.0113* Co40 1L8892 17.1988 Cs-137 4.1739 8.3957 En-154 0.1407 0.0319 Eo64 4.0386 4.00R2 Note: Sold text andicates positively identmed nuchdas 4-69 A-2

e B & W RESULTS 12/15/95 - Saxtoft Samples SYSTEM m = = > 806** MV SWAS m r Total nCl/W 1 11 Total Uramum (ncl/g l I Results Normalised Ma - "' t Notes uCla to 100 % RATIOS "~ ' - - ~ 1 N=3 4 271:-04 a '"""% 'lWHNu===:?!!Il#ilillithu\\%%%%%%%%%W 2 G-f 4 TRS THAN 5.06 !-05 9 aaaa% 'lWe _ iiii#/',f///huMM4M%%W O.0078SI '"~- ' " 3 Fe-m 7.54 i-04 0.13 3 % 4 my-w 3.11 i-03 0.SesFw '#A //####ri/##A f#/hWMhM%%%%%W 5 a v -= LM :-01 45.08e0% i W/#### ril###ml//h%%%%M4%%MW 8 r-es Not ;';eh 0.00 !+00 01 ~ i ifis. __. _if i/tiiAuuhMuiuMM -= 7 t-as 2.35 ;-02 4.1aer= r ^^ 3-1: 1 0.9915SI 8 0a-pe was THAN 3.26 i-05 OA145 % i9 - - - _ _ _ --a jWJ A%%Mhh%MM l 9 M-SS 1.90 i-OS ei i ss &Mn%%ht%MW i, AMMW MMM '-f as ERS THAN 4.30 E-05" 04viw= i 10 - - = _.i r#4 A%EMW MMM 11 ae-ar .FRSTHAN 2.89 !-06u a-12 mu-23s 3.15 i-03 O P to-30 k an-21 0.14053 WMMnMMRW E ff- ~ ~ - ' ' -'22 13 mu same40 7.23 i-03 1. J<.= 14 eu-sei 1.28 !-01 Pu-41 i an-41 5.71439 ! 8A%M%%H M M 15 ~P' m for LESS THAN 5.40 i-os 0.0000 % 'A'!!# - 18 Ei~ ? 2.2 41 i-02 33449% Mi-l 1i Pu-lIl 6 7.11111 AAunMun4%%M 17 LESS THAN 7A2 E-06 0.0000 % 'lWid: '- fa n. gg ,,, w P 1.34 i-04 0 a***% 5 rm' " ~ . MAMMununM riffi: 4As sihM4ME l 337 i-04 0.we -w i 3 3 f.7/244 19 20 o 4 234 1.22 i-06 OA002% i TEF#! i i Eiir 4%EE% i W#5 4=duiWWM 21 i/-mas i _ERS THAN 1A0 l a - -. _,e U' M-: W 1 1- - 9~ ' 22 1-23s t.E5S THAN S.20 ; 0A0019 i Afm=' 01 U' IO 88129 9 E I 23 ,Ta-ap 9.50 !-02 16.8 M rl 24 i u-rey PRETHAN 6.06 i-04 0.10W1. n-- n.- -JAMREM !% i i 25 A r-r;Js n PRSTHAN 5.11 -05 a--- _- L~--n- ' ?sisiNWii as; e- 'm 5RBTHAN l127 i=04 a a**1% 'iinnn.. iMWiaGWU 27' J- 'as iFERTHAN I S.841 ;-05 OA15RI. CS S3-1 f7 1-- - m

s-sr i 2.37 l-02 4.173 %

Co-4I ms-11 iF 4.04841 3gi 291 co-44 MR8 T'.e 1.95 :-04 Or-Cl i-ti FiC;; : 0 0.24713 i u- 'KX tsm87"AN 7.70 ;-04 0.15l PIi d ~.._.. Wi A%MHM1MM - - ~ " 30 ~ T.iAfl&hMMnMMWM 's "..-.a lu-954 7.99 ;-04 0.1=, ie 31 iilW####dlifil A%MME% MEW i O ** Tu.- -'55 2.19 i-@ i l 32 Totske a > 5.55 :-O ' 1N-: I A-3

