ML20236N916

Rev 2 to Wmg 9801-7025, Saxton Reactor Pressure Vessel & Internals Final Characterization

ML20236N916
Person / Time
Site:	Saxton File:GPU Nuclear icon.png
Issue date:	04/30/1998
From:	External (Affiliation Not Assigned)
To:
Shared Package
ML20236N893	List: ML20236N889 ML20236N910 ML20236N916 ML20236N936 ML20236N948 ML20236N963
References
WMG-9801-7025, WMG-9801-7025-R02, WMG-9801-7025-R2, NUDOCS 9807160058
Download: ML20236N916 (64)

Text

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Saxton Nuclear Experimental Corporation Facility Reactor Vessel Characterization Report I

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SAXTON REACTOR PRESSURE VESSEL AND INTERNALS FINAL CHARACTERIZATION i Report WMG 9801-7025 i

April 1998

, Prepared for:

Raytheon Engineers and Constructors Prepared by:

WMG, Inc.

16 Bank Street Peekskill, NY 10566 l i

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i FOREWORD i

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'This report details component characterization work performed by WMG, Inc. to support the decommissioning of the Saxton Nuclear Power Plant by Raytheon. This work was performed by WMG, Inc. under Raytheon subcontract 77581803-49-1.

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i Saxton Reactor Pressure Vessel and internals l

!. WMG Report 9801-7025 1 Finst. Characterization Rev. 2 4/98

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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 3.1.2 Thermal Shield. . . .. . .. .. . ... . ... . . 3-2 3.1.3 Reactor Vessel Cladding . . . . . . . . . . ... . 3-2 3.1.4 Reactor Vessel . .. . . . . ... .. . . . 3-2 3.1.5 Reactor Vessel Insulation . . . . .. . . . .. . . . 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 Barre!.. . .3-3 3.1.12 Upper Core Plate. .

. . .3-3

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i 3.1.13 Upper Core Barre!.. . .. .. . . . 3-4 3.1.14 Reactor Vessel Insulation Can . .. . .. .. . . .3-4 3.2 Radiological Characterization Results.. . . . . . .. .3-4 3.2.1 Neutron Activation Products . . . . . . .. ,3-4 l 3.2.2 Internal Surface Contamination . . . . .. .. . .3-7 3.2.3 Extemal Surface Contamination . .. . . . .. . .3-7 4.0 DOT CLASSIFICATION RESULTS... . . . . . .4-1 4.1 LSA ll1 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 l

' APPENDIX C - REACTOR VESSEL EXTERNAL CONTAMINATION DATA l Saxton Reactor Pressure Vessel and Internals WMG Report 9801-7025 Final Characterization )

Rev. 2 4/98 i ii ,

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LIST OF FIGURES Figure Title Page 2 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 lli 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 1

Saxton Reactor Pressure Vessel and internals WMG Report 9801-7025 Final Characterization Rev. 2 4/98

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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 characterization 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 characterization 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),

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l Saxton Reactor Pressure Vessel and Internals WMG Report 9801-7025 I Final Characterization Rev. 2 4/98 1 -1 i

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2.0 EXECUTIVE

SUMMARY

The SNEC reactor vessel and internals consist of the activated internal components and the activated reactor vessel that surrounds the intact immovable internals components and are shown in Figure 2-1. The vessel and intact internals comprise the 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.

l Final radiological characterization of the SNEC reactor vessel transportation system l yielded a total activity of 1,285 curies as of 8/1/98 with 1,282 curies attributed to l 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) 111 material. Each component within the vessel and the vessel itself have specific activities below the LSA-Ill limit of 2 x 10

A2/g. On average, the vessel and the internals, excluding the grout, has a specific activity of 1.7 x 10 A2/g ( < 0.1 % of the LSA 111 limit) and the worst case component 4

l has a specific activity of 7.1 x 10 A2/g (< 4% of the limit).

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.

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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 activated from neutron exposure during plant operations. These activated metal 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 sur' face contamination decay corrected to August 1,1998 are provided herein. The surface 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 1 quantified separately.

3.1 Physical Description The SNEC reactor is a 23.5 MWt (nominal) experimental 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 i 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 /e inch thick box-like structure immediately surrounding the reactor core. The baffle assembly weighs approximately 1,210 pounds and has a surface area of 1.21E+05 cm2 Saxton Reactor Pressure Vessel and Internals WMG Report 9801-7025 Final Characterization Rev. 2 4/98 3-1 1

3.1.2 Thermal Shield The thermal shield is attached to the inner wall of the reactor vessel. It is a 94 inch tall hollow nght circular cylinder with an inner radius of 24.5 inches and a thickness of 3 inches. The thermal shield has an estimated weight 2

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 '/a inch. The cladding has an inner radius of 28 7/e inches, an estimated weight of 2,040 pounds, and a surface area of 2.47 E+05cm'.

3.1.4 Reactor Vessel The reactor vessel includes a removable closure head and a 143 inch tall laminated plate cylinder flanged at one end and fused to a hemi-spherical 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 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 penetrations for the control rods and coolant flow. The lower core plate has2 a weight of 500 pounds and an estimated surface area of 1.40 E+04 cm.

