ML20116A119
| ML20116A119 | |
| Person / Time | |
|---|---|
| Site: | Saxton File:GPU Nuclear icon.png |
| Issue date: | 07/18/1996 |
| From: | GENERAL PUBLIC UTILITIES CORP. |
| To: | |
| Shared Package | |
| ML20115K054 | List: |
| References | |
| NUDOCS 9607250302 | |
| Download: ML20116A119 (250) | |
Text
{{#Wiki_filter:hwd4m-+= amW e,.s ATTACHMENT 2 REFERENCE 13 SUPPORTS QUESTION 24 RESPONSE
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i i i i I !. SNEC FACILITY SITE CHARACTERIZATION REPORT i } 4 Radiological and Environmental Data Supporting the Development of the ( Saxton Nuclear Experimental Corporation (SNEC), Decommissioning Plan
- Containment Vessel and Remainina External Facilities -
GPU Nuclear Corporation
S:xton Nuclear Experim:ntal Carp rati:n (SNEC) Site Characterization Report lA t
\ \~ l TABLE OF CONTENTS LISTOFTABLES.........................................................................................................................v LIS T O F FIG U R E S .. .. ... . ...... .... .. ....... ........ .. .. .. .. . . . .. . . . . .......... .. . .. ... ... ... ...... .. .... ... .. ............ .. .... ....... ... v iii 1.0 I NTR ODU CTIO N AN D OV E RV I EW .......... .. ...................................... .... .................................. 1 -1 1.1 PURPOSE.... ........ . . . .. . .. . . . . . . . . . . ... ... . . . . . . . . . . . . . .... 1 1 1.2 SCOPE . .. . . . . . . . . .. . . . . . . . . . . . . . . ... .. .. . . . . . 11 1.3 PROGRAM STRUCTURE. . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . .. . . 1-3 1.4 REGULATORY CRITERIA FOR UNRESTRICTED RELEASE . .. . . ... 1-3 1.5 ORGANIZATION AND RESPONSIBILITIES . . . . .. . . .1-4 1.6 CHARACTERIZATION ACTIVITY CONTROL. . .. . . . .. .. ... .. . 1-4 I
2.0 H i STOR I C AL I N FO R M ATI O N.... ................. ................... .. .............................................. ....... 2 1 2.1 SITE DESCRIPTION. . ..... .. . . . . . . . . . . . . . . . . . . . . . . . . . . .. . .2-1 l 2.1.1 Site Layout.... .... . .. ... . . . . . .. .. . . . . . . . . . .. . .. . . .. 2-2 2.1.2 Station Stnxtures.. .. .... . ... . .. .. .. .. . . . . . . . .. .2-3 \ 2.1.3 Facility Operational History... . .. .. .. .. . . . . . . . . .2-3 2.1.4 Historical Radiological Survey Data (1971) . . . . . . .. . 2-6 2.2 REMAINING SITE STRUCTURES. .. . . . . . .. .. .2-7 2.2.1 Area Designations.. . . .. . . . . . . . . . .. . .. . . ... . .2-7 2.2.2 Area Characterization Strategy.. . . . ... .. . . .. . . . . .. . . . . .2-9 3.0 CH ARACTERIZATION DOCU ME NTATION AND R EPORT . ........................ ................................... 3-1 3.1 GENERAL DOCUMENTATION . .... . . . . . .. .. . .. . . . .. 3-1 3.2 DATA PRESENTATION . ... . . . . . . . . . . . . . . . . . . . . . .3-2 3.3 RADIOLOGICAL DATA. .. . . . . . . ... . . .3-2 3.3.1 Unbiased Survey Results. . . . . . . . . .. . . . .3-2 3.3.2 Biased Survey Results. . .. . .. . . . . . . . .. .. .. . . . . . . .3-3 3.4 STRUCTURAL DATA... . ... . . . .. . . . .. . . . 3-3 3.5 SYSTEMS AND COMPONENT DATA.. . . . . . ... . .. . . . . . . .. .3-3 3.6 REACTOR VESSEL AND ADJACENT STRUCTURES - ACTIVATION DATA.. .. .3-3 3.7 ENVIRONMENTAL AND HAZARDOUS MATERIALS DATA. .. . . . .3-4 4.0 C H AR ACTE RIZATION R E S U LTS S U M M ARY ......... . .... . ............ ............................... ................ 4-1 4.1 STRUCTURAL CHARACTERIZATION.. . . .4-2
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! 1 l Sexton Nuclear Experimental C:rp:rati n (SNEC) Site Characterization Report [
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TABLE OF CONTENTS (Continued) 4.1.1 Data Needs.... . . . . . . . .. .. . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . .. .. . . 4 2 4.1.2 Applicable Areas and Surfaces.... . . ..... . . ... .. . .. . . . . . . . . . . . . . . . . ........ 4-2 4.1.3 Analytical Methads.. ......... ... . . . . . . . . . . . . ... . ... . . . . . . . . . . . . . .. .. .. 4-2 4.1.4 Selection of Measurement And Sampling Locations . ... .. .. ... ... ... .... . . . . . . . . . . . . . . .4-5 4.1.5 Estimates of Structural Surface Area (It') . ... . . . . . . . . . . . . . . . .. .. . .. . ... . . 4-6 l 4.1.6 Unbiased and Biased Surveys.. . . . .. . .. .. . . . . . . . . . . . . . . . . . . . . . .4-8 4.1.7 Data Reporting and Analysis.. . . . . . . . . . . . . . . . . .. . . . . . . .. .... 4-9 4.1.8 Surface Contamination Evaluations... ..... . .. . . . . .. . . .... . . . . . . . . 4-9 4.1.9 Conwete Surfaces-General Description.. .. . .. ... . . . . . . . . . . . . . . .. . . . . . . . .. . . . . ... 4- 1 1 4.1.10 Field Measurements Of Concrete Surfaces. . . . . . . . . . . . . ..... .. . . 4 13 4.1.11 Metallic Surfaces.. . ....... ... . ... . . . . . . . .. . . . . . . . . . .. . 4 14 i 4.1.12 Smear Analysis (General Area (GA), Smear Analysis Results).. . . . .. . . . . 4 15 i 4.1.13 Penetration Studies - Concrete Core Bore Sampling Methodology... . . . . . . . . . . . . . . . .. .... 4 19 l 1 4.1.14 Activation Of Concrete Core Bores.. . . . .. . . . .4-25 4.1.15 Fission Product Contaminated Core Bores.. . . . . . . . .. . . . . . . . . . . . . . 4-26 j 4.1.16 Applied Limits.. . . . . . . . . . . . . . .. . . . . .. .. . . . . . .. 4 26 l j
\v) 4.1.17 Conclusions.. . . ..
4.1.18 B&W Analysis Of Core Bore Samples .
... . . 4-27
.4-26 4.1.19 Estimating Concrete Waste Volumes.. .. . . . . . . . . . . . . . . . . . . . . .... . . . . .. . .4-30 j l
4.2 SYSTEMS AND COMPONENT CHARACTERIZATION ... .. . .. ... . ... .4-38 I 4.2.1 Data Needs.. . .... . .. . . . . . . . . . .... . . . . . . . . . . . .4-38 4.2.2 Applicable Areas and Equipment... .. . . ..... . . . . . . . . . . . . . . . . .. . . . .4-39 1 l 4.2.3 Selection of Sampling Locations.. . . . . .4-39 4.2.4 AnalyticalMethods. . . . . . . . .4-39 l 4.2.5 Data Reporting and Analysis.. . .. . . 4-41 1 4.2.6 Radiation Measurements On Equipment And Components.. . . . . . . . . . . .4-41 ) 4.2.7 Radiation Measurements of Piping Systems.. . . . . . ..... . . . . . . . . . . . . . 4-41 4.2.8 Samples OfIntemal/Extemal Contamination .. . . . .. ... .. .. . . . .. . 4-42 4.2.9 Sample Results Relating To Decontamination. ... . . ... ... .. .. . 4-44 4.2.10 Results Of Piping System Analysis.. . . . . . ... .. . . . . . . .4-45 4.2.11 Sample Data Used in Curie Estimates For Components And Structures.. . . .4-47 4.2.12 Curie Estimates For Components And Equipment... . . . . .. . . 4-50 4.3 ACTIVATION OF SNEC REACTOR VESSEL AND SURROUNDING STRUCTURES.. .4 56 4.3.1 Data Needs.. ... . .... . . ... . . . . .4-56 ii
t S;xtan Nucle:r Experim:ntal Ccrp:ratirn (SNEC) l Site Characterization Report l l ! Q{ TABLE OF CONTENTS (Continued) , 4.3.2 Applicable Areas and items... ..... ... ... .. . .. . . . . . . . . . . . . . .......................4-57 I 4.3.3 Analytical Methods........... ..... ........ . ... . . . .. . . . . . . . . . . . . . . . . . . . . . . .. .. . .. 4-57 l 4.3.4 Confirmatory Stsvey Data. .. . . . . . . .. . . . . . .. .. . . . . . . . . . . . . ....... .. .... 4-58 : 4.3.5 Activation Analysis Results Summary..... . .. . . . .. . . . . . . . . . . . ...... 4-67 4.3.6 Reactor VesselIntemal Surface Materials Sampling Results.-.. .. .. . . . . . . . . . . . . . ...... 4-68 4.3. 7 Conclusions .. ....... . . . .... .. . . . . . . . . . . . . . . . . . . . . . .. . . . . . . . . . . . . . . . . . . . . . . . .. 4-70 l 4.4 ENVIRONMENTAL & HAZARDOUS MATERIALS CHARACTERIZATION.... .... . . . . . . . . . . . . . . . . 4-71
- 4. 4.1 Data Needs .... .......... ... .... ... ... . . . . .... .. . .. .... . . . . ... . .. ...................................4-71 4.4.2 Selection Of Sampling Locations... ... . . . ... ..... .. . .. . . . . . . . . . . . . . . . . . . ..... 4-72 l 4
4.4.3 Soil Characterization Repods.. . ... ..... . . . . . . . . . . . . . . . . . . . . . . .. .. . 4-73 \ l 4.4.4 Environmental Characterization Of SNEC Soils.. ... ... .. . . . . . . . . . . . 4 73 l 4.4.5 Applicable Contaminants and Sampling Media in Soil.. . . . . . . . .. . . . . . . . . .. . . 4-74 4.4.6 Air, Sediment And Ground / Potable Water... . .. . . . . . . . . . . . ..... .... . ...... 4-75 4.4.7 AnalyticalMethods... . .. . . . .. . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . ... .4-80 4.4.8 Sample And Analysis Summary Usting For Samples Extemal To SNEC CV.. . . . . . . . . . .. .4-81 4.4.9 Summary Of Environmental Results Extemal To SNEC CV.., . . .... . . . . . . . . . .4-82
/m\ 4.4.10 Data Reporting and Analysis For Hazardous Materials Wrthin The CV.. .. .. .. . . .4-84 l
4.4.11 SNEC Site Characterization Plan Repoding.. . . . . . .. . . . ... . . . . . . . . . 4-85 1 4.4.12 PCB Sampling Results.... . . . . . .. ............ . .. . . .. .. . . . . .4-85 4.4.13 Bunk LeadInventory..... ... . . . .. .. . . .. . . . . . .4-86 4.4.14 Leadin Paint And Other Sample Materials ...... .. .. . . . . . . .. . . . . . . . . .. . . . . . . 4-86 4.4.15 Asbestos Sampling Results... .. . . . ... ... ... . . . .... . . . . . . . . . . ... 4-87 4.4.16 TCLP, Inorganics Sampling Results.. . . . . . . . . . . . . . . . . .. . . . . . . . . . . 4-92 4.4.17 Bulk Cadmium & Mercury.. . . . .. . . . . . . . . ... . . . .. .. . .. 4-93 4.5 DATA TABLES.. . . . . . . . . . . . . . . . . . . . . . . . .. .. . . . . . . . . . . . . . . . 4-94 5.0 HEALTH PHYSICS SURVEYS SUPPOR TING CH AR ACTERIZATlON .............................. ......... ... 5-1 5.1 DOCUMENTATION (GENERAL).. .. . .. . . . . . . . . .. .5-1 5.2 HEALTH PHYSICS SURVEY RESULTS. . . . . . . . .. . . . . . . . . . . . 5-1 5.3 ALARA CONSIDERATIONS.. . . . . . . . . . .. .. . . . . . . . . .. 5-2 5.3.1 Worker Exposure Control... . .... . . . . . . .. . . . . . .. .. . . . . . .5-2 l 5.3.2 Assessing Risks... ... . . . . ... . . . . . .. . ....... 5-2 5.3.3 Special Considerations For Airbome Radioactive Material Generation Activities., . . . . . . .. 5 3 l 5.3.4 Air Sampling Results And Procedures.. .. . .. . . . . . . . 5-4 5 ' \ 5.3.5 Radon in The SNEC CV.. . . .. . . . . . . 5-5 iii l
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l l 1 1 Snxtsn Nucle:r Experimsnt:1 Corparr.tlon (SNEC) l Site Characterization Report J TABLE OF CONTENTS (Continued) i 1 l 5.3.6 Routine Radiological Surveys.. ... . . . . . .. . . . . . . . . . . . . . .... .. . . . .. . S-S l 5.4 CHARACTERIZATION SURVEY METHODS... .. . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . .. .. .. 5-6 l 5.5 INSTRUMENTATION, USE AND PERFORMANCE .. . . . . .. . . . . . . . . . . . . . . . . . . . . . . . . . ...57 l 5.6 HOT SPOT IDENTIFICATION PROGRAM.. . . .. . ........ . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . 5-7 , 1 5.7 PERSONNEL RADIATION EXPOSURES ..... .. .. . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . 5-8 l 1 6.0 QA/ Q C R E Q U I R E M E NTS . .... ........... ..... .......... ......... . ..... . . . ... . . .... .. .. .. .... . ..... . .. ..... ..... ...... ... .. ......... 6-1 i ; l 6.1 DATA COLLECTION... . . . . . . . . . . . . . . . . . . . . . . . . . . .. . . . . .. .. 6-1 6.2 MEASUREMENT QUALITY CONTROL . .. . . . . . . . . . . . . .. .. . .. ...... 6-1 ; 1 6.3 SAMPLE RETENTION . .. .. . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . 6-2 l
7.0 REFERENCES
..............................................................................................................................7-1 I LIST OF APPENDICES O
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l i l S*xt:n Nucl:rr Experimental C:rp:rcti:n (SNEC) Site Characterization Report i (~h ' ! (v) List of Tables Table Title Page 2-1 Historical Radiological Conditions During Operation (1971) 2-6 1 4-1 Radionuclide Inventory for the SNEC Facility 4-1 4-2 Radionuclide Concentrations in CV Tunnel Water and Sediment 4-12 4-3 Direct Frisk Measurements of Concrete Surfaces 4-13 4-4 Metallic Surfaces 4-15 4-5 SNEC Facility Surface Contamination Analysis Results - Composited Smears of January 1995 4-17 4-6 SNEC Facility Surface Contamination Analysis Results - Composited Smears of June 1995 4-18 4-7 Off-site Core Bore Locations and Counting Results 4-21 [] 48 Countiig Results, On-Site Core Bore Locations Outside of CV 4-22 4-9 GNdC CV Concrete Core Bore Sample Locations 4-24 4-10 Composite Results - Concrete Core Bore Sample SX862950119 4-25 4-11 Comparison of SNEC and B&W Concrete Analysis Results (pCi/g) 4-29 4-12 SNEC Concrete Waste Volume Estimate 4-31 l 4-13 SNEC Facility Radionuclide inventory of Contaminated Concrete 4-32 j 4-14 Radionuclide Content of Septic Tanks A & B 4-43 4-15 Systems Samples From SNEC Facility 4-44 4-16 TRU and Ni-63 Content of Piping System Sample Materials 4-45 4-17 Radionuclide Inventory for SNEC Facility Piping 4-46 i 4-18 Radiological Samples and Survey Data Used in SNEC Curie Estimates 4-47 4-19 Summary Listing of Activities and Waste Classifications 4-52 to 4-54 4-20 Radionuclide Inventory for SNEC Facility Components & Systems 4-55 i 4-21 Composite Results- Concrete Core Bore Sample SX862950119 4-65 I / l < 4-22 Core Bore Samples Taken to Determine Neutron Activation 4-66 Q) v
1 l Sixt:n Nucle:r Exp:riment:1 Cnrpnratinn (SNEC) ! Site Characterization Report ! A Table Title Page 4-23 Metal Activation Samples From SNEC Reactor Vessel Region 4-66 l 4 24 Reactor Vessel Curie Estimate - Calculated Values 4-68 ( 4-25 Radionuclide Composition of Safety injection and Reactor Vessel Sample l Materials 4-69 4-26 Summary Table of Contaminants Detected in SNEC Sampling Media 4-75 l 4-27 Ground Water Station Results (Activity in pCi/ liter) 4-76 4-28 Ground Water Station Results (Activity in pCl/ liter) 4-78 l 4-29 Aquatic Sediment Station Results (Activity in pCi/g - Dry) 4-79 4-30 Potable Ground Water Station Results (Activity in pCilliter) 4-80 4-31 CV Pipe Tunnel Results (Activity in pCilliter) 4-80 4-32 General Sampling Analysis Information 4-81 4 33 ERL Source / Sample Activity Tracking Sheet - Typical ERL Detection Limits 4-81 & 4-82 I 4-34 Total Lead Content of Surface Coatings and Resin Materials 4-87 4-35 Asbestos Sampling Requirements Listed in SCP and Assigned Results 4-89 4-36 Asbestos Bulk Sample Results 4-90 & 4-91 4-37 Composite Surface Samples From SNEC CV (CV Areas 1,2,3,4 & 6) 4-91 4-38 TCLP Extraction / Metals Analysis Results 4-92 4-39 Area 1, (5 mR/h General Area) 4 94 4-40 Area 2, ("11 mR/h General Area) 4-95 4-41 Area 3 (3 mR/h General Area) 4-96 4-42 Area 4 (0.2 mR/h General Area) 4-97 4-43 Area 5 (0.01 mR/h General Area) 4-98 4-44 Area 6 (~200 mR/h General Area) 4-99 4-45 Area 7 (0.017 mR/h General Area Around CV Outside) 4-100 4-46 Area 8 (0.014 mR/h General Area Average) 4-101 I l / 'N 4-47 Area 1, Alpha Surface Contamination Results 4 102 l vi
I Sixtsn Nucleir ExperimIntal Carptratien (SNEC) ) Site Characterization Report Table Title Page J 4-48 Area 2, Alpha Surface Contamination Results 4-102 4-49 Area 3, Alpha Surface Contamination Results 4-103 4-50 Area 4, Alpha Surface Contamination Results 4-103 1 4-51 Area 6, Alpha Surface Contamination Results 4-104 4-52 Air Particulate Station Results (Activity in pCl/MD 4-105 to 4-110 l l 4-53 Counting Results, On-site Core Bore Locations Inside CV 4-111 to 4-119 ) 5-1 Typical Health Physics Instrumentation 5-7 l l 5-2 SNEC Hot Spot Identification Log 5-8 )
Saxt:n Nucleir Experim:ntal Crrportti:n (SNEC) Site Characterization Report /^ r List of Figures Figure Title Page 2-1 SNEC Station Location of Environmental Samphng Points 2-10 22 1995 SNEC Facility Post-Soil Removal Configuration 2-11 2-3 SNEC Facility Area Locations Rotary Bridge Crane, Sectional View (Looking West) 2-12 2-4 SNEC Facility Area Locations Containment Vessel, Sectional View (Looking North) 2-13 4-1 SNEC Facility Core Bore Locations Outside of Containment 4-33 4-2 SNEC Facility Area Map and Core Bore Sample Locations - Containment Vessel-Areas by Quadrant - Plan Above Operaty Floor Elev. 765'-8" 4-34 4-3 SNEC Facility Area Map and Core Bore Sample Locations - Containment Vessel-l Areas by Quadrant - Plan Above Operating Floor Elev 781'-4* 4-35 l
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4-4 SNEC Facility Area Map and Core Bore Sample Locations - Containment Vessel- 1 1 Areas by Quadrant - Plan Above Operating Floor Elev 795'-2* 4-36 j ( 4-5 Containment Vessel-Core Bore Sample Locations - Plan View Above Operating Floor Elev 812'-0* 4-37 4-6 SNEC TLD Strings - Extemal to Vessel 4-59 4-7 SNEC Facility RV Exposure Rate - Intemal to RV in Feet Above & Below Core Centerline 4-60 4-8 SNEC Reactor Vessel- Highest Contact Measurements Core Centerline 4-61 j 4-9 Core Bore Sample SX822950037 (Area 2) ~8'-3' East of Hot Leg, Above SW Ops Water Line 4-63 4-10 Reactor Cavity High Neutron Flux Region (AREA 6), Core Bore Sample SX861950119 4-64 5-1 Average Contamination Levels and General Area Exposure Rates, Containment Vessel, Sectional View (Looking North) 5-10 5-2 Average Contamination Levels and General Area Exposure Rates - Plan View Above Operating Floor Elev 765' 8" Containment Vessel (CV) 5-11 a siii
i Saxten NuclIar Experimentil Ccrp2 ration (SNEC) Site Characterization Report b 0 Figure Title Page 5-3 Average Contamination Levels and General Area Exposure Rates- Plan View Above Operating Floor Elev 781'-4" Containment Vessel (CV) 5-12 5-4 Average Contamination Levels and General Area Exposure Rates - Plan View Above Operating Floor Elev 795'-2* 5-13 5-5 Average Contamination Levels and General Area Exposure Rates - Plan View Above l Operating Floor Elev 812*-0 5-14 l l 1 l l l 1x
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Saxt:n Nucirr Experiment:1 Ccrpcration (SNEC) ; Site Characterization Report j j I~
1.0 INTRODUCTION
AND OVERVIEW l This report entitled " Radiological and Environmental Data Supporting the Development of the SNEC Decommissioning Plan", has been prepared from data collected as directed by the Saxton Nuclear Experimental Corporation (SNEC) Facility, Site I Characterization Plan (SCP, Reference 1) and other related procedures and
- documents. The scope of this data collection process extends over areas of the facility j that became intemally or extemally contaminated or activated during the SNEC facdity's j operating history.
1.1 PURPOSE This document provides data summaries and interpretations , and defines methodologies by which areas, structures, systems and the surrounding soil at the SNEC facility were radiologically and environmentally characterized. This data summary is presented for planning purposes, and will be used to determine the effective and l j- appropriate decontamination and dismantling techniques and activities supporting plant f-j . decommissioning and dismantlement. This data is also necessary for planning i radioactive material disposal, assessing potential hazards during decommissioning and decontamination work, determining ALARA controls, and accurately scheduling and ! estimating the cost of overall program objectives. 1 i I 1.2 SCOPE- ! This report is intended for use by SNEC personnel, GPU Nuclear Corporation (GPU I Nuclear) personnel, contractors and subcontractors, and other organizations selected by GPU Nuclear as having a viable interest in fulfilling decommissioning and
- dismantlement objectives. This report includes the SNEC SCP (Appendix A), and l presents the results of the data collection process as outlined in the SNEC SCP. The i scope of this document includes the following charactenzation data:
i
- 1. Structural Characterization (Section 4.1);
1-1 f i 4
.- p y p- y .e.w-, ,-
l S xt:n NucleIr Experim:ntal Ccrpar;tinn (SNEC) l Site Characterization Report j [ 1 V - Locations, areas, and activity levels of structural surfaces (e.g. concrete, structural steel, decking materials, etc.). I 1
- Radioactive materials retained in porous materials.
1
- Depth and activity levels of contaminant penetration into porous or cracked surfaces (e.g. depth of contamination in concrete floors, wa!!s, and ceilings).
- 2. System and Component Characterization (Section 4.2);
- Location, surface areas and activity levels in piping systems and components.
- Location, dimensions, volume and activity levels of contaminated equipment,
- ducts, fixtures, etc.
- 3. Reactor Vessel Activation Characterization (Section 4.3);
- Calculation and confirmation of activity levels induced by activation in the
( k reactor vessel, associated components, and adjacent structures. 1
- 4. Environmental and Hazardous Materials Characterization (Section 4.4);
- Waste classifications of contaminated materials.
- 5. Soil Characterization (Section 4.4.2);
- Location, volume, and activity levels of contaminated soil.
- 6. Health Physics Surveys Supporting SNEC Site Characterization (Section 5.0).
- 7. QAIQC Additions, and or modifications to specific survey and or sampling requirements listed in the SNEC SCP were made with the concurrence of the SNEC Project Consultant, the SNEC facility Site Supervisor and the Manager Decommissioning Projects.
1-2 l l
S*xton Nuclear Experim:ntal Ccrporation (SNEC) Site Characterization Report 1.3 PROGRAM STRUCTURE 4 SNEC facility site characterization was conducted under the guidance of the SNEC Site Characterization Plan (SCP, Reference 1). This plan was originally developed by and in cooperation with TLG Services Incorporated'. Upon completion of the initial draft of the SCP, a fully reviewed, approved and proceduralized version was issued. The SNEC Site Characterization Report, is divided into seven (7) sections. They are the introduction, Historical Information, Data Presentation, Characterization Results, Health Physics Surveys Supporting SNEC Site Characterization, QA/QC Requirements and References. The Characterization Results section is further divided into Structural Characterization, System and Component Characterization, Activation of SNEC Reactor Vessel and Surrounding Structures, Environmental and Hazardous Materials Characterization and Data Tables. In addition, a photographic series has been included in Appendix B. (
\ Whenever possible, tables and figures are included within the section where they were discussed. However, several large tables have been placed in Section 4.5.
1.4 REGULATORY CRITERIA FOR UNRESTRICTED RELEASE The Nuclear Regulatory Commission (NRC), as the licensing authority for the SNEC facility, has the authority to define requirements for releasing materials for unrestricted use. The NRC has published Regulatory Guide 1.86, which details acceptable surface contamination levels below which a radioactive material license may be terminated. This is a guide, and is subject to interpretation contingent upon NRC approval on a case-by-case basis. NRC Regulatory Guide (RG) 1.86 Criteria, shown in Appendix C, is assumed to contain reasonable criteria for categorization of surface contamination on plant structures and within plant systems. When soil contamination is detected, the dose rate at 1 meter will be determined and compared to the 5 uR/hr level suggested by NUREG-0586, " Final Generic Environmental impact Statement on Decommissioning of 1 TLG Services, Bridgewater, CT. 1-3
SIxt:n Nucletr Experim:ntal Corp:ratian (SNEC) Site Characterization Report Nuclear Facilities". Determination of the final release criteria is not intended to be a part of this program. Rather, this program provided information that may be utilized later along with RG 1.86 or NUREG-0586, to separate survey data into three (3) easily defined categories: (1) Undetectable (<MDA or <MDL) (2) Detectable, but less than RG 1.86 or NUREG-0586 criteria (3) Greater than RG 1.86 or NUREG-0586 criteria Final criteria for unrestricted release will be specified at a later date. 1.5 ORGANIZATION AND RESPONSIBILITIES > The objectives of the SNEC SCP were achieved using implementing documents such as SNEC Work Instructions (SWI), SNEC Procedures or from other guicance provided by Standing Order Memorandums (SOM) and GPU Nuclear procedures. In addition to (D i these documents, SNEC facility management personnel provided "on the spot" clarification and interpretation of site characterization issues and data collection requirements, as needed. A listing of references is included in Section 7.0. To ensure that all facets of the plan were property implemented, the site characterization process required the combined efforts and expertise of various personnel and other organizations within GPU Nuclear. This effort required extensive coordination and planning by managers assigned specific responsibilities. The SNEC Decommissioning Plan (Reference 2), details their positions and briefly describes these responsibilities. In addition, Nuclear Safety Assessment personnel provided inspection, observation and valuable input during the implementation of SCP activities. 1.6 CHARACTERIZATION ACTIVITY CONTROL Characterization Activities were controlled as follows: (1) Personnel performing site characterization work met all applicable requirements of I
, existing SNEC training and qualification procedures and programs.
( 1-4 1
, SSxtan Nucl;&r Experim:ntal Ctrp: ration (SNEC)
Site Characterization Report I (2) Appropriate Health Physics procedures and work instructions were used to obtain and record all applicable radiological controls data. 3 (3) Environmental data were co!!ected under appropriate Environmental Controls i procedures. 1 (4) Contractors used to obtain concrete core bore sections were fullyqualified through l previous experience and/or training. Contractors met all SNEC or GPU Nuclear required training objectives. (5) Off-site contractors used for sample analysis were pre-approved by GPU Nuclear through a quality assurance program. ] (6) When necessary, data reduction and analysis was performed under the SNEC Technical Evaluations procedure (or equivalent) (Reference 35), which requires that a two (2) party calculation / verification process be used to evaluate data. 2 Industry standard computer codes e.g. Microshield and Lotus 1-2-3 were utilized ) as needed for data evaluation, organization and presentation.
^
(d l l 2 Microshield is a product of Grove Engineering, Inc., Rockville, MD. 3 Lotus 1-2-3 is a product of Lotus Development Corp., Inc., Cambridge, MA. O 1-5
Saxt:n Nucirr Experimental Csrptrati:n (SNEC) Site Characterization Report f\ 2.0 HISTORICAL INFORMATION 2.1 SITE DESCRIPTION The Saxton Nuclear Experimental Corporation (SNEC) facility, is a deactivated pressurized water reactor (PWR), which was licensed to operate at 23.5 megawatt thermal (23.5 MWTh). It is owned by the Saxton Nuclear Experimental Corporation and supported by GPU Nuclear Corporation. The SNEC reactor facility is maintained under a Title 10 Part 50 License and associated Technical Specifications. In 1972 the license was amended to possess but not operate the SNEC reactor. The license expires on February 11,2000 or upon expiration of the SNEC corporate charter, whichever occurs first. I I The facility was built from 1960 to 1962 and operatea from 1962 to 1972 primarily as a research and training reactor. After shutdown in 1972, the facility was placed in a I condition equivalent to a status later defined by the NRC as SAFSTOR. Since then, A it has been maintained in a monitored condition. The fuel was removed from the j b Containment Vessel (CV) in 1972 and shipped to the Atomic Energy Commission ! (AEC) facility at Savannah River, SC., who remain as owner of the fuel. As a result, neither SNEC or GPU Nuclear have any responsibility relative to the spent fuel from the SNEC facility. In addition, the control rod blades and the superheated steam test loop were shipped offsite. Following fuel removal, equipment, tanks, and piping located outside the CV were removed. The buildings and structures that supported l reactor operations were partially decontaminated from 1972 through 1974. After the formation of the GPU Nuclear Corporation in 1980, SNEC formed an agreement with GPU Nuclear to use GPU Nuclear and its resources to maintain, repair, modify, or dismantle SNEC facilities as may be required. Both SNEC and GPU Nuclear are subsidiaries of the same parent company, General Public Utilities Corporation, (GPU). While SNEC remains the owner of the facility, a license amendment has been submitted to designate GPU Nuclear as a co-license holder. GPU Nuclear will then have responsibility to comply with the license and technical 21
l Sext:n Nucle:r Experiment:1 Corp:rttinn (SNEC) Site Characterization Report O specifications. GPU. nuclear will cany out the SNEC facility decommissioning on f behalf of the site owner, SNEC. l The radiological condition of the facility following shutdown was documented in a report titled " Decommissioned Status of the SNEC Reactor Facility" forwarded to the l United States Nuclear Regulatory Commission (USNRC) on February 20,1975. i i j Decontamination and removal of reactor support structures and buildings was performed in 1987,1988, and 1989, in preparation for demolition of these structures. f This inc'uded the decontamination of the Control and Auxiliary Building, the i Radioactive Waste Disposal Facility, yard pipe tunnel, and the Filled Drum Storage Bunker, and the removal of the Refueling Water Storage Tank. A comprehensive final release survey was conducted from October,1988 to June,1989, to verify that residual contamination was within USNRC guidelines for unrestricted use. Details of the decontamination activities and final survey are provided in the report: " Final O Release Survey of Reactor Support Buildings" (Reference 6). These buildings were demolished in 1992 after acceptance of the final release survey by the USNRC,. In November 1994, the SNEC Soil Remediation Project was completed. This was a comprehensive project involving soil monitoring; sampling, excavation, packaging and shipment of slightly contaminated site soil. This program successfully reduced radioactive soil contamination levels below NRC current and presently proposed levels required to meet site cleanup criteria for unrestricted use.
'2.1.1 Site Layout The site is located about 100 miles east of Pittsburgh and 90 miles west of Harrisburg in the Allegheny Mountains, three fourths of a mile north of the Borough of Saxton in Liberty Township., Bedford County, Pennsylvania. The site is on the >
north side of Pennsylvania Route 913,17 miles south of U.S. Route 22, and about 15 miles north of the Breezewood Interchange of the Pennsylvania Tumpike. Figure ( 2-1, identifies the location of the site relative to the landmarks of the local area. 2-2
i 2 Sixten Nuclear Experimintal Corporatien (SNEC) Site Characterization Report 'O i V
- The SNEC facility was built on the east side of and adjacent to the Saxton Steam Generating Station (SSGS) of Pennsylvania Electric Company. This station was
- located on the east bank of the Raystown Branch of the Juniata River as shown on property map, Figure 2-2. The property comprises approximately 150 acres. The f
SSGS was demolished in 1975.
- The SNEC facility site comprises 1.1 acres containing the yard and buildings from the nuclear plant. An additional 9.6 acre area is fenced in around the electrical 1 switchyard and buildings still in use by Penelec (Pennsylvania Electric Company, a subsidiary of GPU). This area adjoins the Raystown Branch of the Juniata River, f Portions of the SNEC facility site as well as a portion cf the Penelec area and the surrounding uncontrolled lands are in the 100-year floodplain of the Raystown Branch and Shoup Run. Additional large areas of low lying fields and scrub tree growth surround the site.