8 & W RESULTS 11/3/95 - Saxton Samples SYSTEM = m a > Sat lrtyInrecson Pionna - Ploe Seccan Benenen MVand Check Vaho. I r2806 0000I en Total TMu content (ncilg) 1 I 41 4400 ll Total Uransum content (nCI/g) se e i e n n on 4 le e int. Results Nortnalized Nuclide Notes uCl/o to 100 % Nuclides RATIOS LESS THAN i 8.37E-03 0.0040% f//##//#//##//#/u/#/////#t/Awwwwunggg% ggt 1 M-# GSSTHAN 1 1.195-021 0.0057% f#///H///##//#///#/U/H/#/HA%%%%%uush%%%\\ 2

- 94 i

3 re-3 i 15.27E + 001 2.5318% I Fe-es/Co-80 I c.147211 1 1.14E+001 0.5477% f##////######//##/#/#/uAnwungu%gstwn) 4 Mi-l 8 5 Ms-t 3 19.12E e 011 43 813eur#######1/#/#/#InnunuA9sungun%%%%%1 e 'r-i ls i LESS THAN N/R I 0.0000% f###//##########//#/uaguggg%g%% gung ] 7 'r-1 0 5.91 E-011 0.2830% # Sr-e0/Cs-137 0.71205I1 i e J i b-14 1.65E-021 0.0079% f##//######////##/##/#/A%%%%%g%%%%%m 9 7c-P LESS THAN 5.86E-04 : 0.0003% r/#/#######////#////#/#ua%%%gggg%%gh%1 1 10 J-f. LESS THAN 1.41 E-031 0.000F% f//////#########//####/An%%%%%%%%%%%1 I 11 No. LESS THAN 3.80E-011 0.t esst r#////#//#####I#/##1//#//agungsgu%wwww1 j 2.27E +001 1.0008%d Pu-2SS/Am-241 0.292908 l 12 Pu, ' f 13 Pu-23 0 40 4.99E+001 2.3973% f#1/##//######1/////#1/#/An%%%gg%%%%%%1 j 14 .pu, of 5.72 E +011 27.4790%4 Pu-241/Am-241 6 7.380654 15 ru, er LESS fMAN 2.35E-021 0.0113% F//////#1/###////##/#/////#/A%%%u%%%%%%%%1 i i is s m. '4 r i 7.7tE+00! 3.7232%d Am-241/Pu-23s i 3.414104 l 17 Am -; 'es LESS THAN I 1.52E-021 0.0073% t###/#/#//#//#//#I////#H//Au%%%%%%%%%nM1 l 18 cm, 54 2 LESS THAN I S.11E-021 0.03m)% t///#///##//#/#///#/#/#/#//An%%%ww%nhunst 1 19 car -2< L /T44 L :SS THAN 1 9.62Hi- 021 0.04Et% f#/#/#//###//##///##//##An%g%ghn%%%%%1 1 20 1. ', d L :S$THAN 1 2.36! !- 021 0.0113% f#//////##//#//#//#/#/#/#//Anh%%%%%%%%%%1 4 L [$$ THAN 4 6.54' :- 031 0.0031% F#//////##//###/////##///#/An%%%B%%%%%%%1 <1. 21 L :$$THAN I 1.1 31 !- 021 0.0054% I U-233/Asn-241 1 0.001444 i/. ', 22 685.840711 .,o -1 1,I i 3.581 !+011 17.1900%i Am-241/U-2SS 23i i 241 I u-t4 ir u :SSTHAN I 1.341!- 011 0.0644% l'###1/#####//#//#///////#n%%%%%%%%%%%%1 25! J. 7-70 Ie i L $$THAN i 1.4 51 !- 021 0.0070 % W//#Il#////#/#/#####///#a%%%%%%%MWW\\%1 26 b-f,