3.1.7 Lower Guide 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 '/,e inches on a side with a 3'/ is inch diameter hole in the center. Two adjacent sides of the square plate have a 3 inch tall '/e inch thick sheet attached to help align fuel assemblies. The guide blocks weigh a total2 of 150 pounds and have a total estimated surface area of

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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 l

_ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ __ . _ _ _ 1

3.1.8 Shroud Tubes c

t 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 portion of the tube is a 38 inch long right circular cylinder 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' l

3.1.9 Lower Support Tie Rods L

' 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, l

3.1.10 Balance of the Lower Support Assembly j

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 l

The lower. core barrel separates the core inlet 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 cm'.

3.1.12 Upper Core Plate i 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 Saxton Reactor Pressure Vessel and internals WMG Report 9801-7025  !

- Final Characterization Rev. 2 4/98 3-3

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 banel (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 E+04 cm 2 3.1.14 Reactor Vessel Insulation Can This % inch thick,100 inch tall stainless steel right circular cylinder of 77

/s inch 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 cm2 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 documented 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 <

characterization results. The results from neutron activation are combined with l the internal surface contaminants and external contamination estimates provided I herein to determine the total activity by radionuclides which is summarized in l

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 l

dimensional neutron transport calculations with the ANISN (Ref. 4) program and activation analysis with the ORIGEN2 (Ref. 5) program.

This ANISN/ORIGEN2 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

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methodology. As shown in the table, the total activity of the components l l is 1,282 curies as of 8/1/98. These activation analysis results have been l 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 i 3-4

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Table 3-1 ,

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Radiological Characterization Results Summary l

i intemal Extemal Total Activation Coreiamination Contamination Activity f 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 Co40 4.52E+02 4.39E-01 9.25E-04 4.52E+02 i 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 l 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 l Tc-99 2.23E-03 LLD i .99E-08 2.23E-03 l 1-129 LLD LLD LLD Cs-137 8.05E-02 1.16E-03 8.16E-02 i Sr-90 7.76E-02 2.96E-06 7.76E-02 l Pu-238 3.03E-02 3.99E-07 3.03E-02 l

Pu-239/40 6.91 E-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 L

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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 sainples l obtained in April 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. 1 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 Report. Comparison of the isotopic distribution 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 f and was assumed representative of the quantitative contamination levels

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over the wetted surface of the areas of low flow within the reactor vessel.

This yields an estimated contamination level of 9.2 uCi/cm2 (as of 8/1/98). l l

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 cm2 and the internal surface of the reactor vessel is estimatea at 2.47E+05 cm2 as documented in GPUN calculation 5340 95-011 enclosed in Appendix B.

This surface area in conjunction with the 9.2 uCi/cm2 y elds a total activity 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 l 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 aerived 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

?

. o

}

Three swipe samples were obtained at different locations on the internals with sample areas varying from 100 to 450 cm2 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/ cm2 to 4.98E-04 uCi/ cm 2 2

with an average of 4.7 E-04 uCi/ cm . 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 IE 85-46 and2 a fixed to removable ratio of 10). The Co-60 'value of 4.7 E-04 uCi/cm was used in conjunction with Co-60 based scaling factors derived from the average nuclide distribution described 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+05 cm2 2 2

[1.13E+06 2 cm (total) - 8.42c+04 cm (thermal shield exterior)- 2.47E+05 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 activities,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 vessel showed 120,000 DPM/100 cm 2 This worst case contamination level was used in conjunction with representative isotopic sample data from Area 6 in the " Site Characterization Report" (Ref.1 Table 4-6), an estimated surface area of 392,000 cm2 and an assumed swipe removal 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. This results in a total exterior contamination activity of level of 21.2 millicuries.

The supporting data for this estimate is enclosed in Appendix C l

l i

Saxton Reactor Pressure Vessel and internals WMG Report 9801-7025 l

Final Characterization Rev. 2 4/98 l 3-8 C_________________

4.0 DOT CLASSIFICATION RESULTS The SNEC vessel and internals characterization results were classified according to the following DOT criteria.:

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 111 leach testing requirements of 49 CFR 173.468 The DOT classification results are summarized in Table 4-1. As shown in Table 4-1, the material 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 Ill )'

determination, below, unshielded dose rates and LSA Ill leach test are discussed separately 4.1 LSA Ill Determination I

l As discussed in Section 3 above the activities of individual components vary I considerably due to their location relative to the active core region during operation. Therefore, the radioactivity can not be considered to be " distributed throughout" the material. The material must be considered as a " collection of solid objects" for classification as LSA 111. The worst case specific activities from l

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 l

component is well within the concentration limits of 49 CFR 173.403 at < 4% of l 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 (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 l Final Characterization i Rev. 2 4/98 4-1 L l

Table 4-1  :.

DOT Classification Summary Total A2 A2 (1) RQ RQ (2) Fissile Nuclide Curies Curies Fraction Curies Fraction Grams (3)

Fe-55 2.17E+01 1.08E+03 2.01 E-02 1.00E+02 2.17E-01 ---

Nb-94 1.31 E-02 1.62E+01 8.07E-04 1.00E+01 1.31 E-03 ---

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 8.01 E+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 5.41 E+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 5.41E+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 ---

1129 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 5.41 E-03 5.60E+00 1.00E-02 3.03E+00 1.78E-03 Pu-239/40 6.91E-02 5.41 E-03 1.28E+01 1.00E-02 6.91 E+00 1.11 E+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-07. 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 RQ fraction is greater than 1 rnaking it a Reportable Quantity of radionuclides (3) The grams of fissile material is less than 15 making it fissilo excepted.

l j

i Saxton Reactor Pressure Vessel and Internals WMG Report 9801-7025 Final Characterization Rev. 2 4/98 4-2 l

f,, ,

Table 4-2 ,

I

\

LSA I!! Determination Based on Lower Guide Block Data

, Total A2 A2 A2/ gram LSA ll1 Nuclide Curies Curies Fraction Fraction 3

\

l l Fe-55 2.89E+00 1.08E+03 2.68E-03 3.88E-08 1.94E-05 )

l Nb-94 1.22E-03 1.62E+01 7.56E-05 1.09E-09 5.47E-07  !