2.1.2 Station Structures The only remaining SNEC facility buildings and structures are the Containment Vessel (CV), the concrete shield walllocated around the NW and NE quadrant of the CV, the tunnel sections that are immediately adjacent to the outer circumference of the CV and portions of the septic system, weirs, and associated underground discharge piping. Concrete barrier walls have been installed to isolate the open ends of the tunnel that were connected to the Control & Auxiliary Building, the Radioactive Waste Disposal Facility, and the Steam Plant. Portions of the Steam Plant Tunnel still exist downstream of where it is blocked off. This area will also be verified to meet unrestricted release criteria following dismantlement. 2.1.3 Facility Operational History Operational history of plant systems and support buildings and facilities is an important component of determining radiological status of the facility. The following
\
2-3
l S xt:n Nucirr Experim:ntil Ccrpor:ti:n (SNEC) Site Characterization Report is a chronology of major operational events at the Saxton Nuclear Experimental Corporation since start of operation in 1962. Event Date l Construction Authorization.. ... . . . . .. ... .. ... ... .. . .. .... . .... February 11,1960 Initial License issued...... . . . . . . . . . . . . . . . . . . . . . .......... .... . ..... November 15,1961 Initial Criticality........ . . . . . . . . . . ........ . . . . . . . . . . . . . . . ... .. April 13,1962 1 First Electricity Generated... . . . ..... . ... . . . . . . . . . .. November 16,1962 Liquid Spill Outside St Pump House (~1 gal.-10uCi's) November 26,1968 Storage Well Leaks, Possibly Resulting In Extensive Contamination Of Internal CV Concrete Structures.. . .1968 Through Layup i of SW
/ Experiments With Mixed Oxides Fuel, And Fuel Cladding q \ Intentionally Failed" (Last Fuel, Cycle).. . . . . .. .. .. . . . . .. December 11,1969 i May 2,1972 Unclanned Gas Releases I 7.32 Curies.... . ....-... .. . . . . . . . . . . . . . . . . . . . . .. . . . .. .....May 14,1970 ,
0.034 Curies.. . ... .. ........ ... . . .. . . . . . . . . . . . . . . . . . . . . . . . . .... ... August 26,1970 80.2 Curies... . . . . . . . . . . . . . . . . . . . . . . . . . . . ... . .. . .. . .. ... ... November 29,1971 ) 1 19.7 Curies..... . . . .... . . . . . . . . . .. . .... . .. December 15,1971 l l Final Shutdown..... . . . .. .. . .. . ..May 1,1972 Nuclear Fuel and Other Removable "Special Nuclear Materials" Shipped Off Site.. . . . . . . .. . . . . . . . . . . . . . . .. . .. July - November, 1972 : 1 By-Product Material Removed From Site (With Exception of Materialin Exclusion Areas)... .. . . . . . . . . . . November,1972 - ( Early 1974 ( 2-4
i S:xton Nuclerr Experimintal Ccrp: ration (SNEC) Site Characterization Report j Event Date i j Facility Placed in A " Mothball" Condition..... ........ .... . .. .. . ........ .... February 1975 4 i Groundwater Removed From RWDF and Yard Pipe Tunnel. ..... . .. .. . . ... . .. ... ...... .... .... ..... ... .. .... . ... 1. ate 1986 - Early , 1987 , 1
- Decontamination of Control & Auxiliary Building '
l RWDF, RWST, and Yard Pipe Tunnel.... .... . ... .. ... ... . .. .. 1987 & 1988
- Final Release Survey of C&A Building, i RWDF, RWST, and Yard Pipe Tunnel..... .. .. .. . . . ...... . . .. October 1988 -
- June 1989 Pennsylvania State University Soil Characterization.... .... ... .. .. . . . . . . . . . ... . . . . . .. . .. ..... December,1988 &
January,1989 EG&G Measurements Aerial Suntey.. .... .. .. . . . .... .. .. . . . . . July 1989 (' Comprehensive Radiological Survey of CV (Scoping Survey).. .... .
. . . . . . . . . . . . ... . . . . . . . . . . .1991 l
Demolition of C&A Building, RWDF, RWST I and Yard Pipe Tunnel.. . . . . . . .. . . ........... 1992 Soil Remediation Project.. .. ... ... . . . . . . . . . . . . . . . . . .... June - November 1994 Site Characterization of CV & Remaining Facilities.... . ... .1995 s 2-5
. .- - =
l S*xtan Nucirr Experimintal Corp: ration (SNEC) Site Characterization Report rs iv) 2.1.4 Historical Radiological Survey Data (1971) Radiological surveys were routinely performed during SNEC facility operation. The following information was collected September 13-17,1971, and shows typical operational radiological conditions that existed within the containment vessel (CV). l Table 2-1, Historical Radiological Conditions During Operation (1971) RADIATION LEVELS LOCATIONS (mR/h) DPM/100 cm' l Operation Level CV (AREA 4) 1 ~280 Storage Well Bridge (AREA 4) I ~160 J Auxiliary Level (AREA 3) <1 ~110 Auxiliary Level (AREA 3) <1 ~210 C. C. Pumps (AREA 3) <2 < 100 Storage Well Pumps (AREA 3) <2 ~ 250 Letdown Vahe (LRC-21V, AREA 3) 200 ~2,550 CV Basement G.A. (AREA 1) 10 ~420 l ( Discharge Tank (AREA 1) 260 - 8,450 Shutdown Cooling Pumps (AREA 1) 9 ~ 270 Discharge Tank Pumps (AREA 1) 10 -450 Control Rod Room (AREA 1) 35 ~270 Lower Primary Level (AREA 2) 80 ~1,130 l Regen. Heat Exchanger (AREA 2) 1,750 ~790 Non-Regen. Heat Exchanger (AREA 2) 75 ~760 Upper Primary Level (AREA 2) 100 ~ 1,130 Main Coolant Pump (AREA 2) 325 ~1,450 Pressurizer Bottom (AREA 2) 110 - 110 Pressurizer Top (AREA 2) 5 ~1,900 Valve HIC-27 (AREA 2) 100 ~ 1,170 Steam Generator Bottom (AREA 2) 210 ~300 Reactor Grating Level (AREA 6) 250 ~ 107,640 Reactor Head Level (AREA 6) 260 ~ 100 4 l G 2-6
7 __ . )
, Sixt*n Nucl:ar ExperimIntal Corp:r..tlon (SNEC)
Site Characterization Report 2.2 REMAINING SITE STRUCTURES 2.2.1 Area Designations 4 The remaining Containment Vessel facilities consist of eight principle structures and locations (see Figures 2-3 & 2-4, end of this section), designated as the following:
- AREA 1 - S,outhwest, Southeast, and Northwest Quadrants of Contairment Vessel
$ between elevations 765'-8" (concrete floor) and El. 777'-8" (concrete 1 I ceiling) in southwest quadrant, El. 779'-0" (steel platform) in southeast quadrant, and El. 775'- 2" (concrete ceiling of the " Rod Room") in northwest quadrant. The area a!so includes the three and one half foot deep sump located in the floor, and a 4 foot wide concrete ledge (El. 768'-3") extending around the circumference of the area. AREA 2 - The Southwest quadrant of the Containment Vessel between elevations 779'-8" (Concrete Floor) and 814'-6" (Concrete Ceiling). Three (3) steel ( platforms are installed. One platform, El. 789'-4", is located beneath the Steam Generator, Pressurizer and Primary Coolant Pump. The second, El. 795'-2", essentially extends over the entire area of the quadrant and the third, El. 807'-0"is located around the upper heads of the Steam Generator and Pressurizer. AREA 3 - The Southeast quadrant of the Containment Vessel between elevations 781'- 4 (steel platform) and 810'-0" (concrete ceiling). One additional steel platform is installed at elevation 795'-2". Both steel platforms extend over the entire area of the quadrant, each containing an 8 foot l by 8 foot open hatch. There is a similar opening in the concrete ceiling. AREA 4 - The concrete operating floor of the containment vessel (CV) (El. 812'- 0"), and surfaces up to the top of the CV dome. The area includes: 1) three access hatches (equipment, personnel, and escape); 2) the concrete walls and platform located in the southwest quadrant 2-7
S:xten NucleIr Experimintil Ccrpsriti:n (SNEC) Site Characterization Report containing the steam generator, pressurizer and reactor coolant pump;
- 3) the polar bridge crane; 4) a steel platform (El. 818'- 0"); and 5) the ,
movable bridge over the reactor vessel and fuel storage area. AREA 5 - The below grade concrete tunnel which wraps around the outer. circumference of the containment vessel between azimuths 35 degrees and 270 degrees. AREA 6 - This area is comprised of the Northwest and Northeast quadrants (reactor compartment and storage well, respectively) of the Containment Vessel between elevations 765'-8" (concrete floor) and 807'-0" (concrete ceiling). All concrete surfaces of this area, including the surrounding concrete walls are lined with a Series- 300, four-coat, catalyzed phenolic protective lining. The reactor compartment contains O the reactor vessel, associated piping, and the reactor vessel "can", which was used as a barrier to keep the storage well water off the piping and reactor vessel surfaces. There are two aluminum platforms installed in the reactor compartment. One removable platform at elevation 793'- 2", and a second platform at elevation 800'-6". The storage well contains
-_the movable fuel storage racks (up/down direction), three demineralizer vessels (Storage Well, Boric Acid, and Purification), which are mounted at elevation 776', and a concrete pad for storage of the fuel shipping cask (above El. 781'-4").
AREA 7 - The outside of the steelliner of the Containment Vessel (CV) extending from Grade Level (El. 811') to the top of the CV Dome. This area also includes the concrete shield structure located around the circumference of the CV between azimuths 270' and 35' (El. 811' to about 822'). AREA 8 - This area includes the tanks and enclosures which received the sanitary discharge from the C&A Building, processed the sewage, and 2-8
-1 Srxton Nucl:ar Experimental Corporation (SNEC)
Site Characterization Report O V discharged the effluent to the Raystown Branch of the Juniata River. The system consisted of: 1) two concrete septic tanks installed in series, Tank "A" and Tank "B"; 2) a concrete pump well contained within a concrete block pump house (no above grade structures remaining); 3) a concrete Chlorine Contact Tank / Weir unit and a storm drain system. 2.2.2 Area Characterization Strategy i All of the areas designated in section 2.2.1 above, were evaluated with respect to : their potential hazard. Comprehensive radiological surveys, equipment and material ! inventories, component inspections and hazardous materials analysis were performed which included some destructive sample collection and removal. l 1 Comprehensive characterization activities were not considered important for some materials and/or locations were early removal possibilities existed, and where these 1 i materials and items retain little radiological or hazardous materials. Other areas in section 2.2.1 above were more extensively characterized to provide detailed l information that will be used to " map-out" a strategy for safe removal and disposal. l This " graded" approach to characterization is appropriate and has provided the best ) balance between the time available for work activities, ALARA considerations and reducing the expense of conducting the characterization effort and associated ! activities. l 4 4 4 O 2-9 m-y
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' S xten Nuclear Experlm:ntal C:rptrati:n (SNEC) 4 Site Characterization Report
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Saxt:n Nucirr Experimental Ctrpor;tirn (SNEC) Site Characterization Report
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, Snxton Nucle r Experimintal Ccrper:tien (SNEC)
Site Characterization Report ( 3.0 CHARACTERIZATION DOCUMENTATION AND REPORT 3.1 GENERAL DOCUMENTATION Proper documentation of every aspect of the characterization program included accurate mapping of the survey data, with its relationship to the site. In addition, instrument measurements and analytical results were documented to include the i following items as applicable.; (1) Location of measurements or samples taken (diagram, map, etc.).
]
- (2) Date or dates measurements or samples were taken or collected. l (3) The measured concentration of the specific nuclides in bulk samples as l applicable.
(4) Measurements of contamination were reported as removable alpha, l j beta / gamma activity levels (dpm/100 cm2 , uCi/ unit area, etc.). (5) The analytical error was reported for all laboratory analysis. l ! Note: Outside vendor analysis results are included in Appendix D for all 1 samples collected. j
- (6) Name of surveyor (s), sampler and analyst, if applicable.
- (7) Analysis date.
i (8) Instrument Manufacturer, Model No., Serial No. and reference to calibration i i data. l (9) Reviewer's name. Other data and information is documented according to j approved SNEC procedures referenced in Section 7.0. i i f f 3-1
S*.xt:n Nuclear Experimint:1 Ctrp:rati n (SNEC) Site Characterization Report 3.2 DATA PRESENTATION Data reduction and analysis results calculated from information collected during the Site Characterization Program have been presented in the appropriate segment of section 4.0 and the Appendices. The specific areas, data acquisition methods, sampling techniques, and calculational methods employed are described in some detail in the SNEC facility SCP, Section 4 (Appendix B),' or are presented in detail within this report or references listed in Section 7.0. 3.3 RADIOLOGICAL DATA The sensitivity of the measuring and sampling techniques used depends on the purpose of the survey. Direct measurements taken in " unbiased" surveys require the use of techniques having a Minimum Detection Level (MDL) less than the maximum permissible levels identified in Table 1 of USNRC Reg. Guide'1.86 or more recent regulatory documents. When background radiation levels precluded practical achievement of desired sensitivity, attemate methods of establishing contamination or radiation levels were used. Measurement methodology is described for each technique used. Some data collection efforts outlined or required by the SNEC facility SCP were deferred or performed under a reduced scope because of the difficulties in obtaining the data without simultaneously performing some dismantlement operations. In other cases, characterization efforts may have been postponed or deferred to accommodate weather conditions, decommissioning plans or ALARA considerations. 3.3.1 Unbiased Survey Results For purposes of this document, unbiased surveys are surveys that do not preferentially select measurement points which have a higher probability of contamination than other locations not selected. The SNEC Facility Site Characterization Plan (SCP), has specified AREA 7 as a candidate for unbiased surveys. However, a limited number of survey points were indicated for AREA 7 in the SNEC SCP. A more applicable use of unbiased survey methods is presented O 3-2
S xt:n Nucirr Experimental Corporation (SNEC) Site Characterization Report ( within soil release surveys conducted externally but in the vicinity of the SNEC CV I (AREA 7). No locations within the SNEC Facility CV were chosen for the application . i of unbiased survey measurement methodology. l 3.3.2 Biased Survey Results i Most of the survey data collected at the SNEC Facility is collected using a biased survey technique. This information is presented in a tabular presentation, employing , a graphic depiction of the area and location on maps or diagrams. These locations I are clearly marked and identified. i 3.4 STRUCTURAL DATA
- Structural data is compiled in tables within the applicable sections. Concrete 1 i
surfaces, steel and aluminum grating and platforms, comprise the majority of this information. Calculated information is presented along with the reasons and I ( assumptions used in determining the estimated values. ( 3.5 SYSTEMS AND COMPONENT DATA Systems and component survey data are compiled in tables and summarized for ease of review. The tables include the gross total and removable surface contamination level estimates. In addition, nuclides detected during sampling I evolution's are listed along with their abundance normalized to one (1) curie. Actual survey data is presented on SNEC facility radiological surveys referenced in section 7.0 and are summarized within this report. 3.6 REACTOR VESSEL AND ADJACENT STRUCTURES - ACTIVATION DATA The results of the activation verification analysis are presented. Confirmatory radiological survey measurements performed by SNEC facility Radiological Controls and GPU Nuclear engineering personnel are included to show how the calculated radiation measurements compare to actual measurements. In addition, thermoluminescent dosimeter (TLD), data is presented and compared to calculated
/ values. The expected waste volume produced by the activation of the reactor vessel 3-3
l Saxtsn Nucinr Experimental Crrpor;tien (SNEC) Site Characterization Report and surrounding support materials are also presented. The benchmarking process used is described and summarized. 3.7 ENVIRONMENTAL AND HAZARDOUS MATERIALS DATA The environmental samples and information acquired under the SNEC Site Characterization Plan include the following: Location of susoected asbestos containina materials (paint. insulation. etc.). l Results include type of asbestos, presence of non-asbestos 6bers (%), presence of non-hbrous materials etc. Location and concentration of lead in oainted surfaces and other materials l Results reportedin ppm, mg/kg etc. , l p - Location of bulk lead materials (shieldina etc.) i Results reported as weight (Ibs) and volume (ft').
- Location of PCB's and substances identified for TCLP. inoraanic analysis.
Results reported in ppm, mg/kg or mg/l as appropriate. i Test parameter sited with detection limit.
- Location of and cuantities of mercury and cadmium within the CV.
Results reportedin volume (cc's, ft', etc.), and or weight (Ibs, gms.,etc.) \ 3-4
S xt:n Nucleir Experimental C:rporatirn (SNEC) Site Characterization Repop ( ( 4.0 CHARACTERIZATION RESULTS
SUMMARY
Approximately 93% of the SNEC facility radionuclide inventory is retained within the steel structures of the reactor vessel. The remaining radionuclides (~7% of the total) are located within/on components, system piping, concrete and facility structural surfaces. The estimate of record for the SNEC facility is presented in the following 1 table. This listing represents greater than 95% of the total calculated facility radionuclide inventory (nuclides >1% of total are listed in bold text). Table 4-1, Radionuclide Inventory For The SNEC Facility ; i Radionuclide Curies % Of Total I Ag-108m 3.73E-02 0.0024 Am-241 2.05E+00 0.1316
- Am-243 3.79E-03 0.0002
- Ag-39 2.34E-03 0.0002 Ce-144 1.81E-01 0.0116 l Cm-242 2.14E-02 0.0014 l Cm-243/Cm-244 7.14E-02 0 0046 Co-60 6.04E+02 38.7270
/" Cs-134 4.53E 02 0.0029 l
(' Cs-137 C-14 8.47E+00 7.37E+00 0.5 6 1 0.4727 Eu-152 5.47E-01 0.0351 Eu-154 1.89E-01 0.0121 l Eu-155 8.21E-02 0.0053 Fe-55 3.84 E+01 2.4606 H-3 4.15E+00 _ 0.2661 l I-129 2.10E-01 0.0135 Nb-94 1.16E+00 0.0745 Ni-59 7.26E+00 0.4659 Ni-63 8.56E+02 54.8803 Np-237 1.44E-01 0.0092 Pu-238 6.57E-01 0.0422 Pu-239/Pu-240 3.15E+00 0.2022 Pu-241 2.29E+01 1.4659 Pu-242 6.98E-03 0.0004 Ru-106 3.71E-01 0.0238 Sh-125 1.13E-01 0.0073 Sr-90 1.98E+00 0.1270 Tc-99 8.77E-02 0.0057 U-234 4.llE-02 0.0026 U-235 4.58E-03 0.0003 U-238 3.35E-02 0.0022
,v) Totals => ~1560 -100 %
41
Sixt n NucleIr ExperimIntil C:rp;rati:n (SNEC) Site Characterization Report l 2 in the following sections of this report, data collected from each area of the facility is ) examined. The methods used to determine the curie estimates for components, system piping and structural surfaces is presented. In all cases, radionuclides are assumed to be present at the detection limits provided by the analytical laboratory. . 4.1 STRUCTURAL CHARACTERIZATION
- 4.1.1 Data Needs )
Structural characterization results contained within this report, include a determination of the:
- - Amount of surface contamination (fixed & loose), on structural materials.
- - The distribution of radionuclides present on contaminated surfaces.
- The depth of contamination which penetrated porous surfaces.
4.1.2 Applicable Areas and Surfaces i O G Structural characterization included the interior and exterior surfaces of the Containment Vessel, the adjoining below grade tunnel, and portions of the underground septic system. A detailed listing of the structural locations examined is provided within APPENDIX A. Typical structural characterization subjects are presented below: Floors - areas of potential spills; areas of heavy traffic Walls - less than 2 meters high - potential for splashes from nearby equipment Walls - greater than 2 meters high - settling of dust and sprays Horizontal Surfaces - gravitational settling of dust and dirt Ceilina Surfaces - sprays or steam leaks 4.1.3 Analytical Methods The following describes the analytical techniques which were used to characterize existing SNEC structures, and the criteria under which these methods were implemented. v . 4-2
S:xtire Nuclear ExperimIntal Ctrp rati:n (SNEC) Site Characterization Report
\
A. Surface Contamination Determinations (fixed & loose, a-p-7): 1 Measurements of fixed and loose a-p-y radioactivity were performed on representative surfaces within each AREA and from each survey unit : specified in the SCP. ; Appropriate radiological instrumentation and methodology were employed. See Section 5.0 for types and characteristics of typicalinstrumentation used. Alpha analysis of smear samples was performed in accordance with existing i procedures. A minimum of one alpha measurement was made in each survey unit. i Fixed Contamination Fixed contamination is defined as the subsurface contamination and activation l products plus surface contamination that is tightly bound or adherent. Subsurface i m \ contamination and activation products may be evaluated from core bore analysis (V or some other destructive methodology. For purposes of this report, tightly bound + or adherent materials on the surface of the concrete is defined as the fixed l surface contamination component. ~ Loose Surface Contamination Loose surface contamination is that amount of contamination that can be easily removed by simple physical or mechanical actions. For the purposes of this report, smear contamination surveys measure this component. The removal efficiency for a tip!TI smear is assumed to be ~10% out will depend on surface and contaminant characteristics (Reference 14 and GPU Nuclear Radiological 2 Engineering / Instrumentation Information). A typical smear samples 100 cm of surface area. O. y/ 4-3
d
. Srxt:n Nucint Experimental Corporation (SNEC)
Site Characterization Report B. Contamination At Depth: Determinations of the amount of contamination penetrating porous structural f materials were performed in locations where significant surface contamination was found and there was a likelihood of contaminants penetrating the surface. j The following method was employed to determine the depth of contamination in i concrete surfaces. > l l l Core Drillina (core bores) I 1 The purpose of core drilling is to determine the extent 'of contamination or l activation of materials where the potential existed for contamination or activation to exist beneath the exposed surface of the structure. Core drilling was performed by utilizing a hollow-bit concrete coring drill, 3 inches in outside diameter, and drilling out the specified depth of material. Coring apparatus consisted of electrically driven, water lubricated, heavy duty drills attached to j masonry hole saws. The core bores removed were 2 3/4" in diameter. All V sampling was performed in accordance with the direction contained within SNEC 1 Work Instruction SWl-94-001 (Reference 56). C. Bulk Sampling and Analysis: Bulk Samples were collected and analyzed to determine the radioisotope concentrations and relative distributions of these isotopes in sediment, scrape samples, filter media and core bore sections. A representative selection of l sample types were analyzed to determine the presence of hard to detect radionuclides, and in some cases, were also analyzed to determine potential hazardous materials concentrations. Nuclide Identification High resolution gamma spectroscopy was used to identify gamma emitting isotopes on site. In addition, many samples were also analyzed at off-site facilities for typical 10 CFR 61 type radionuclides (shown on following page). A U 4-4
l l S::xtsn Nuclear ExperimInt:1 Ccrp:ritinn (SNEC) Site Characterization Report i 1 [ Ag-108m Am-241 C-14 Co-60 Cm-244 Cs-134 Cs-137 Eu-152 l Eu-154 Eu-155 Fe-55 Fe-59 l H-3 1-129 Nb-94 Ni-59 Ni-63 Pu-238 Pu-239 Pu-240 Pu-241 Sb-125 Sr-90 Tc-99 U-234 U-235 U-238 D. AdditionalClassifications: In addition to the above analytical requirements, the results of the structural evaluations will be used to classify surfaces according to the following two criteria:
- Less than or eaual to the aoolicable release criteria
- Greater than the aoolicable release criter;a k" 4.1.4 Selection of Measurement And Sampling Locations Sampling locations for biased surveys were selected from areas or locations that represented the highest contamination levels (whenever possible), with the following exceptions:
(1) Situations where contamination levels in a selected area were extremely low. No discernible levels above background were detected with standard measurement techniques. (2) Areas where general area contamination or radiation levels were extremely high (mrad smearable and/or high mR/h general area exposure rates). Therefore, maintaining personnel exposures ALARA was considered along with data collection requirements.
/'
k 4-5
Saxt:n Nucle r Experimental Carporati:n (SNEC) Site Characterization Report O V 4.1.5 Estimates of Structural Surface Area (ft') Surface contamination results were tabulated for various elevations and surfaces within the eight (8) areas previously designated at the SNEC facility. Estimates of surface area for major structural components are tabulated below. AREA 1 l 1 Location: LowerLevel of CV, Southwest, Southeast and Northwest Quadrants l SURFACE / MATERIALS ESTIMATED SURFACE AREA i Concrete Floor (El. 765'-2") 705 ft l 2 Concrete Ledge (El. 768'-3") 350 ft Exposed (Steehlined) Sump (El. 765'-2") 64 ft2 Concrete Ceiling (El. 775'-2") 733 ft2 j Concrete Walls (El. 765'-8" to 781'-4") 2543 ft' Steel Platform Ceiling (El. 781'-4") 370 ft: rh I I AREA 2 V Location: Middle Level of CV, Southwest Quadrant SURFACE / MATERIAL ESTIMATED SURFACE AREA Concrete Ficor Surface 475 ft2 l Concrete Ceiling Surface 475 ft2 Concrete Wall Surfaces 2843 ft2 2 l Steel Platform (El. 789'-4") 89 ft 2 Steel Platform (El. 795'-2") 325 ft 2 Steel Platform (El. 807'-0") 89 ft AREA 3 Location: Middle Level of CV, Southeast Quadrant SURFACE / MATERIAL ESTIMATED SURFACE AREA Steel Platform (El. 781'-4") 414 ft2 2 Steel Platform (Ei. 795'-2" 414 ft 2 Concrete Ceiling Surface 381 ft Concrete Wall Surface 2405 ft2 s 4-6
l a Sextrn Nuclicr Experimintal Ctrpcraticn (SNEC) Site Characterization Report fh i 1 AREA 4 i i Location: Operating Level of CV, and Above ! SURFACE / MATERIAL ESTIMATED SURFACE AREA Concrete Operating Floor (El. 812'-0") 1202 ft2 Concrete Platform (El. 818'-0") 669 ft2 Steel Platform (El. 818'-0") 75 ft* , 2 i Concrete Walls (El. 812'-818') 341 ft
- CV Steel Walls (El. 812'-818') 666 ft
I CV Steel Walls (El. 818'-845'-2") 4267 ft2 Movable Bridge (El. 812'-0") 28 ft: 2 Polar Bridge Crane (El. 845') 196 ft
]
) 2 CV Upper Dome (El. 845'-2" to 870'-2") 3926 ft Equipment Access Hatch (El. 814'-5") 33 ft2 Personnel Access Hatch (El. 823'-9-1/2") 71 ft* j Escape Access Hatch (El. 815'-6") 20 ft2 AREA 5 Location: Tunnel Outside of CV SURFACE /MATERIAc ESTIMATED SURFACE AREA Concrete Floor (El. approx. 803') 953 ft2 Concrete Wall (El. 803' to 811') 2014 ft2 Concrete Ceiling (El. 811') 953 ft2 Concrete Surface Above Grade Outside Tunnel 953 ft* f% O 4-7
l \ l l Saxt n Nucle:r Experimintal Ccrporatian (SNEC) Site Characterization Report V AREA 6 Location: Reactor Compartment and Storage WeII of CV SURFACE / MATERIAL ESTIMATED SURFACE AREA Concrete Floor Surface 217 ft2 Concrete Ceiling 574 ft2 Concrete Wall Surface 4025 ft' Platform (El. 793'-2") 206 ft' j Concrete Ledge (El. 768'-3") 818 ft' Concrete Ledge (El. 779'-8") 113 ft' l Platform (El. 800'-6") 234 ft' I l l AREA 7 l Location: Outer Walland Dome of CV 1 SURFACE / MATERIAL ESTIMATED SURFACE AREA 2 Steel Walls (< 2 meter high) (El. 811' to 817') 615 ft 2 Steel Wall (> 2 meter high) (El. 817' to 845') 4424 ft Steel Dome (El. 845' to 870') 3927 ft' 2 Concrete Shield Wall (F:orizontal Surface) 173 ft 2 Concrete Shield Wall 327 ft AREA 8 Location: Sanitary Sewage Treatment Facilities SURFACE / MATERIAL ESTIMATED SURFACE AREA 1000 gallon concrete Septic Tank "A" 173 ft: 600 gallon concrete Septic Tank "B" 107 ft: 1000 gallon concrete Pump Well 173 ft' Concrete block Pump House (below grade) 348 ft2 2 Concrete Chlorine contact Tank / Weir unit 192 ft 4.1.6 Unbiased and Blased Surveys Most structural surfaces within the SNEC CV retain some surface (loose or fixed), contamination. Radiological survey personnel routinely included areas of higher ! t 4-8
I i Sixton Nucl;;r ExperimIntal Carporati:n (SNEC) j Site Characterization Report b contamination levels whenever possible. Therefore, no unbiased surveys were performed for areas within the CV interior. Unbiased surveys were limited to locations extemal to the SNEC facility CV. 4.1.*/ Data Reporting and Analysis Structural surfaces that have been characterized were sorted by total and removable surface contamination results (i.e., as less than the applicable release criteria or greater than or equal to the applicable release criteria). Results of structural surveys ) were interpreted, analyzed and the average values and isotopic distributions of radionuclides on these surfaces were calculated from smear analysis results. 4.1.8 Surface Contamination Evaluations Several methodologies were used to determine the loose contamination components , on structural surfaces at the SNEC Facility. Smear surveys were used to evaluate the loose or removable surface contamination component. Direct frisk l p measurements (when possible), were used to determine fixed surface contamination levels. Smears Standard smear survey techniques were used to evaluate loose contamination on surfaces. One hundred square centimeters of sampled surface area per smear is the standard smear area sample size. Direct Surface Measurements Standard Geiger-Mueller, thin window, pancake type detectors were used to estimate the fixed surface contamination levels at the SNEC facility. Typical detection equipment of this type is a HP-210T probe attached tc an Eberline instrument model number E-140N. This unit reads out in counts per minute (CPM). Field measurement requirements specified a probe to surface distance of about 1/2". The probe is 2 sensitive to a surface area of approximately 15 cm r% 4-9
i j: [ Stxton Nucizr ExperimIntal Carpcraticn (SNEC)
- Site Characterization Report lo Special Surface Measurements Typical frisker probes respond to the beta radiation component predominately, in
! SNEC facility radionuclide combinations. For purposes of this report, beta radiation is !- assumed to be a surface phenomena since Cs-137 beta particles are stopped in l about a millimeter of concrete and Cs-137 is the dominant isotope found in surface contamination at the SNEC facility. Standard frisker probes, with and without 1/4" l-j thick Plexiglas covers over the sensitive area, held at fixed distances from the surface to be surveyed, were used to provide both background and surface contamination levels (in counts per minute (CPM)). With Plexiglas covers removed { ("open window" readings), the total response of the instrument to background and f surface contamination may be estimated. With Plexiglas covers installed (closed l j window), readings provide only the background gamma and x-ray radiation j response, since almost all beta radiation is shielded by the cover materiai. Then the i _ difference between an open and closed window frisker measurement is the [( estimated surface beta radiation component. Standard Health Physics HP-210T detector efficiencies for the nuclide mixtures present on SNEC facility structural j surfaces are established by the GPU Nuclear Radiological Instrument Department i as -10%. Using the frisker probe sensitive area of ~15 cm 2, and the detector i i efficiency, the total, fixed and loose surface activities were estimated for affected s j surfaces. j Surface Decontamination Experiment l Radiologically contaminated surfaces were hand wiped to remove the majority of f loose surface contamination present after initial open and closed window readings were taken using a pancake frisker type probe with a 1/4" Plexiglas cover. A second
- set of open and closed window readings were then taken to determine the remaining surface contamination levels after simple decontamination. This remaining surface I radiation component is a reasonable approximation of the " fixed" contamination component for surface deposited materials. These values do not include the 4-10
i
. S:xtan Nuclerr Experim:ntal Ccrporation (SNEC)
Site Characterization Report
/3 contamination that may have penetrated porous surfaces below about one millimeter, such as into concrete.
This approach provides an estimate of the total (loose and fixed), surface contamination present on SNEC facility structural materials. In addition, the results also give an indication of the fraction of surface contamination that may be easily removed with standard non-destructive cleaning techniques (hand wiping). 4.1.9 Concrete Surfaces-General Description The majority of concrete surfaces within the SNEC facility CV have been painted. In the storage well (AREA 6), many cracks are visible in the walls and floors, and the liner of the storage well is separating from the concrete (see photographic section in Appendix B). The walls in the storage well are believed to have been penetrated to a significant depth by storage well water containing soluble fission products. The majority of this contamination occurred when the Storage Well was filled to about the (O
%/
) 791' elevation during reactor operations and greater levels during fuel transfer evolutions. Core bore samples from below the operational water line have shown significant penetration of contaminants below the exposed surface area. Potentially contaminated concrete will be re-surveyed and re-evaluated as necessary during decommissioning. Any concrete found to be above USNRC approved release criteria will be removed and packaged as radioactive waste. This area is classified as affected and generally above release criteria.
In AREAS 1,2 and 3, low to moderate levels of contamination have been detected at depth. However, most of this concrete may be below release criteria with the exception of the floor in AREA 1 and the four and five foot thick concrete divider walls that separate AREAS 1, 2 and 3 from the Storage Well. The remaining concrete surface is only lightly contaminated near the surface with some locations already below Regulatory Guide 1.86 release limits. v 4-11
Saxtin Nuclerr Experim:ntti C rpsrati:n (SNEC) Site Characterization Report
\- In AREA 4, the surface of the concrete is lightly contaminated. However, some of the exposed surface area in this region of the CV is at or below applicable release limits.
The one exception is the shield blocks over the reactor cavity. Samples from these 20 ton concrete blocks have shown activation products at some depth in their < underside (towards the reactor). This is an affected area with some surfaces at or below applicable release limits. In AREA 5 (tunnel outside of CV), a limited number of concrete core bore samples have been taken, but these samples have shown almost no contamination above applicable release criteria and almost no depth penetration. However, most of the surface area in AREA 5 is currently under water. An off-site analysis by B&W has detected Cs-137 and Co-60, which are the only positively detected radionuclides found in tunnel water or sediment. The water is slightly radioactive but near release limits for radionuclides. The sediment retains higher levels of contamination (see
/S Table 4-2).