I L !SSTHAN I 3.1 71 !- 021 0.0132%lW//////#//////###////##///#d%%MhWWWW\\%%%1

.;s - f, o U!SS THAN

2.3 31 i=021 0.0112 %i Co-184dCo-137 0.020074 27 l

i;;s - f, T I S.301 i- 011 0.3957%i Co-40/Cs-137 48.132533 23 23 < e-f 44 LESS THAN t 4.341 - 021 05E4hi Cs-187/Co-GO 0.02318i 30 u-f s2 LESS THAN I 1.07E-011 0.0514 %J#//#/#///#////#//#//#/#///b%%%%%%%%%%%%1 u-754 I 6.85E-02t 0.0319% i%%%%%%%%%%%%%%%%%%%%g%%%%%%%%n1 31 i 32l D-f 55 i 1.70E-021 0.E-2%VHHH///UlHHHHHIHH/HHH/Mu%%%%%%44%%1 Totatsa s > I 2.03E+02p 100.0D%I N/M - Not Reported .A-4

1 i i SNEC Facility BVInternalSmear Analysis On 4/1/98 three (3) smears were taken from within the SNEC Facility RV. The smears l were collected in a region centered approximately six (6) to eight (8) feet below the instrumentation entrance pon marked as N-6 on drawing 646J830. Smear number 1 (SX865980198) was taken from the South West section of the intemal wall of the upper i j core support barrel assembly. Smear number 2 (SX865980196) was taken from the west l internal wall of the of the upper core support barrel assembly. Smear number 3 (SX865980197) was taken from an internal vertical instrument tube or instrument string. Estimates of the three surface areas samoted were made They art l Smear 1 (SX865980198)- ~450 cm 2 2 L Smear 2 (SX865980196)--100 cm 4 2 Smear 3 (SX865980197)- ~200 cm l Gamma scans of the smears produced the following activity results: Sample No. Co40 (uCTs) Cs 137(uCrs) An-241 (uCPs) Arve Sampled la cm' SX8659801% 4.41944E-02 3.94695E-03 1.34386E-03 100 200 SX865980197 9.40322E 02 2.23344E-03 SX865980198 2.23932E-01 1.56626E 03 4.43920E-03 450 l Activity per unit area (cm ) is estimated in the following table: 2 Sample No. g[c ) Cs 137 (uCVcm ) Ass-241 (uClhan') 8 SX865980196 4.41944E-04 3.94695E-05 1.34386E 05 SX865980197 4.70161E 04 1.!!672E 05 SX865980198 4.97627E 04 3.48058E 06 9.86489E-06 i Average uCi/cm is presented below: ) 2 s 8 Sompie No. l g g Cs-137(uCVca ) Aan-241(uCUcm ) AVERAGE 4.699E 04 1.804E 05 1.165E 05 In the following Table the sctivity of the RV for each result is provided assuming 1% of the total activity was removed (10% ofloose representing 10% of total). The RV 2 internals have ~1.13E K)6 cm of surface area. RV Total For Co4e(uCO Cs-137 (uCI) An-241 (uCI) AVERAGE 5.31E44 2.04E+03 1.132E+03 i On the following page the above values are used to scale in the total activity for the SNEC RV based on internal isotopic distribution from SI & RV samples. h - 5~