Co-60. 5.06E+01 ~ 1.08E+01 4.68E+00 6.77E-05 3.39E-02 Ni-59 8.90E-01 1.08E+03 8.24E-04 1.19E-08 1.94E-06 5.96E-06 9.69E-04 "

{

Ni-63 1.09E+02 8.11E+02 1.34E-01 Eu-154 5.50E-06 1.35E+01 4.08E-07 5.90E-12 2.95E-09 Eu-155 1.25E-06 5.41 E+01 2.31E-08 3.34E-13 1.67E-10 H-3 1.68E-01 1.08E+03 1.56E-04 2.25E-09 1.13E-06 I 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 I l-129 0.00E+00 --- --- --- --- l 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 j Pu-238 6.61 E-05 5.41 E-03 1.22E-02 1.77E-07 8.83E 05 Pu-239/40 1.50E-04 5.41 E-03 2.77E-02 4.01 E-07 2.00E-04 ,

Pu-241 1.78E-03 2.70E-01 6.58E-03 9.52E-08 4.76E-05 l Pu-242 3.18E-04 5.41 E-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 l Cm-243/24 4.85E-07 8.11 E-03 5.98E-05 8.65E-10 4.32E-07 6.14E-06 l

U-234 2.29E-05 2.70E-02 8.49E-04 1.23E-08  !

I Totals 1.63E+02 4.93E+00 7.13E-05 3.57E-02 Weight Grams = 6.91E+04 l

l Saxton Reactor Pressure Vessel and intemals WMG Report 9801-7025 Final Characterization Rev. 2 4/98 4-3

. _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ -___a

b a 4.3 LSA 111 Test Requirements Since 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 111 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 distribution of this activity results in contamination levels of 3 uCi/cm 2 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 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 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.

I Saxton Reactor Pressure Vessel and Internals WMG Report 9801-7025 Final Characterization Rev. 2 4/98 4-4 i

L.__ _ _ _ _ _ _ _ _ . . _ _ _ _ _ _ . _ . _ _ _ . _ _ _ - - - - --

e 'O 5.0 NRC CLASSIFICATION RESULTS Component specific activities were compared to the waste classification limits of 10 CFR 61.55 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, there is one component, the lower guide blocks, which exceeds the Class C limits if considered separately. Although the lower core guide blocks are an integral part of the -

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 1 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 the BTP criteria is summarized in Table 5-2 and shows that the concentration averaging criteria are satisfied.

The concentration averaging BTP 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 ,

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 {

i applicable.

The maximum, minimum and volume averaged concentrations for each controlling '

l 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

By averaging the component that exceed Class C limits wi';h the lower core, the Class C limits are satisfied within the concentration averaging guidance of the BTP.

Therefore, 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 criteria for concentration averaging and are acceptable for disposal as low level waste.

i .

l l

Saxton Reactor Pressure Vessel and Internals WMG Report 9801-7025 Final Characterization Rev. 2 4/98 5-2

l Table 5-1 .' .

10 CFR 61 Classification Averaged Over All Components I Pan 61 Pan 61 i 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  !

Co-60 l 4.52E+02 7.30E+01 , l Ni-59 6.57E+00 1.06E+00 i 220 0.005 i Ni-63 i 8.01E+02 1.29E+02 i 7000 0.018 Eu-154 2.59E-03 4.18E-04 Eu-155 5.88E-04 9.50E-05 H-3 1.49E+00 2.41 E-01 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 Cs-137 ' 8.16E-02 1.32E-02 4600 0.000 Sr-90 7.76E-02 1.25E-02 7000 0.000 Pu-238 i 3.03E-02 6.18E-01 100 0.006 Pu-239/40 6.91 E-02 1.41E+00 l 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 l 1.44E-01 2.94E+00 ' 100 0.029 Cm-242 5.86E-06 1.20E-04 l 20000 0.000 i Cm-243/244' 1.27E-03 2.58E-02 ! 100 0.000 U-234 ' 2.29E-05 i Totals 1.29E+03 2.07E+02 2.17E+01 0.073  ! 0.018 Total Metal Volume (m3)= 6.19E+00 f Total Weight (gm) = 4.90E+07 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

1. Saxton Nuclear Experimental Corporation

• Site (SNEC) j Characterization Report, GPU Nuclear,1996.

2. The Saxton Facility Reactor Vessel, Internals, Ex-Vessel Lead, Structural Steel and Reactor Compartment Concrete Shield Wall Radionuclides 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.

l A. A Users Manual for ANISN, A One Dimensional Discrete Ordinates l 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.

I

7. (Drawing) Saxton Reactor Plant Lower Core Guide Block Assy..

881D181, Westinghouse Electric Corporation, February 1961.

l 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.