This structure will be re-surveyed during or prior to facility decommissioning, when i the water has been removed. AREA 5 is an affected area with the possibility of having significant portions of structural materials at or below release criteria. Table 4 2, Radionuclide Concentrations in CV Tunnel Water And Sediment Sample Number Cs-137 Concentration Co-60 Concentration SX856950167-SD (Liquid) 3.44E" uCi/ml 1.16E" uCi/ml SX856950167-SD (Solids) 2.94E" uCi/g 6.39E* uCi/g AREA 7 concrete is already below applicable release criteria, with the exception of one concrete pad. This is a lightly affected area. The AREA 8 sewer system, has not shown any contaminated concrete surfaces. These structures will be removed atmost entirely during or prior to decommissioning. 3) a 4 12
Sixt:n Nucle:r Experimental Corporation (SNEC) Site Characterization Report t O) (' There are a few small pipe sections that connect the sewer system components and one ten inch pipe that originally discharged effluent into the river (now closed off). Limited information exists conceming these pipe sections. During or prior to dismantlement, this system will be re-examined, removed or released as unaffected. l 4.1.10 Field Measurements Of Concrete Surfaces , The following table presents the results of measurements of concrete surfaces using
- the method described in Section 4.1.8. While these values are approximations, they l
. provide an indication of the degree of contamination that exists on concrete surfaces t
- throughout the SNEC facility containment vessel. In addition, they satisfy the requirements set forth in the SNEC Site Characterization Plan (Appendix A). In 2
l 1 Section 4.1.11, the same measurement method is used to estimate the degree of loose and fixed contamination present on metallic surfaces within the SNEC CV. Table 4-3, Direct Frisk Measurements Of Concrete Surfaces O
- Note: Assumes an HP-210T probe is sensitive to-15 cm2 of surface area.
Radiological sNEC CV Location Pre-Decon Post-Decon % 1mse % Fixed uCi/100cm 3 Survey No. Net CPM" Net CPM
- IAose concrete Fimr 4 817-95-0327 4,400 1,400 -48% ~32% 9.01E AREA 1 811-95-0481 Concrete Floor 10,800 6,800 ~37% -63% 1.20E" AREA!
conacteFimr 4 817-95-0327 6,200 4,200 -32% -68% 6.01E AREA 1 Concrete Flmr 4 821 95-0328 37,000 16,000 -57% -43% 6.31E AREA 2 821-95-0328 Concrete Floor 4,400 1,000 ~77% ~23% l.02E4 AREA 2 843-95-0330 Concrete Floor 750 500 -33% -67% 7.5 ) E AREA 4 843-95-0330 Concrete Floor 2,700 2,100 4 22 % ~78% 1.80E AREA 4 concrete wall 4 817-95-0327 4,400 350 -92% -8% 1.22E AREA 1 836-95-0480 concrete wall 300 300 None None <MDA AREA 3 836-95-0480 concrete wall 13,400 11,800 -12% -88% 4.80E4 AREA 3 concrete wall 4 821-95-0328 24,000 23,000 -4% -96% 3.00E AREA 2 836-95-0482 concrete wall 800 ~ 400 ~50% ~50% 1.20E 4 AREA 3 conecte wall 4 834-95-0325 24,600 22,800 ~7% -93% 5.41E AREA 3 824-95-0329 concrete wall 600 300 ~50% ~50% 9.01 E AREA 2 843-95-0330 Concrete Head Stnd. 600 400 ~33% -47% 6.01E-3
, AREA 4 k
w 4-13
l l S xt:n Nuclar Experimental Ccrp ration (SNEC) j Site Characterization Report j [) i V i , 4.1.11 Metallic Surfaces Metallic surfaces include gratings, platforms, upper containment walls, component exterior surfaces, piping extemal surfaces, and other miscellaneous structural ' materials. Some of these surfaces at the SNEC facility have been painted. Table 4-4 provides the information collected from direct frisk measurements of metallic surfaces within the SNEC facility CV. These measurements were performed in ! accordance with the method described in Section 4.1.8. 1 Data Tables, Section 4.5 ) Among the data tables presented in Section 4.5, are radiological sun /ey results from l 1 characterization surveys of most metallic surfaces, equipment and components within the SNEC facility CV. The mean and range of contamination levels present on j components and surfaces, is provided. The number of smears taken (In most j cases), is equal to or greater than the number specified by the SCP, for the item or surface listed therein. In some cases, items not specified by the SCP survey program are presented in these tables as additional information. For some equipment, it is assumed that the extemal contamination levels are the same as the average contamination level present on other similar miscellaneous surfaces or equipment, within the respective area where the item is located. In all cases the data presented in Section 4.5 is reasonably comprehensive and representative of exterior surfaces of components, equipment and structural surfaces within the SNEC facility CV. v 4-14
Sixt:!n Nucle:r Experirn:ntil Ccrp:rati:n (SNEC) Site Characterization Report
\'
Table 4-4, Metallic Surfaces ! ~ 2
- Note: Assumes a friskeris sensitive to ~15 cm of surface area.
Radiological SNEC CV Pre-Decon Post-Decon % Ieose % Fixed uCi/100cm' l Survey No. Iecation Net CPM
- Net CPMa Imose 3
817-95-0327 SteelTank 7,500 4,800 ~36% ~64% 8.11E AREA 1 817-95-0327 Pump Motor 10,800 8,800 -l9% -81% 6.01E'3 AREA 1 817-95-0327 SS CRDM'S 4,000 2,000 -50% ~50% 6.01E 3 l' AREA 1 d 817-95 4 327 SS Pump Drain 25,800 16,200 -37% -63% 2.88E l Tnk. AREA 1 833-95-0326 StL Fir. Graing 11,000 9,000 -18% -82% 6.01E'3 AREA 3 i 833-95 4326 Stt chiller 3,500 800 ~77% ~23% 8.llE-3 AREA 3 l 833-95-0326 StL Flt. Grating 4,600 2,600 -44% ~56% 6.01E'3 l AREA 3 3 833-95-0326 SS Piping 4,400 1,600 - .64% ~36% 8.4IE AREA 3 821-95-0328 Sit vent Hm 11,000 8,000 -27% ~73% 9.01E 3 1 AREA 2 ' l 821-95-0328 Heat Exchanger 12,000 6,000 -50% ~50% 1.80E" AREA 2 834-95-0325 StL Pump Basin 2,100 900 ~57% ~43% 3.60E'i AREA 3 834-95 4 325 StL ficer Gratmg 4,000 2,800 ~30% ~70% 3.60E'3 l AREA 3 834-95-0325 Steel Hanger 2,100 450 ~79% ~21% 4.95E.2 AREA 3 3 834-95-0325 Stt Floor Grating 2,000 1,600 -20% -80% 1.20E AREA 3 824-95-0329 Stt Floor Grating 5,400 3,400 -37% ~63% 6.01E2 AREA 3 824-95 0329 StL vlave #27v 12,000 8,000 -33% ~67% 1.20E" AREA 2 824-95-0329 StL Elect. Bx. 7,000 4,600 -34% -66% 7.21E'3 AREA 2 2 843-95-0330 Stt Eq. Hatch 1,200 600 -50% ~50% 1.80E Door AREA 4 3 843-95-0330 StL Tool Rack 2,890 2,000 -29% ~71% 2.40E AREA 4 d 843-95-0330 StL Eq. Trolly 800 600 ~25% ~75% 6.01E AREA 4 4 4.1.12 Smear Analysis (General Area (GA), Smear Analysis Results) Off-site analysis of groups of smears from AREAS 1,2,3,4 and 6 was performed by Teledyne Brown Engineering. These composited smears were collected and analyzed in January and June of 1995. Smears were taken from randomly selected surfaces within each of the AREAS of the CV. These surfaces included concrete, steel platform grating, components (extemal locations) and other miscellaneous exposed surfaces. Each composite smear group from each of the above areas, 4-15
. . . - - - - - - - - - - - ^^~^~~" ~ ~ ~ ~~ ~^ ~
j_,_. i i l Stxten Nucleir Experimant:1 Corpor tion (SNEC) ! Site Characterization Report i j contained between twenty (20) and sixty (60) smears. The June 1995 sample group included another set of AREA 2 smears, and provides confirmation of the January 1995 analysis results for AREA 2. See Tables 4-5 and 4-6. I i 1 1 1 I J-I
\-
4-16
l Sixt:n NuclIrr Experimintil Ccrp:ratian (SNEC) Site Characterization Report i O) t V January 1995 Smear Results I
- The following sample information was reported by Teledyne Brown Engineering for l l
the January 1995 analysis of smears from AREAS 1,2,3 & 4 . See Appendix D for t Teledyne data sets. ; TABLE 4-5 i l l SNEC Facility Surface Contamination Analysis Results l Composited Smears Of January 1995 NUCLIDES AREA 1 % OF AREA 2 % OF AREA 3 % OF (uCi's) TOTAL (uCI's) TOTAL &4 1 TOTAL (uCi's) C-14 3.0E-5* 0.0081 2.0E-5* 0.0242 2.0E-5* 0.2319 Ni-59 3.0E-4
- 0.0814 3.0E-4
- 0.3628 3.0E-4* 3.4781 ;
Sr-90 6.8E-4 0.1845 1.0E-3 1.2094 3.0E-5 0.3478 Fe-55 5.0E-4
- 0.1356 40E4' O.4838 3.0E4' 3.4781 Tc-99 4.0E-5
- 0.0109 3 0E-5* 0.0363 4.0E-5* 0.4637 I-129 5.0E-5
- 0.0136 4.0E-5* 0.0484 7.0E-5* 0.8116 Co40 2.87E-3 0.7786 8.31E-4 1.0050 2.59E-4 3.0028 Zn45 3.0E 4* 0.0814 8.0E-5* 0.0%8 1.0E-5* 0.1159 Ru-106 3.0E-3* 0.8139 1.0E-3
- 1.2094 9.0E-5* 1.N34 h Cs 134 2.0E-4* 0.0543 4.0E-5* 0.N84 6.0E-6* 0.0696 i j Cs-137 3.56E-1 2,0E-3*
96.5780 7.66E-2 92.6432 6.26E-3 72.5768 Ce-144 0.5426 5.0E-4
- 0.6047 4.0E-5* 0.4637 H-3 5.0E-4
- 0.1356 5.0E4* 0.6047 8.0E4
- 9.2750 Ni43 1.2 E-3 0.3255 5.4 E-4 0.6531 8.9E-5 1.0318 l Pu-238 4.6E-5 0.0125 3.1E-5 0.0375 4.0E4 0.0464 l U-234 1.lE4* 0.0003 1.0E-6
- 0.0012 1.lE-6* 0 0128 U-235 1. l E-6* 0.0003 1.0E-6
- 0.0012 1.lE-6* 0.0128 l U-238 1. lE-6
- 0.0003 1.0E-6
- 0.0012 1 lE-6* 0.0128 Am-241 1.8 E-4 0.0488 1.3 E-4 0.1572 1.2E-5 0.1391 Cm-242 1.3E-6* 0.0004 2.6 E-6 0.0031 1.3E4* 0.0151 l Cm-244 2 2E-6* 00006 1.0E-6* 0.0012 9.5E-7
- 0.0110 l Pu-239 1.0E-4 0.0271 8.3 E-5 0.1004 8.6E4 0.0997 Pu-241 6.1E-4 0.1655 5.5E-4 0.6652 2.8E4' 3.2462 Pu-242 9.9E-7
- 0.0003 1.2E4' O.0015 1.2E-6
- 0.0139 TOTAIS 3.69E-1 100 % 8.27E-2 100 % 8.63E-3 100 %
- Reported as "Less Than" values l Note : Because of similar nuclide compositions, smear results from AREA 3 and 4 t
(Table 4-5), were combined prior to analysis. l O m 4-17
. Saxtan Nucler Experimsntil Corp:rrtian (SNEC)
Site Characterization Report b i V June 1995 Smear Results l The following sample information was reported by Teledyne for the June 1995 analysis of smears from AREAS 2 and 6. See Appendix D for the actual data sets. TABLE 4-6 SNEC Facility Surface Contamination Analysis Results Composited Smears Of June 1995 1 NUCLIDES AREA 2 (uCi's) % OF TOTAL AREA 6 (uCi's) % OF TOTAL l C-14 1.1E-5 0.0017 1.1E-5 0.0001 Ni-59 1.0E-3' O.1585 1.7E-3 0.0096 Sr-90 1.2 E-2 1.9021 2.3E-2 0.1304 i Fe-55 5.0E-4
- 0.0793 2.5E-2 0.1418 4 Tc-99 3.2E-5 0.0051 6.2 E-4 a0035 I-129 2.0E-5* 0.0032 2.0E-5* 0.0001 Mn-54 3.0E-4
- 0.0476 4.0E-2* 0.2268 Co-60 1.49E-2 2.3618 7.18E+0 40.7143 Zn-65 7.0E-4' O.1110 1.0E-l* 0.5671 Nb-94 3.0E-4* 0.0476 5.0E-2' O.2835 Ru-106 5.0E-3* 0.7925 4.0E-1
- 2.2682 Ag-110m 4.0E-4
- 0.0634 7.0E-2' O.3%9
( Sb-125 3.0E-3* 0.4755. 1.0E-l' O.5671 Cs-134 3.0E-4' O.0476 5.0E-2
- 0.2835 Cs-137 5.63E-1 89.2394 9.03E+0 51.2048 Ce-144 3.0E-3* 0.4755 3.0E-l* 1.7012 i Np-237 1.0E-3* 0.1585 8.0E-2' O.4536 I H-3 6.0E-4
- 0.0951 2.0E-3 0.0113 N143 1.9 E-2 3.0116 1.3E-1 0.7372 Pu-238 3.5E-4 0.0555 3.1E-3 0.0176 U-234 1.0E-6' O.0002 9.0E-7* 0.0000 U-235 1.0E-6' O.0002 9.0E.7' O.0000 U-238 1.0E-6' O.0002 9.0E-7* 0.0000 Am-241 8.5E-4 0.1347 1.3E-2 0.0737 Cm-242 7.4 E-6 0.0012 1.0E-4 0.0006 Cm-244 1.9 E-5 0.0030 3.0E-4 0.0017 Pu-239 8.9E-4 0.1411 8.2 E-3 0.0465 j Pu-241 3.7E-3 0.5865 2.8 E-2 0.1588 l Pu-242 4.8E-6 0.0008 2.4E-5 0.0001 TOTAIS 6.31E-1 100 % 1.76E+1 100 %
- Reported as "Less Than" values (values listed in Bold were positively identified).
Nuclides with half-lives of < 100 days or naturally occurring isotopes e.g. K-40, Ra-226 and Th-228, were not included in the percent of total columns. These nuclides are not present in sufficient quantity to be significant. "Less than" values are assumed to be valid for calculations related to curie evaluations. 4-18
Sixt n Nucl rr ExpirimIntil Ccrpuratian (SNEC) Site Characterization Report Use These results are useful in estimating typical contamination levels for nuclides commonly found on surfaces throughout the SNEC facility. In addition, these values will provide information useful to health physics personnel when estimating the hazard from potential airbome radioactivity levels and nuclide compositions during future decommissioning work. 4.1.13 Penetration Studies - Concrete Core Bore Sampling Methodology As described in Section 4.1.3-B, core bore samples were removed under the guidance of SNEC Work Instructions (Reference 34). Core bores were then subdivided for further examination including contamination penetration studies and bulk sample 10 CFR 61 type analysis. A masonry saw was used to slice the core bore samples into -13 mm (1/2") thick slices, starting with the top or exposed portion k and continuing through each core bore, until MDA values were encountered for detectable SNEC generated nuclides. The contamination levels of concrete core p bore sections was determined by gamma-ray spectroscopy. Calibration of the I g G a l HP(Ge) detection system was performed by use of a commercial standard prepared l with approximately the same density and thickness as a typical 1/2 inch concrete slice. SNEC Work Instruction 94-006 (Reference 61), govemed the core bore cutting and analysis operation. All core bore slices were analyzed from both sides and the average of the two results was used to represent the activity concentration for the individual slice. A suitable background slice was counted prior to each potentially affected sample and repeatedly throughout the core bore analysis effort. Typical background MDA values for Cs-137, Co-60 and Eu-152 were less than 0.30 pCi/g for the unaffected blank. The complete data set for ah the concrete sample counting effort is contained in SNEC Technical Evaluation Calculation 6575-95-0032 (Reference 81). Sample Location Assignments l Core bore sample locations were originally specified in the SNEC facility Site Characterization Plan (SCP). Additions and deletions to these locations were made 4-19 l l
Sixten Nuctur Experim:nt'.I Carper" tion (SNEC) Site Characterization Report l under the guidance of SNEC and GPU Nuclear management personnel after considering current radiation survey results, available funding, ALARA concems and facility structural integrity . Figures 4-1 to 4-5 show the final locations selected and the core bore sample numbers assigned by SNEC management personnel. In addition to the core bore locations selected from within the SNEC CV, core bore samples were also taken from surrounding Penelee property and from locations in and around the town of Saxton. These last two groups of samples are believed to be from the same construction time period as that of the SNEC facility, and are considered to be representative of concrete background for relevant nuclides. The following table presents sample locations and results for these two groups. o I
, l l
l 4-20
M Sixt:n Nucl:cr Experlm:ntil Ccrp:r ti:n (SNEC) p Site Characterization Report 4 (' Table 4-7, Off-Site Core Bore Locations and Counting Results. l CORE BORE CORE pCi/g pCl/g pC1/g l SAMPLE NO. LOCATION LENGTH Co-60 Cs-137 Eu-152 SX895010090 Saxton IGA Supermarket Sidewalk Slice 1 West Side 8.7cm <0.31 <0.31 <0.32 Slice 2 (same as above) - <0.26 <0.27 <0.28 ] SX895010091 Saxton IGA Supermarket Sidewalk
- Slice 1 East Side 8.7 <0.30 <0.28 <0.26 4
Slice 2 (same as above) - <0.34 <0.21 <0.29 SX895010092 Tussey Mountain High School, Front Sidewalk, West Side Slice 1 9.8 <0.26 <0.25 <0.29 i l Slice 2 (same as above) -- <0.23 <0.26 <0.27 ; i SX895010093 Tussey Mountain High School, Front Sidewalk, East Side i Slice 1 8.4 <0.30 <0.29 <0.33 Slice 2 (same as above) - <0.28 <0.26 <0.25
- SX895010095 South East Corner of Penelec Propeny, Concrete Slab # 1
[ Slice 1 14.6 <0.25 <0.27 <0.28 0 Slice 2 (same as above) - <0.27 <0.23 <0.28 i SX895010096 South East Corner of Penelec Property, Concrete Slais #2 Slice i 10.8 <0.25 <0.27 <0.27 Slice 2 (same as above) - <0.33 <0.25 <0.28
.l SX895010097 Old Westinghouse Air Sample Station Pad, Penelec Propeny-NE j Slice 1 30.5 <0.26 <0.40 <0.26 Slice 2 (same as above) - <0.37 <0.27 <0.29 SX895010098 Penelec Line Shack, Front Sidewalk Slice 1 North Side 7.9 <0.28 <0.29 <0.26 Slice 2 (same as above) - <0.26 <0.24 <0.28 SX895010099 Penelec Line Shack, Front Sidewalk Slice i South Side 7.6 <0.28 <0.27 <0.29 Slice 2 (same as above) - <0.27 <0.21 <0.28 Concrete sample counting results from on-site, but c/Lside of the containment building are O reported in the following table.
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l S xtsn Nuclerr Experimentrl Cerp:rati:n (SNEC) Site Characterization Report G Table 4-8, Counting Results, On-Site Core Bore Locations Outside Of CV. I CORE BORE CORE pCl/g pCi/g pCi/g SAMPLE NO. LOCATION LENGTH Co40 Cs-137 Eu-152 ( SX871950001 Concrete Slab Below Personnel Airlock At CV Slice 1 11.9 cm <0.32 1.09 i0.30 <MDA" Slice 2 (same as above) - <0.29 <0.24 <MDA SX872950002* Concrete Snield Wall External <0.28 <0.30 <0.27 ) Subsurface Slice To CV(Used As Bkgnd Blank) -10 cm (Avg.) (Avg.) (Avg.) l SX871950003 Concrete ledge Adjacent I n fonheast Side Of CV Slics I i1.6 cm <0.23 2.00t0.40 <MDA i i j Slice 2 (same as above) - <0.29 <0.26 <MDA SX871950004 Concrete Pad WNW Of CV - Former HEPA/ Vent Slice 1 System 15 cm <0.91 156.lil9.9 <MDA I Slice 2 (same as above) - <0.30 <0.38 <MDA SX881950101 Concrete Core From Weir Floor Slice 1 15.3 cm <0.34 <0.34 <MDA Slice 2 (same as above) - <0.36 <0.28 <MDA , SX882950102 Weir,Through Concrete Divider Wall (West End) Slice 1 19.4 cm <0.32 <0.33 <MDA Slice 2 (same as above) - <0.27 <0.24 <MDA Through Concrete Divider Weir Wall (East End) Slice 8 - <0.30 <0.25 <MDA Surface Slice Slice 9 (same as above)
- <0.34 0.73 10.24 <MDA SX852950103 Through Concrete Support Wall In Tunnel (West Slice 1 End) 23.8 cm <0.34 8.51 il.22 <MDA Slice 2 (same as above)
- <0.26 0.37 10.16 <MDA l
I Slice 3 (same as above) - <0.31 <0.27 <MDA Through Concrete Support Wall In Tunnel (East End) - <0.36 <0.26 <MDA Slice 7 Slice 8 (same as above) -- <0.30 <0.27 <MDA Surface Slice Slice 9 (same as above) - <0.34 2.72 i0.51 <MDA 1 l 1 ( i 4-22 l
S2xt:n Nuclear Experimental Ccrporati:n (SNEC) Site Characterization Report O Table 4-8, Counting Results, On Site Core Bore Locations Outside Of CV (Continued) CORE BORE CORE pCi/g pCi/g pCi/g i SAMPLE NO. LOCATION LENGTH Co.40 Cs-137 Eu-152 i SX853950104 Through Concrete Ceiling OfTunnel 1 Slice 1 (Top End-Outside) 37.3 cm <0.33 12.56 *1.76 <MDA 4 j Slice 2 (same as above) - <0.33 <0.37 <MDA Slice 3 (same as above) - <0.31 <0.28 <MDA Through Concrete Ceiling OfTunnel Slice 7 (Bottom End-Inside) - <0.29 <0.22 <MDA i j Slice 8 (same as above) - <0.23 <0.22 <MDA i Surface Slice
- Slice 9 (same as above) - <0.29 0.58 10.20 <MDA
< SX852950105 Tunnel Wall At South Entrance Hatch j Slice 1 15.9 cm <0.31 1.35 10.31 <MDA j Slice 2 (same ns above) - d).26 <0.27 <MDA $ Slice 3 - (same as above) - <0.27 <0.24 <MDA j' SX801950106 Westinghouse Concrete Pad Slice 1 Below SNEC Office Trailer 6.8 cm <0.35 <0.37 <MDA ^ Slice 2 (same as above) - <0.34 <0.31 <MDA ' \,V/ SX801950107 Westinghouse Concrete Pad Slice 1 Below SNEC Office Trailer 9.7 cm <0.29 <0.29 <MDA Slice 2 (same as above) - <0.28 <0.25 (MDA j SX871950108 Shield Wall Adjacent To CV, NW Side Slice I rHigh n-Flux Region) 18.8 cm <0.30 <0.33 <MDA Slice 2 (same as above) - <0.30 <0.30 <MDA ! Slice 9 i Taken At A Depth Of-17.5 to (same as above) - <0.23 <0.17 (MDA j- 18 8 cm Below Exposed Surface
- This core was used as a background blank sample and counted frecuently throughout the core bore
. analysis program. , " AII <MDA values are approximately the same as that obtained for the blank or background core slice for this isotope. l Table 4-53 (Section 4.5), provides the remaining core bore results for the core bore samples taken from within the SNEC CV. The depth of the concrete slice is referenced to the starting point of the drilling operation, and is the bottom side of the slice listed. The following table presents all of the SNEC CV concrete sample p locations. 4-23
i l l Saxt:n Nuclerr Experim:ntal Carp:rati:n (SNEC) Site Characterization Report p i ! Table 4-9, SNEC CV Concrete Core Bore Sample Locations. SNEC Sample Number Location SX841950008 CV Floor,818' El., West Side SX841950009 CV Floor,818' El., East Of 008 SX841950012 CV Floor Plug,818' El. SX842950013 Wall, Between 812' & 818' El., East Side SX842950014 Wall Between 812* & 818' El., By Steps SX842950015 Wall Between 812' & 818' El., North Side SX841950016 CV Floor,812' El., West Side Under Platform SX841950017 CV Floor,812' El., NE Of Reactor Cavity SX841950018 CV Floor,812' El., West Side By Platform SX832950019 SW Wall,~799* EL, Above Water Line SX832950020 CV Outer Wall, ~802" El. SX832950021 CV SW Wall, ~791'6" El., At Water Line SX832950022 North Wall,~788* El. SX832950023 North Wall, ~786* El. SX812950024 CV Outer Wall, Rod Room ~768' El. SX811950025 CV Floor, Below RV, ~765'-8" El. SX811950026 CV Floor, SE Corner Of Sump, ~765'-8" El. SX811950027 CV Floor, Between Drain Pumps, ~765'-8" El. SX811950028 CV Ledge, West Of Drain Tank, ~768' El. SX812950029 CV Outer Wall, West Of Drain Tank, ~770' El. /O SX812950030 Ledge / Wall, South Of Filter Cubicle, ~766' El. ( SX822950031 SX821950032 SW Wall, ~10' East Of Hot Leg Penetration, ~789' El. CV Floor Under S/O,~779' El. SX821950033 CV Floor Under S/O, ~779' El. SX822950034 Wall East Of Non-Regen Heat Exchanger, ~782' El. SX822950035 CV Outer Wall NW Of S/O, ~798' El. SX822950036 SW Wall Next To Hot Leg, ~790'-8" El., At Water Line SX822950037 SW Wall 8'-3" East Of Hot leg, ~799' El. SX861950056 Storage Well Floor, ~765' 8" El. SX861950057 Storage Well Floor, '765'-8" El. SX862950058 Storage Well, Outer Wall, ~771' El. SX862950059 Storage Well, Outer Wall, ~771' El. SX862950060 Stora ge Well, inner Wall, East Of Rx, ~768' El. SX862950061 Storage Well, inner Wall, East Of Rx, ~768* El. SX862950062 Reactor Cavity Area, South Wall,797* El. SX862950063 Reactor Cavity Area, South Wall,797' El. SX862950064 Reactor Casity, Shield Wall North West Of RV,784' El. SX862950065 Reactor Cavity, Shield Wall North East Of RV,784' El. SX861950066 Reactor Casity Floor, ~779'-8" El. SX862950119 CV Outer Wall, NE Of RV, in Hi-Flux Region, ~795' El. SX841950120 To SX841950125 Reactor Cavity Shield Blocks,1 To 6, ~812' to 807' El. l 4 l 1 l j g v 4-24
S2xt:n Nuclerr ExperimIntal Corporati:n (SNEC) l Site Characterization Report l 4.1.14 Activation Of Concrete Core Bores , i Activation products have been detected and predicted in concrete material located in the reactor compartment near the reactor vessel head, above the operational water line (~791' El.). The predominant gamma emitters detected are Co-60 and Eu-152, with measurable quantities of Eu-154. A later discussion of the analysis results from activated concrete is presented in Section 4.3, An analysis by B&W of SNEC concrete core bore sample SX862950119, taken from the suspected high flux area of the CV containment wall, has provided the i following information and respective radionuclide mix. This is a composite result for slice numbers 1,2,6,9 & 13. Slice number 6 contained a 57.74 gram sample of steel reinforcement material (rebar). Table 4-10, Composite Results - Concrete Core Bore Sample SX862950119. Isotope As % Of Total Activity , Ag.108m 0.0266 Am-241 0.0464 Cc-141 0.0924 Co-60 2.4443 Cs-134 0.0213 Cs 137 7.5581 l C 14 2.3263 Eu-152 3.5327 Eu-154 0.1996 Eu-155 0.0489 H-3 63.5044 1-129 0.4071 Nb-94 0.0204 , Ni-59 0.7042 Ni-63 18.6422 Ru-106 0.1540 Sb-125 0.05 % Tc-99 0.2206 Note: Isotopes listed above in bold text were positively identif;ed. V 4-25
i l , Srxt:n Nuclear Experimentil Ctrp rati:n (SNEC) Site Characterization Report i V 4.1.15 Fission Product Contaminated Core Bores The predominant fission product detected by SNEC personnel was Cs-137. This nuclide represents greater than 95% of the activity detected by SNEC facility personnel within and on concrete surfaces. Removal of the first 1 to 2 inches of concrete materials will remove the majority of this isotope. 4.1.16 Applied Limits Current regulations set forth requirements that limit total dose to populations from released materials. A release limit for radionuclides present at the SNEC facility, that will result in a dose of $15 mrem per year (residential scenario) to the population, will be maintained as a requirement during decommissioning. To achieve this, typical SNEC radionuclide concentrations present within and on structural materials were input to the RESRAD' computer code (Reference 8). Calculated concrete values from the Reactor Vessel activation analysis and surface contamination radionuclide N mixtures were selected as the limiting cases. The resulting, and limiting isotopes were: Cs-137, Co-60 and Eu-152. Conservative isotopic concentrations based on appropriate modeling were selected to meet the $15 mrem per year dose to the population requirement. GPU Nuclear will adjust the release criteria for concrete or other materials to be within the regulatory requirements, but reserves the right to recalculate the applicable site release limits during the decommissioning process to applicable limits, as required. For purposes of this report, the methodology described above was used to conservatively estimate the concrete waste volumes that may be generated during dismantlement. 4.1.17 Conclusions Cs-137 is the most significant radionuclide detected in concrete sampled during the charecterization program. The Cs-137 activity concentration decreases dramatically beyond the first 1/2" slice, with only a small fraction of the total activity detected beyond the first slice (> 1/2"). The bulk of the contaminated concrete is found in the l fN l 4 RESRAD is a US DOE computer code used to develope compliance with residual radioactive material
's \ guidlines.