l Total (bened on Ce40) Tetsf (based on Cs-137) Total (based on M1) Isotope s 'falue % of Total ucr. ucri ucrs Fe-55 729E-02 0.74 % 2977.7 640.5 220.7 N b-94 124E-03 0.01 % 50 6 10.9 38 Co-60 1.30E+00 13 12 % 53099 9 114222 3934 8 Ni-59 6.08E-02 0.61 % 2483 4 534.2 1840 Ni-63 4 84E@ 48 83 % 197695 0 42525 6 14649.6 Eu-154 7.53E43 0.08 % 307 6 66.2 22.8 Eu-155 1.71E 03 0 02 % 69 8 15.0 52 H3 414E-03 0.04 % 169 1 36 4 12.5 Cs-137 2.32E-01 2.34 % 9476 3 2038 4 702 2 Sr-90 224E-01 226% 9149.5 1968 1 678.0 Pu-238 9 04E-02 0.91 % 3692.5 794 3 273.6 Pu-239/240 2.05E-01 2.07 % 8373 4 1801 2 6205 Pu-241 2.43E@ 24.52 % 99256.0 21350.7 7355.1 Am-241 435E-01 4 39 % 17768 0 3822.0 1316 6 U 234 6 63E-04 0.01 % 27.1 58 2.0 Cm-242 4.21E-05 0.00 % 1.7 04 01 Cm-243 5 89E-03 0.06 % 240.6 51 8 17 8 S U M => l 9.91 E +00 l 100.00% l 4048381 87084l 29999l Total Based Os Co40 (Ci) 1 Total Based Os Cs-137 (Ci) Total Based On Am-241 (Ci) 4.048E-01 8.708E-02 2.999E-02 These numbers do not represent what is present at the bottom of the vessel but instead are more representative of vertical surfaces in the high flow area (probably a lower bounding limit). Barry Brosey x8330 pl # _ g IL Holmes Approved R. D. Holmes x 8637 l f~

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Cmear Nuclice I,ibrary Used: 'C:\\ GENIE 2K\\CAMFILCS\\150,NLE . tDENTIFED NUCLIDES Wet ir!e Id Saergy Yield . Activity' Activity Na:re Confid.ence ('av) [t) D.ici/UN ) Uncertainty . CO-60 0.995 1173.22*' 100.00 4.4 *i377.E.002 4.080925-003 3332rd9*' 300.00. 4.119445-002 4.44036C-003 QS-137

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Smear. Wuclide Library Usedt.C:\\ GENIE 2K\\Cset'ILE$\\150.sGa IDENTIFIED NUCI. ids Muc.U de 2d' Energy Yield Actietty Activity Name Confidence (kgVI. [t). (uC1/UN ) Uncertainty 50-60 0.995 ida.22*. I00.00 2.19674G-001 1.931715-002 '3.U2.49*

  • 00.00 2.23932E-001 2.1826?E -002 CE-13'1 0,994 66't.65*

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l t l l 1 l l I l APPENDIX B l l REACTOR VESSEL l WETTED SURFACE AREA l l CALCULATION i l l l l l l l Saxton Reactor Pressure Vessel and Intemals WMG Report 9801-7025 Final Characterization Rev. 2 4/98 i

I i l l APPENDIX B REACTOR VESSEL WETTED SURFACE AREA CALCULATION l l [ i Saxton Reactor Pressure Vessel and Intemals WMG Report 9801-7025 Final Characterization Rev. 2 4/98

I Supplemental Area Calculations 3.1.6 & 3.1.7 Lower Core Plate & Guide Blocks This item includes The Lower Core Plate AND Guide Blocks. The Lower Core Support Plate Surface Area of 2176.2 in' calculated per 5.31, Sheet 4 of 13-Appendix B does not include the 32 guide blocks attached to it. The dimensions for appended sketch were obtained from Ref. 7. The area calculation in a format similar to Appendix B for this item is shown below. A = 1 Side of the Block - Hole Area + 2(1 Side of the Shroud) 2 2 A= (5.0625 ) -(n/4 3.5625 ) + 2[(5.0625+.375+.125) 3.6875) A = (25.63 - 9.97) + 75.7 A = 91.36 in' per Guide Block Thus the total area is: A = Lower Core Plate + 32 Guide Blocks A = 2176.2 + 32(91.36) 2 2 A = 5099.7 in = 3.29+04 cm 3.1.8 Lower Support Shroud Tubes Per Section 3.1.7 the portion of these tubes to which the activation calculations correspond is 36 inches tall. Therefore, the length of the Large Tube (6.5 inch diameter) must be adjusted to 25 vice 38 inches, while the length of the Small Tube (3.5 inch diameter) remains 11 inches for a total unit length of 36 inches The Lower Support Shroud Tube Area (9 Tubes) is thus: A = (2 x Large Tube + 2 x Small Tube 4 Interface Area ) x 9 Tubes 2 2 A = [(2 x 6.5 25) + (2 ' n 3.5 11) + 2 n/4 (6.5 - 3.5 )]9 = 11,790 in' = 2 7.61E+04 cm 3.1.9 Lower Support Tie Rods Per Section 3.1.8 the portion of these rods to which the activation calculations correspond is 36 inches tall. Therefore, the length constituent of the Tie Rod Surface Area (5.3.4, Sheet 6 of 13, Appendix B) must be adjusted to 36 vice 43.5 inches. Referring to 5.3.4, Sheet 6 of 13, Appendix B, the Lower Support Tie i Rod Area is thus: 8(n 1 36) = 905 in' = 5.84E+03 cm 2 Saxton Reactor Pressure Vessel and Internals WMG Report 9801-7025 Final Characterization Rev. 2 4/98 l L------_-----____--_-------------------------------------------B - - - - - -