0, GPU Nuclear. l l

l l <

l l

)

Saxton Reactor Pressure Vessel and Internals WMG Report 9801-7025 I Final Characterization Rev. 2 4/98 6-1

APPENDIX A REACTOR VESSEL INTERNAL CONTAMINATION DATA l

Saxton Reactor Pressure Vessel and Internals WMG Report 9801-7025 Final Characterization Rev. 2 4/90

i

{

I

)

I APPENDIX A REACTOR VESSEL ,

INTERNAL CONTAMINATION DATA 1

I

)

J i

Saxton Reactor Pressure Vessel and Intemals WMG Report 9801-7025

l. Final Charactenzation Rev. 2 4/98 l'

Saxt:n Nucirr Experim:ntal Corporation (SNEC)

Cite Charact:rizati:n R: port 4 Table 4-24, Reactor Vessel Curie Estimate-Calculated Values ,

Components Weight in Ibs. Curies 24 Year After Shotdown Core Baffle 1,210 642.6 Thermal Shield 10,491 132.7 Reactor Vessel Clad 2,043 4.185 Reactor VesselWall 87,313 4.737 RV Taenlanan Can (1) 4,994 1.078 RV Support Can(1) 5,442 3.034 Imer Core Gmde Blocks 152 182.9 Iseer CorePlate 500 226.7 Inner Support ShroudTubes (2) 338 8.323 Iower Support Tie Rods (2) 54 2.113 Balance aflewer Support Annernhly (3) 106 0.0744 Imwer Core Barrel (4) 1,067 58.46 Upper Core Plate 264 165 Upper Core Banel(4) 454 19.71 hinne aflewer Core Banel Assembly (5) 1,823 0.2593 hinn* of Upper Core Barre! Assembly (6) 832 0.1183 Notes: (1} Includes weight of component between 3* above and 3* below the actrve core.

(2) includes weight of component down to 3' below the active core.

(3) Weight of lower support ehroud tubes, lower support lie rods and associated components more that 3' below the active core.

(4) includes weight of component up to 3' above the active core.

(5) Weight of lower core barrel and ===~r=**d wT.pe.wt more than 3' above the active  ;

core.

Weight of upper core barrel and associated w. T .por .t more than 3' above the active (6) core, i

L 4.3.6 Reactor Vessel Intemal Surface Materials Sampling Results A sample of the intemal surface deposited materials was collected on a swab. The swab sample was taken of surface materials present in the RV interior, 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, approximately 9* from the center of the vessel horizontally (see Figure 4-8). An additional sample was taken from a section of the safety injection piping system  ;

i between the vessel end installed check valve. t3oth of these materials are of similar 4-68 l I p-I l ,

i 1

Saxton Nuclear Experirnental corporation (SNEC)

Site characterization Report composition. The safety injection materials ar's similar in relative radionuclides l )

composition, but have a higher specific activity, making their inclusion a more conservative choice. The combined materials from both samples (as % of total) were used to aid in determining the RV waste classification. The complete sample results 1 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.inte_ mal,s,urface 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 SI & RV sample results and calculated RV intemal surface areas.

Table 4-25, Radionuclides Composition Of St And RV Sample Materials RV SAMPLE Si SAMPLE RADIONUCLIDES  % OF TOTAL  % oFTOTAL H-3 0.0858 0.0040' Fe-55 0.1328 2.5318 Ni-59 0.5477 0.5477 Ni 63 45.0849 43.8136 Sr-90 4.1387 0.2839 Nb-94 0.0145* 0.0079 Tc-99 E0033. 0.0003*

h-238 0.5548 1.0905 W-239/240 1.2733 2.3973 Pu-241 22.5424 27.4796 Am-241 3.9449 3.7232 Cm-242 0.0236 0.0390*

Cm-243/244 E0682 0.0462*

U-234 0.0002 0.0113' Co 60 16.8892 17.1988 Cs-137 4.1739 8.3987 Eu-154 0.1407 0.0319 Eu-154 0.0386 0.0082 Note: Bold text indicates pos:tively identifiec nuclides 4-69 A~d

8 & W RESULTS 12/15/95 - Saxt3n Ocmpl33

~

~ ~

SYSTEM = = = > Sob's AVSWA8 1

m se umeer m Der Total ncilg) 1 - 11 -

Total Uranium inCilg)

[- ' 'Bample Type i_ GUN SWAB B Results Normalized NMWe Nm uCl/a to 100 % Nuit tites RATIOI:

1 N-3 4 min-os 0.09EE% R#il##ffGfavillHilVHH)Aus%%%%%%huW 2 c- f 4 WRSTHAN 5.06E-05 0.0080 % WE/#!#v ' M##M#AAnn%WMHRWY 1 l.svisi i 3 Ur- i 7.54E-os- 0.1 = Ft ' *; S O.5esiw Wo 2* Yiiiih%%%%%B4%E 4 m -; i 3.11 E-03 ' .