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Srxt:n Nucleir Experim:ntal Corp: ration (SNEC) Site Characterization Report Storage Well/ Reactor cavity (AREA 6). A large part of the total concrete waste volume presented in this report, is the result of the expected need for removal of concrete material from this area. The four and five foot walls that separate AREA 6 from AREAS 1,2 and 3 are contaminated to well below surface depth. In addition, the Reactor Cavity containment walls in the NE and NW sectors, and above the I original operating water level (El. 791'), were exposed to neutron streaming. The most affected section is in the north east sector (high flux area). Sample number SX862950119, containing a section of rebar, was taken in this region. The elevated Co-60, Eu-152 and Eu-154 concentrations found in the rebar and concrete indicate that a significant amount of concrete will have to be removed from the containment l wallin this region. The shield blocks above the reactor vessel were also affected by the increased neutron occurrence above the operational water line. No neutron activation of concrete was observed below the operational water line in the vicinity of the Reactor Vessel. See Table 4-12 for estimated concrete volumes. (% 4.1.18 B&W Analysis Of Core Bore Samples A select group of concrete core bore slices were sent to B&W for analysis that included hard to detect nuclides as well as typical gamma emitters. B&W positively identified Cs-137, Co-60, Eu-152 and Eu-154 as well as Cs-134, Ni-59, Ni-63, H-3 & l Nb-94. All other isotopes were reported as "less than" values. Comparison of the SNEC and B&W Results l The following table presents the analytical results from SNEC facility and B&W analysis of concrete materials. These two analysis results were obtained by use of different counting geometry's and preparation methods. As previously mentioned, , SNEC personnel counted slices of concrete in total. B&W pulverized the slices, dissolved them, and then took an aliquot from the solution for the analysis. GPU Nuclear personnel were informed by B&W after the analysis results were reported, that there was a possibility that the concrete crusher used to pulverize the concrete slices was inadvertently cross contaminated with low !avel activities of Cs-137. This (G makes the B&W results somewhat more uncertain for at least the low level Cs-137 4-27
Sixt:n Nuclar Experirnantal Corp:raticn (SNEC) Site Characterization Report \ activities reported herein. GPU Nuclear believes that the analysis performed by SNEC personnel, identifying Cs-137, Co-60, and Eu-152 has incorporated a smaller total error, except for the core slice that contained steel rebar. In part, the reason for this is that the whole slice was counted intact, rather than a small portion of the whole. The B&W concrete analysis report is provided in Appendix D. SNEC And B&W Concrete Results Summary The SNEC facility counting results average about 111% of B&W results for Co-60, with a standard deviation of i 34%. SNEC facility Cs-137 results average ~63% of B&W reported values, with a standard deviation of 34%. SNEC facility Eu-152 counting results are more consistant with B&W values but show a negative bias. SNEC facility results for this isotope average 65% of the B&W results with a standard deviation of 6%. These overall differences between SNEC facility results and B&W results have been discussed with B&W personnel. The SNEC facility did p not perform analysis for non-gamma emitting radionuclides in concrete. \.] Use The results of the B&W analysis of concrete core bore sections were, and will be used to provide radionuclide concentrations for hard to detect nuclides in relation to other gamma emitters. As a fraction of the total radionuclide mix, they were input to the RESRAD computer code to determine what level of remediation may or should be perfomled. However, for the majority of concrete at the SNEC facility, Cs-137 is the important radionuclide. Removal of this radionuclide to concentrations compatible with current release criteria will in most cases remove or reduce other radionuclides to acceptable levels. The exception is the area above the water line in the reactor cavity where neutron activation has occurred in excess of the results predicted by the reactor vessel activation analysis (Section 4.3). These results will also provide additional waste stream information for waste packaging and shipping purposes. Om 4-28
- Saxt
- n Nucl2ir Experimintil C:rp3ratien (SNEC)
Site Characterization Report (U a Table 4-11, Comparison Of SNEC and B&W Concrete Analysis Results (pCilg) No. Core Bore SNEC Co-60 Cs-137 Eu-152 Eu-154 Co-60 Cs-137 Eu-152 Eu-154 i Sample No. Number SNEC SNEC SNEC SNEC B&W B&W B&W B&W I 0091 0127 <0.29 <0.24 <0.27 N/R <0.01 0.12 <0.11 <0.04 2 0026,0027 0128 <0.96 1717 <l .87 N/R 0.5 1480 <0.35 <0.09 l 3 0032,0033 0129 8.14 5836 <2.56 N/R 6.37 4690 0.73 <0.13 4 0021 0130 <0.84 12.16 <0.65 N/R 0.26 58.6 <0.14 <0.15 j 5 0021 0131 2.07 6.87 <0.38 N/R 1.32 14.2 <0.44 <0.10 6 0021 0132 <0.31 0.36 <0.29 N/R 0.03 1.24 <0.18 <0.06 7 0016,0017,0018 0133 <0.61 15.01 <0.65 N/R 0.1 13.5 <0.14 <0.05 8 0103 0134 <0.31 3.05 <0.3 N/R 0.03 3.% <0.13 <0.05 9 0056,0057 0135 24.69 10668 <4.29 N/R 31.9 13800 <4.49 <0.51 10 0056,0057 0136 <l.17 665.2 <2.34 N/R 2.15 M5 4.64 <0.14 i i 11 0025 0137 3.25 6074 <6.7 N/R 1.9 7180 2.16 <0.23 12 0025 0138 <0.51 1419 <3.04 N/R 0.29 1870 2.61 <0.10 l 13 0025 0139 <0.38 18.34 <0.47 N/R 0.21 31.4 1.18 <0.15 i 14 0025 0140 <0.36 2.82 <0.33 N/R 0.05 11.5 <0.48 <0.15 q 15 0031 0141 <0.65 108.8 <l .28 N/R 0.16 127 <0.17 <0.05 l I 16 0031 0142 14.8 195.8 <l.08 N!R 14.5 246 <0.89 <0.15 l 17 0031 0143 <0.4 6.55 <0.37 N/R 0.12 13.8 <0.18 <0.07 18 0031 0144 <0.33 4.77 <0.32 N/R <0.03 8.98 <0.18 <0.06 19 0037 0145 <0.65 106.4 <0.79 N/R 0.09 132 <0.15 <0.05 x' 20 0037 0146 3.41 29.22 7.09 N/R 2.67 27.8 9.31 0.59 21 0037 0147 <0.36 17.28 <0.53 N/R 0.1 32.1 0.33 <0.08 22 0037 0148 <0.99 7.63 5.1 N/R 0.88 14.9 7.57 0.37 23 0125 0149 1.4 3.43 5.91 N/R 1.47 6.47 10.7 0.64 24 0125 0150 <0.68 0.39 6.01 N/R 0.92 1.62 9.05 0.61 3 25 0125 0151 <0.33 0.19 2.22 N/R 0.32 1.08 3.3 <0.14 1 26 0125 0152 <0.23 0.23 <026 N/R 0.08 1.89 0.48 <0.04 27 0066 0153 14.97 498.6 <2.49 N/R 11.8 513 <0.78 <0.13 28 0066 0154 <0.32 1.06 <0.28 N/R 0.13 3.66 <0.26 <0.09 29 0062 0155 11.22 127.3 5.67 N/R 8.81 119 8.4 0.59 } 30 0062 0156 0.97 35.86 6.45 N/R 1.25 46.7 10.2 0.37 31 0063 0157 25.72 211.2 4.92 N/R 15.7 261 7.81 0.56 32 0063 0158 1.12 0.41 6.8 N/R 1.7 3.94 10.6 0.53 33 0065 0159 <l .23 36.82 <0.79 N/R 1.09 37.4 <0.16 <0.05 34 0065 0160 <0.38 0.6 <0.27 N/R 0.12 1.39 <0.23 <0.07 4 35 Oil 9 0161 11.53 279.9 22.99 <3.27 9.02 233 31.5 1.94 36 0119 0162 53.39 0.83 25.89 <2.12 92.896 5.67 46.99 3.34
; 37 0119 0163 4.02 0.69 37.37 <2.25 4 42 2.19 60.8 2.67 38 0119 0164 2.12 0.56 22.56 < l .% 2.59 1.66 32 1.32 i Note: SNEC number 0162 above, is the composite results from the rebar and concrete sample +
materials. Presence of rebar affected the SNEC counting results for this sample (slice 6, SX862950119). A rebar locator was used to avoid rebar in all other sampling locations. N/R - Not Reported. A 4-29
S:xton Nuclerr Experimint:1 Corporati:m (SNEC) Site Characterization Report O 4.1.19 Estimating Concrete Waste Volumes Core bore sample locations were selected after reviewing radiological survey results I and specifications of the SCP. Whenever possible, samples were selected from locations that contained higher contamination levels, but it was not possible to sample (core bore), every concrete surface in or around the SNEC CV. While a significant number of surfaces were sampled, few samples were taken from ceiling areas. This was because of the lower level of contamination typically found on ceiling areas and the difficulty encountered with safely mounting core bore equipment to ceiling surfaces. Therefore, a minimum removal thickness of 0.25 inches was used for purposes of estimating concrete volumes generated from scabbling ceiling and low contamination level concrete surfaces. In instances where several concrete core bore samples were taken from the same surface, average activity concentrations were developed for slices taken from at or near the same depth. Conservative assumptions about the depth of penetration were made when radionuclide O concentrations appeared to continue beyond the length of the core bore samples taken from a selected location. For instance, the 5 foot concrete wall separating the primary and reactor compartments (based on thme (3) through-wall core bore samples), was assumed to be in need of complete rerloval. This represents about 45% of the total concrete volume estimate for the SN2C facility. Table 4-12 presents the estimate of tr a concrete. waste volume for the SNEC facility. This estimate will be refined as necessary to remain within the current or future regulatory release guidelines as more information becomes available during or prior to decommissioning. O 4-30
Sixt2n Nuclear Experimental Ccrp2ratten (SNEC) Site Characterization Report ! i V Table 4-12, SNEC Concrete Waste Volume Estimate SNEC LOCATION Square Depda Volianne Weight Weigh AREA Feet On) ft* Obs) (tens) 1 765' 8' El Floor 705 6.8 399.5 56862 28.4 1 768' 8* EL Ledge 350 1.3 37.9 5397 2.7 1 765' 8' El. Walls 2543 4.3 911.2 129700 64.9 1 775'2" EL Ceiling
- 733 0.25 15.3 2174 1.1 2 779' 8* El. Floor 475 5.1 201.9 28734 14.4 2 5'7 hick Pnmary Compartment Wall (See Note 1) %8 59.9 4833.9 688027 344.0 2 Pnmary Compartment Walls, Other Than 5' Section 1875 5.3 827.9 117845 58.9 2 814' 6" EL Concrete Ceiling
- 475 0.25 9.9 1409 0.7 3 Auxiliary Compartment Walls (All) 2405 2.7 541.1 77020 38.5 3 810' El. Ceiling
- 381 0.25 7.9 1130 0.6
- 4 818' El Operating Floor 669 0.6 33.5 4761 2.4 4 812' El Operating Floor, Includes Top Or Shield Plugs 1202 1.8 180.3 25663 12.8 4 4 812' El Operating Fir., Interior Walls 341 0.6 17.1 2427 1.2 5 CV Tunnel Walls, Below Grade 2014 0.4 67.1 9555 4.8 5 CV Tunnel Ceiling, Below Grade 953 0 00 0 0.0
- 5 CV Tunnel Floor, Below Grade ( Assume 812' Ops. Floor Depth) 953 1.8 143.0 20347 10.2 5 CV Tunnel Ceiling, Exterior (Grade level) 953 0.6 47.7 6782 3.4 6 765' 8' Storage Well Floor 1035 5.6 483.0 68747 34.4 6 Storage Well Walls-SW Side 2465 2.8 575.1 81855 40.9 j 6 Rx Cavity Side Walls, < Operating Level Water Line (NO 5' WALL) 387 1.6 51.5 7337 3.7
( 6 Rx Cavity Side Walls, > Operating Level Water Line (NO 5' WALL) 545 18 817.5 116358 58.2
\ 6 Rx Cavity Side Ledge At 779' 8" Elevation 113 1.9 17.9 2547 1.3 6 807 El. Ceiling OrStorage Well (Shield Blocks etc.) 574 12.8 612.3 87146 43.6 7 Concrete Walls And Pads Outside Of CV 500 0.2 8.3 1186 0.6 8 Sanitary Sewage Treatment Facilities, Weir, Tanks etc. 993 0 0.0 0 0.0 TOTALS => 246 % N/A 10841 1543009 772
- Volume Based On A Minimum Removal Depth Of 0.25 Inches.
3 Note: 10841 ft* a 307 M of Concrete Waste Materials. Significant assumptions concerning several wall and floor areas were made in assembling the above information. One assumption is that the 5 foot wall between the primary and reactor compartments will be removed entirely. A second assumption is that all concrete around the inside of the reactor compartment, adjacent to the outer CV steel liner and above the operational water line must be removed. The elimination of both of these walls in their entirety may not be necessary or prudent. In addition, the presence of the reactor vessel and associated support structures makes in place measurements impossible. Only detailed surveys and sampling during the decommission phase will determire additional needs regarding concrete removal. b(" 4 31
Sixt n Nucle:r Experirnental Ctrp;rati n (SNEC) Site Characterization Report O The fol!owing table summarizes the concrete activity concentrations with the addition of the hard to detect radionuclides obtained from the B&W report. Note that with the addition of the activation products and surface contamination, Cs-137 is ~50% of the nuclide mix with H-3 at about 27% of the totalinventory. Many of the isotopic totsis are the result of including "less than" values from analytical reports. Surface contamination is about 40% of the total. 2 Table 4-13, SNEC Facility Radionuclide inventory Of Contaminated Concrete i ISOTOPE ACTIVITY (Ci's) % of Total Ag-108m 4.36E-04 0.09 Am-241 8.24E-05 0.02 Ce 144 1.87E-03 0.40 Co-60 4.6E-02 9.75 Cs-134 3.12E-04 0.07 Cs-137 2.38E-01 50.42 C-14 5.74 E-03 1.22 l Eu-152 1.42E-03 0.30 l Eu-154 4.09E-04 0.09 Eu-155 1.60E-04 0.03 f 'N Fe-55 1.61E-04 0.03 Fe-59 6.23E-04 0.13
\
H-3 1.29E-01 27.35 I 129 9.01E-04 0.19 Mn-54 2.49E-04 0.05 Nb-95 3. I 1E-04 0.07 Ni-59 1.67E-03 0.35 Ni-63 3.93E-02 8.34 Np-237 4.98E-04 0.11 Pu-239 5.24E-05 0.01 Pu-241 1.84E-04 0.04 Ru-106 2.50E-03 0.53 Sb-125 6.25 E-04 0.13 Sr-90 1.59E-04 0.03 Tc-99 5.22E-04 0.11 Zn-65 6.23E-04 0.13 Totals => 4.72 E-01 100 % Nuclides greater than 1% of total are in bold text. p) t v 4-32
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l 4.2 SYSTEMS AND COMPONENT CHARACTERIZATION l Systems characterization activities determine the radiological characteristics of known or potentially contaminated systems. The scope of these activities include: ! I (1) Determining the contamination levels on the interior surfaces of components. (2) Performing gamma survcys adjacent to components (used to estimate internal curie content). (3) Quantitative and qualitative isotopic determinations of representative samples, as applicable. Systems were characterized by selecting sections of piping systems for direct measurements, and sampling these .same systems (or like systems), to obtain representative radioactive contamination samples. Components that were selected were opened to expose the interior surfaces for measurements, or sections of system piping or equipment were removed as deemed appropriate, for analysis on-site or at off-site laboratories. 4.2.1 Data Needs This section of the SNEC Facility Site Characterization Report (SCR), sets forth the methodology by which potentially contaminated systems and equipment at the , SNEC facility were characterized. The data that was obtained as part of the system and component characterization efforts includes: Data for estimating the total radionuclide inventory within each system. Determinations of isotopic identity and concentration ratios of radionuclides that l l comprise the radioactive contaminants within selected systems. j 1
- Comparisons of characterization data with applicable release criteria to determine decommissioning work scope.
Waste stream identification and classification. o U 4-38 i
, S xt:n Nucle:r Experim:ntil Ccrporati:n (SNEC)
Site Characterization Report U 4.2.2 Applicable Areas and Equipment Equipment and systems within the Containment Vessel that could conceivably contain radioactive materials were reviewed. These systems and components were then identified and selected based on the following criteria: Systems or equipment which carry fluids or gases that may have circulated through the reactor vessel, radioactive waste systems or spent fuel storage systems. Systems or equipment with indications of contamination as determined from previous surveys and maintenance records, as well as gamma extemal exposure rate scans performed specifically for the systems and equipment characterization program (scoping measurements). 4.2.3 Selection of Sampling Locations C Component characterization is on a system by system basis. Specific items within each system were chosen for characterization and selected to represent the overall condition of that system. The distribution of radioactive contaminants within a system were determined and a realistic curie estimate was prepared. 4.2.4 Analytical Methods The following describes the analytical techniques that were used to characterize intemal systems contamination, and the criteria under which they were implemented at sample locations. A. Removable Contamination Determinations When practical, smear or scrape samples were taken from intemal surfaces of equipment or components opened for characterization. All samples collected were gamma scanned. If appropriate, the samples were counted for gross beta O 4 39
. . - . . . . . - . - . - . .- - - . - --.. . - - - . - - . ~..- .- - .- _ _ .- .
I Saxtsn Nucisir Experiment:1 Corpsr: tion (SNEC) ] Site Characterization Report l 1 and alpha contamination. In most cases, samples were sent to off-site analytical l laboratories for 10 CFR 61 type analysis. (! i B. TotalInternal Surface Contamination Determinations l Total intemal surface contamination levels were estimated for component j surfaces expected and not expected to be intemally contaminated. 'These f estimates were made from measurements using appropriate radiological ! detection instrumentaSon, and/or from results provided by on-site or off-site i i analyticallaboratories. L i' i C. Bulk Sampling and Analysis i l Sediment deposits were collected on a case by case basis for isotopic analysis. Types of analysis specified are listed in SCP (Appendix A). Typical subjects for this type of sampling was the CV sump and the tunnel extemal to the CV. j D. Piping Systems Samples 1
- Samples from individual piping systems representing potentially different waste l streams were gathered by cutting out and removing portions of these systems, Each piping system was first identified and radiologically surveyed to verify that a f
f reasonable sample could be ot'tained without exposing cutting team personnel to excessive radiation hazards during the cutting operation. The piping sections were typically less than 30 centimeters (-12"), long and from about 4 to 10 centimeters (~1.5" to ~4"), in diameter. Pipe sections from various systems were removed, capped and sent to off-site laboratories for analysis that in all cases included a 10 CFR 61 type radionuclide determination. O 4 40
{ .Srxton Nuclxr Experimentil Corporation (SNEC) Site Characterization Report i ! s j 4.2.5 Data Reporting and Analy. sis r i Equipment and components that have been characterized are presented in tables 3 along with relevant radiological survey information (Section 4.5). Results of gamma l sumeys were interpreted and analyzed and the total intemal surface contamination l results are estimated. The " average" isotopic distribution of radionuclides on intemal l piping and components is then presented. l 4.2.6 Radiation Measurements On Equipment And Components The data tables presented in section 4.5 provide results of general radiation and contamination measurements of extemal surfaces of plant structures and components. When available, smear and direct frisk measurements are also shown. Measurements of radiation from components were taken at one foot intervals along their lengths and insulation thickness was measured. This was done in four locations approximately 90 degrees apart around each component. All the results were then averaged and input to a computer radiation shielding code. l 4.2.7 Radiation Measurements Of Piping Systems l Piping systems that were evaluated, are listed within the SCP. Radiation surveys were specifically designated. Surveys were conducted along the length of a section of system piping. Radiation measurements were taken at one foot intemals from the top and bottom of horizontal pipe runs, and from two opposite sides of vertical pipe sections. Measurements were taken on contact with piping insulation (when present), j and at one foot from pipe section centerlines. Insulation thickness was determined for each pipe section and the average radiation measurement result was then calculated for the length of pipe surveyed. This was done for each system evaluated. All of these parameters were input into a radiation shielding computer code along with the appropriate sample data. Activity per unit area within each pipe section was calculated using the output of the shielding code. The calculated intemal activity per Oy unit area was then applied to the remainder.of the unsurveyed piping system. The ' 4-41
Srxt n Nuclear Experimintal Corporation (SNEC) Site Characterization Report [ complete piping system curie estimate is presented within SNEC Technical Evaluation Number 5830-95-023 (Reference 79). An engineering walk-down was performed early in the characterization process to determine all of the SNEC facility piping system pipe lengths and sizes. This information is presented in Appendix E. 4.2.8 Samples Of Internal / External Contamination The SNEC SCP provides a listing of systems and components that are to be characterized. Samples from these systems and components were collected during characterization efforts in 1995. Samples included ~17 sediment samples from sumps and tanks, ~17 pipe sections from system piping, ~8 scrapings from air handling units and instrument racks etc., and smears from intemal surfaces of components. From these samples, about 30 were sent to B&W for analysis. The complete sample log from the SNEC facility can be seen in Appendix F. All samples j were given a SNEC facility sample number and logged into the sample log. l Several systems samples that were specified by the SCP, could not be obtained I
- because of high radiation levels present in the vicinity of the, intended sampling I locations. However, reasonable substitutes were found. The septic system and associated piping and drainage system had limited accessibility during the 1995 j characterization effort. This was partially due to elevated ground water levels from inclement weather and winter snow and ice buildup. Several samples were collected
, from Septic Tanks A & B. Radioactivity content of the septic tanks is presented below (only positively identified radionuclides are shown). , l s t V 4-42
Saxt n Nucle:r Experimsnt'l Carptrati:n (SNEC) Site Characterization Report 'v Table 4-14, Radionuclide Content Of Septic Tanks A & B ITEM ISOTOPES ACTIVITY Septic Tank A(Liquid Phase) Ni-63 1.79E-05 uCi/ml Septic Tank A (Solid Phase) Ni-63 8.49E-05 uCi/g Co.60 1.07E-06 uCi/g Cs-137 2.llE-05 uci/g Septic Tank B (Liquid Phase) Ni-63 8.83E 05 uCi/ml Sr-90 1.00E-06 uCi/mi Septic Tank B (Solid Phase) Ni-63 5.04E-05 uCi/g Pu-238 5.56E-06 uCi/g Co 60 1.23E-06 uCi/g Cs-137 3.20E-05 uCi/g The complete isotopic analysis results of Septic Tank materials is shown in Appendix D. A planned drainage line inspection will be performed prior to decommissioning. / 'N information from this effort will be available at a later time. \ The following table presents the SNEC facility system samples that were collected and sent off-site for analysis during the characterization program. These samples, while not all inclusive, are representative of the SNEC facility systems internal contamination. The radionuclide concentrations indicated from these sampling results and surface contamination data were directly applied within computer shielding code calculations to develop the curie estimate of record for the facility. J 4-43
S"xt:n Nucl:ar Experim:ntal C:rp:ratinn (SNEC) Site Characterization Report Table 4-15, Systems Samples From SNEC Facility item Sample No. Sample Description (Sample Type in Bold Text) 1 SX815950005 CV Sump Sediment 2 SX815950006 Scrape Sample From Instrument Rack No.1 3 SX815950007 Scrape Sample From Instrument Rack No. 2 4 SX815950010 Vents & Drains System, Pipe Section Between CV Sump Pumps & CV Wall 5 SX815950050 Control Rod Room Air Handler Scrape Sample 6 SX815950068 Discharge Tank Sediment 7 SX815950168-SD Condensate Tank Sludge / Sediment 8 SX825950011 Pipe Section from Make-Up Demin Water Line 9 SX825950038 Pnmary Air Handler Scrape Sample 10 SX825950040 Purification System, Pipe Section Between MC Piping & Purification Demin. I1 SX825950041 Pressure Control & Relief, Pipe Section Between MC Cold Leg & PZR Spray Line 12 SX825950042 Shutdown Cooling. Pipe Section Between MC Piping & Safety Injection Piping 13 SX825950043 Pressure Control & Relief, Pipe Section Between MC Piping Hot Leg & PZR - Surge Line 14 SX825950044 Pressure Control & Relief, Pipe Section Between PZR Relief Valves & DMege Tank 15 SX825950045 Component Cooling Pipe Section, Betwten CC Heat Exchanger & Discharge HaA r Of Pumps 16 SX825950046 Storage Well System. Pipe Section Between Inlet Of SW Pumps & Discharge Point In SW 17 SX825950047 Vents & Drains System, Pipe Section From Vent & Drain Header Exitmg To Discharge Tank 18 SXR25950086 S/G Bottom Blowdown Line Pipe Section 19 SX825950087 S/G Surface Blowdown Line Pipe Section 20 SX825950111 S/G Bowl Smear, Cold Leg Side [N 21 SX835950039 Scrape Sample From Auxiliary Air Handler ( 22 23 SX835950071 SX835950085 Scrape Sample From Secondary Side Of Steam Generator Component Cooling River Water Pipe Section 24 SX856950167-SD CV Tunnel Sediment / Water 25 SX865950053 Safety Injection Pipe Section, Between RV & Check Valve 26 SX865950069 Safety Injection Pipe Section, Between Check Valve & CV Wall 27 SX865950118 Make-Up Demin Line Pipe Section 28 SX865950165-SM Reactor Vessel Swab / Smear OfInternal Surface Activity 29 SX885950048 Septic Tank A Sediment 30 SX885950049 Septic Tank B Sediment 4.2.9 Sample Results Relating To Decontamination Off-site analysis of systems samples included an attempt to remove the materials from the piping intemals. Acid leaching operations for periods of three to four hours, using a 30% HNO3/5% HCL acid solution netted anything from greater than 90% removal of intema' contamination to almost no removal (less then 1%). This suggests that a field effort designed to remove internal contamination "in-place', may encounter mixed success. Additionally, the intemal contamination of some pipe sections, has a considerable transuranic nuclide content (~73,000 nCi/g), with [ b] Pu-241 as high as 75% of the total mixture for at least one sample. Ni-63 is also a 4-44
Sixton Nucleir Experim:nt:1 Corp 3rsti:n (SNEC) Site Characterization Report l ()
\
consistent contributor to the mix of individual nuclides, reaching as much as 90% of J the mix for one sample. This information is summarized for fifteen SNEC pipe section samples in the table presented below. l Table 4-16, TRU & Ni-63 Content Of Piping System Sample Materials j System nCi/g TRU Grams Ni43 as a % Co40 Cs-137 1 Removed of Total Activity % Activity Removed % ActMty Removed Vents & Drains 1.9 163 61 77 77 MUP Demin. H2O 1 10 3 <1 65 I Purification 1349 10 3 <1 -0 Pressure Control & Relier 213 10 48 17 -0 l shutdown Cooling 1404 11 8 21 64 Pressurt: Control & Re:ier 74 53 89 9 -0 Pressure Control & Relier 35 2 16 43 32 Component Cooling 74 <1 8 -O 30 l Safety Injection 72.800 <1 44 86 65 River Water Cooling .04 134 4 -0 17 i Resin Sluice 29 2 27 -0 -0 S/G surface Blowdown .06 34 23 -0 52 l Vents & Drains 107 69 49 97 92 7' Note: Several samples from the same system are shown in the above table. However, each {s sample was taken at significantly different points within the respective system. i in Table 4-16 above, the Cs-137 and Co-60 values were obtained from extemally scanning the pipe section, both before and after acid leaching the intemal surfaces. The percent activity removed from the leaching process is shown. It is believed that leaching the pipe sections with strong acid removes sample material more evenly and provides more accurate informat;on relating to the relative mix of radionuclides than haphazard scraping operations. All sample data analysis results from sampling SNEC facility systems is provided in the Appendix D. Included with the original analysis reports are the normalized isotopic concentrations (as a percent of total activity) for each sample. 4.2.10 Results Of Piping System Analysis The following summary table presents the curie totals as reported for SNEC facility piping systems. With daughter radionuclides added, this listing is represents greater Q D than 95% of the calculated totalinventory for the SNEC facility piping systems. 4-45
-. . - = . . - . . .
4 Sixten Nucl:ar Experimental Ccrpsraticn (SNEC) Site Characterization Report Table 417, Radionuclide Inventory For SNEC Facility Piping Radionuclide Curies % Of Total Ag-108m 1.83E-03 0.0157 , Am-241 2.88E-01 2.4679 Am-243 2.71E-04 0.0023 Ce-144 7.24E-03 0.0620 i Cm-242 1.86E-03 0.0159 Cm-243/Cm-244 1.34E-02 0.1146 Co-60 1.26E+00 10.8019 ) Cs 134 2.51E-03 0.0215 J Cs-137 1.35E+00 11.6564 C-14 2.75E-02 0.2356 1 Eu-152 2.75E-02 0.2356 Eu 154 3.17E-02 0.2715 ( Eu-155 1.04E-02 0.0890 i Fe-55 1.45E-01 1.2513 . Fe-59 2.16E-05 0.0002 i H3 3.25E-02 0.2783 1129 4.12E-03 0.0353 Nb-94 6.10E-03 0.0523 , Ni-59 5.31E-02 0 4547 , Ni-63 2.77E+00 23.6983 i
/]/ Np-237 Pu-238 Pu-239/Pu-240 1.06E-02 1.15E-01 5.66 E-01 0.0908 0.9815 4.8549 Pu-241 4.52 E+00 38.7335 Pu-242 5.24E-04 0.0045 Ru-106 1.78E-02 0.1527 Sb-125 5.15E-03 0.0441 Sr-90 3.91E-01 3.3503 Tc-99 2.53E-03 0.0217 U-234 3.82E-04 0.0033 U-235 1.32E-04 0.0011 U 238 1.37E-04 0.0012 Totals => 11.67 100 %
Note: Bold text indicates major contributors. l Worth noting, is that almost six (6) curies of TRU materials are calculated to be present in the piping systems. The majority of this material is expected to be in the Safety injection system, Reactor Primary Coolant lines, Shutdown Cooling and Purification system piping. The waste classification for the SNEC facility piping system (collectively), is expected to be Class C waste. However, several pipe sections are greater than Class C individually. O) 5 G 4-46
Saxttn Nuclear Experimental Ccrpor:ti:n (SNEC) Site Characterization Report l / 'i 4.2.11 Sample Data Used in Curie Estimates For Components And Structures The following table presents the assigned sample results for each major component I within the SNEC facility. In addition, the radiological survey number from the l radiation or contamination survey used to evaluate the listed items is provided. Table 4-18, Radiological Samples And Survey Data Used in SNEC Curie Estimates COMPONENT SAXTON SAMPLE No. RADIOLOGICAL SURVEY No. AREA 1 Shutdown Down Cooling Heat Exchanger SX825950042 895-95-0371 Shutdown Down Cooling Pumps SX825950042 224A-910338,0339 Condensate Return Pump SX815950168SD, 89B-95-0652 SX825950038 Condensate Return Unit Tank SX815950168SD, 89B-95-0652 l SX825950038 Pressurizer Discharge Tank SX815950068 810-95-0262 Dtscharge Tank Pumps SX815950068 225A-91-0343 CV Sump Pumps SX815950005 214-95-0647 Storage Well Demn Suction Firer SX825950046 214-95-0648,893-95-0755 Purification Filter SX825950040 241-95 0648 Storage Well Demm. Discharge Filter SX825950046 214 95-0648,893-95-0755
]
yj CCW Pump Filter SX825950040 241 95-0648 Rod Room Air Handler SX815950050 898-96-008 Rod Room Vent Fan SX815950050 218/219-9l-0229,218-91-0274 ControlRod Drives Smear Composite 1 218/219-91-0229,81Z-95-0671 Instrument Rack 1 SX815950006, SX815950007 818-95-0742 Instrument Rack 2 SX815950007, SX815950006 818-95-0742 ) New Fuel Storage Racks Smear Composite 1 214/216-91-0282 l Reactor Head Insulation Ring Smear Composite 1 214/216-91-0280 Pump Leak-off Drain Tank Smear Composite 1 89B-95-0756 Containment Vessel Sump Volume SX815950005 Volumetric Analysis AREA 2 Steam Oenerator SX825950lli,071,053 891-95-0397,0344, 220/221-91 0347 Pressunzer SX825950043 892-95-0381 Main Coolant Pump SX825950111 891-95-0396 Regen Heat Exchanger SX825950040 893-95-0456 Non-Regen Heat Exchanger SX825950040 893-95-0455 Instrument Rack Smear Composite 2/6 82Z-95 0672 Air Handler In Overhead SX825950038 898-95-0757 Pnmary Air Handler SX825950038 898-85-0757,210-9l-0251,0230 t O s 4-47
l 1 l l Sixt:n Nucirr Experim:ntil Ccrp:rati:n (SNEC) ' Site Characterization Report
/ -
( ) l COMPONENT SAXTON SAMPLE No. RADIOLOGICAL SURVEY No. AREA 3 Storage Well Heat Exchanger SX825950N6 897-95-N45,2278-91-0336 Storage Well Pumps SX825950N6 227A 91-0335 Component Coolmg Heat Exchangers SX825950N5 894-95-0670 Component Cooling Storage Tank SX825950N5 83Z-95-0668 4 Component Cool Pumps SX825950045 830-95-0105 Incore Instrument Drives Smear Composite 3/4 206-91 0231,212 91-0216 Auxiliary Air Handler SX835950039 898-95-0757,898-95-0757 Operating Floor Air Handler SX835950039 898-95-0757 AREA 4 Inside Upper Dome Smear Composite 3/4 842-95-088 Steel walls, El. 812'To 818' Smear Composite 3/4 201 91-0149,0189,0191 Steel walls, EL 818' To 845'-2" Smear Composite 3/4 201 91-0217,201/204-91-0197 Equipment Hatch Smear Composite 3/4 201-91-0189 Personnel Hatch Smear Composite 3/4 201-91-0169,0217,202 91-0182 Emergency Escape Hatch Smear Composite 3/4 201-91-0191,84D-95-060 Polar Crane Smear Composite 3/4 203-94 & 23,0385 Movable Refueling Bridge Smear Composite 3/4 204-91-0163,849-95-046 Small Contanunated Pump Smear Composite 2/6 89B-95-0651
,- m Large Contammated Pumps Smear Composite 2/6 89B-95-0651
[ l Electrical Distnbution Box Smear Composite 3/4 201-91-0169,201-91 0192
\ Equipment Hatch Trolley Smear Composite 3/4 84Z-95-0667 l Tool Rack Smear Composite 3/4 2N-91-0179 l Fuel Handling Tools Smear Composite 3/4 Not Surveyed j Teleflex shields (Steel Section) Smear Composite 3/4 201-91 0194 l Fuel Trnsfer Cask Skid Smear Composite 3/4 203 91-0162 l
Head Stand Smear Composite 3/4 204-91 0163,204-91 0179 l Lights Smear Composite 3/4 Not Surveyed l Air Circulatmg Fan Smear Composite 3/4 Not Surveyed l AREA 6 Reactor Vessel Calculated Value TLG 867-95@28, N 34,0228 Reactor VesselInsulation Can Calculated Value TLG 867-95-N28,0434,0228 Reactor Vessel Support Can Calculated Value TLG 867-95-0428,0434,0228 Spent Fuel Rack Smear Composite 6 862-95-0228 Boric Acid Demineralizer Vessel SX865950118 869-95-0429 Storage Well Deneralizer Vessel SX865950ll8 869-95M29 Purification Demmeralizer Vessel SX865950118 869 95-N29 Spent Fuel Pool Elevator Smear Composite 6 Not Surveyed Support Stand Smear Composite 6 Not Surveyed Internals Rigging Fixture Smear Composite 6 862-95-0228 Reactor Vessel Head Lift Rigging Smear Composite 6 Not Surveyed Super lieated Test Equipment Tank SX825950111 & SX865950053 862-95-0427 ( L 4 48
Sr. con Nucl:ar Experiment:1 Corpor;tian (SNEC) l Site Characterization Report l l OTHER MISCELLANEOUS ITEMS SNEC SAMPLE No. SURVEYS OR METHODS
- All Orating (Steel) Smear Composites 1,2,3/4 or 6 Average Smear Results Using
( (All Areas) Surface Area Estimate ! All grating (Aluminum) Smear Composites 1,2,3/4 or 6 Average Smear Results Using (Reactor Compartment) Surface Area Estimate Step Gratmg Smear Composites 1, .2,3/4 or 6 Average Smear Results Using j (All Areas) Surface Area Estimate Site Concretc SNEC Gamma Scans Many Surveys & Core Bore Analysis Site Piping Misc. SNEC Samples Numbers Many Surveys RV Internal Surface Contamination SX865950053, Surface Area Estimate With Sample SX865950165SM Results I j l l t l l l ? l l l i L 'T 4-49
._. _ ._ _ ._ _ _ _-____ - - _ .- _ _ _ _m _ ._
l L I Saxt:n Nucirr ExperimIntIl Ccrper; tion (SNEC) Site Characterization Report At the time of this report preparation, not all of the items in the table above have been evaluated for total activity. However, all major components, surfaces and piping systems have been analyzed. The items that do not have curie estimates prepared as of this reporting date are items such as the CV lighting and other items listed above, still unsurveyed or evaluated for extemal surface area. In all cases, these unevaluated components are surface contaminated objects (SCO's) with l small surface areas representing only uCi quantities of radioactive contamination.
. These items represent less than 1/10 of 1% of the total activity of the SNEC facility.
During decommissioning these items will be evaluated prior to disposal. 4.2.12 Curie Estimates For Components And Equipment l_ The following three pages present the tables originally presented in the "SNEC l Facility Decommissioning Plan" (DP) (Reference 2). These tables have been updated with the latest information with respect to the estimated curie content for components and structural materials. The original data included daughter nuclides such as Ba-137m, Y-90 and Te-125m. etc. The daughter nuclides have been l ! removed from the tables which may now show a slightly different value then original;y i presented. In addition, some round-off errors have been corrected. Concrete activity now includes the estimate for H-3 and other nuclides not assayed for by SNEC personnel. Tne activated concrete materials have been scaled to the hi-flux core l l to detect bore sample (SX862950119). The remainder of the concrete hard nuclides were scaled to a composite of other core bore samples from the SNEC facility. ; Several items within the following tables are not grouped with respect to their waste l classification. These items are listed as "To Be Determined" (TBD). At the time of this report, these items are under evaluation. However, several conservative 4 assumptions were used to estimate the curie loading of these ite is. These
- assumptions provided a higher curie estimate for these components than n. be
- b. necessary. The solution to this problem may be simply re-evaluating the components V
! 4-50 l 4 1
S xt:n Nucl=r Experiment:1 Ccrparitirn (SNEC) Site Characterization Report O V in question to obtain a more realistic curie estimate. The following approach will be l used to re-evaluate these items on an as needed basis:
- 1. A review of the assigned waste stream will be performed.
- 2. Radiation measurements performed around the component (General Area Measurements) will be reviewed to estimate the affects of background radiation levels on the component radiation 'neasurement results.
! 3. Radiation shielding code input will be reviewed and a more realistic shielding code model may be developed. ,
- 4. Additional radiation surveys may be taken to verify existing measurements.