3.1.10 Lower Support Balance 1 The summary of results of Appendix B states the surface area inside the Saxton reactor vessel that is routinely wetted dunng operations is 1.13E+06 cm2 Although not representative of the actual surface area of such items as the control rod spider. the difference between the total area (1 13 E+06 cm ) and the 2 sum of the other areas inside the Saxton reactor vessel listed below is included with this item as calculated below Total Area inside Vessel Routinely Wetted 1.13E+06 Core Baffle -1 21 E+05 Thermal Shield -168E+05 Reactor Vessel Cladding -2.47E+05 Lower Core Plate & Guide Blocks -3.42 E+04 Lower Support Shroud Tubes -7.03E +04 Lower Support Tie Rods -5 84E+03 Lower Core Barrel -4.82E+04 Upper Core Plate -5.70E+03 Upper Core Barrel 7 71E+04 Lower Support Balance 3.53E+05 3.1.11 Lower Core Barrel Per Section 31.10 the portion of the barrel to which the activation calculations correspond is 36 inches tall. Therefore, the sloped and upper cylinder of the Lower Core Support Barrel (5 5, Shest 9 of 13, Appendix B) are ignored and only the lower cylinder (and cut-outs) is considered. Referring to 5.5. Sheet 9 of 13, Appendix B, the Lower Cc.e Barrel Area is thus: A= 7643.1 - 173.2 = 7,470 in' = 4.82E+04 cm 2 3.1.12 Upper Core Plate The Appendix B surface area calculations do not identify the Upper Core Plate. It was assumed that the surface area to weight ratio of the upper and lower core plates is constant. Therefore the surface area of the Upper Core Plate can be calculated from the surface area of the Lower Core Plate (without guide blocks) and a ratio of their weights as shown below. A = 2176.2 in' x 264 lbs (Upper Core Plate) + 651 lbs (Lov.er Core Plate & Guide l Blocks) A = 883 in* = 5.70E+03 cm2 Saxton Reactor Pressure Vessel and Internals WMG Report 9801-7025 Final Characterization Rev. 2 4/98 B-2

3.1.13 Upper Core Barrel Per Section 5.13 of Ref. 2 the portion of the barrel to which the activation calculations correspond is 36 inches tall. Therefore, the sloped and upper cylinder of the Lower Core Support Barrel (5.5, Sheet 9 of 13. Appendix B) are considered. Referring to 5.5, Sheet 9 of 13, A'ppendix B, the Upper Core Barrel Area is thus: A = 9693.5 + 2256.8 in: = 11950.3 in' = 7.71E+04 cm' 3.1.14 Reactor Vessel Insulation Can The exterior surface area of 100 inch tall. 77 '/e diameter right circular cylinder is calculated as shown below: A=n OD Height A = n 77 375 100 2 2 A = 24,310 in = 1.57E+05 cm l l Saxton Reactor Pressure Vessel and Internals WMG Report 9801-7025 Final Characterization Rev. 2 4/98 8-3