5 Mem i f * *aE-01 45.0849 % WFI/#E l 5 5 ###P##AiWM%%%%%%%WL s .

r-i t P Not Reported 0.00 i+00 Ouuuv= in= ..iiiii//AAn%%%%%%%%W (

2A5 i-02 9.99156 7 . :r-i i1 4.13B7% lBr-ii,k . s - 11 17 s JiD- 4 ESS THAN 8.26l E-05 0.0145 % WJf#faf4Ui 1###J'ar#AA%%%%%%%%%%%

9 *c- P 1 ACE 0.0033 % WJ#v . ##; id#iiAt%%%%%%%%%%

10 '- f. JR$THAN '

4 ADE-05 i OA070% #; fiiJ L 'J i '; U###daf#AAn%%%%%%%%%% -

11 Ao, l r7 JRSTHAN 2.88E Omuuin #4 idF_ _ J#JTuiiihd%ius%%%%sh 12 res%  ; e 3.15E-03 0.5548 % Fu ' l l W so ' 41 0.14083 13 Pu-23 W240 7.2SE-03 1.2F33% '####a d a uff#1 F///h%%%WiWM%%%%1 14 Pu . 4f 1.2SE-01 22.5424 % Pu-l 41 Am . 41 15.71429 15 Pu . '42 WSSTHAN 5.40E-08 OD000% W###L 7#diinuu1 ##/hB%%%%%%%%%%

16 Am . Wf 2.24E-02 SA449% Am-;M,1/Pu-l SS 7.11111 -

17 Am . *i GSS THAN ' 7 **8-08 0.0000 % W////###1f#####u##A6%%%%%%%%%%%

16 cm-; 'd ' I 134E-04 0== ######lft/##////##//AA%%%%%%%%%%%

to cm-26 /; 44 SATE-os 0.0882 % '#1H/###4f##iiuii###At%%%%%%ut%%%

20 u. '. 4 1 **8 -06 0.0002 % #1/####Jf########/JA%%%%%%%%%%% i 21 U. 2; i WSS THAN 1 ADE-07 0.0000 % WY#iiiiiii/#/niiiiiiiiiiAA%1%%%%%%%%%

D U-; TI I WSS THAN 6.20E-07 0.0001 % J-23RIAm-241 0.00003 23 Co-< k1 i 9.59E-02 16ABB2% Am-941/U-238 SS129.03228 24 Mu-f 6 '

I l PRSTHAN 6.05E-04 0.1085 % W/#/#########I#///AA4%%%%%%%%%%

25 Aa- fc im i JSSTNAN 5.11E-05 0.0000 % W//#####hiiiiiii/####A%%%%%%%%%%W p 26 150- f. 5 JSS THAN 1.37E-04 04241% '///#/#####//#////#//##A%%%%%%%%%%%

27 Gs-fad ESS THAN 6 8.64E-05 0A158% Cs-184/Cs-137 0.00373 28( Cs- f D7 1 2.37E-02 4.1739 % Co-80/Cs-137 4.04641 29l Ge- f 44 GSS THAN I 1.96E-04 0.0340 %^

Cs-187/Co-80 0.24713 301 Eu- f 52 mRS THAN I 7.70E-04 0.1. J^ W######I///###/##//)A%%%%%%%%%%W 31 Eu- f 54 i 4729E-os 0.1= vin '/#///#####///#I#/f#/#1A%%%%%%%%%%%

32. Eu- f 55 l- 1 2.19E-os 0 0^ - - ,. W/#nn####/##In/#inAnnu\wwwwwuWW1 1 Totals a = > 1 5.68 E - 01il 100.00 % i A-3 1

B G W RESULTS 11/3/95 - Saxton Samples SYSTEM = = = > Saletv /meccon Piomg - Pice Seccon Between RVand Check Vahe. '

E Saston i B& W I E 72806.0000ll Isample Numaen Samose Numoer Total TRU Content (nCl/g)

I SX865950053 I 9509065-01 I 41 440011 Total Uranium Content (nCi/g)

I Samose Type e Pipe Section i I Descnotion a SCP. Item 6. Int. (

Results Normalized Nuclide Notes uCi/c to 100 % Nuclides RATIOS 1 H-3 i ESS THAN 8.37 E-03 : 0.0040% F#/#/###/##/#/#/####/#Au%n\%\%Bunuuu) 2 C- 7 4 i LESS THAN i 1.19E-021 0.0057% F#######/###/#####/#A%%%%%%%%%%%B1 3 Fe-55 4 1 5.27 E+001 2.5318 % I Fe-55/Co-60 i O.147211 .

4 Ni-59 4 1 1.14 E+001 0.5477% F/#/##//#//##/##/#/#/#/#ABh%%%%%%%%%uV 5 Ni-os 8 i 9.12E+011 43.8136% f######/#/##/#/#/#//##Auh%%%%%%%%%%j 6 Sr-89 i LESS THAN I N/R I 0.0000% F####///#//#/#/#/##//##/A%%%%%%%%%%%Bt 7 Sr-90 l 1 5.91 E-01 1 0.2839%e Sr-90/Cs-137 i O.712051i 8 NO-94 1 1.65 E-021 0.0079% F#/#/#//##//##//##///#////An%%%%%%%n%%nt 9 Tc-99 4 GSS THAN 5.86 E-041 0.0003% f/#////###///#####//###/Aun%%utuhnw%nt to 1-f29 i LESS THAN I 1.41 E-031 0.0007% F/#/#####//####/#/##/#A%%%%%%%%%%%%)

11 ha-237 1 LESS THAN 3.80E-011 0.1825% F/#####/#/###//#/N#/#//Athhnhuuhwn%M 12 Pu-238 1 2.27 E+001 1.0905%il Pu-238/Am-241 1 0.292901 13 Pu-239/240 1 4.99 E+00 l 2.3973% i/###/##/##////#////#/#///AWhhhhuuhn%%%)