- 5. Additional samples may be collected.
C 6. An engineered solution may be used. Conclusions ! The majority of SNEC facility systems, components and equipment have been
- reviewed with respect to total curies and waste classifications. Problem areas have be;n identified that will be addressed prior to the start of the decommissioning process. "TBD" items will be managed or handled appropriately, within the framework of SNEC procedures or work instructions.
l
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( ) Tcbl -19 Summary Listing of Activities and Waste Clatsifications Page 1 of 3 AVERAGE AVERAGE ACTIVITY uCi's PER CONCENTRATION 100 cm^ 2 ESTIMATED VOLUME IN WASTE PONENT (uCi/cc) FOR 890 CURL _ES_C_UBl_Q_ METERS _ CLASSIFICATION _ l COM_TA A 1 -, .. N3
- t.3 EE0E@iE E Eih 03 M2@s?n E N Ti:te - m i W m 7/ZIMI
~ BHOTDQWWCQoDNGREATExCRAREr 4.05E i 60 N/A 8.63E-01 0.21 TBD SHUT DOWN COOLING PMPS 5.77E-01 N/A 2.45E-01 0.42 Class C C6NDEN5hTETIETURN PUMP 1.02E-03 N/A 3.40E-05 0.03 Class A CONDENSATE RETURN TANK 5.16E-04 N/A 4.57E-05 0.09 Class A DISCHARGE TANK 7.25E-02 N/A 3.36E-01 4.63 Class A DISCHARGE ', ANK PMPS 2.32E-02 N/A 4.80E-03 0.21 Class A SUMP PUMPS 1.43E-01 N/A 3.59E-03 0.03 Class A STORAGE WELL DEMIN. SUCTION FILTER 2.82E-02 N/A 2.40E-03 0.09 Class A PURIFICATION FILTER 1.20E-01 N/A 5.42E-04 0.005 Class A STORAGE WELL DEMIN. DISCHARGE FILTER 5.96E-02 N/A 1.88E-04 0.003 Class A CCW PUMP FILTER 1.29E-01 N/A 2.76E-03 0.02 Class C ROD ROOM AIR HANDLER 4.76E-04 2.3E-02 1.18E-03 2.48 Class A ROD ROOM VENT FAN N/A 2.1 E-02 TBD 0.03 Class A CONTROL ROD DRIVES N/A 1.2E-01 TBD 0.05 Class A INSTRUMENT RACK 1 1.4'.E-02 N/A 1.90E-02 1.35 Class A INSTRUMENT RACK 2 3.91 E-03 N/A 3.69E-03 1.23 Class A NEW FUEL STORAGE RACKS N/A 3.7E-02 TBD 5.05 Class A REACT ~R TIEADlNDUffTi6N RlNG N/A 3.6E-01 TBD 5.45 Class A PUMP LEAKOtLDRAIN TANK 1.45E-02 N/A 3.43E-03 0.24 Class A CO_NJA!NMENT VESSEL S_ UMP VOLUME 3 95E-02 N/A 2 09E--02 0.53 Class A AREA 2 I #'^ '
- !? . 3' >' ~h 's M' E Y '
STEAiWGENERATbR S 23G+00 N/A 3 72E+01 7.f2 ClasiiFC PRESSURIZER 3.84E+00 N/A 2.46E+01 6.41 Class C MAIN COOLANT PUMP 6.20E-02 N/A 3.28E-02 0.53 Class A REGEN HEAT EXCHANGER 3.73E+00 N/A 3.02E + 00 0.81 TBD NON-REGEN HEAT EXCHANGER 2.08E-01 N/A 4.85E-02 0.23 Class C INSTRUTENT_ RACK 2.79E-03 N/A 6.47E-03 2.32 Class A AIR HANDLER IN OVERHEAD (see Primary)_ 1.39E-03 1.1 E+00 4.99E-03 3.60 Class A PRIMARY AIR HANDl.ER l 1.77E-03 1.1 E+00 6.39E-03 3 60 Class A L TBD - To Be Determined At A Later Time. SCO - Surface Contamina'ed Object. REVO 4-52
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Table 4-10 Summary Listing of Activities and Waste Classifications Page 2 o. . AVERAGE AVERAGE ACTIVITY uCI's PER CONCENTRATION 100 cm^ 2 ESTIMATED WASTE CO_MP_ONENT (uCi/cc) ISCO) CURIES CUBIC METERS CLASSIFICATION AREA 3 =mm- , mL&, ' m mmmmmmggynygg
~-
STORAGE WELL HtIi EXCHANGER 2.88E-02 PT/A 2.89E-03 0.10 Class A STORAGE WELL PUMPS 2.91E-03 N/A 1.45E-03 0.50 Class A COMPONENT COOLING HEAT EXCHANGER 8.93E-03 N/A 8.94E-03 1.00 Class A
- COMPONENT COOUNG STORAGE TANK 1.25E-02 N/A 1.24E-03 0.10 Class A COMPONENT COOUNG PUMPS 8.15E-03 N/A 3.46E-03 0.42 Class A IRC655iS5T5Uid$AfDillVES N/A 1.8E-01 TBD 0.M Class A A_UXILIARY AIR HANDLER 4.00E-04 3.6E-01 2.96E-04 0.74 Class A OPERATING FLOOR AIR HANDLER 4.81E-04 6.3E-01 2.70E-03 5.62 Class A AVERAGE AVERAGE ACTIVITY uCi's PER CONCENTRATION 100 cm^ 2 ESTIMATED VOLUME IN WASTE CO. MPONENT (uC. i/cc) (SCO) CURIES CUDIC METERS CLASSIFICATION
. _ . - . . . .my ,,,m. .. g .
pgg , INSIDE UPPERD-QUE N/A 4.5ET06 9.62E 05 418 Class A STEELWALLS, EL 812' TO 818' N/A 1.2E-06 4.36E-06 0.77 Class A STEELTATALLS, El. 018' TO 845-2" N/A 1.3E-06 2.96E-05 3.47 Class A EQUIPMENT HATCH N/A 1.3E-06 2.36E-07 3.29 Class A PERSONNE_L HATCH N/A 2.4E-06 9.21 E-07 8.55 Class A EMERGENCY ESCAPE HATCH N/A 7.7E-06 8.33E-07 1.56 Class A POLAR CRANE N/A 5.9E-06 6.33E-06 113.27 Class A REFUEUHdfDMIDGE N/A 3.9E-06 5.98E-Of 0.68 Class A SMALL CONTAMINATED PUMP 1.09E-03 N/A 4.24E-05 0.04 Class A LARGE CONTAMINATED PUMPS 4.08E-04 N/A 1.31 E-04 0.32 Class A ELECTRICAL DISTRIBUTION BOX N/A 1.6E-02 TBD 0.91 Class A EQU'PMENT HATCH TROLLEY N/A 4.2E-02 TBD 4.58 Class A TOOLRACK N/A 8.4E-02 TBD 0.19 Class A FUEL HANDUNG TOOLS N/A 8.4E-02 TBD 0.10 Class A TELEFLEX SHIELD (Steel Section) N/A 1.5E-03 TBD 4.25 Class A FUEL TRANSFER CASK SKID N/A 1.8E-03 TBD 2.98 Class A
- - ~ HEAD STAND N/A 1.6E-02 TBD ~~ 1.16 Class A ~
S~ ~ ' N/A 8.7ET02' TBD 0.70' Class A AIR CIRCULATING FAN N/A 5.0E-01 TBD 0.51 Class A TBD - To Be Determined At A Later Time. [SCO - Surface Contaminated Object. REVO 4-53
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Table 4-10 ()' I V) . Summary Listing of Activities and Waste Classifications Page 3 of 3 AVERAGE AVERAGE ACTIVITY uCi's PER CONCENTRATION 100 cm^ 2 ESTIMATED VOLUME IN WASTE COMPONENT (uCiLcc) (SCO) CURL _ES CUBIC __ METERS _CLASSIF_ICARON_. ARFA6 4*Mewawrawemmeasuma--- -;meh.s_e - RDCTDTWEB~SA 1.13E+D2 N/A 1.45E+03 12.85 Class C REACTOR VESSELINSULATION CAN 1.21E-01 N/A 1.08E+00 8.95 Class A R54CTdRV$55$L 50#1@RTCAN 1.66E-01 N/A 3.03E+00 18.26 Class A SPENT FUEL RACK 4.09E-04 6.0E-01 4.60E-03 11.25 Class A BORIC ACID DEMINERALIZER VESSEL 3.80E+01 N/A 1.59E+01 __ 0.42 TBD STORAGE WELL DEMINERAltZER VESSEL 5.65E+00 N/A 8.03E-01 0.16 TBD PURIFICATION DEMINERAUZER VESSEL - 1.04E+00 N/A 4.35E-01 0.42 TBD 5PCNT FUs[~PDDLEL5V4TUR N/A 6.8E-01 TBD 2.72 Class A SUPPORT STA.ND N/A 6.8E-01 TBD 0.45 Class A INTERNALS RIGGING FIXTURE N/A 6.8E-01 TBD 0.51 Class A REACTOR VESSEL HEAD UFT RIGGING N/A 6.8E-01 TBD 0.51 Class A ZsOP55Ill%i$Dj5T EQUIPhi5Si_~ TANK 1.79E-02 N[A 6.365 -03 0.36 Class A 111111]~Il1111111111I111
-~ IlllLIllHillllHillllllllillllllll11111111111111111111111l11111111!IlllIlullllklimmillmt[IIIMlI[IlllMll mmimmillimmmt 'ImimimmmlimilMI GMTINGfall areas) ~ ': o' _ ~~ m: ,, ,. 0.004 agging
.a~? gag @ E TNs@s
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^
- - k J ' . 'J s/ M ^3/Fr ^2
~'?^"?'? ~"}j 'vs i ,23i[i' -
""'"~'#^C'"""(Steel NT^q Only)T*% m ~ 2.9Ea~o1 - 837E-03 5WL Class A e ^~ c ? * ' >
^
v' - 'i _ Average __ Total Total M ^3 4M#4.akikWM GRATING (Rx Compartment) - m: v n ". ' , , ,,G - E s.M: 0.004
$^ C , ' J- ~ ' '
s
<s " ' ? ' , 2 ': < 'C-M ^&Tt ^2 IW$$iM$ ggs ig$$g#@$y$ifiT$$
! ',"'Ty^^P'(Afuminum """':p?'. , ' ' "
; - ' ,x: .: Only) 'T~"*7
,'j q ' _,<. ,1 >p't i kj,i 8 SEZO11 7 62E-63 1.56i Class A
- " - Average l Total Total M ^3 \ ""'- A ' 4 '.t . ~
STEP GRATING (all areas) (Steel Only)
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Average Total Total M ^3
- ~
MsL e *
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~~~~~ SiTG CONCRETG ~~ ~-' 1.546; 03 - C 3 72E-01 30 M Class A SITE PtPING 168E+00 " ^ c 1.17E+01 7.0 Class A& C INSULATION TBD ' ' > TBD 17.2 Class A PROCESS WASTE GENERATED TBD , -a included TBD TBD RV INTEP_NAL_ SURFACE CONTAMINATION N/A '- '
1.13E+01 N/A Included With RV TOTALS ~ K s '~- - ' < > <' 1559 604~
; 'a , '? 7 TBD - To Be Determsd A%tteTTme.
SCO - Surface Contaminated Object. Note: The Vohrme Estmate Takes No Credt ForDecontambation, Vohme RedJction OrPackaghg Arrangements. REV0 4-54
I l i ! l l S xtan Nucle r Experimental Corporation (SNEC) j Site Characterization Report I
~
l [ ') \ t s
'j 1 The summarized results of the component and system curie estimate, with the RV radionuclides removed are shown below. l Table 4-20, Radionuclide inventory For SNEC Facility Components & Systems Radionuclide Curies % Of Total Ag-108m 3.65E-02 0.0381 j Ag-110m 4.95E-04 0.0005 Am 241 1.63E+00 1.6982 j l
Am-243 3.37E-03 0.0035 Ce-144 1.78E-01 0.1860 ) Cm-242 1.79E-02 0.0187 ) Cm-243/Cm-244 5.88E-02 0.0613 i Co-60 7.42E+00 7.7476
]
Cs-134 4.38E-02 0.0457 1 Cs-137 8.22E+00 8.5788 l C-14 6.21E+00 6.4861 l Eu-152 3.66E-01 0.3823 Eu-154 1.80E-01 0 '875 , Eu-155 7.95E-02 0.0830 , l p Fe-55 1.15E+00 1.2022 l ( Fe-59 7.65E-04 0.0008
l H-3 2.74E+00 2.5733 I I-129 2.10E-01 0.2192 Mn-54 2.95E-04 0.0003 Nb-94 1.15E+00 1.1984 Ni-59 6.54E-01 0.6823 i Ni-63 3.98E+01 41.5772 Np-237 1.34E-01 0.1395
- Pu-238 5.65E-01 0.5896 Pu-239/Pu-240 2.74E+00 2.8595 Pu-241 2.01E+01 20.9517 Pu-242 6.33 E-03 0.0066 Ru-106 3.62E-01 0.3777 Sb-125 1.llE-01 0.1158 Sr-90 1.74E+00 1.8165 Tc-99 8.60E-02 0.0897 l U-234 4.05E-02 0.0422 U-235 4.40E-03 0.0046 U-238 3.32E-02 0.0347
; Zn-65 7.13E-04 0.0007 Totals => 95.82 100 %
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v ! 4-55 I I l
S xt:n Nuclear Experim3ntil Corp: ration (SNEC) Site Characterization Report 4.3 ACTIVATION OF SNEC REACTOR VESSEL AND SURROUNDING ! STRUCTURES This section provides a summary of the results from the activation analysis of the SNEC facility reactor vessel, reactor internals, reactor compartment structural supports and surrounding concrete. Radiation measurements and sample results j supporting this effort are provided. In addition, an estimate of the amount of surface I contamination present on the SNEC facility reactor vessel internals is included. The neutron-induced radioactivity and surface contamination levels of these components are then used to assist in determining their 10 CFR Part 61 waste classifications and subsequent Low Level Radioactive Waste (LLRW) volumes to be disposed of during decommissioning. Additionally, these analyses provide the data required to assist l ALARA planning for work activities such as reactor vessel removal. ! The activation analysis of the SNEC reactor vessel was performed by TLG Services5 at the request of GPU Nuclear Corporation. The analysis scenario considered was a l (] final shutdown of the facility on April 2,1972, after three (3) cycles of both power and experimental operation, with an ensuing decay period of 24 years to July 1,1996. l l l The information included in + section is taken directly from the analysis report and other supporting survey data supplied by GPU Nuclear (Reference 4). The survey information is used to confirm and validate the ana;ytical effort, and provides assurance that a solid basis exists for presenting the Reactor Vessel and surrounding structures as acceptable for disposal at cornmercial buriel facilities. l This information is the result of efforts by GPU Nuclear and SNEC management personnel to provide reasor able and representative information about the SNEC J Reactor Vessel and its surroundings. l 4.3.1 Data Needs l The following characterization information was required by the SNEC SCP: 5 TLG Services, Bridgewater, CT l 4-56
Sixt:n Nuclear Experimental Corporation (SNEC) Site Characterization Report O From Activation Analysis: ; Provide the curies per gram radionuclide concentrations for all RV intemals and the reactor vessel. Provide LLRW classifications and waste volume estimates. From Bulk Sample Analysis As Required By The Site Characterization Plan: Obtain primary system component (s) samples representative of the reactor and ! l reactor primary system. I i Gamma Radiation Surveys As Required By SCP: i In the Reactor compartment, extemal to the water seal can, survey a minimum of four (4) locations,90 degrees apart, at the core centerline. Survey primary system j piping, as best shielded from other nearby sources, using a reproducible geometry that may be input to a point kemel radiation shielding computer code for source term j confirmation calculations. j 4.3.2 Applicable Areas and items D The Reactor Support Structure, Insulation Can, Water Seal Can, surrounding concrete and structural materials all require characterization. The neutron activation 1 t analysis was pe.rformed to support SNEC facility site characterization of the reactor pressure vessel and intemals. This analysis models these components. l 4.3.3 Analytical Methods The Saxton Nuclear Experimental Corporation has enlisted the support of TLG Services, to perform the activation analysis and thereby estimate the neutron-induced radioactivity contained in the reactor pressure vessel and intemals. The activation analysis was performed using a one-dimensional neutron transport computer code (ANISN) to calculate neutron flux levels radially and axially extemal { to the reactor core. A point neutron activation and depletion computer code i (ORIGEN2) was used to determine the total radioactivity induced in the irradiated components. Input to this analysis includes: i
. /" ; 1. Plant operating history and total effective full power hours.
4 ( 4-57 { i t
4 Sott:n Nucle:r Experimintal Ctrporttirn (SNEC) Site Characterization Report v
- 2. Plant operation parameters.
- 3. Nuclear fuel characteristics and use, including fuel bum-up.
- 4. Geometry and material composition of the reactor pressure vessel, intemais, and surrounding components and concrete structures.
Where plant specific information was not available, reasonable and conservative assumptions were used. Available literature was researched and referenced as appropriate. From the above input, a one-dimensional neutron transport computer model of the reactor intemals, vessel wall and irradiated portions of the support and concrete structures was created. Radial and axial neutron fluxes at full reactor power were subsequently calculated. Point neutron activation and depletion calculations were
. L/ run for each component at each location of interest, using the neutronics data generated and the plant operating history. Output from these calculations is in curies per gram for radionuclide concentrations, which is required for the determination of 10 CFR Part 61 waste classifications and LLRW volumes estimates. Additionally, this data may be used to estimate personnel exposure rates during other decommissicning activities.
Verification of the activation analysis has been performed by comparing exposure rate measurements to calculated values based on predicted neutron activation isotopic distributions. 4.3.4 Confirmatory Survey Data Several- confirmation surveys have been performed by SNEC/GPU Nuclear personnel. As showil below:
- g k
4-58
Sixt:n Nuclear Experirnentil Csrp;rati:n (SNEC) Site Characterization Report v + 4 strings of thermoluminescent dosimeters (TLD's) were hung in four , quadrants of the SNEC reactor (external to the vessel),90 degrees apart, and at known distances from the RV horizontal centerline. l The following figure presents the results of the TLD study. The TLD's were placed at , one foot intervals, starting from one foot below the reactor head platform (~793' El.). l f SNEC TLD STRINGS - EXTERNAL TO VESSEL l one string Per ouadrant 400 e String 1 4 NW8HE
+ "92 300 ouadrants h
- String 3 e String 4
(- !- 100 n,- . _t- - . . O ' ' ' ' ' ' ' ' ' ' ' ' ' ' ' ' ' ' ' ' ' 792.17 790.17 788.17 786.17 784.17 782.17 780.17 Distance From Platform At Elevation 793.17'(In Feet) l Figure 4-6 The TLD information presented above, shows the asymmetric core exposure rate i profiles. Depressions in the exposure rate values at the center of strings 1 & 4 are believed to be due to obstructions that shield the TLD's from the reactor vessel. The obstructions are structural supports that once retained the ex-core instrumentation monitors at core centerline. Strings 2 & 3 show reduced gamma exposure rate ! profiles due to non-uniform core loading. The distance from the TLD strings to the core horizontal centerline was between 74* and 76". O O 4-59
S:xt:n Nucle r Experimental Ctrporati:n (SNEC) Site Characterization Report N.)\
- A single TLD string was inserted into the RV, at near the center of the core (horizontally & vertically).
SNEC Facility RV Exposure Rate -Internal to RV in Feet Above & Below Core Centerline 1378 R/h Max.
.5 rw etcm i - N 3 Instnament Port
, , . _._. s o
O
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o O ,,,,,,,,,,,,,,,,
'o 200' a 600 800 20 1200 in R/h
( Figure 4-7 TLD String Placement Internal To SNEC RV At Core Centerline. l Figure 4-7 above shows the maximum possible exposure rate at the core centerline. This value is probably overstated (from 10 to 20%), because of a problem encountered during the removal of the TLD assembly. The calculated value from the activation analysis for the same location is ~1098 R/h (TLG). Radiation measurements were performed at the perimeter of the support structure on contact with the support can and at selected distances around the core centerline. b O 4-60
! Sixt:n Nucle:r Experimental C:,rp:r;ti:n (SNEC) l Site Characterization Report
\ l l Radiological controls personnel surveyed the reactor vessel exterior. The following data was assembled from radiological survey information.
NOIEH Top View TLD STRING 1 4,000 mR/h N 4 TLD STRING 4 A / ,400 mR/h
\ 5,000 mR/b 4 y ~p /
l
/ ! N 2,700 mR/h N '
/ ','
4 s y 2,400 mR/h
/
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\
i Y $hy; . \ [ - - -- M$ahqB
- m. ~ -- --- O\$3ggg---
g > M mR/h / k i
/ N W mR/h x l N-3 Instrument ort 400 mR/h ,,, Y \ y _/
3,000 mR/h
-N N j TLD STRING 2 (( TLD STRING 3 k Highest Contact Measurements 500 mR/h Core Centerline SNEC Radiological Controls 800 mR/h 891-95 4426 SNBCREACTOR VRWRL i 0
867-95 4434 867-9544*8 Figure 4-8 Radiological survey information confirm that the highest exposure rate region around the vesselis located in the north east quadrant, at the vertical centerline of the core region. This inforrrtation supports the activation analysis and augments the results from the ex-vessel TLD measurements. Surface Contamination Analysis: The external surface contamination present on the Reactor Vessel and structures 2 extemal to the reactor vessel is typical of AREA 6 surfsce contamination (see Section 4.1). The Reactor Vessel internal surface contamination has been estimated
/]
v 4-61
S:xt::n Nucle:r Experimental Corporation (SNEC) Site Characterization Report 1 from contamination levels present in safety injection piping between an installed check valve and the Reactor Vessel. The estimated intemal surface area of the l 8 SNEC Reactor Vessel is ~1.13X10 square centimeters. The safety injection piping 2 is estimated to contain approximately 10.4 uCi/cm of ntemal contamination. The [ current estimate of ~11.8 curies of RV intemal contamination assumes that: j 1. No significant quantities of fuel related materials remain intemal to the RV and, ! 2. Only a surface coating of materials similar in composition and activity per unit area to the materials present in the SI system remain within the RV. ; 1 b l i items one and two above, have been confirmed by reactor vessel intemal video l inspection during February 1996. No significant deposits of debris were found during j the inspection effort that included a ~360 degree scan of the near bottom region of 4 inside of the vessel. In addition, the intemals of the reactor vessel were observed to be mostly clear of surface deposits with most metallic surfaces appearing unsoiled ! and effulgent. Item two above, is believed to be a reasonable assumption since a j sample from the reactor vessel retained similar isotopic concentrations as a sample from the safety injection system, and intemal surfaces appear to be relatively clear. Activated Concrete and Structures Eight (total) concrete core bore samples were obtained from wall surfaces above and below the reactor compartment flood level. These samples were obtained by penetrating the five (5) foot wall separating the reactor compartment from the steam generator compartment and from the one and one-half (1.5) foot thick CV wall. Concrete corc bores were also obtained from the reactor shield plugs above the RV, which form the operating floor on elevation 812', and from below the reactc,r (control rod room floor) on elevation 765'. The analyses results from these activated concrete core bores were compared, and benchmarked to the predicted activation anslysis results for concrete. Review of this data indicates that there is little or no activation of 4-62
1
- I Srxton Nucle r Experimental Ccrporati
- :n (SNEC) l Site Characterization Report concrete below the operational water line. The concrete above the operational water l line (~791' El.), is neutron activated.
4 I The following core bore profiles have been assembled from samples from this region. However, as discussed in section 4.1.17, the removal of the majority of these concrete walls above the water line (around the vessel) is expected, making this information important to the waste packager and shipper during dismantlement. I 4 CORE SAMPLE SXB22950037 (AREA 2),-8* 3" EAST OF HOTLEG, ABOVE SW WATER LINE 1D00 25.0 ]
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Pnmary Compartmert Sde j 1 00 0.0 13 29 4.5 61 #2 35 3 250 120 12 2 127 S7 144 14 9 15 1 15 2 Depth in Centimeters From Starting Point. Through. Wall Core Figure 4-9 The above chart demonstrates increased activation product concentrations (Co-60 and Eu-152), from <MDA values (on left), to greater than MDA values nearer the j reactor vessel. The coring operation started from the primary compartment (on left). The following chart shows concrete activation in what is believed to be the high flux area of the containment vessel adjacent to the reactor vessel. The core bore results i suggest that near complete removal of the CV interior concrete containment wall in this location may be necessary. I f t l l 4-63
i a S:xt:n Nuclear ExperimIntil Ct rpurati::n (SNEC) Site Characterization Report {
\
i 's i REACTOR CAVITY HIGH NEUTRON FLUX REGION (AREA 6)
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, l: n aciaremnymaswince DE PT H (cm.) hinytenime ciar Q Figure 4-10 l l An analysis by B&W of SNEC concrete core bore sample SX862950119, taken from O 4 sy ' the suspected high flux area of the CV containment wall, has provided the ! 1
- information presented in Table 4-21. This is a composite result for slice numbers 1, j 2,6,9 & 13. Slice number 6 contained a 57.74 gram sample of steel reinforcement material (rebar). Note that when taken together, the largest percentage of activity is
- provided by H-3 and Ni-63.
3 4 N f J 4-64
Saxt:n Nucl:ar Experlm:ntal Ccrp:r:ti2n (SNEC) Site Characterization Report v Table 4-21, Composite Results - Concrete Core Bore Sample SX862950119. Isotope As % Of Total Activity Ag-108m 0.026o j Am-241 0.0464 Ce-144 0.0924 Co-60 2.4443 Cs-lM 0.0213 ; 1 Cs-137 7.5581 l C-14 2.3263 Eu-152 3.5327 Eu-154 0.1996 Eu-155 0.0489 H-3 63.5044 I-129 0.4071 Nb-94 0.0204 S Ni-59 0.7042 i ) Ni-63 18.6422
%/ 1 Ru 106 0.15.
Sb-125 0.0506 2 Tc-99 0.2206 Note: Isotopes listed above in bold text were positively identified. All other results are *Less Than' values. All radionuclide concentrations (in pCi/g) for concrete core bore samples have been presente,d in Section 4.5 (Deta Tables). j TLG Services believes that up to 15 inches of concrete materials will need to be removed from the walls around the reactor vessel above the operating level water i line. This is reasonably consistent with SNEC sampling results. Additional samples l will be needed during decommissioning to minimize the amount of LLRW generated. l The following is a listing of core bore samples taken to determine the presence of j I activation products. All sample results are presented in Sections 4.1 and 4.6. O 4-65
Sixt:n Nuclur Experimental Carp:r:ti:n (SNEC) Site Characterization Report O
\'j Table 4-22, Core Bore Samples Taken To Determine Neutron Activation SNEC Sample Number Location Length (cm)
SX811950025 CV Basement Floor, Below RV la Rod Room 36 SX822950031 nrough Pnmary Compt. Wall (S), Below Ops. Water Level 152.4 SX822950036 nrough Pnmary Compt. Wall (S), Below Ops. Water Level 152.4 SX822950037 Through Pnmary Compt. Wall (S), Above Ops. Water Level 152.4 SX862950062 5' Thick South Wall, RV Head Area (SW), Above Water Level 13.8 SX862950063 5' Thick South Wall, RV Head Area (SE), Above Water Level 8.8 i SX862950064 1.5' Thick CV Wa'l (NW),-Core Centerline, Below Water IAvel 15.2 ] SX862950065 1.5* Thick CV Wall (N),-Core Centerline, Below Water Level 14.1 ~ SX861950066 Floor At Base Of RV Support Structure ~779'-8"El. (NE) 11.5 SX841950120 Shield Block 1, Above RV 137.3 SX841950121 Shield Block 2, Above RV 134.3 SX841950122 Shield Block 3, Above RV 135.3 SX841950123 Shield Block 4, Above RV 124 SX841950124 Shield Block 5, Above RV 131.4 SX841950125 Shield Block 6, Above RV 152.4 Steel Structures Steel samples from the reactor vessel area were taken to aid in benchmarking the G activation analysis. The results for these samples indicate the following specific _. activities for Co-60. Table 4-23, Metal activation Samples From SNEC RV Region Item SNEC Sample No B&W Reported TLG Calculated Description uCi/g-Co-60 uCi/g-Co-60 SI Pipe Section (SW) SX865950053 1.43E-04 7.9E-05 (Above Water Level) SI Pipe Section (East) SX865950069 1.82E-04 2.7E-05 (Above Water Level) RV steel Support SX866950116 6.08E-04 1.17E-03 (Below Water Level) v 4-66
Saxt:n Nuclear Experimental Carparati:n (SNEC) Site Characterization Report
/ 'i U Activation analysis results of materials from below the water line by the vessel, appear to be reasonably close to calculated values for selected isotopes. Above the operational water line the activation of materials is higher than expected. These higher than calculated values are believed to be caused by neutron streaming. This is a reasonable explanation for the disparity.
4.3.5 Activation Analysis Results Summary ' The analysis indicates that approximately 1,452 curies of neutron activation products, mainly Fe-55 (37.06 Cl), Co-60 (594.5 Ci) and Ni-63 (810.9 Ci), will be present in the reactor vessel wall and clad, internals, insulation can and support can, 24 years after shutdown (July 1,1996). This estimate does not include the activity from intemal and external surface contamination, which is a small percentage of the total activity for this component. The following table presents the curie estimate for the SNEC reactor vessel and intemais. Based on the analysis report the SNEC ' facility reactor vessel may be shipped as a Class C waste material. Materials such as
' lead, structural steel and concrete materials outside the vessel will be Class A radioactive waste. The activation analysis report provides specific details about selected radionuclide distributions and components within the vessel or exterior structural materials (calculated results), and will be used to assess attemate handling methods.
/~ (w 4-67
I Saxt:n Nucirr Experimental Corporation (SNEC) Site Characterization Report { (- Table 4-24, Reactor Vessel Curie Estimate-Calculated Values a Components Weight in abs. Curies 24 Year j _ After Shutdown j Core Baffle 1,210 642.6 j Thermal Shield 10,491 132.7 Reactor Vessel Clad 2,043 4.185
- Reactor Vessel Wall 87,313 4.737 RVInsulation Can (1) 4,994 1.078
- RV Support Can(1) 5,442 3.034 Lower Core Guide Blocks 152 182.9 Lower Core Plate 500 226.7 lower Support Shroud Tubes (2) 338 8.323 Lower Support Tic Rods (2) 54 2.113 i Balance of Lower Support Assembly (3) 706 0.0744 Lower Core Barrel (4) 1,067 58.46 i Upper Core Plate 264 165 Upper Core Barrel (4) 454 19.71
{ Balance of Lower Core Barrel Assembly (5) 1,823 0.2593
- Balance of Upper Core Barrel Assembly (6) 832 0.1183 i f Q Notes: (1) includes weight of compor.ent between 3' above and 3' below the active core, j (2) includes weight of component down to 3' below the active core.
(3) Weight of lower support shroud tubes, lower support tie 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 associated components more than 3' above the active core. (6) Weight of upper core barrel and associated components more than 3' above the active core. 4.3.6 Reactor Vessel Internal Surface Materials Sampling Results A sample of the intemal surface deposited materials was collected on a swab. The swab sampb was taken of surface materials present in the RV interior, at the bottom of the vessel. The swab was attached to a iong 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 between the vessel and installed check valve. Both of these materials are of similar 4-68
Saxt:n NucI:ar Experim:ntal Carp:ratirn (SNEC) Site Characterization Report composition. The safety injection materials are similar in relative radionuclide 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 for these two samples are presented ln Appendix D. The following table contains the positive values from both B&W analysis rasults for these two sample materials, except where marked with an asterisk ("les s than values"). The internal surface area of the reactor vessel was determined frcm 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, Radionuclide Composition Of St And RV Sample Materials RV SAMPLE SI SAMPLE 1 RADIONUCLIDE % OF TOTAL % OF TOTAL H3 0.0858 0.0040* I \ Fe-55 0.1328 2.5318 i / D Ni-59 0.5477 0.5477 Ni43 45.0849 43.8136 Sr-90 4.1387 0.2839 Nb.94 0.0145' O.0079 Tc-99 0.0033 0.0003* Pu-238 0.5548 1.0905 Pu-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 0.0682 0.0462' U-234 0.0002 0.0113* Co 60 16.8892 17.1988 Cs-137 4.1739 0.3987 Eu-154 0.1407 0.0319 Eu-154 0.0386 0.0082 Note: Bold text indicates pos:tively identified nuclides t 4-69
. . . . ~ ..
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)
Sixtsn Nucle:r ExpIrimsntal Corpsr:tisn (SNEC) 1 1 Site Characterization Report
- I l
l
\
4.3.7 Conclusions 1 The TLG Services supplied reactor vessel activation analysis was reviewed by GPU Nuclear personnel familiar with analysis of these type. GPU Nuclear believes the analysis assumptions and approach used were appropriate and correctly applied. ! Field data largely support the calculated results from the activation analysis. For j l i j structural materials above the operational water line (~791' El.), by the reactor !, vessel, calculated results differ from field measurements. This can be explained by j neutron streaming which is difficult to model with one dimensional neutron transport l analysis techniques. Additional field measurements during decommissioning will be l 4
- used to monitor removal and class 1y structural materials, i I
l Activation of lead materials from near the RV is calculated to be about ~50 uCi/Kg. l ! Neutron irradiated lead will be Class A waste materials. { l I i i* l 3 l i 1 i ! i 4 1 i l t 4 70 1 1
S xt::n Nucle:r Experimental Ccrp ratisn (SNEC) Site Characterization Report , 4.4 ENVIRONMENTAL & HAZARDOUS MATERIALS CHARACTERIZATION , Environmental characterization activities have determined the radiological characteristics of potential contamination in soil and other materials around and
; within the SNEC site, not covered previously. The scope of these characterization activities are in addition to the work performed in accordance with the SNEC Soil Characterization Plan (Reference 9). This section of this report also provides information on principal hazardous materials concerns and includes the l methodology for sampling for the presence of materials which would affect the health and safety requirements of the work force during decomrnissioning (i.e. asbestos, l lead, or other hazardous materials).