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Ref Dwgs MV-80016 1 Dimension Feet inches em ! Radius.cm Rx Vs! 10 4.833 58.000 147.321 73.66 Lower Canning ID 6.406 76.8751 195.26l Lower Canning Thickness 0.021 0.2501 0.64l i Lower Canning OD 6.4481 77.375i 196.531 98.27 Lower Canning Length, Support Skirt Flange to below seam 8.323 99.875 253.88 Shot Sheeld OD 5.167 62.000-157.481 78.74 Ref G01-1192-003. 5.16 Support Can ID 6.7501 81.000) 205.741 Support Can Thickness 0.0421 0.5001 1.271 Support Can OD 6.8331 82.0001 208.28: 104.14 Support Skirt Flange to Closure Bolt 6.750 81.000.- 205.74 a Area Calculations it' in e,2 Lower Canning Cylinder 168.5951l 24277.69 156629.9 Shot Shield Hemisphere 41.931541 6038.141-39955.671 Lower Connecting Disk ! 11.68766 1683.022. 10858.19i Upper Cylmder 144.906 20066.46 1346221 j Closure Head Hemisphere 36.69555i 5284.159 1 34091.281 ) Upper Connecting Disk 18.325961 2638.938! 17025.371 TOTAL OUTER RV AREA 1 422.14171 60788.411 39218L51 i l C-5 l L______________.____________________________..._________________________________

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l Saxton Vesact Extemcl Crntaminatign Estimate Survey Survey Location B-G . Alpha Date Number dpm/100 cm2 dpm/100 cm2 i 06/21/95 863-95-252 15 8690 57 16 16300 69 f 17 4900 0 l 18 2880 0 l 19 18840 126 20 11440 44 21 22600 44 Avg 12236 49 10/12/95 862-95-638 25 10000 13 l 10/19/96 ?? 6 120000 320 l 10/21/97 231-97-1545 7 14000 31 l' Max dpm/100 cm2 120000 320 i dpm/cm2 1200 3 10% 10% Swipe Removal Efficiency Adjusted dpm/cm2 12000 32 uC1/cm2 5.41E-03 1.44E 05 (, _ Estimated Surface Area cm2 3.92E+05 4.70E+09 1.25E+07 Total dpm 7.84E+07 2.09E+05 Total dps Total Curies 2.12E-03 5.65E-06 /l/', / i o A /.d ( Prepared By-V /@ C Reviewed By-Page 1 of 2 c7 1

Distnbution Data From Site Characterization (Area 6 - Table 44) Activity. A2 A2 Sample Pert:ent Curies. Curies Fracbon H-3 2.00E 03 0.012 % 2.58E-07 1.08E+03 2.39E-10 C-14 1.10E-05 0.000 % 1.42E 09 5.41E+01 2.62E-11 Tc-99 6.20E-04 0.004 % 7.99E 08 2.43E+01 3.29E-09 l l-129 <2.00E-05 l Fe-55 2.50E-02 0.152% 3.22E-06 1.08E+03 2.98E 09 l Ni43 1.30E-01 0.791 % 1.68E 05 8.11E+02 2.07E 08 l Sr-90 2.30E 02 0.140% 2.96E 06 2.70E+00 t10E 06 Co40 7.18E+00 43.665 % 9.25E-04 1.08E+01 8.57E 05 l Cs-137 9.03E+00 54.916 % 1.16E43 1.35E+01 8.62E 05 Pu-238 3.10E-03 0.019 % 3.99E-07 5.41E 03 7.38E-05 l Pu-239/240 8.20E-03 0.050 % 1.06E 06 5.41E 03 1.95E 04 l Pu-241 2.80E 02 0.170 % 3.61E-06 2.70E 01 1.34E-05 i Pu-242 2.40E-05 0.000 % 3.09E 09 5.41E-03 5.72E47 l Am-241 1.30E-02 0.079 % 1.68E-06 5.41E 03 3.10E-04 l Cm-242 1.00E-04 0.001 % 1.29E-08 2.70E41 4.77E 08 Cm-243/244 3.00E44 0.002 % 3.87E 08 8.11E 03 4.77E 06 ) l 1.64E+0i 100.00 % 2.12E-03 7.71E 04 l Page 2 of 2 l C-8

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