14 Pu-247 1 5.72 E+01 1 27.4796% e Pu-241/Am-241 1 7.380651 15 Pu-242 I GSS THAN 2.35 E-021 0.'0113% r//##/####/#///#//#####A%%%%%%%%%%%u1 16 Am-24f 4 I 7.75E+00: 3.7232%I Am-241/Pu-238 I 8.414101 17 1 Am -243 1 MSS THAN i 1.52E-021 0.0073% 7//////####//####///#//#//A%%%%%%%%%%%Mi 18 Cm - 242 i LESS THAN i 8.11 E-021 0.0390% fl##/#//##/#////##//#//#//ABuuhhuhB%%%%1 19 Cm -243/244 i LESS THAN i 9.62 E-02 i 0.04a2% F#/#//#/##//#####/####A%%%%%%%n%%%%1 20 l U-234 i LESS THAN I 2.36E-021 0.0113% f#/#//#######/######/AMuunt%uunnhut 21 U-235 LESS THAN I 6.54E-031 0.0031 % F#//#/#//##/#/#////#//###A\u%%\B%%BMWWW1 22 U-238 LESS THAN I 1.13E-021 0.0054% s U-238/Am-241 1 0.00146 u 23 Co-60 1 4 3.58E+011 17.1988% 8 Am-241/U-238 t 685.84071 l 241 Ru-706 I LESS THAN i 1.34E-011 0.0644% f#/###!/#/#####/#/#//#/Anunnu%%%%num 231 Aa- f oam 8 LESS THAN I 1.45E-021 0.0070% F#/##/#/#/##I#//##//##//A\%%\uB%%%%%%M 261 Sb- f 25 i LESS THAN I 3.17E-021 0.0152% fl#/##//#####/##/####/A\n%%%%%%%%%\M 27' Cs- f 34 i LESS THAN i 2.33E-021 0.0112 % I Cs-134/Cs-137 i 0.028075 26 Cs-f37 I I 8.30E-011 0.3957 % I Co-60/Cs-137 43.132533 29 Ce- f 44 i LESS THAN I 4.34E-021 0.020e% s Cs-137/Co-60 0.023181 ,

30 Eu- f 52 i GSS THAN i 1.07E-011 0.0514% 7###//#/##/#######/#//A%\uuh%\%%%%MM 31 Eu- T54 I I 6.65E-021 0.0319% %%%%%%%%%%%B%%%%%%WWWWW\%%%%%%M j 32u Eu-f55 8 e 1.70E-021 0.0082% f#/#//##/#/////#Il#/####/1%umn"MWVnVM  !

l Totats = = > W 2.08E+02k 100.00% #

N/R - Not Repo1ed I

SNEC Facility RVInternalSmear Analysis -

On 4/1/98 three (3) smears were taken from within the SNEC Facility RV. The smears were collected in a region centered approximately six (6) to eight (8) feet below the instntmentation entrance port 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 core support barrel assembly. Smear number 2 (SX865980196) was taken from tiie west 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 ramoled were made They are:

2 Smear 1 (SX865980198)-~450 cm 2

Smear 2 (SX865980196)-~100 cm 2

Smear 3 (SX865980197) ~200 cm Gamma scans of the smears produced the following activity results:

2 Sampie No. Co40 (uCI's) Cs-137 (uCts) Am-241 (uCi's) Area Sampled in cm SX865980196 4.41944E-02 3.94695E-03 1.34386E-03 100 SX865980197 9.40322E 02 2.23344E-03 - 200 SX865980198 2.23932E-01 1.56626E-03 4.43920E-03 450 l Activity per unit area (cm 2) is estimated in the following table:

C 2 2 Sample No. g,e Cs 137 (uCFcm ) Am-241 (uC1/cm )

)

SX865980196 4.41944E-04 3.94695E-05 1.34386E-05 SX865980197 4.70161E-04 1.Il672E-05 -

SX865920198 4.97627E-04 3.48058E-06 9.86489E-06 Average uCi/cm2is presented below:

2 Sample No. Cs-137 (uCi/cm') Am-241 (uCi/cm )

CF ')

AVERAGE 4.699E-04 1.804E-05 1.165E-05 In the following Table the activity 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+06 cm of surface area.

l RV Total For Co40 (uCI) Cs-137 (uCI) Am-241 (uCi)

I AVERAGE 5.31EH)4 2.04E+03 1.132E+03 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.

& - 5'

. e + . ,, _,

Total (based on Co.60) ToWI(based on Cs.137) Total (based on Am441)

Isotope s Value  % of Total ucr. ucrs ucrs Fe-55 729E-02 0.74 % 2977.7 640.5 220.7 N b-94 124E-03 . 0.01% 50.6 10.9 3.8 Co-60 1.30E40 13.12 % 53099.9 114222 3934.8 NL59 6.08E-02 0.61 % 2463.4 534 2 184.0 Ni-63 4.84E40 4.83 % 197695.0 42525.6 14649.6 Eu-154 7.53E-03 0.08 % 307.6 662 22.8 Eu-155 1.71E-03 0.02 % 69.8 15.0 52 H-3 4.14E-03 0.04 % 169.1 36.4 12.5 Cs-137 2.32E-01 2.34 % 9476.3 2038.4 7022 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-233/240 2.05E-01 2.07 % 8373.4 18012 620.5 Pu-241 2.43E+00 24.52 % 99256.0 21350.7 7355.1 Am-241 4.35E-01 4.39 % 17768.0 3822.0 1316.6 U-234 6.63E-04 0.01 % 27.1 5.8 2.0 Cm-242 421E-05 0.00 % 1.7 0.4 0.1 Cm-243 5.89E-03 0.06 % 240.6 51.8 17.8 S U M=> l 9.91 E400 l 100.00% l 404838l 87084l 29999l Total Based On Co40 (Ci) Total Based On 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).