4.4.1 Data Needs l The following information is addressed within in this section: Determinations of the contaminant activity levels in the soil. l Determinations of the isotopic identity and concentration ratio of the l O (gi radionuclides that comprise any radioactive contamination. l l l Data for and estimates of the total radionuclide inventory in the soil. l Comparison of characterization data with applicable release criteria to determine decommissioning work scope. The location and type of asbestos used as thermal system insulation l The location and type of asbestos used in surfacing material (e.g. coatings or coverings) A Total Metals Analysis to determine the presence and concentration ratio of Lead, Chromium, and Cadmium metal contaminants in painted surfaces Complete TCLP, inorganic analysis on sediment and sludge samples. O O 4-71
i i Saxton Nuclear Experimental Corporation (SNEC) , Site Characterization Report i f ( A 4.4.2 Selection Of Sampling Locations For the initial soil characterization work conducted in 1993 and 1994 the SNEC site was divided into two areas as follows: 1) Affected Area (biased) - The 1.1 acre area
- comprising the area surrounding the CV, yard area and the Westinghouse area
- 2)
Unaffected Area (unbiased) - The 9.6 acre area which is fenced in around the electrical switchyard and building still in use by Pennsylvan?a Electric Company. Based upon the methods in NUREG/CR-5849 (Reference 13) the affected area was surveyed into 10 x 10 meter grid systems where systematic scanning surveys (radiological) and grab sampling were conducted. Random surveys were conducted in the unaffected area. Locations of well monitoring, air and sediment sampling were determined based on past demolition work, soil remediation work, direction of groundwater flow, liquid offsite release points and environmental sampling program requirements. p ( Hazardous materials sample locations within the SNEC CV, have been selected for characterization based on experience that the item or location may be contaminated with asbestos or hazardous metal contaminants (TCLP, Inorganic Analysis). These areas include:
- Insulation on piping, heat exchangers, and other process components (asbestos)
- Coatings on installed equipment and structural surfaces (metal contaminants, composite samples etc.)
- Intemal surfaces of selected systems known to have contained metal contaminants (composite samples)
Coatings on pools or liners (asbestos) v 4 72
Saxt:n Nuclear Experimental Corp:rati:n (SNEC) Site Characteriration Report Components likely to contain PCB contaminants (e.g., liquid filled electrical l transformer, or lighting ballast's). 4.4.3 Soil Characterization Reports Saxton Nuclear Experimental Corporation conducted soil characterization and installed ground water monitoring wells at the SNEC Facility as part of the decontamination and decommissioning efforts associated with the Control and Auxiliary (C&A) Building, the Radioactive Waste Disposal Facility (RWDF), and Drum Storage Bunker, Undergroun' nt Resin Storage Tank, and interconnecting tur.cels. Documents which report on wn characterization include: SNEC Soil Characterization Plan, Document 6575-PLN-4724.05.
" Report on Drilling and Radiometric Analysis of Samples Collected at Sites of Spent Resi:1 ond Liquid Waste Tanks", SNEC Facility, Saxton, PA, prepared by Pennsylvania State University, January 16,1989.
p -
" Geologic, Chemical, Radiometric and Geotechnical Studies of Samples from V Eleven Drill Holes in Surficial Materials", Saxton Nuclear Facility, Saxton, Pennsylvania, prepared by Peru ylvania State University, December 20, 1988.
"An Aerial Radiological Survey of the Saxton Nuclear Experimental Corporation Facility and Surrounding Area", Saxton, Pennsylvania, prepared by EG&G Energy Measurements, July 1989.
4.4.4 Environmental Characterization Of SNEC Soils In November 1993, soil characterization was performed at the SNEC site to determine the contaminant activity levels in the soil. Subsequent analyses were performed during soil remediation work between July and October 1994. Survey procedures and methods to assess soil radioactivity and hazardous material content were based upon NUREG/CR-5849 (Reference 13) and 40 CFR 261 (Reference 17). Soil characterization and disposal work are described respectively in procedures 6575-PLN-4742.05, "SNEC Soil Characterization Plan" (Reference 38) and 6575-PLN-4542.07,"SNEC Soil Disposal Plan" (Reference 50). Results of all the analytical m 4-73
f l l ! Sixt:n Nucle:r Experimental Carpor;tian (SNEC) l Site Characterization Report
\-
data are descriced in the "1994 Saxton Soil Remediation Project Report" (Reference !
- 9) dated May 11,1995.
On average, as a result of the soil remediation work, the predominant nuclides ! remaining in the SNEC soil are Cs-137 and Co-60 at concentrations less than 1.0 pCi/g and 0.1 pCi/g respectively. Although 56,266 ft' (11 millicuries) of the soil was disposed of at Envirocare and Bamwell, some residual Cs-137 still remains in areas described in the soil remediation report. Most of these are confined to grids (G-9 and l F-9), located outside and immediately north and northeast of ths containment vessel. These grids contain localized hotspots near or in remnants of old structures and piping buried underground. Appendix 1 and Figure 3 of the soil remediation report provides the pre and post Cs-137 concentrations in the SNEC Facility soil. It is estimated for soil that approximately 352 ft' containing 5.4 millicuries of Cs-137 and 27 ft* containing 0.026 millicuries of Co-60 will need to be removed from these grid
/'7 locations to meet the currently proposed NRC cleanup standards. Attainment of average soil concentrations of <2 pCilg Cs-137 and <0.5 pCilg Co-60 will meet the residential scenario ALARA dose of 3 mrem guideline per year in accordance with NUREG 1500.
4.4.5 Applicable Contaminants and Sampling Media in Soll A hazardous material analyses was performed on SNEC soil. The soil was found to be nonhazardous. Although most of the affected soil has been removed from the site, these analyses are also representative of the soil remaining. Chemical compounds such as cyanides, sulfides, copper, zinc, and tetrachloroetherie (TCE) were found per the Mountain States Laboratory analysis results. Since the SNEC site was adjacent to a former coal fired power plant, the discovery of these compounds, which are products common in fossil fuel combustion waste (i.e., flyash and bottomash) is expected. Sulfides were also present in past mine operation drainage that took place in the area of Saxton. Several EPRI reports such as EA-5321_ (1987) (Reference 18) and CS-5355 (1987) (Reference 19), describe the 4-74
1 Saxtan Nucle:r Experimental Carp: ratisn (SNEC) Site Characterization Report O chemical characteristics and TCLP analyses of utility waste. The characteristics of Pennsylvania coal generally contain approrfmately 10-15 mg/kg, Cu; >50 mg/kg, Zn; 1000 mg/l, sulfide and trace quantitiet of cyanide. In addition, common Pennsylvania soil contains quantities of zine and copper in the 10-300 ppm and 10-100 ppm ranges respectively. The presence of TCE was found in a few of the samples of soil prior to sending this j soil to Envirocare. Based on interviews with former SNEC employees it is possible, but cannot be verified, that chemicals containing TCE may have been used as part of the site's operation. The nonhazardous determination of the SNEC soil was based on the TCE limit of 0.7 mg/l. For soil containing TCE to be considered a ; hazardous waste, it would have to be derived from a spill of spent solvent containing greater than 10% TCE, a spill of unused commercial grade TCE or TCLP leachate containing >0.7 mg/L TCE. The exact source of the TCE in the SNEC soil is not known. However, to the best of cur knowledge, no spills of either spent or unused
- solvents or chemical mixtures containing TCE occurred at the site. The soil samples all contained less than 0.7 mg/l TCE in the TCLP analyses. Therefore, based on the l appropriate determination protocol, the SNEC soit does not constitute a hazardous waste.
4.4.6 Air, Sediment And Ground / Potable Water l Air, sediment and ground / potable water sampling have been and continue to be conducted on the site. Table 4-52 (Section 4.5), provides the analytical results from j the past year (1994/1995), for air particulate monitoring. The following summary table shows the type of contaminants detected in various sample media extemal to the SNEC CV. Table 4-26, Summary Table Of Contaminants Detected in SNEC Sample Media ) SAMPLE MEDIA CONTAMINANT COMMENTS ! ' Soil Cs-137 ; Ground Water. Non-detectable Well GEOS shows periodic l levels of H3 (200-600 pCi/L) j ! Potable Water Non-detectable ! Air Particulate Non-detectable l i Sediment Cs-137 4-75 -l l (
I Sixt2n Nucl:cr Experimental C rporati:n (SNEC) l Site Characterization Report i O G l l Reference the "1994 Saxton Soil Remediation Project Report" (Reference 9), for contaminants in soil, and the following tables for radiological data related to ground water, potable water, and sediment. Table 4-27, Ground Water Stations Results (Activity in pCilliter)
- GPUID Station Date Co-60 Cs-134 Cs-137 H-3 077163 SX-GW GEO-1 07/13S4 <5 <4 <4 < 170 078537 SX-GW-GEO-1 10/0694 <3 <3 <3 <180 079885 SX GW GEO-1 01/12 S 5 <4 <3 <3 < 190 081186 SX4W GEO-1 G4/05S5 <4 <3 <3 < l80 081958 SX-GW GEO-1 05/30/95 <$ <4 <4 <180 082223 SX-GW-GEO-1 06/13S 5 <4 <3 <3 < 190 082744 SX-GW GEO-1 07/13/95 <7 <6 <6 < l60 083323 SX GW-GEO-1 08/17/95 <3 <3 <3 < 180 0&3816 SX GW-GEO-1 09/15/95 < 20 < 20 < 20 < !90 077164 SX-GW GEO-2 07/13/94 <3 <3 <3 < 170 1
078538 SX-GW-GEO 2 10/06/94 <4 <3 <4 < 180 < 079886 SX-GW-GEO-2 01/12/95 <3 <3 <3 < !90 081137 SX-GW-GEO-2 04/05S 5 <6 <4 <6 < 180
\ 081959 SX-GW GEO-2 05/30 S 5 <3 <3 <3 < 180 082224 SX-GW.GEO-2 06/13/95 <3 <3 <3 < 180 l l 082745 SX-GW-GEO-2 07/13SS <6 <4 <4 < 160 l I 083324 SX-GW GEO-2 08/17/95 <3 <3 <3 < 180 l 083817 SX-GW-GEO 2 09/15/95 < 15 < 14 < 13 < 190 077165 SX-GW GEO-3 07/13/94 <9 <6 <7 < 170 078539 SX-GW GEO-3 10/06/94 <2 <2 <2 < 180 l 079887 SX-GW-GEO-3 01/12/95 <6 <4 <5 < 190 081188 SX-GW4EO-3 04/05 S 5 <7 <6 <6 < l80 081960 SX-GW GEO-3 05/3065 <6 <5 <5 < 180 082225 SX-GW-GEO 3 06/13/95 <7 <5 <5 < 190 l 082746 SX-GW-GEO-3 07/13/95 <4 <3 3 < 160 l 083325 SX-GW GEO-3 08/17/95 <6 <5 <5 < l80 075682 SX-GW-GEO-4 04/06/94 <4 <3 <3 < l80 077166 SX-GW.GEO 4 07/13/94 <6 <6 <7 < l70 078540 SX-GW-GEO 4 10/06 S 4 <3 <3 <3 < 180 081189 SX-GW-GEO 4 04/05/95 <3 <2 <2 < !80 081962 SX-GW GEO4 05/3065 <4 <3 <3 < 180 l
082226 SX-GW GEO4 06/13/95 <6 <6 <6 < 190 082747 SX-GW-GEO4 07/13/95 <3 <3 <3 < 160 083326 SX-GW GEO4 08/1765 <7 <6 <6 < 180 , 083818 SX4W-GEO4 09/15/95 <3 <3 <3 < l90 077167 SX GW GEO 5 07/13/94 <2 <2 <2 < 170 ) 078542 SX-GW4EO-5 10/06/94 <3 <3 <3 560
- 130
] 079888 SX GW GEOA 01/12/95 <8 <6 <7 < 190 081190 SX GW-GEO-5 04/05/95 <3 <3 <3 < 180 4-76
Srxtsn Nucirr Experimental C:rp;ratirn (SNEC) Site Characterization Report
\
GPUID Station Date Co40 Cel34 Ce137 H-3 081 % 3 SX4W-GEO 5 05/30/95 <6 <5 <5 270
- 120 082227 SX GW-GEO-5 06/1305 <2 <2 <2 370
- 130 082746 SX-GW-GEO-5 07/13S 5 <6 <5 <5 370
- 110 083327 SX GW-GEO-5 08/17 S 5 <3 <2 <2 390
- 130 l 083819 SX-GW GEO-5 09/15S 5 <3 <3 <3 410
- 130 07/168 SX4W-GEO4 07/13/94 <3 <3 <3 < 170 j 078543 SX-GW GEO 6 10/06/94 <8 <6 <6 < 180 l 079889 SX4W4EO6 01/12 S 5 <6 <6 <6 < 190 f 081190 SX-GW GEO4 04/05S 5 <2 <2 <2 < 180 l
081964 SX GW-GEO4 05/30/95 <7 <6 <6 < 180 082228 SX-GW-GEO 6 06/13/95 <2 <2 <2 < 190 082749 SX4W GEO4 07/13/95 <6 <6 <6 < !60 083328 SX-GW GEO4 08/17# 5 <4 <4 <4 < 180 l 083820 SX-GW-GEO4 09/1565 < 16 < 15 < 14 < 190 0 7/169 SX GW-GEO-7 07/13/94 <2 <2 <2 < !70 078544 SX-GW.GEO-7 10/06/94 <3 <3 <3 < 180 079890 SX-GW.GEO 7 01/12/95 <3 <3 <3 < 190 081192 SX4W-GEO 7 04/05S5 <3 <3 <3 < !80 081 % 5 SX-GW GEO-7 05/30/95 <2 <2 <2 < !80 l 082229 SX GW-GEO-7 06/13/95 <3 <2 <3 < 190 l 082750 SX-GW GEO-7 07/13 S 5 <3 <3 <3 < 160 ! 083329 SX-GW-GEO 7 08/17S 5 <3 <2 <2 < 180 t 083822 SX-GW GEO 7 09/15/95 <3 <3 <3 < 190 l 081193 SX-GW-GEO-8 04/05/95 <6 <4 <4 < 180
=
082751
**=2l =l :'
SX GW-GEO 8 07/1365 <8
<6 ll
<6 ll:
< 160 l
t I l l l l l l l l ( i O 4-77
l SIxtrn Nuclear Experimental C:rpsratirn (SNEC) Site Characterization Report
/ h \ >
U Table 4 28, Ground Water Stations Results (Activity in pCilliter).
- GPUID Statkm Date Co 60 Cs-134 Cs-137 H-3 Sr-90 j 075685 SX-GW-MW-1 04S664 <7 <6 <7 <180 <1.0 077170 SX-GW-MW 1 07/13/94 <l.6 <l .4 <l .6 <170 <l .0 l
l 078545 SX GW-MW-1 1006/M <3 <2 <2 <l80 <l .0 079330 SX-GW-MW-1 11/29/94 <l.8 <1.5 <l.6 <170 <1.0 079891 SX4W-MW-1 01/12/95 <2 <2 <2 <!90 <0.7 l 081194 SX4W-MW-1 04/05/95 <2 <1.6 <l.8 <!80 4.8 081 % 7 SX GW-MW-1 05/30/95 <l.3 <1.1 <l .1 <180 <l .0 l 082231 SX GW-MW-1 06/13/95 <3 <3 <3 <190 <l .0 082752 SX-GW-MW-1 07/13 S 5 <l .3 <l .1 <l .2 <!60 <l .0 083330 SX GW-MW-1 08/17/95 <3 <2 <2 <180 <1.0 j 083823 SX-GV/-MW-1 09/15/95 <4 <3 <3 <190 <l .0 075686 SX GW-MW-2 04/06/94 <l .2 <l.1 <l .2 <180 <l .0 077171 SX GW-MW-2 07/13/94 <l .3 <l.1 <l .2 <!70 <l .0 O 078546 SX-GW-MW-2 1006/94 <l.5 <l.5 <l .4 <180 <l .0 l 079892 SX-GW-MW-2 01/12/95 <1.4 <l .4 <l .6 <!90 <0.9 081195 SX-GW-MW-2 04S5SS <4 <3 <3 <180 <0.8 081 % 8 SX GW-MW *. 05/3065 <1.4 <l .4 <l .4 <!80 <l .0 082232 SX-GW-MW-2 06/1365 <3 <2 <2 <190 <l .0 082753 SX GW-MW-2 07/1365 < l.6 <1.3 <l .5 <!60 4.6 083331 SX4W-MW-2 08/17/95 <1.1 <1.1 <l .3 <180 <l .0 083824 SX-GW.MW-2 09/15/95 <3 <2 <3 <190 <l .0 l l l l l
.U 4-78
S:xt::n Nucirr Experimental C rporati:n (SNEC) , Site Characterization Report ' p (j
\.
i Table 4-29, Aquatic Sediment Stations (Activity in pCilg - dry). 1 l I GPUID Station Date Co40 Cs-134 Cs-137 K-40 Ra-226 Th-232 U-235 J 074596 SX-SD-Al-1 01/1364 2.l
- 0.2 4.840.7
< 0.05 < 0.04 2.l
- 0.8 1.1
- 0.2 < 0.2 )
075689 SX-SD Al-1 04/06S4 < 0.06 <0.04 2.1
- 0.2 5.8
- 0.7 2.6
- 0.6 1.2
- 0.2 < 0.2 077133 SX SD Al 1 07/1364 < 0.03 < 0.02 23
- 0.2 5
- 0.5 23 *0.6 0.8 0.11 < 0.16 078533 SX-SD-Al-1 10/06/94 < 0.04 < 0.04 0.2
- 0.04 6.4
- 9.8 2.7
- 0.8 1.3
- 0.2 < 0.2 079895 SX-SD-Al-1 01/12S 5 0.071
- 0.031 <0.02 33 *0.3 5.6 e 0.6 2.2
- 0.6 1.l
- 0.1 < 0.15 l
1 081198 SX-SD-Al-1 04/05 S 5 0.027
- 0.023 < 0.02 2.6*03 5.8 e 0.6 2.4
- 0.7 1
- 0.1 < 0.16 ;
082739 SX-SD Al 1 07/1365 < 0.04 <0.03 1.8
- 0.2 8
- 0.8 2.9
- 0.8 1.3
- 0.2 < 0.2 .
I 074597 SX-SD Al 4 01/13 S 4 < 0.02 <0.02 0.74
- 0.07 8.6
- 0.9 1.4
- 0.5 0.65
- 0.I < 0. I4 075690 SX SD-Al 4 0406S4 < 0.03 < 0.03 0.39 0.04 7.4
- 0.7 13 0.5 0.58
- 0.13 < 0.15 077159 SX-SD-Al-4 07/13/94 < 0.015 < 0.014 0.21 0.02 8.8
- 0.9 2.1
- 0.4 0.8
- 0.09 < 0.11 078534 SX SD Al-4 10/06/94 0.022
- 0.012 <0.01 0.28
- 0.03 7.6
- 0.8 I
- 0.2 0.65
- 0.07 < 0.08 0798 % SX-SD-Al-4 0I/12S5 <0.013 <0.011 0.22
- 0.02 6.9
- 0.7 13 *03 0.64
- 0.06 < 0.09 081199 SX-SD-Al-4 0445/95 <0.011 < 0.01 0.15* 0.02 7.8
- 0.8 1.3
- 0.3 0.66
- 0.07 < 0.08 l
082740 SX-SD-Al-4 07/13/95 < 0.015 < 0.013 0.1620.02 7.1
- 0.7 1.1
- 0.3 0.57 0.07 < 0.11 074598 SX-SD C14 01/1364 <0.05 < 0.04 1.5
- 0.1 8.5
- 0.9 2.8
- 0.8 1.2
- 0.2 < 0.2 l
l f 075692 SX-SD-Al4 04/06 S 4 < 0.02 <0.02 1.5 s 0.1 9.3
- 0.9 2.6
- 0.5 1
- 0.1 <0.15 077162 SX4D-A14 07/13/94 < 0.03 < 0.02 1.9 s 0.2 9.7
- I 3.2
- 0.6 1.2
- 0.1 < 0.18 078535 SX-SD-A14 1046/94 0.064
- 0.027 <0.02 2.1
- 0.2 10
- 1 2.7
- 0.6 1.1
- 0.1 < 0.14 079897 SX-SD-A14 01/12/95 <0.02 < 0.017 1.2
- 0.1 11
- I 2.8 0.4 1.2
- 0.1 <0.12 081200 SX-SD A14 044565 < 0.03 < 0.02 1.5
- 0.1 11
- I 2.9
- 0.6 1.2
- 0.1 < 0.16 l
l 082742 SX-SD A14 07/13/95 <0.08 <0.05 2
- 0.2 9.1
- 1.2 33*1 13 *0.2 <03 j 07569i SX-SDQ1-2 04/06/94 <0.015 < 0.012 0.099
- 0.017 6.2
- 0.6 1.4
- 03 0.64
- 0.07 <009 0 77160 SX-SD-Q1-2 07/13/94 < 0.013 < 0.011 0.26
- 0.03 7.5
- 0.8 1.1
- 0.2 0.59
- 0.07 < 0.1 ,
l ! 078536 1006/94 < 0.015 < 0.013 8
- 0.8 1.5
- 0.4 0.72
- 0.07 SX SD41-2 0.18 e 0.02 < 0.12 i 079898 SX-SD Q1-2 0I/I2/95 <0.011 < 0.01 0 086 3 0.014 7.6
- 0.8 1.2
- 0.2 0.7
- 0.07 <0.09 081202 SX-SD-Ql-2 040595 < 0.014 < 0.012 0.049
- 0.016 7.4
- 0.7 1.2
- 03 0.72
- 0.07 < 0.11 082743 SX-SD.Q12 07/13/95 < 0.018 < 0.016 0.073
- 0.022 7.5
- 0.7 1.7
- 0.4 0.74
- 0.1 < 0.13
{ i i l l l I' l (O) v i 4-79 l l
e I l l S'xt2n Nucinr Experimintal C:rp;rati:n (SNEC) Site Characterization Report \ O Table 4-30, Potable Ground Water Stations (Activity in pCilliter). GPUID Station Date Co40 Cs-134 Cs-137 H-3 074599 SX-GW-El-1 01/13/94 < l .8 < l.7 < l.7 < 150 075687 SX-GW-El.1 04/06/94 < l.5 < l.4 < l.5 < 180 0 7/173 SX GW-El-1 07/13/94 < l.3 < l .4 < l.4 < 170 078531 SX4W-El.1 10/06/94 <3 <3 <3 < 180 079893 SX GW-El 1 01/12/95 <3 <2 <2 < 190 081196 SX-GW-El-1 04/05/95 < l.7 < l .6 < l.6 < 180 OILi _ SX-GW El 1 07/13/95 <3 <2 <2 <!60 075688 SX-GW Gl.1 04/06/94 < l.3 < l.2 < .: ? < 180 0 7/172 SX GW Gi-1 07/13/94 < l.6 < l.3 < l.4 < 170 078532 SX GW Gi-1 10/06/94 < l.2 < l.1 < l.1 < 180 0798M SX-GW.Gl-1 01/12/95 < 1.6 < l.4 < l.5 < 190 l 081197 SX-GW Gi-1 04/05/95 <3 <2 <3 < 130 082754 SX-GW-G1 1 07/13/95 < l.3 < 1.2 < l.3 < 160 Table 4-31, CV Pipe Tunnel Results (Activity in pCl/ liter) ! GPU ID Station Date Co40 Cs-134 Cs-137 H-3 075362 SX4W-Gi 6 03/16/94 < l.5 < l.4 7.6 i 1.8 < 180 076959 SX-GWG14 06/15/94 <3 <3 12
- 3 <130 078354 SX-GWG14 09/20/94 <4 <3 <4 < 180 079731 SX-GW Gi-6 12/15/94 <3 <2 16
- 3 < 190 081185 SX-GW-G14 03/24/95 < l.8 < 1.5 5.1
- 1.6 < 180 082212 SX.GW-G14 06/06/95 <3 <2 8.l
- 2.6 < l90 4.4.7 Analytical Methods l Bulk soil and water samples were collected and analyzed using methods described by the SNEC Radiological and Environmental Monitoring Program (Reference 31).
Typically one liter (water) or 1 kilogram (soil and/or sediment) were obtained to meet minimum sensitivity requirements related to gamma / alpha spectroscopy (gamma emitting nuclides and transuranics), liquid scintillation counting (tritium) and proportional counting (strontium). Air particulate samples were run to achieve a volume of approximately 500 cubic meters and analyzed for gamma emitting
, nuclides.
t o 4-80 t l
Saxt n Nuclzr Experirnent:1 C:rp r:ti:n (SNEC) Site Characterization Report ! [, \ l 4.4.8 Sample And Analysis Summary Listing For Samples External To SNEC CV
- The total number of samples obtained extemal to the CV, and the type of analyses that were reported are listed in the following table.
l Table 4-32, General Sampling Analysis information l Sample Type Number of Samples Analysis Type l Site Characterization (soil) 538 Gamma Spec /Transuranics/ 22 Strontium / Gamma Spec Soil Remediation (soil) 381 Gamma Spec Hazardous Material Profile 15 TCLP Soil Density Profile 1 Compactiblity/ Moisture Content Air Particulates 96 Gamma Spec / Gross Alpha / Gross Beta Potable Water 19 Gamma Specffritium/ Gross Beta Groundwater 87 Gamma Spec / Tritium Sediment 27 Gamma Spec l Total Samples 1186 The minimum sensitivities for typical analysis procedures are listed in the following g) ( tables. Table 4-33, ERL Source / Sample Activity Tracking Sheet - Typical ERL Detection Limits. Water Aquatic Biota Isotope Method LLDs MDLs Method Air LLDs MDLs Method LLDs MDLs pCi/L pCi/m^3 pCi/KG (wen Mn-54 A 1.9 1.2 J 9 5 Fe-59 A 4 2.4 J 20 12 Co-58 A 2 1.2 J 9 5 CM>0 A 2 1.2 J 11 7 Zn-65 A 4 2.4 J 20 12 Zr-95 .A 4 2.4 J 16 10 Nt-95 A 2 1.2 J 10 6 Cs-134 A 1.8 1.1 F 0.0019 00012 J 8 5 Cs-137 A 2 1.2 F 0.002 0.0012 J 10 6 Ba-140 A 9 5.5 La-140 A 4 2.4 Gr-Alpha B 0.9 0.5 0 0.0009 0.0005 Gr Beta C 1.6 1.0 H 0.003 0.0018 ( I-l31 D 0.3 0.2 1 0.016 0.0097 11-3 E 130 79 K 0.0001 0.00006 i s / <
's )
v l l 4-81 l l l
r , Srxt n Nucle:r Experimental Ccrp:rati:n (SNEC) Site Characterization Report O Table 4-33, Continued Isotope Method Milk MDLs Method Terrestrial MDLs F.lethod Sediment & MDLs LIES pCi/L Vegetation / Soil LLDs Meat LLDs pCi/KG pCI/KG(n et) (drv) Mn-54 A 1.9 1.2 L 15 9 Fe-59 A 4 2.4 L 40 24 Co.58 A 2 1.2 L 15 9 Co40 A 2 1.2 L 17 10 Zn45 A 4 2.4 L 40 24 Zr-95 A 4 2.4 L 30 18 Nb-95 A 2 1.2 L 20 12 Cs-134 A 1.8 1.1 L 10 6 L 13 8 Cs-137 A 2 1.2 L 12 7 L 20 12 Ba-140 A 9- 5.5 l La-140 A 4 2.4 Gr-Alpha Gr-Beta I-131 M 0.3 0.18 L 18 11 H-3 Methods Leoend i A 3.5 Liters for 50K Seconds i B 0.4 Liters for 100 Minutes with 0.2 cpm bkgnd and 25% Efficiency. l l C 0.4 Liters for 100 Minutes with 1.3 cpm bkgnd and 40% Efficiency. l l D 3.5 Liters for 60K Seconds l E 7 mis for 220 Minutes with 4 cpm bkgnd and 27% Efficiency. j ! F 2700 Cubic Meters for 4K Seconds G 680 Cubic Meters for 100 Minutes with 0.5 cpm bkgnd and 27% Efficiency. . ! H 680 Cubic Meters for 20 Minutes with 3.5 cpm bkgnd and 47% Efficiency. l l 680 Cubic Meters for 1K Seconds J 1 KG for 10K Seconds i K 7g for 240 Minutes with 4 cpm bkgnd and 27% Efficiency, t L 1 KG for 14K Seconds l M 3.5 Liters for 60K Seconds with 95% recovery l 4.4.9 Summary Of Environmental Results External To SNEC CV The results of soil sampling and analyses are discussed in Sections 4.4.4 - 4.4.5 and in related and referenced documents. Air particulate stations are put into service whenever any site radiological operations l l are being performed. The particulate filter is analyzed for gross alpha, gross beta and gamma scanned, and only the positive activities detected are listed in this report. Three environmental air sampling stations are located around the site and 4-82
f Saxton Nuclear Experimental Corpor; tion (SNEC)
- Site Characterization Report l l )
U one control station is located 10 miles from the site. In general, the weekly trends of l l gross alpha and gross beta activity at all stations are similar. Gamma-emitting radionuclides related to the SNEC Facility operation were not detected on any of the samples. As expected, all of the samples contained naturally occurring Be-7. l Concentrations detected are similar on both indicator and control filters. l Ten environmental monitoring well stations plus two potable ground water stations i 1 are sampled at least quarterly. Periodically, low levels of tritium (200-600 pCi/L) have been detected in one of the environmental groundwater monitoring wells (GEO-5). Beneath this area is the former site of the Radiological Waste Disposal Facility tunnel. It is likely that this positive result is due to residual tritium from activity in this area arising from plant operation. Nevertheless, tritium concentrations from this
~
station are wei, below the United States Environmental Protection Agency Primary Drinking Water Standard of 20,000 pCi/1. Gamma scans from this station, as well as i O' all other groundwater well stations, have always resulted in less than detectable limits. Low levels of Cs-137, and at times Co-60, have been detected in aquatic sediment samples collected near the SNEC facility (SX-SD-A1-1 and SX-SD-C1-1) and at the discharge to the river (SX-SD-A1-4). Cs-137 was also detected in the control station (SX-SD-Q1-4), which is secured upstream from the discharge. The results are not surprising since these radionuclides have a propensity to attach to sediment. Levels l have been much lower since the completion of the soil remediation project. K-40, Ra-226 and Th-232 are naturally occurring isotopes and present in all sediment. The discharge to the river was plugged in 1993. The CV Pipe Tunnel (SX-GW-G1-6) station is deemed a radiological as opposed to an environmental station. This sample is monitored for trending purposes. Positive l Cs-137 results are typical for this station due to its close proximity to the containment vessel. (v) i 4-83 i i
Saxten Nucl::Ir Expsrirnantal Ccrpor tion (SNEC)
- Site Characterization Report I \
V 4.4.10 Data Reporting and Analysis For Hazardous Materials Within The CV A. Asbestos Containina Material The results of the asbestos analysis include: (1) the sample identification number and location; (2) the asbestos type and approximate amount; (3) the presence of non-asbestos fibers and approximate amount; and (4) the , presence of non-fibrous materials (listed in source data files). B. Full TCLP,Inoraanic Anaivsis The results of the contaminant analysis include: (1) the sample identification number and location; (2) the materials tested; (3) the result in mg/ liter or ug/gm (or similar units); and (4) the detection method and limit. 1 C. Total Metals Analysis l Total Metal Analysis was used as a screening test. The metals selected f)T
\' represent typical metals found in paints. The results are reported in mg/kg.
l When the Total Metals Analysis exceed any of he below listed action levels I l ! for the target metals, full TCLP, inorganic analysis were performed. Action Levels 100 mg/kg for Lead l 20 mg/kg for Cadmium 100 mg/kg for Chromium D. PCB Containing Material The results of the PCB analysis include: 1) the sample identification numoer I and location: 2) the material tested; and, 3) the results in ppm (or similar units). ( 4 o 4 84
Sixten Nuclear ExperimIntal C rporation (SNEC) Site Characterization Report 4.4.11 SNEC Site Characterization Plan Reporting The following are the suggested data collection requirements as presented within the SNEC SCP. Asbestos
- Location of sample
- Type and % of asbestos fibers present
-Type and % of non-asbestos fibers present
- Presence of non-fibrous materials Total Metals Analyses
- Location of sample
- Test / parameter (metal tested) - Results (mg/kg)
O TCLP. Inoraanic Analysis: (Metal Contaminants)
- Location of sample
- Test / Parameter (metal tested)
- - Results (mg/l) l - Detection limit l
l l PCB's
- Location
- Results (ppm) 4.4.12 PCB Sampling Results A review of components inside the SNEC facility CV, indicates that the only polychlorinated biphenyl's (PCB's) present are the lamp ballast's for the mercury lamps in the overhead of the CV. There are eight lamps with ballast's manufactured by the Jefferson Electric Company, of Bellwood, Illinois. These items are catalog number
! 232-441. O 4-85
S xt:n Nuclear Experim2nt:1 Ccrpor; tion (SNEC) Site Characterization Report i Research indicates these definitely contain PCB laden materials and that sampling to confirm this is more expensive than disposal as PCB bearing material. Ballast units are totally sealed and are surface contaminated only. A sample ballast has been decontaminated to current free release criteria suggesting that all ballast units may be successfully free released and disposed of as simple hazardous waste and not as mixed waste 4.4.13 Bulk Lead inventory A bulk lead inventory of the SNEC facility was conducted during the characterization program. The results of the lead inventory have been documented within SNEC Technical Evaluation 6575-95-001 (Reference 100). A summary of the results indicate that there is approximately 14,000 pounds of lead within the SNEC CV. This material is largely lead bricks and sheet lead. Additionally, there are some lead blankets of various sizes and some instrument shields of solid lead casting that comprise the remainder of the inventory. Some of the lead materials (bricks) have been laying in water within the Storage Well. The result is that some leaching of lead compounds has occurred into water and resin material present within this area. 4.4.14 Lead in Paint And Other Sample Materials Since the Saxton Nuclear Experimental Corporation facility was built in the early 1960's, it was presumed that a majority, if not all paint used during the construction of the facility contained a certain percentage of lead. A sample of paint taken from the exterior of the building has show some lead content (~1.67% lead). Several composite samples of paint and surface coating materials were obtained and sent to B&W for a total lead content analysis and a Toxicity Charactenstics Leaching Procedure (TCLP). The results showed that lead materials are present in surface coating from within the SNEC CV. Lead is also present in resin materials that cover the Storage Well floor area within the CV. The reason for this is probably due to leaching of lead compounds from lead shielding remaining within the CV. Additional testing for lead in paint materials will be performed during or prior to decommissioning. All paint will be assumed to contain lead until determined otherwise. TCLP results are reported in section 4.4.16. []
\
_/ 4-86
. - -.-.. . . - - - . - . - . - . . . - . - . . - ~ . - . - . . _ - _ _ - . - - _ _ _ . _ _
a {' . Sixt:n Nuclear Experimintal Ccrparatien (SNEC) Site Characterization Report l i When piping or other components have to be burned into sections for removal purposes, the suspected lead based paint will be removed at the bum location to reduce the airborne concentration of lead. Removing the lead based paint would eliminate the t
- need to comply with the OSHA lead standard. Removal of the lead based paint can be .
iicrsrplished by the use of paint strippers. If paint strippers cannot be used, then the 1 OSHA lead standard would be applicable due to potentially elevated airbome r concentrations of lead produced during the buming process. The table below presents total lead analysis results for several SNEC waste materials. i Table 4-34, Total Lead Content Of Surface Coatings And Resin Materials i SNEC Sample No. - Sample Materials AnalyticalMethod Results Pamt M=. AREAS 1,2,3 & 4 SW8466010 34.200 mg/Kg ! SX8C6950083 Pamt Chips, AREA 6(SW) SW846 6010 16.4 mg/Kg SX8C6950084 Storage WellResm Water SW846 6010 29.7 mg/L SX866950115 (Liquid) StorageWellResin Solids SW8466010 1880 mg/Kg SX866950115(Solids) 4 4.4.15 Asbestos Sampling Results Since the Saxton Nuclear Experimental Facility was built in the early 1960's, it was presumed prior to sampling, that a majority if not all insulation used during the ) construction of the facility contained a certain percentage of asbestos. During May 8 through the 10 of 1995, bulk insulation samples were taken of various 1 l components, piping systems and vessels throughout the SNEC containment building. 4 The location of the samples taken were identified by a specific number for that sampling l f_ point (example BAsb-001-95). This sampling point number was attached to the exact location (when practicable), where the core insulation sample was taken. The sample point locations and sample numbers were entered into the computer data tracking system developed by GPUN called NREX. The piping and component system analysis results are identified within table 4-35 following this section. The bulk samples collected were analyzed on-site by Polarized Light Microscopy (PLM). i The analysis was conducted by a lab technician from the GPUN System Laboratory in 4 ' 4-87 E
-, - , - ew, - an. ~ ,
r Saxt n Nucle r Experimental Corporation (SNEC)
- l. Site Characterization Report
\ /
Reading, PA. The System Lab is certified by the National Voluntary Laboratories Accreditation Program (NVLAP) for bulk insulation identification for asbestos. The analyst is certified by McCrone Research Institute for PLM. During this sampling period, a total of 61 bulk insulation samples were taken. In previous studies, nine additional bulk insulation samples were also taken. Out of these 72 total samples, it appears that the only insulation that does not contain any asbestos is the insulation on the straight pipe runs for the HVAC system. These straight pipe runs j contain fiberglass. However, the elbows, valves and tees for the HVAC systems contain troweled-on insulation containing asbestos. All bulk insulation samples that were identified to contain asbestos consist of either chrysotile or amosite asbestos fibers. Any piping system that needs to have the insulation removed, and has not been samoled, will be presumed asbestos containing material (PACM) and treated accordingly until l determined otherwise. l (O) xJ The following standard practices are possible ways to handle the asbestos insulation without increasing the potential for airbome release to unprotected workers during , demolition.
- 1. Using Pennsylvania certified asbestos workers, remove all asbestos insulation before l any other demolition activities commence. This would remove the concems of asbestos exposure by accidental contact with the asbestos insulation during pipe, component and vessel removal.
- 2. Commission a consulting firm that has experience in safety and industrial hygiene and is able to do analysis for PLM and Phase Contrast Microscopy (PCM) on site. The safety and industrial hygiene experience should be in demolition and asbestos abatement, confined space entry requirements, and scaffolding use. ,
\
- 3. Commission an asbestos abatement firm and contractor that has Pennsylvania certified workers and has experience in large scale asbestos abatement projects.
l i
- 4. Remove the non-asbestos insulation prior to removing the asbestos insulation.
m) s l 4-88
e S-J:n Nucimr Experim7ntal Corp;ratian (SNEC) l Site Characterization Report i p 1 b Table 4-35, Asbestos Sampling Requirements Listed in SCP And Assigned Results SYSTEM LOCATION MA'ITRIAL Main Coolant System Steam Generator ACM Main Coolant Pump ACM Main Coolant Piping ACM Fittmg ACM j Pipe ACM ' Pressure Control & ReliefSystem Pressunze ACM l Piping ACM l Fitting ACM l j Pipe ACM l l Purification System Regenerative Heat Exchanger ACM Non-Regenerative Heat ExcFanger ACM Piping ACM Fitting ACM
- Pipe ACM i
Component Cooling System Component Cooling Heat Exchanser ACM I Piptng ACM l Fitting ACM Pipe ACM i Sampling &Imak Detection Swtem Pipmg PACM Fitting PACM Pire PACM Shutdown Cooling Svstem Shutdown Cooling Heat Exchanger ACM Piping ACM l g Fitting ACM ! 'N Pipe ACM l Safety Injection System Piping PACM i Fitting PACM Pipe PACM Storage Well Svstem Storaec Well Heat Exchanger ACM i Piping ACM Fitting ACM Pipe ACM i Air Handlers l Cooling Heating & Ventilation - Auxiliary Compartment PACM
. Primary Conmartment PACM
- Rod Room PACM
- Operating Floor PACM
- Piping Elbows. Tees. Valves & Y's Contain Asbestos.
- Steam Supply & Returns Strainht Runs Of Pine Are
-Water Suppiv & Returns Fiberalass l
Secondary Side Systems Main Feedwster Pipme PACM Fitting PACM l Pipe PACM l Main Steam Piping PACM Fitting PACM Pipe PACM Blowtlown Pipe PACM Fitting PACM i Pme PACM ! ACM Ad==** Contanng Matenal i PACM Presumed Asbestcs Containing Material The following table presents the analytical results from the asbestos sampling program. i ! 4-89
Sixt:n Nuclxr Experim:ntal Ccrpsratisn (SNEC) Site Characterization Report O
-('
Table 4-36, Asbestos - Bulk Sample Results sample Number Location Asbestos Type NonAbestos Materials
& % of Total & % of Total BASB 000195 CC Hx Exchanger-M EL Chrysotile 3% MineralWool 50%
BASB 0002 95 Vahe Etxw V-1552, M EL Chrysotile 5% Mineral Wool 10% BASB 0003 95 CC Hx Exhanger M EL Chrysotile 3% Mineral Wool 30% BASB 0004 95 Valve Body V-1572,795' EL Chrysotile 5% Cellulose 2%, MineralWool 5% BAS 8 0005 95 V-167, By Hx Exhanger, M El Not identified Cellulose 65%, I l MineralWool 20% l BASB 0006 95 Unkage Transmeer, V-1540,795' EL Chrysottle 5% Mineral Wool 12% ; BASB 0007 95 White Cloth V-167,795' EL Chrysotile 5% Mineral Wool 12% ' BAS 8 0008 95 V-1549,795' EL Arnosite 10% Not identfed BAS 8 0009 95 AJeary Steam Une, By CC Pump i Amosite 10% Not IdentMed BASB 0010 95 Under Steps, M EL Amosite 8% Mineral Wool 2% BASB 001195 3' Atxw Valve V 1602, M EL Chrysotile 5% Mineral Wool 7% l BASB 0012 95 Condensate Retum By Pump 1 Amosite 10% Not identMed BAS 8 0013 95 Above Valve V-2170,781' EL Chrysotile 3% Mineral Wool 15% l BAS 8 0014 95 Betw Valve, V-1586 Chrysotile 10% Mineral Wool 5% l BASB 0015 95 By Transmeer,781' EL Chrysotile 5% Mineral Wool 10% l BAS 8 0016 95 Storage Wet Hx Exchanger, 781' EL Chrysotile 2% MineralWool 25% ! BASB 0017 95 Storage Wee Hx Exchanger, 781' EL Chrysotile 3% Mineral Wool 10% l BASB 0018 95 Etxw Of V 1553, 781' EL Chrysotile 3% Mineral Wool 12% b BASB 0019 95 Vafwe Body of V-1585, 781' EL Chrysotile 15% Mineral Wool 2% k BASB 0020 95 90' Cooling Water Lines, 765' EL Chrysotile 5% Cellulose 15%, Mineral Wool 30% BAS 8 002195 Upstream Of Valve V 1550, 765' EL Amosite 12% Not identified BASB 0022 95 Body of Vahe V-1140, 765' EL Chrysotile 5% Mineral Wool 10% BAS 8 0023 95 Dacharge Of Pumf 765' EL Amosite 15% Not identrfied BAS 8 0024 95 Between V-1139 & V-1440, 765' EL Amosite 15% Not identified BASB 0025 95 Pipe Etxw Off Pump, M EL. Chrysotile 3% Cellulose 25%, MineralWool10% ( BASB 0026 95 Shutdown Cochng "T*, 765' EL Chrysotile 5% Mineral Wool 10% j_ ~BASB 0027 95 Equipment Drain Tank, 765' EL Chrysotile 5% Mineral Wool 10% l BASB 0028 95 Equipment Drain Tank, 765' EL Amoste 15% Not Identified ! BAEB 0029 95 Dacharge Tank Pump 2, UC 21V Top Amosce 20% Not identified BASB 0030 95 LRC 21V, 781' EL Chrysotile 3% Mineral Wool 20% ' BASB 003195 TIC 21V, 781' EL Amoste 20% Not identified BASB 0032 95 HVAC Unit Piping Elbow, 781' El. Chrysotile 5% Cellulose 25%, MineralWool10% BASB 0033 95 From LRC 21V To V-400, 781' EL Amosce 60% Not identified i BASB 0034 95 Green Hx Exchanger,781'El Amosite 15% Cellulose 20% I BASB 0035 95 To LRC 21V Bekw V 401,781' EL Amosite 60% Cellulose 5% BASB 0036 95 Shutdowwi Coahng Une,781' EL Amosite 30% Not identified BASB 0037 95 RV 60, Left Of Door ~6' Up,781' EL Amoste 60% Not identified BASB 0038 95 Under V,57, ~4' Up,781' EL Chrysotile 7% Mineral Wool 12% BASB 0039 95 HIC 31V Shutdown Coohng,781' EL Chrysotde 3% MineralWool 25% BASD 0040 95 PZR Rehef & Safetes,795' EL Amosite 60% Not identified [A
\
BASB 004195 BASB 0042 95 RC Pump,795' EL PZR Spray Line. 795' EL Amosite 10%, Chrysotile 3% Amostte 60% Cellulose 20%, Mineral Wool 2% Not identified 4-90
S:xt:n Nucl:ar Experimental C:rp:ratirn (SNEC) Site Characterization Report
/ %
)
Sample Number Location Asbestos Type Non-Asbestos Materials
& % of Total & % of Total BASB 0043 95 RC Une Pump,795' EL Amosite 60% Not identified BASB 0044 95 HTC-27V inlet Vane,795' EL Chrysotile 7% MineralWool10%
BASB 0045 95 PZR Relief & Safety Vanes,795' EL Amosite 60% Not identified BASB 0046 95 Une Off Pump, '6' Up, M EL Amosite 60% Not identified BASB 0047 90 Charging Pump Retum Une , M EL Amosite 60% Not identified BASB 0048 95 Sde Of PZR ,795' EL Amosite 40% Not identified BASB 0049 95 Steam Generator ,800' EL Amosite 60% Ce!! ulose 15% l BASB 0050 95 Une From PZR,795' EL Amosite 35% Not identified BASB 005195 UC 2 Tank By PZR ,795' EL Amosite 60% Not identified BASB 0052 95 Steam Generrter White Cloth,800' EL Chrysotile 60% Synthebe 40% BASB 0053 95 Une By VM Top Of PZR,795' EL Chrysotile 5% Mineral Wool 15% BASB 0054 95 V-21U Above S/G,800 EL Amosite 60% Not identified BASB 0055 95 One Frorn S/G To RCP, M EL Chrysotile 7% Mineral Wool 15% BASB 0056 95 V-2456 Bf PZR Top,800' EL Amosite 25% Cellulose 20% BASB 0057 95 1.re From RV To S/G,79S EL Amosite 60% Not identified BASB 0058 95 Bottom Of PZR,795' EL Amosite 60% Not identfied BASB 0059 95 V-409 Top Of PZR, SOU EL Amosite 60% Not identfied l BASB 0060 95 Top Of PZR, Cloth Covered ,807 EL Amosite 60% Not identified l BASB 006195 By Storage WaA, Ckth ,800' El Not identified Cellulose 50% BAsb-194-91 Flange On CC Hx Exenanger Chrysottle 30% Not identified BAsb 195-91 RWerWater Ppe Elbow Chrysotile 5% Glass Fibers 15% /O BAsb-196-91 RherWater For HVAC Chrysotile 35% Not Identified ( ) BAsb-197-91 RegeneratNe Heat Exchanger Chrysotile 8% Glass Fibers 15% BAsb-198-91 Storage Wee Heat Exchanger Amosite 40% Not identified BAsb-199-91 Storage Wed Heat Exchanger Not identiried Cellulose 10%, Glass Fibers 10% BAsb-200-91 Chater Ppng MetalJacket Amosite 15% Glass Fibers 5% BAsb-202-91 Main Fee &eter Ppng Not Identfied Glass Fibers 20% BAsb-203-91 Heat Exchanger Amosite 25% Not identified in addition to the asbestos insulation samples listed above, composite paint chips and surface coatings were analyzed for the presence of asbestos materials. The following table lists these analysis results. No asbestos materials have been found in this material.
}
Table 4-37, Composite Surface Samples From SNEC CV (CV AREAS 1,2,3,4 & 6) Sample Numbers Weight (e) Asbestos Fibers ? Total % Asbestos % Other Materials SX8C69500083 35.973 No 0% Cellulose 1-2% AREAS 1,2,3 & 4 9509066-02AA* SX8C69500084 19.94 No 0% Cellulose <1% AREA 6 9509066-03AA*
, *B & W Sample Number (V) 4-91
r S:xten Nuclear Exp2rimental Corporation (SNEC) l Site Characterization Report LO 4.4.16 TCLP, Inorganics Sampling Results Toxicity Characteristics Leaching Procedure (TCLP) criteria for inorgani,s were tested on samples from the SNEC facility Discharge Tank, composite surface l coating samples from various locations, and resin from the Storage Well. Based on the analysis, paint / coating materials contain no toxic metals above the TCLP limits. Sediment removed from the Discharge Tank also contained no metals above TCLP L levels. Therefore, paint / coatings removed from within the reactor building and sediment removed from the Discharge Tank do not incur any additional hazard from toxic metals. However, the Storage Well resins contain lead (7.2 mg/L) above the I TCLP limit and would be considered mixed low-level hazardous waste if disposed of without additional treatment. Treatment of resins to remove the EPA hazardous waste characteristic includes several options. Since the original sample contained both resins and water from the Storage Well, and the lead is contained principally in the water phase, the resins can be separated or washed to remove the hazardous properties prior to disposal. The l water can be processed with other aqueous radiological wastes. If the lead is actually attached on the resin, regeneration may remove the lead. The resin or resin / water mixture may also be solidified to prevent lead from leaching from the waste. This will render the waste nonhazardous. Additional TCLP sampling will be needed to further define SNEC facility waste prior to disposal, and to develop options for the Storage Well resins. Table 4-38, TCLP Extraction / Meta!s Analysis Results Analyte DischargeTank AREAS 1,2,3 & 4 AREA 6 AREA 6 Sediment Paint / Coating Paint / Coatings . Resin Limit SX815950068 - SX8C6950083 SX8C6950084 SX866950ll5 (mg/L) Arseme <0.5 mg/L <0.5 mg/L <0.5 mg/L <0.5 mg/L 5.0 Barium <10.0 mg/L <10.0 mg/L <10.0 mg/L <10.0 mg/L 100.0 Cadmium 0.40 mg/L <0.1 mg/L <0.I mg/L 0.12 mg/L 1.0 Chromium <0.5 mg/L <0.5 mg/L <0.5 mg/L <0.5 mg'L 5.0 2 Lead <0.5 mg/L <0.5 mg/L <0.5 mg/L 7.2 mg/L 5.0 Mercury <0.021 mg/L 0.055 mg/L <0.021 mg/L <0.021 mg/L l 0.2 Selemum <0.16 mg/L <0.16 mg/L <0.16 mg/L <0.16 mg/L 1.0
/ Silver <0.5 mg/L <0.5 mg/L <0.5 mg/L <0.5 mg/L 5.0 4-92
Sixten NuclIar ExperimintIl Ccrporation (SNEC) Site Characterization Report i V 4.4.17 Bulk Cadmium & Mercury Cadmium: During the lead inventory, a sheet of cadmium was uncovered which had been employed in a failed fuel detector experiment. This sheet, measuring approximately 1 20 inches by 18 inches by 0.028 inches thick, is a surface contaminated object which will be decontaminated and free released for recycle or disposed of as a hazardous waste. l l 1 Mercury: j During interviews with Northem States Power Corporation, personnel who had been involved in the decommissioning of the Pathfinder Nuclear facility, it was uncovered that some of the instrument transmitters in the SNEC facility CV may contain small amounts of elemental mercury. This mercury, if present, would need to be removed prior to disposal of these instruments. Contacts made with some of the transmitter , manufacturers confirmed that they are likely to contain small quantities of mercury in t , ( sealed bellows and or bulbs. Devices which are suspected of containing mercury will i be disassembled to confirm its presence. Any mercury is likely to be contained within sealed components and hence should not be radiologically contaminated. This will be verified prior to disposal and or recycle. There are approximately 55 devices in this category. l l l 4-93
l
)
Sixt:n Nucle:r Experiment;l Carp rati:n (SNEC) Site Characterization Report 4.5 DATA TABLES Table 4-39, AREA 1,(-5 mR/h General Area)
- Denotes Average Contact Exposure Rate Values Unless Otherwise Noted COMPONENTS & MEAN SMEAR RANGE OF SMEAR
- CONTACT DIRECT FRISK LOCATIONS DPM/100CM2 DPM/100CM2 m R/h CPM 765' 8" El. -33,100 -2,100 to -120,000 Not Taken -24,000 CONCRETE FLOOR 768' 3" El. Concrete ~15,000 ~2,300 to -45,000 Not Taken -10,000 Ledge 765* 2" El. -107,000 -50,000 to -160,000 -2.5 Inside N/A Sump 775' 2" El. - 1,100 1,000 to -2,000 Not Taken -770 Concrete Ceiling Concrete Walls -6,700 -250 to -67,000 Not Taken -2,300 781' 4" El. -1,400 1,000 to -2,000 Not Taken ~5,400 Steel Platform Other Surfaces -41,000 -360 to -480,000 Not Taken -12,000 instrument Racks 1 & 2 -64,000 -50,000 to ~75,000 -16 Not Taken Shutdown Cooling Heat Not Taken Not Taken -18.4 Not Taken Exchanger Shutdown Cooling -654,000 -500,000 to -796,000 -12 Not Taken Pumps Cond. Return Pump -500 MDA to -880 0.2 Not Taken Cond. Return Tank -2,600 MDA to -11,600 0.2 Not Taken PZR Discharge Tank -7,000 Not Taken -63.4 Not Taken One Smear Discharge Tank ~10,600 Not Taken -5.9 -10,800 Pumps One Smear Sump Pumps Not Taken Not Taken -2.5 in Sump by Not Taken Pumps SW Demin. Suction Fitter Not Taken Not Taken -3.2 Not Taken Punfication Filter -130,000 Not Taken -2.6 Not Taken One Smear SW Demin. Discharge Not Taken Not Taken -6.5 Not Taken Filter Rod Room Air Handler -880 -360 to -1,400 < 0.2 -750 Rod Room Vent Fan -2,100 ~750 to -3,500 Not Taken -700 CRDM's -3,000 -1,000 to 6,600 Not Taken -4,000 One Smear Fuel Storage Rack -8,300 -8,100 to -8,400 Not Taken Not Taken O
b 4 94
S2xt:n Nucle:r Experimental C:rpor tion (SNEC) l Site Characterization Report t n
' Table 4-40, AREA 2,(~11 mR/h General Area) l
! COMPONENTS & MEAN SMEAR RANGE OF SMEAR
- CONTACT DIRECT FRISK j LOCATIONS DPM/100CM2 DPM/100CM2 m R/h CPM 770' 8" El. -55,000 -1,800 to -1.1E6 Not Taken -21,000 Concrete Floor 814' 6" El. -800 -240 to -1,000 Not Taken ~800 Concrete Ceiling l Concrete Walls -1,800 -110 to -24,000 Not Taken -10,600 l
789' 4" El. -107,000 -1,500 to ~300,000 Not Taken Not Taken Steel Platform 795' 2" El. ~8,700 -6,200 to -11,000 Not Taken ~7,700 Steel Platform ! 807'El. -3,500 -2,000 to -6,000 Not Taken Not Taken Steel Platform Other Surfaces ~75,000 -1,150 to -570,000 Not Taken -7,800 Steam Generator ~5,300 -2,000 to -13,500 -23.2 Not Taken Pressurizer -9,100 -980 to ~33,400 -2. 7 Not Taken Reactor Coolant Pump ~55,210 -4,500 to -180,000 -2.6 Not Taken l Regenerative Heat -22,200 -1,100 to -100,000 -101 Not Taken Exchanger Non-Regenerative Heat -4,000 -1,300 to -9,100 -6. 3 -12,000 Exchanger l Instrument Rack ~571,000 -32,000 to ~750,000 -0.6 Not Taken Over Head Air Not Taken Not Taken <1.0 Not Taken Handler Pnmary Air Handler -15,100 -2,700 to -27,000 -4 -?5,000 l l l l 4-95 I l
l Saxt:n Nuclear Experimental C:rptrati:n (SNEC) Site Characterization Report b G Table 4-41, AREA 3,(~3 mR/h General Area) COMPONENTS & MEAN SMEAR RANGE OF SMEAR
- CONTACT DIRECT FRISK LOCATIONS DPM/100CM2 DPM/100CM2 mR/h CPM 781' 4" El. ~8,100 -1,200 to -47,500 Not Taken -7,800 Steel Platform 795* 2" El. -2,000 -860 to -4,400 Not Taken -3,000 Steel Platform Concrete Ceiling ~790 -230 to -1,000 Not Taken Not Taken f
Concrete -610 -230 to -4,900 Not Taken -6,600 Walls ! Other Surfaces -6,100 -460 to ~59,000 Not Taken ~5,900 Storage Well Heat Not Taken Not Taken 1.4 Not Taken Exchanger Storage Well Not Taken Not Taken 0.9 Not Taken Pumps Component Cooling Heat -3,800 -1,700 to -8,600 0.2 -500 j Exchanger ! Component Cooling < 1,000 < 1,000 Not Taken 950 Storage Tank ( , Component Cooling -16,500 -6,900 to -38,400 0.2 -2,100 l Pumps l k Auxiliary Air Handler
~7,000 -1,900 to -12,000 1.0 -12,000 Operating Area -13,500 -1,000 to -26,000 Not Taken -21,000 l Air Handler I
l l l l l 4 I \,s l 4-96
Sixt:n Nucliar Experimintal C rp: ration (SNEC) Site Characterization Report c Table 4-42, AREA 4,(~0.2 mR/h General Area) COMPONENTS & MEAN SMEAR RANGE OF SMEAR
- CONTACT DIRECT FRISK LOCATIONS DPM/100CM2 DPM1100CM2 m R/h CPM 812' El. Ops. Floor Concrete ~360 -250 to -1,500 Not Taken -1,700 818' El. Ops. Floor Concrete -1,030 -1,000 to -1,500 Not Taken -450 818' El. Steel -1,000 si,000 Not Taken -3,500 Platform Concrete Walls -270 -270 to -1,000 Not Taken -380 812' to 818' El. Steel Walls -270 -250 to -370 Not Taken -100 818' to 845' ~280 -260 to -520 Not Taken Not Taken El. Steel Walls MovaMe Bridge -870 -270 to -2,000 Not Taken -800 Polar Crane -1,300 -280 to -3,500 Not Taken -2,900 CV Upper Dome 1,000 51,000 Not Taken -60 Equipment Access Hatch -290 -250to~370 Not Taken -1,200
, Personnel Access Hatch ~530 -260 to -9,100 Not Taken -200 Escape Hatch -1,700 -790 to -8,900 <02 Not Taken i
Other Surfaces -1,200 -250 to -5,400 Not Taken -1,400 l Ventilation -870 -270 to -2,200 Not Taken Not Taken Cavity Eq. Rails Contaminated ~294,000 -1000 to ~4.8e6 -0. 3 Not Taken Pumps (3 ea) Electrical Distnbution Box ~3,500 -1,040 to -5.300 Not Taken Not Taken ToolRack -1,500 Not Taken Not Taken -2,800 One Smear Fuel Handling Tools Not Taken Not Taken Not Taken Not Taken Teleflex Shield ~330 -260 to -1,100 Not Taken Not Taken Fuel Transfer Cask Skid -400 -250 to ~520 Not Taken -600 Head Stand -290 -270 to -350 Not Taken -600 Lights Not Taken Not Taken Not Taken Not Taken Air Circulation Fan -6,400 -1,000 to -12,000 Not Taken ~16,800 f O 4 97
S:xt:n Nucitar Experirnental Ccrpcration (SNEC) Site Characterization Report O Table 4-43, AREA 5,(-0.01 mR/h General Area) COMPONENTS & MEAN SMEAR RANGE OF SMEAR
- CONTACT DIRECT FRISK LOCATIONS DPM/100CM2 DPM/100CM2 mR/h CPM 803* El. Concrete Floor Water Covered Water Covered Water Water Covered Covered 803* to 811' El. Concrete < 220 s220 Not Taken ~30 Walls Concrete Ceiling < 220 8220 Not Taken ~10 Exposed Steel Not Taken Not Taken Not Taken Not Taken l
l l 4 4-98
l Sixt:n Nucle:r Experim;ntal Ccrporation (SNEC) Site Characterization Report O t U Table 4-44, AREA 6,(~200 mR/h General Area) COMPONENTS & MEAN SMEAR RANGE OF SMEAR ' CONTACT DIRECT FRISK LOCATIONS DPM/100CM2 DPM/100CM2 m R/h CPM 765' 8" El. Concrete Floor ~2.7E6 -2.4E6 to ~3E6 Not Taken N/A l l 807' El. Concrete Ceiling Not Taken Not Taken Not Taken Not Taken l l Concrete Walls Reactor -31,600 -2,000 to -100,000 Not Taken N/A Side ! Concrete Walls -470,000 -20,000 to ~4.8E6 Not Taken N/A Storage Well Side 793' 2" El. -138,000 -55,000 to ~323,000 Not Taken N/A Steel Platform 768' 3" El. -1.74E6 -420,000 to ~2.4E6 Not Taken N/A Concrete Ledge l 779' 8" El. -200,000 Not Taken Not Taken N/A l Concrete Floor One Smear 800' 6" El. -240,000 -20,000 to -1.1E6 Not Taken N/A ! Steel Platform Reactor Vessel ~8.5ES -60,000 to -2.4E6 -5,000 N/A (Highest) Reactor Vessel -29,600 -2,900 to -140,000 Not Taken N/A Upper Support R"E'l;r Vessel Piping Not Taken Not Taken ~52 N/A , '\. Reactor Vessel Support 8.5E5 -60,000 to ~2.4E6 -5,000 N/A Structure (From RV) RV Head Lift Rigging Not Taken Not Taken Not Taken N/A l l Internals Rigging Not Taken Not Taken Not Taken N/A Support Stand Not Taken Not Taken Not Taken N/A Spent Fuel Pool Elevator -150,000 -100,000 to ~200,000 Not Taken N/A Super Heated Test Not Taken Not Taken -12 N/A Equipment Tnpod Not Taken Not Taken Not TaKen N/A l Spent Fuel -134,000 -39,000 to -424,000 Not Taken N/A Storage Racks Bone Acid ~1.8E6 Not Taken -8,333 N/A Demineralizer One Smear Storage Well ~4.8E6 Not Taken -282 N/A Demineralizer Ono Smear Purification ~60,000 Not Taken -242 N/A l Deminerarizer One Smear l Other Surfaces -150,000 -100,000 to -200,000 Not Taken N/A l f v i L ! 4-99 l
Stxtsn Nuctur Experimintal Ccrporation (SNEC) Site Characterization Report t \ Table 4-45, AREA 7,(~0.017 mR/h General Area around CV outside) COMPONENTS & MEAN SMEAR RANGE OF SMEAR
- CONTACT DIRECT FRISK LOCATIONS DPM/100CM2 DPM/100CM2 mR/h CPM 811'to 81T El. <1,000 s1,000 Not Taken -BKGND Steel Walls 81T to 845'
<224 s224 Not Taken -70
! El. Steel Walls l 845' to 870'
<224 s224 Not Taken -230 El. Steel Dome Concrete Wall <1,000 s 1,000 Not Taken -BKGND (Horizontal)
Concrete Wall <1,000 s 1,000 Not Taken -BKGND (Vertical) l l , O \ ! l I i ) v 4-100
Ssman Nuclerr Experim:ntal C rpor: tion (SNEC) Site Characterization Report ! Table 4-46, AREA 8,(-0.014 mR/h General Area average) COMPONENTS & MEAN SMEAR RANGE OF SMEAR
- CONTACT DIRECT FRISK LOCATIONS DPh/100CM2 DPM1100CM2 m R/h CPM Septic Tank "A" 0.023 G.A in Tank Concrete Floor Sediment Covered Sediment Covered Sediment N/A Covered Concrete Walis Not Taken Not Taken Not Taken Not Taken Concrete Ceiling Not Taken Not Taken Not Taken Not Taken ,
Concrete Manway Not Taken Not Taken 0.11 ~BKGND Septic Tan,M 0.012 GA in Tank Concrete Ftoor Sedimeni Oovered Sediment Covered Sediment N/A Covered Concrete Walls Not Taken Not Taken Not Taken Not Taken ; 1 Concrete Ceiling Not Taken Not Taken Not Taken Not Taken Concrete Manway Not Taken Not Taken 0.065 -BKGND
\
Pomo Well 0.010 GA in well Concrete Floor Sediment Sediment Sediment N/A Covered Covered Covered i Concrete Walls Not Taken Not Taken Not Taken Not Taken Concrete Ceiling Not Taken Not Taken No* Taken Not Taken Concrete Manway Not Taken Not Taken 0.035 ~BKGND Concrete Block House Not Taken Not Taken Not Taken Not Taken f.!99.C Concrete Chlonnator/ Weir Concrete Floor 239 s239 Not Taken ~BKGND Concrete Walls 239 s239 0.010 -BKGND
\
V 4-101
, S xt:n Nucl:ar Experimental Carp;rttiin (SNEC)
Site Characterization Report Table 4-47 to Table 4-51, Alpha Surface Contamination Results l The following tables present typical AREA by AREA alpha surface contamination smear results (in DPM/100 cm*), for the SNEC Containment building structural components (AREAS 1, 2, 3, 4 & 6). Table 4 47, AREA 1 Mean Smear Range of Smea7 Structural Surface DPM/100cm2 DPM/100cm2 765' 8" El. -30 8 to -90 Concrete Floor 768' 3" El. -20 8 to ~30 Concrete Ledge 765' 2" El. -70 -20 to -130 CV Sump 775' 2" El. -5 Not Taken Concrete Ceiling One Smear Concrete Walls -10 8 to -80 ] i I 781' 4" El. -5 Not Taken a , Steel Platform One smear ! Other Misc. Surfaces -40 8 to ~570 , t Table 4-48, AREA 2 Mean Smear Range of Smear Structural Surface DP34/100cm2 DPM/100cm2 779' 8" El. ~30 8 to -120 Concrete Floor , 814' 6" El. -10 5 to -10