Bamf Brosey x8330 MM, & & Holmes Approved R. D. Holmes x 8637 Q~N

(/02/98 10: 12:11 AM Prge 3

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.................... IDTE ITIED }10CL1 DES ....................

Energy Yield Activity Activity Nuclida Id Uncertainty Name Confidence (kev) {t) (uci/UN }

0.995 1175.22' 200.00 9.40977E-002 S.38583E-003 CD-60 3332.4** 100.00 9.40322E-002 9.26195E-003 CS-U7 0.995 6M .65* SS 12 2.23344E-003 4.92721E-004

' - Energy line fcond in the spectrma.

9 = Energy line not used for Weighted Mean Activit.y Envrgy Toler& ace : 2.000 kev Nuclide confidence index threshold - 0.30 Errors quoted at 2.000 signa g.a . e r ci e - ste t- 5 /n g< \dt.md Pips 1

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• 4702/98 9:57:22 Mi Page 5

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............. ...... . TOENTIFIED NUCLIDES

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Nactide Id Eaefcgy Yield . Activity' - Activity Na:r:e Confidence (2ev) {t) DICi/UN ) , Uncertainty

..Co-60 0.995 1173.22" 100.00 . 4.d372E .002 4.080925-003 1332 49+~ 300.00 4.41944E-002 4.44036E-003 C3-137 0. 995 661.65' B5.12 '3. 9469hE-00?, 5.7 622.1E-O'04

• ' N1-241 C.996 5 9. 5'P .

25.90 1.34306C-003 4;S61695-0.04

• Energy line found.in the spectrt:r.

S =; Energy line -cot' used for Weighted.Mean Activity Energy Tolerance : 2.000 key Nuclide confidesce inde:< thre. : hold = 0.30 l Errors quoted at 2.000 sigtr.a - -

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.................... IDENTIFIED NUCIID55 ....................

2d Etjergy YielCL Activtty Activity .

NUC.U de Name Confidence (kev) (t). {uCi/UN l Uncertainty C0-60 0.995 1373.22* . 100.00 2.196740-001- 2.93171B-002

'U.32. 4.9* 100.00 2.23933E-001 2.18267E- 002 -

CS-137 0,9.04 661. 65* 85.12 1.56626E-003 6.42512E-DO4

. AM-241 D.99S $9.53*' 35.90 4.09205-003 1.05368E-003

• = Energy line found in the spectrum.

• Pe Eneriff line not used for Weighted Mean Activity -

. Energy 'rolerance : 2.000 kev -

0.30 Muclide' confidence index t.hreshold ==- .

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gg O T o l Tfi- sM

APPENDIX B i

REACTOR VESSEL WETTED SURFACE AREA CALCULATION i

l 1

l

)

Saxton Reactor Pressure Vessel and Internals WMG Report 9801-7025 Final Characterization Rev. 2 4/98 i

f. _ _ _ . _ _ _ . . _ _ . _ _ _ _ _ . _ _ _ . _ _ _ _ _ _ _ _ _ _

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

_______ -__ - _ - _ N

Supplemental Area Calculations 3.1.6 & 3.1.7 Lower Core Plate & Guide Blocks This item includes The Lower Core 2 Plate AND Guide Blocks. The Lower Support Plate Surface Area of 2176.2 in calculated per 5.3.1, 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 +

A= 2 2 2(1 Side of the Shroud)

(5.0625 )

A = (25.63 - 9.97) + 75.7 -(n/4 3.5625 ) + 2[(5.0625+.375+.125) 3.6875]

A = 91.36 in2 per Guide Block Thus the total area is:

A = Lower Core Plate + 32 Guide Blocks A = 2176.2 + 232(91.36)

A = 5099.7 in = 3.29+04 cm2 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: 1 A = (2 x Large Tube + 2 x Small Tube + Interface Area ) x 9 Tubes A = [(2 n 6.5 2' 25) + (2 x ' 3.5 ' 11) + 2 n/4 (6.5 2- 3.5 2)]9 = 11,790 in2 =

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  ;

l 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 Rod Area is thus: l 2

8(n

• 1 ' 36) = 905 in = 5.84E+03 cm2 i Saxton Reactor Pressure Vessel and Internals WMG Report 9801-7025 Final Characterization j Rev. 2 4/98 B-1

_ -_-_- a

r l- .

3.1.10 Lower Support Balance The summary of results of Appendix B states the surface area inside the Saxton reactor vessel that is routinely wetted during 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 cm2 ) and the 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 -1.68E+05 Reactor Vessel Cladding -2.47E+05 Lower Core Plate & Guide Blocks -3.42E+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 -7.71 E+04 Lower Support Balance 3.53E+05 3.1.11 Lower Core Barrel i Per Section 3.1.10 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, Sheet 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 Core Barrel Area is thus:

4 A= 7643.1 - 173.2 = 7,470 in2 = 4.82E+04 cm2 l 3.1.12 Upper Core Plate

The Appendix B surface area calculations do not identify the Upper Core Plate.

I It was assumed that the surface area to weight ratio of the upper and lower core plates is constant. Therefore the r,urface area of the Upper Core Plate can be calculated from the surface area of the Lower Core Plate (without guide blocks) l and a ratio of their weights as shown below.

2 A = 2176.2 in x 264 lbs (Upper Core Plate) + 651 lbs (Lower Core Picte & Guide Blocks)

A = 883 in' = 5.70E+03 cm 2

Saxton Reactor Pressure Vessel and Internals WMG Report 9801-7025 i

Final Characterization Rev. 2 4/98 B-2 l i

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, Appendix B, the Upper Core Barrel Area is thus:

2 2 A = %93.5 + 2256.8 in = 11950.3 in = 7.71E+04 cm' 3.1.14 Reactor VesselInsulation Can The exterior surface area of 100 inch tall, 77 '/s diameter right circular cylinder is calculated as shown below A=x OD ~ Height A = n 77.375 100 A = 24,310 in2 = 1.57E+05 cm 2 i

i I

Saxton Reactor Pressure Vesse! and Internals WMG Report 9801-7025 Final Characterization Rev. 2 4/98 B-3 E________.____________________-.._

suo,,gg Cate No. Rev.No. SheetNo.

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l APPENDIX C REACTOR VESSEL EXTERNAL CONTAMINATION DATA l

Saxton Reactor Pressure Vessel and internals WMG Report 9801-7025 Final Characterization Rev. 2 4/98 i

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Ref Dwgs MV-60016 Dimension Feet inches cm Radius.cm 4.833 58.000 147.32 73.66 Rx VslID Lower Canning ID 6.4061 76.8751 195.261 Lower Canning Thickness 0.0211 0.2501 0.641 Lower Canning OD 6.4481 77.375 196.531 98.27 Lower Canning Length, Support Skirt Flange to below seam 8.323 99.875 253.68 Shot Shield OD 5.167 82.0001 157.481 78.74 Ref G01-1192-003,5.16 Saaaart Can ID 6.7501 81.0001 205.74 ;

Snaaart Can Thidmess 0.0421 0.5001 1.271 Support Can OD 6.8331 82.0001 208.281 104.14 Support Skirt Flange to Closure Bott 6.7501 81.000 305.74 2

Area Calculations ft2 in' cm Lower Canning Cylinder 168.5951 24277.69 186629.9 Shot Shield Hemisphere 41.93154 6038.141 38955.67 Lower Conneding Disk 11.68766 1683.022 10858.19 Upper Cylinder 144.906' 20066.46 134622 Closure Head Hemisphere 36.695551 5284.159 34091.281 Upper Connecting Disk 18.325961 2638.9381 17025.371 TOTAL OUTER RV AREA 422.14171 60788.411 392182.51 i

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Saxton Vessel Extem:I Contamination Estimate B-G '. Alpha Survey Survey- Location dpm/100 cm2 dpm/100 cm2 Date Number 57 l 15 8690 06/21/95 863-95-252 16300 69 16 4900 0 17 2880 0 18 i

18840 126 19 11440 44 20 22600 44 21 12236 49 Avg 10000 13 862-95-638 25 10/12/95 120000 320

?? 6

_ 10/19/96 14000 31 231-97-1545 7 10/21/97 120000 320 Max dpm/100 cm2 1200 3 dpm/cm2 10% 10%

Swipe Removal Efficiency 12000 32 i

Adjusted dpm/cm2 5.41E-03 1.44E 05 uCi/cm2 3.92E+05 l

Estimated Surface Area cm2 4.70E+09 1.25E+07 Total dpm 7.84E+07 2.09E+05 Total dps I

2.12E-03 5.65E-06 Total Curies r ,

/A/,/2/n

/

Prepared By:

Reviewed By: / V h

l Page 1 of 2 C-1

j Distribution Data From Sit] Charact:rization (Area 6 - Table 44) f Actmty , A2 A2 i Sample Percent Curies . Curies Fraction l *

. \

2.00E-03 0.012 % 2.58E-07 1.08E+03 2.39E-10 .

H-3 1.10E-05 0.000 % 1.42E-09 5.41E+01 2.62E-11 C-14 Tc-99 6.20E-04 0.004 % 7.99E-08 2.43E+01 3.29E-09 l-129 <2.00E-05 -

Fe-55 2.50E-02 0.152 % 3.22E-06 1.08E+03 2.98E-09 Ni43 1.30E-01 0.791 % 1.68E45 8.11E+02 2.07E-08 Sr-90 2.30E-02 0.140 % 2.96E46 2.70E+00 1.10E 06 Co40 7.18E+00 43.665 % 9.25E44 1.08E+01 8.57E45 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 Pu-239/240 8.20E-03 0.050 % 1.06E-06 5.41E-03 1.95E-04 Pu-241 2.80E-02 0.170% 3.61E46 2.70E 01 1.34E-05 Pu-242 2.40E 05 0.000 % 3.09E-09 5.41E-03 5.72E47 Am-241 1.30E42 0.079 % 1.68E 06 5.41E 03 3.10E44 Cm-242 1.00E-04 0.001 % 1.29E 08 2.70E-01 4.77E-08 Cm-243/244 3.00E-04 0.002 % 3.87E-08 8.11E43 4.77E-06 1.64E+01 100.00% 2.12E-03 7.71E 04 Page 2 of 2 i

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