. Concrete Ceiling Concrete Walls -8 -8 to -10
~
789' El. Steam ~670 8 to 4700 4 Generator Platform 9 795' El. Steam ~24 24 to -24 Generator Platform 870' El. Steam -10 5 to -25 Generator Platform Other Misc. Surfaces -10 -O to -350 1 4-102
S:xt:n Nucirr ExperimIntal Ccrptr:tien (SNEC) Site Characterization Report Table 4-49, AREA 3 Mean Smear Range of Smear Structural Surface DPM/100cm2 DPM/100cm2 781' 4" El. -10 8 to -50 Steel Platform 795' 2" El. -10 8 to -20 Steel Platform - Concrete Ceiling 5 Not Taken ! One Smear l Concrete walls -8 8 to -10 Other Misc. Surfaces -10 8 to -60 ; fable 4-50, AREA 4 Mean Smear Range of Smear Structural Sur' ace DPM/100cm2 DPM/100cm2 812' El. Opers ting 8 8 Floor Coner ate 818' El. Conccete 5 5 Platform ( 818' EL Steel 5 5 ( Platform Concrete walls 5 5 Steel Walls 8 8 Polar Crane -5 8 Other Misc. Surfaces -8 8 to -10 I 1 4-103
Saxt:n Nucle r Experimtntal Ccrporation (SNEC) Site Characterization Report l Table 4-51, AREA 6 Mean Smear Range of Smear i Structural Surface DPM/100cm2 DPM/100cm2 765* 8" El. Storage ~560 -280 to -840 Well l Concrete Floor l Concrete Walls -130 -6 to -870 Reactor Side l Concrete Walls -460 -30 to ~3,500 Storage Well Side 793' 2" El. -1,400 -80 to -4,800 Steel Platform 768' 3" EL SW -880 -230 to -1,500 Concrete Ledge Reactor Vessel ~380 -10 to -2,700 Support Can Fuel Storage Rack ~1,100 ~700 to -1,900 1 800' 6" El. Reactor -110 -60 to -200 l Side Steel Platform l l l l i I I 4-104
Saxton Nuclear ExperimIntal C rporati:n (SNEC) Site Characterization Report k Table 4-52, Air Particulate Station Results (Activity in pCi/MS ) GPUID Sample Description Stop Date & Time Nuclide Positive Result 77618 SX-AP-Al -2 072894@1335 Be-7 0.1400
- 0.0500 Gross Alpha 0.0017
- 0.0006
~
Gross Beta 0.0180
- 0.0030 77711 SX-AP-AI -2 080494 Be-7 0.1700
- 0.0400 Gross Alpha 0.0014
- 0.0008 Gross Beta 0.0210
- 0.0020 77830 SX-AP-A1 -2 081194 Be-7 0.1200
- 0.0500 Gross Alpha 0.0014
- 0.0008 Gross Beta 0.0140
- 0.0030 78073 SX-AP-Al -2 081894 @ 1000 Be-7 0.1100
- 0.0600 Gross Alpha 0.0021
- 0.0008 Gross Beta 0.0210
- 0.0030 78078 SX-AP-Al -2 082594 @ 1520 Be-7 0.2000
- 0.0800 Gross Alpha 0.0020
- 0.0008 Gross Beta 0.0140
- 0.0030 78164 SX-AP-A1 -2 090294@0930 Be-7 0.1600
- 0.0400 Gross Alpha 0.0018
- 0.0007 Gross Beta O C50
- 0.0040 78355 SX-AP-Al -2 090994@0845 Be-7 0.1400
- 0.0600 Gross Alpha 0.0016
- 0.0007 Gross Beta 0.0170
- 0.0030 78359 SX-AP-Al -2 091294 @ l130 Be-7 0.3000
- 0.1300 Gross Alpha 0.0040
- 0.0015 Gross Beta 0.0560
- 0.0090 78551 SX-AP-Al -2 100194 @ 0740 Be-7 0.1000
- 0.G400 Gross Alpha 0.0010
- 0.0005 Gross Beta 0.0100
- 0.0030 78900 SX-AP-Al -2 100794@1440 Be-7 0.1000
- 0.0600 Gross Alpha 0.0019
- 0.0008 Gross Beta 0.0140
- 0.0040 78901 SX-AP-Al -2 100794 @ 1440 Be-7 0.1300
- 0.0500 Gross Alpha 0.0014
- 0.0008 Gross Beta 0.0180
- 0.0G40 78905 SX-AP-Al -2 101494@1100 Be-7 0.1200
- 0.0500 Gross Alpha 0.0015
- 0.0007 Gross Beta 0.0160
- 0.0030 78909 SX-AP-Al -2 102194 @ 1250 Be-7 0.1500
- 0.0500 Gross Alpha 0.0017
- 0.0007 Gross Beta 0.0230
- 0.0040 81923 SX-AP-AI -2 051895 @ l105 Pe-7 0.0790
- 0.0600 Gross Alpha 0.0005
- 0.0003 Gross Beta 0.0100
- 0.0030 82000 SX-AP-A1 -2 052695 @ l400 Be-7 0.0860
- 0.0810 Gross Beta 0.0079
- 0.0026 82001 SX-AP-A1 -2 052695@l400 Be-7 0.I400
- O G400 Gross Beta 0.0120
- 0.0030 N 82826 SX-AP-Al -2 061695@0930 Be-7 0.0130
- 0.0800 Gross Alpha 0.0022
- 0.0011 4-105
S xt:n Nucle:r Experim2ntal Corporaticn (SNEC) Site Characterization Report l GPU ID Sample Description Stop Date & Time Nuclide Positive Result ! I Gross Beta 0.0093
- 0.0023 82829 SX-AP-Al -2 062295@1215 Be-7 0.1700
- 0.0900 Gross Alpha 0.0052
- 0.0015 Gross Beta 0.0260
- 0.0040 82832 SX-AP-A1 -2 062995@l115 Be-7 0.0860
- 0.0390 Gross Beta 0.0066
- 0.0024 82835 SX-AP-Al -2 070695@1000 Be 7 0.1700
- 0.0500 Gross Alpha 0.0029
- 0.0012
]
Gross Beta 0.0140
- 0.0030 J 82871 SX-AP-Al -2 071395@1200 Be-7 0.0530
- 0.0340 l Gross Alpha 0.0024
- 0.0011 i Gross Beta 0.0090
- 0.0025 78549 SX-AP-A1 -2 092394@0810 Gross Alpha 0.0022
- 0.0008 l Gross Beta 0.0120
- 0.0030 82214 SX-AP-A1 -2 060195@l120 Be-7 0.0750
- 0.0430 Gross Beta 0.0059
- 0.0032 82218 SX-AP-Al -2 060895@1545 Be-7 0.1000
- 0.0400 Gross Alpha 0.0008
- 0.0005 )
i Gross Beta 0.0140
- 0.0030 l 77622 SX-AP-D.1 072194@2340 Gross Alpha 0.0040
- 0.0025
'~
Gross Beta 0.0130
- 0.0070 SX-AP-DI -1 Gross Beta 0.0250
- 0.0100 m 77708 070 94 (@ 0906 ,
lf I b 77832 SX-AP-D1 -1 0811.'4 @ 1435 Be-7 Gross Alpha 0.1200
- 0.0700 0.0019
- 0.0009 l
l Gioss Beta 0.0130
- 0.0030 78074 SX-AP-D1 -1 081G4 @ 0911 Be-7 0.1300
- 0.0900 Gross Alpha 0.0044
- 0.0019 Gross Beta 0.0360
- 0.0070 78079 SX-AP-D1 -1 082594@l530 Be-7 0.1300
- 0.0600 Gross Alpha 0.0017
- 0.0008 Gross Beta 0.0160
- 0.0030 78165 SX-AP-D1 -1 090294@0950 Be-7 0.I100
- 0.0500 Gross Alpha 0.0022
- 0.0008 Gross Beta 0.0250
- 0.0040 78358 SX-AP-D1 -1 090994 @ 1000 Be-7 0.1300
- 0.0700 Gross Alpha 0.0018
- 0.0007 Gross Beta 0.0160
- 0.0030 78362 SX-AP-D1 -1 091694 @ 1410 Be-7 0.0910
- 0.0570 Gross Alpha 0.0015
- 0.0006 Gross Beta 0.0260
- 0.0040 78553 SX-AP-D1 -1 100194@1040 Be-7 0.0790
- 0.0570 Gross Alpha 0.0009
- 0.0005 Gross Beta 0.0095
- 0.0026 78903 SX-AP-D1 -1 100794@1510 Be 7 0 1300
- 0.0700 Gross Alpha 0.0014
- 0.0008 i Gross Beta 0.0120
- 0.0030 A 78908 SX-AP-D1 -1 101494 @ l400 Be 7 0.1300
- 0.0700
[ f g Gross Alpha 0.0021
- 0.0008 I Gross Beta 0.0160
- 0.0030 78910 SX-AP-D1 -1 102194 @ 1140 Be-7 0.1500
- 0.0600 i 4-106 i
l
Saxtrn Nucle:r Experimental Ccrp2ratian (SNEC) Site Characterization Report A U GPU ID Sample Description Stop Date & Time Nuclide Positive Result Gross Alpha 0.0015
- 0.0007 Gross Beta 0.0220
- 0.0040 81924 SX-AP-D1 1 051895@l100 Be-7 0.1900
- 0.0900 Gross Alpha 0.0016
- 0.0011 Gross Beta 0.0170
- 0.0080 82003 SX-AP-D1 -1 052695 @ 1345 Be-7 0.1300
- 0.0300 Gross Alpha 0.0008
- 0.0006 Gross Beta 0.0120
- 0.0030 82827 SX-AP-DI -l 061695 @ 0835 Be-7 0.1500
- 0.0700 Gross Alpha 0.0027
- 0.0011 Gross Beta 0.0120
- 0.0020 82830 SX-AP-D1 -1 062295@1200 Be-7 0.2500
- 0.I100 Gross Alpha 0.0057
- 0.0015 Gross Beta 0.0270 0.0040 82833 SX-AP-D1 -1 062995@l110 Be-7 0.% 30
- 0.0440 Gross Alpha 0.0022
- 0.0011 Gross Beta 0.0057
- 0.0024 82836 SX-AP-D1 -1 070695@1010 Be-7 0.0930
- 0.0570 Gross Alpha 0.0020
- 0.001I Gross Beta 0.0130
- 0.0030 82872 SX-AP-D1 1 071395@1550 Be-7 0.1300
- 0.0600 Gross Alpha 0.0032
- 0.0011 82872 SX-AP-D1 -1 071395 @ 1550 Gross Beta 0.0160
- 0.0030
\d 83145 SX-AP-D1 1 07:095 @ 0940 Be-7 0.1600
- 0.0500 Gross Alpha 0.0031
- 0.0013 Gross Beta 0.0310
- 0.0030 83148 SX-AP-D1 -1 072795 @ 0930 Be-7 0.I400
- 0.0700 Gross Alpha 0.0023
- 0.0011 Gross Beta 0.0250
- 0.0030 78548 SX-AP-D1 -1 092394 @ 0830 Be-7 0.1100
- 0.N00 Gross Alpha 0.0015
- 0.0007 Gross Beta 0.0160
- 0.0030 82215 SX-AP-DI -1 060195@l115 Gross Beta 0.0100
- 0.0040 82219 SX-AP-DI -1 060895@1535 Be-7 0.1100
- 0.0600 Gross Alpha 0.0008
- 0.0005 Gross Beta 0.0150
- 0.0030 77620 SX-AP-G10-1 072894 @ 0915 Be-7 0.1500
- 0.0500 Gross Alpha 0.0009
- 0.00N Gross Beta 0.0160
- 0.0030 77621 SX-AP-G10-1 072894 @ 0915 Be-7 0.1100
- 0.0800 Gross Alpha 0.0014
- 0.0005 Gross Beta 0.0140
- 0.0030 77710 SX-AP410-1 080494 @ 0900 Be-7 0.1400
- 0.0400 Gross Beta 0.0210
- 0.0020 77831 SX-AP-G10-1 081194 @ 1530 Be-7 0.1400
- 0 0800 Gross Alpha 0.0015
- 0.0008
( 78076 Gross Beta 0.0130
- 0.0030 --
SX-AP-G 10-1 081894 @ 1610 Be-7 0.1000
- 0.0700 Gross Alpha 0.0022
- 0.0008 Gross Beta 0.0190
- 0.0030 4-107
Sixt:n Nucirr Experiment:1 Ccrpsratisn (SNEC) Site Characterization Report GPU ID Sample Description Stop Date & Time Nuclide Positive Result 78077 SX-AP-G10-1 Be-7 0.1300
- 0.0600 082]4@1445 Gross Alpha 0.0018
- 0.0008 Gross Beta 0.0190
- 0.0030 78167 SX-AP-G101 090294 @ 1505 Be 7 0.1100
- 0.0400 Gross Alpha 0.0018
- 0.0007 Gross Beta 0.0230
- 0.0030 78357 SX-AP-G10-1 090994 @ 0900 Be-7 0.1500
- 0.0600 Gross Alpha 0.0021
- 0.0007 Gross Beta 0.0130
- 0.0030 78363 SX-AP-G10-1 091694 @ 1320 Be-7 0.1900
- 0.0600 Gross Alpha 0.0016
- 0.0007 Gross Beta 0.0230
- 0.0040 78550 SX-AP-G10-1 092394 @ 0940 Be-7 0.0960
- 0. % 70 Gross Alpha 0.0013
- 0.0006 Gross Beta 0.0160
- 0.0030 78554 SX-AP-G101 100194@l500 Be-7 0.0950
- 0.0350 Gross Alpha 0.0014
- 0.0006 Gross Beta 0.0120
- 0.0030 78904 SX-AP-G10-1 100794 @ 1615 Be 7 0.1600
- 0.0500 Gross Alpha 0.0011
- 0.0007 Gross Beta 0.0099
- 0.0034 78907 SX-AP-G10-1 101494 @-1340 Be-7 0.1100
- 0.0500 l Gross Alpha 0.0015
- 0.0007 Gross Beta 0.0130
- 0.0030 )
78911 SX-AP-G10-1 102194 @ IG40 Be-7 0.2400
- 0.0700 l Gross Alpha 0.0013
- 0.0007 Gross Beta 0.0200
- 0.0040 81922 SX-AP-G10-1 051195@-1400 Be-7 0.1400
- 0.0500 Gross Alpha 0.0014
- 0.0005 Grou Beta 0.0130
- 0.0030 81925 SX-AP-G10-1 051895 @ 0840 Be-7 0.1400
- 0.0700 Gross Beta 0.0072
- 0.0025 82004 SX-AP-G10-1 052695@l345 Be-7 0.1300
- 0.0500 Gross Beta 0.0077
- 0.0026 82216 SX-AP-G10-1 060195 @ 0845 Be-7 0.1000
- 0.0600 Gross Beta 0.0100
- 0.0040 82220 SX-AP-G101 060895 @ 1430 Be-7 0.1300
- 0.0800 Gross Beta 0.0130
- 0.0030 82221 SX-AP-G10-1 060895 @ 1430 Be-7 0.1200
- 0.0500 Gross Alpha 0.0013
- 0.0006 Gross Beta 0.0150
- 0.0030 82828 SX-AP-G10-1 061595 @ 1015 Be-7 0.1100
- 0.0600 Gross Alpha 0.0031
- 0.0013 ;
Gross Beta 0.0140
- 0.0030 82831 SX-AP-G10-1 062295 @ 1045 Be-7 0.2200
- 0.0500 Gross Alpha 0.0041
- 0.0012 Gross Beta 0.0210
- 0.0030 t 82834 SX-AP-G10-1 062995 @ 1320 Gross Beta 0.0050
- 0.0023 82837 SX-AP-G10-1 070695 @ 0900 Be-7 0.0890
- 0.0560 Gross Alpha 0.0020 i 0.0012 4-108 I
l l S:xt:n Nuclarr Experiment-l Ccrptrati n (SNEC) i n Site Characteriz.ation Report ! ( i
\
GPU ID o-.inple Description Stop Date & Time Nuclide Positive Result Gross Beta 0.0140
- 0.0030 1 82873 SX-AP410-1 071395@ll20 Gross Beta 0.0130
- 0.0050 83146 SX-AP-G10-1 072095@0800 Be-7 0.1300
- 0.0400 l Gross Alpha 0.0023
- 0.0012 l Gross Beta 0.0250
- 0.0030 83149 SX-AP-G10-1 072795@0900 Be-7 0.1200
- 0.0400 Gross Alpha 0.0019
- 0.0011 l Gross Beta 0.0220
- 0.0030 77619 SX-AP-J1 -2 072894@1047 Be-7 0.2700
- 0.1000 l Gross Beta 0.0150
- 0.0090 I
! 77709 SX-AP-Jl 2 072994@l454- Gross Beta 0.0340
- 0.0080 77833 SX-AP-J1 -2 081194@1420 Be-7 0.1800
- 700 Gross Alpha 0.0014
- 0.0008 4 Gross Beta 0.0130
- 0.0030 l
, 78075 SX-AP-J1 -2 0818 N @ 1030 Be-7 0.1100
- 0.0600 I Gross Alpha 0.0023
- 0.0008 Gross Beta 0.0230
- 0.0W0 78080 SX-AP-Jl -2 082594@l540 Be-7 0.0980
- 0.0570 Gross Alpha 0.0019
- 0.0008 l Gross Beta 0.0170
- 0.0030 78081 SX-AP-Jl -2 082594@1540 Be-7 0.1600
- 0.0400 Gross Alpha 0.0014
- 0.0007 '
[ Gross Beta 0.0180
- 0.0030 i
( 78166 SX-AP-Ji -2 090294@1000 Be-7 0.1300
- 0.0800 l Gross Alpha 0.0020
- 0.0008 Gross Beta 0.0220
- 0.0030 78356 SX-AP-J1 -2 090994@0905 Gross Alpha 0.0011
- 0.0006 ;
Gross Beta 0.0047
- 0.0026 78360 SX AP-JI -2 091694@1415 Gross Alpha 0.0016
- 0.0010 Gross Beta 0.0260
- 0.0060 78361 SX-AP-Jl -2 091694@1415 Be-7 0.1300
- 700 Gross Alpha 0.0021
- 011 Gross Beta 0.0260
- 060 78547 SX-AP Jl -2 092394@0840 Gross Alpha 0.0011
- 0.0006 Gross Beta 0.0140
- 0.0030 78552 SX-AP-Jl -2 100194 @ 0800 Be-7 0.1000
- 0.0600 Grt,ss Alpha 0.0009
- 0.0005 Gross Beta 0.0085
- 0.0026 K-40 0.0970 A 0.0500 78902 SX-AP-J1 -2 100794@1525 Be-7 0.1200
- 0.0500 Gross Alpha 0.0018
- 0.0008 78902 SX-AP-J1 -2 100794 @ 1525 Gross Beta 0.0140
- 0.0N0 78906 SX-AP J1 -2 101494 @ 1410 Be-7 0.1600
- 0.0400 Gross AIM;a 0.0024
- 0.0008 Gross Beta 0.0180
- 0.0030 78912 SX-AP-Jl -2 102194@l145 Be-7 0.0830
- 0.0570 Gross Alpha 0.0018
- 0.0007 l f]i
( Gross Beta 0.0230
- 0.0MO 81926 SX-AP-J1 -3 051895@1045 Gross Alpha 0.0006
- 0.0003 Gross Beta 0.0110
- 0.0030 4 109
I 1 l 1 l l SaxtIn Nucirr Experimental C:rpar;tirn (SNEC) Site Characterization Report
- O \
{ \ l GPU ID Sample Description Stop Date & Time Nuclide Positive Result l 82002 SX-AP-Jl 3 052695@l330 Be-7 0.1300
- 0.0500 l Gross Beta 0.0100
- 0.0030 f 83147 SX-AP-Jl -3 072095@0930 Be-7 0.1900
- 0.0500
{ Gross Alpha 0.0025
- 0.0012 i Gross Beta 0.0290
- 0.0030 83150 SX-AP.Jl -3 072795@l315 Be-7 0.1300
- 0.0600 Gross Alpha 0.0021
- 0.0011 l Gross Beta 0.0270
- 0.0030 82217 SX-AP-11 -3 060195@l100 Be-7 0.0960
- 0.0640
- Gross Beta 0.0064
- 0.0033
! 82222 SX-AP-J1 -3 060895@1530 Be-7 0.1200
- 0.0500 Gross Alpha 0.0008
- 0.0005 Gross Beta 0.0150
- 0.0030 i
f l l N
\ 'm) l l
l l i l l
#\
U 4-110
i Saxton Nuclear Experimental Corporation (SNEC) Site Characterization Report i Q Table 4-53, Counting Results, On Site Core Bore Locations inside CV Sample Coee Average oft /B Average Of T/B Top to Bottom Average Of T/B Depth j N enher Since Counting Resuks Counting Results Ratio (Cs-137) Counting Results (cm) l l (Note 6) Co.60 pC1/g Cs-137 pCl/g (Note 1) Fm.152 pG's (Note 4) l 410008 1 <0.50 28.45 1 3.69 1.87 < MDA(See Note 2) 13 410008 2 <0.26 <032 N/A < MDA 2.9 l 410008 3 (0.33 <032 N/A < MDA 4.5 410009 1 <0.56 10.12 1 1.58 2.09 < MDA 13 410009 2 <0.25 <0.24 N/A < MDA 2.9 410009 3 <034 <0.27 N/A <MDA 4.5 410012 1 <0.89 13.07 1 2.19 1.90 < MDA 13 410012 2 <035 <034 N/A < MDA 2.9 420013 1 <0.82 3.7010.92 2.52 < MDA 13 420013 2 <0.29 <0.24 N/A < MDA 2.9 420014 1 <0.66 5.03 i 1.08 2.34 < MDA 1.3 1 420014 2 <033 <0.29 N/A < MDA 2.9 { 420015 1 <0.62 2.1910.65 1.89 < MDA 13 ! 420015 2 <0.29 <0.23 N/A < MDA 2.9 I' l 410016 1 <0.29 12.59 2 2.11 2.00 < MDA 13
% 410016 2 <0.31 0.2210.14 1.39 <MDA 2.9 410017 1 < l.08 53.47 1 7.18 1.62 < MDA 13 l
( l 410017 2 <0.36 0.7910.24 2.14 < MDA 2.9 410017 3 <0.31 <032 N/A <030 4.5 i 410017 4 <0 26 <0.23 N/A <0.27 6.1
~fOl7 4 5 <0.28 <0.29 - N/A <0.26 7.7
~
410018 I <0.74 21.76 3.25 1.83 < MDA 13 410018 2 <034 <0.33 N/A < MDA 2.9 { 320019 i < l.58 80.40 1 10.66 1.78 <MDA 1.3 320019 2 <0.49 2.5610.62 1.33 < MDA 2.9 320019 3 <034 1.511035 0.78 < MDA 4. 5 320019 5 <0.24 2.021034 0.89 < MDA 8.9 320019 7 <0.23 1.75 0 31 0.94 <0.20 26 320019 10 <0.24 1.5220.28 1.12 <0.52 94.9 !' 320019 12 <0.26 1.961033 0.95 0.97 0.21 101.9 320019 13 <0.29 1.90 1033 1.13 1.1910.24 103.5 j 320019 I4 (0.35 2.601041 0.89 1.5320.27 105.! 320019 16 <0.57 1.45 035 1.02 3.6110.59 114.6 ! 320019 17 <0.63 1.1010.32 0 87 4.001065 116.2 i i 320019 19 0.7210.30 <0.4 3 N/A 3.811063 120.6 320019 20 <0.92 19.02 1 2.56 l.99 5.4910.85 122.2 320020 1 <039 3.9410.65 2.05 < MDA 1.3
\ 320020 2 <0.26 0.35 0.17 1.03 < MDA 2.9
, \ 320020 3 <0.31 <032 N/A < MDA 4.5 } 1 320021 1 < l.40 24.16 1 3.61 2.20 < l .03 13 l i 4-111 1 I L i
Sixt:n Nucle r Experimintal C rpor tion (SNEC) Site Characterization Report hyle Ceee Average oft /B Average ort /B Top te Bottom Average ort /B Depth Number Slice Commeing Remuks Counthig Results Ratio (Cs-137) Counting Remuks (cm) (Nede 6) C+44 pC1/g Cs-137 pCVs (Note 1) Eu-152 pCl/g (Note 4) 320021 2 <031 0.7510.24 1.03 <0.28 2.9 320021 3 <0.34 <0.53 N/A <0.34 4.5 320021 4 <0.31 <0.60 N/A <031 6.1 320021 5 <033 <0.48 N/A <031 7.7 320021 15 <031 03610.17 1.09 <0.29 100 320021 17 <030 <034 N/A <0.29 117.1 320021 18 <0.28 <039 N/A <030 118.7 320021 19 <034 <0.37 N/A <031 120.3 320021 20 3.8420.75 13.54 2 1.87 2.05 <0.45 121.9 320022 1 <0.69 313.40 1 38.75 1.72 < MDA 13 320022 2 <0.25 13010.32 0.89 < MDA 2.9 320022 3 <0.28 0.8010.26 1.03 < MDA 4.5 320022 4 <0.30 0.7610.23 136 < MDA 6.1 320022 5 <0.27 0.5310.20 0.68 < MDA 7.7 320022 6 <0.26 0.5510.18 0.90 < MDA 93 320023 1 6.5010.97 1367.0 1 16835 1.57 < MDA 13 320023 2 <0.29 4.7710.75 1.64 < MDA 2.9 320023 3 <0.28 0.8820.25 1.I3 < MDA 4.5 320023 4 <0.28 <0.40 N/A < MDA 6.1 120024 1 <0.94 !!.8821.99 139 < MDA 13 120024 2 <0.28 8.6911.16 1.06 < MDA 2.9 120024 3 <0.41 7.0321.05 1.02 < MDA 4.5 120024 4 <0.47 6.5320.99 0.98 < MDA 6.1 120024 5 <0.42 7.86 i 1.15 1.08 < MDA 7.7 120024 6 <0.48 6.5611.00 1.02 < MDA 9.3 110023 1 3.2511.14 6074.0 ! 748.0 1.29 < MDA 13 110025 2 <0.51 1419.0 1 174.8 1.09 <3.04 2.9 110025 3 <0.68 227.10 t 2830 I 85 < MDA 4.5 110025 4 <038 1834 t 237 1.03 <0.47 6.1 110025 5 <0.85 5.7021.17 1.29 < MDA 7.7 110025 7 <037 2.2610.44 0.84 < MDA 18.3 110025 9 <036 2.8210.51 1.13 < MDA 36 110026 1 <2.11 6583.0 1 810.5 1.47 < MDA 13 110026 2 <039 222.5 1 27.55 1.11 < MDA 2.9 110026 3 <0.40 46.61 1 5.91 1.14 <MDA 4.5 110026 4 <0.37 10.87 i 1.50 1.14 < MDA 6.1 110026 5 <033 43210.70 1.08 < MDA 7.7 110026 6 <032 3.1410.57 0.85 < MDA 9.3 % 110027 1 < l .02 213.25 1 26.84 1.94 <2.20 13 110027 2 <035 2.7210.51 1.01 <032 2.9 110027 3 <035 16710.37 1.18 <MDA 45 110027 4 <0.26 1.261031 1.22 < MDA 6.1 4-112
S3xt:n Nucle'r Experimental Ctrporatirn (SNEC) ) Site Characterization Report ! I [' t l Sample Cm Average Of T/B Average Of T/B Top to Bottose Average oft /B Depth Nimeber Slice Counting Results Counting Results Ratio (Cs.137) Counting Results (cm) (Note 6) Co.60 pCl/g Cs-137 pC1/g (Note 1) Eu-152 pC1/g (Note 4) 110027 5 <034 1.561034 1.03 <0.29 7.7 110027 6 <036 1.4820.33 0.91 <031 93 110028 1 <0.42 121.87 t 15.18 1.52 <MDA 13 110028 2 <0.36 3.5020.60 i 00 < MDA 2.9 110028 3 <0.29 1.432034 1.01 < MDA 4.5 110028 4 <037 1.461035 0.91 < MDA 6.1 1 110028 5 <0.29 1.27 2030 1.24 < MDA 7.7 110028 6 <0.26 1361032 1.13 < MDA 93 110028 7 <0.26 1.1220.29 1.04 < MDA 10.9 120029 1 3.6510.69 438.95 1 54.19 1.51 < MDA 13 120029 2 <030 8.1611.19 1.13 < MDA 2.9 120029 3 <0.36 33020.59 1.04 < MDA 4.5 120029 4 <038 1.8810.40 130 < MDA 6.1 120029 5 <0.28 1.531033 1.23 < MDA 7.7 I 120029 6 <0.29 0.7610.23 0.91 < MDA 93 120029 7 <0.29 0.7110.22 1.10 < MDA i0.9 1200.30 1 <0.84 1136.8 t 140.5 1.28 < MDA 1.3 C' 120030 2 <0.63 56.417.46 132 < MDA 2.9 l g 120030 3 <0J0 9.00 t 1.27 0.87 < MDA 4.5 120030 4 <0.20 4.45 0.70 0.99 < MDA 6.1 120030 5 <0.20 33610.57 0.90 < MDA 7.7 120030 6 <0.29 2.7210.49 1.00 < MDA 9.3 220031 1 <0.97 189.80 1 23.95 1.74 <2.06 13 220031 2 <033 28,91 2 3.74 1.06 <0.51 2.9 220031 3 <036 30.52 1 3.95 1.00 <0.53 4.5 220031 4 <031 31.16 1 4.03 1.01 <0.53 6.1 220031 5 <0.27 29.02 1 3.76 0.98 <0.47 7.7 220031 6 <0.31 31.60 1 4.07 0.89 <0.49 93 220031 8 <0.24 38.12 1 4.87 1.02 (0.55 15.7 220031 10 <034 253713.29 0.99 <0.4 5 32.8 l 220031 11 <0.40 6.5510.99 0.95 <037 120.1 220031 13 <0.33 4.77 t 0.76 0.92 <032 123.3 220031 15 <0.31 3.8310.65 1.23 (035 135.4 220031 16 <0.29 3.6010.61 0.97 <033 137 220031 17 <0.34 4.2010.70 0.96 <037 138.6 220031 19 7.9521.15 42.211539 0.91 <0.62 148.4 220031 20 22.08 1 2.57 358.8514435 1.28 <!.56 151 210032 1 32.54 4.44 24435.0 t 3007.5 1.47 < MDA 1.3
<0.33 22.95 1 2.99 1.04 <MDA 2.9 (g 210032 210032 2
3 <0.26 17.18 2 2.29 0.91 <MDA 4.5 i 210032 4 <0.25 13.98 2 1.88 1.00 <MDA 6.1 ! 210032 3 <0.28 143011.93 1.02 (MDA 7.7 4-113
i l i 1 l l Saxttn Nuclerr Experim:ntti C::rotratien (SNEC) I
- Site Characterization Report l f%
k Semiple Core Average Of T/B Average Of T/B Top to Bottom Average Of T/B Depth Namiber Slice Comethig Results Counting Results Ratio (Cs-137) Counting Resuks (cm) (Note 6) Co-60 pCVs Cs.137 pCl/s (Note 1) Eu-152 pCl/g (Note 4) 210032 6 <0.34 13.55 2 1.84 0.93 < MDA 93 210033 1 <0.83 201.4012539 2.01 < MDA 13 210033 2 <0.27 8.7311.24 1.05 < MDA 2.9 210033 3 <0.29 5.9010.89 1.09 < MDA 4.5 210033 4 <0.3 4 5.1220.79 1.00 < MDA 6. I 210033 5 <037 5.7420.89 1.07 < MDA 7.7 210033 6 <030 5.4510.84 1.17 < MDA 93 ) I ! 220034 1 <!.13 705.218736 1.82 < MDA 13 220034 2 <0.63 11.18 1 1.90 0.97 <MDA 2.9 l 220034 3 <0.40 8.5311.22 1.08 < MDA 4.5 220034 4 <036 8.13 i 1.18 1.04 < MDA 6.1 220034 5 <0.27 7.07 t 1.05 0.94 < MDA 7.7 220034 6 <031 63610.95 0.90 < MDA 93 220034 7 <0.31 7.5311.10 1.02 <MDA 10.9 ; 220035 1 <0.65 22.03 2 3.29 1.64 < MDA 13 220035 2 <0.29 6.8511.01 1.01 < MDA 2.9 , 220035 3 <0.35 4.9910.79 1.07 < MDA 4.4 i 1
?
(' 220035 4 <035 36320.62 0.99 < MDA 6
\ 220035 5 <0.29 3.56 i 0.61 1.01 < MDA 7.6
- 220035 6 <0.28 5.2510.83 1.09 < MDA 9.1 i
- 220035 7 <031 44110.76 1.08 < MDA 10.7 220036 1 15.96 1 2.64 371.65 t 46.35 1.82 < MDA 13
! 220036 2 <0.64 34.83 1 4.88 1.07 < MDA 2.9 220036 3 <0.28 23.85 2 3.12 1.08 < MDA 4.5 l 220036 4 <033 193822.56 1.03 < MDA 6.1 220036 6 <034 15.04 1 2.02 1.07 < MDA 15.6 l j 220036 8 <030 15.65 t 2.10 0.92 <0.42 32.7 220036 9B <0.21 13.33 1 1.71 0.91 <0.27 54.9 [ 220036 9D <0.20 11.24 t 1.47 0.93 <0.26 79.1 720036 10 <0.26 83221.20 1.00 (038 106.4 I 220036 12 <0.26 9.05 1.29 1.05 <0.36 126.1 220036 14 <0.23 5.9210.90 0.99 <0.28 143.5 220036 15 <0.27 5.10 t 0.80 1.02 <0.28 145.4 220036 16 <030 4.8220.77 1.01 <035 147 j 220036 17 . <034 4.0110.67 1.06 <030 148.6 220036 18 <0.27 3.9510.66 1.08 <030 !$0.2 220036 19 <0.30 4.1010.68 1.07 <031 151.8 220036 20 < ! .01 37.24 t 533 1.62 <!.16 153.4 220037 <0.89 149.20 1 18.97 136 < MDA 13 g 1 [ 220037 2 <0.41 64.26 2 8.09 1.05 < MDA 2.9
< 220037 3 <0.28 31.20 1 4.03 1.02 < MDA 4.5 220037 4 <0.37 33.54 t 434 0 97 < MDA 6.1 4-114
Saxt:n Nucle!.r Experimental C:rparatisn (SNEC) Site Characterization Report i
\
Sample Core Average Of T/B Average Of T/B Top to Bottom Average oft /B Depth Number Slice Commeing Results Counting Results Ratio (Cs-137) Counting Residts (cm) ! (Note 6) Co.60 pC1/g Cs-137 pCl/g (Note 1) Eu-152 pC1/g (Note 4) l 220037 6 <032 23.72 1 3.09 1.00 <MDA 18.16 l l 220037 8 <034 18.91 2 2.51 1.03 <0.42 35.26 l l 220037 12 <036 17.28 1 2.32 1.05 <0.53 108.6 l 220037 13C <0.26 12.57 1 1.64 1.07 <0.73 !!7.2 l 220037 13B <0.26 10.00 t 1.33 1.02 0.8010.21 1I88 220037 14 <0.31 1034 i 1.M 1.06 1.5510.29 125.8 220037 16 <4.99 7.6311.13 1.05 5.1010.76 137.9 l- 220037 18 1.181037 6.4010.99 0.92 7.40 i 1.03 150 l 220037 19 < l.22 6.8421.05 1.04 7.20 t 1.02 151.6 l 220037 20 5.4920.89 50.48 2 6.41 0.48 6.9811.02 152.2 l 6100 % 1 16.14 1 2.67 18210.0 1 2241.0 1.61 <4.56 13 6100% 2_ <!32 874.45 2 107.75 1.03 < l.70 2.9 l 6100M 3 1.071036 693.20 2 85.46 0.99 < l .88 4.4 6100% 4 1.1910.39 721.65 1 38.95 1.00 < l.78 5.9 6100M 5 <0.67 729.15 1 89.84 0.94 <130 7.5 6100M 6 <0.59 711.65 1 87.7 0.99 <l.72 9.I j 6100% 7 0.8210.29 683.9 2 84.3 1.01 < l.76 10.6 l 6100% 8 <0.65 723.05 1 89.12 1.00 < l.60 12.2
\ 6100M 9 <0.61 683.25 1 84.21 0.99 < l.60 13.7 l 610057 1 81.151934 23305.0 1 2867.5 1.27 <8.63 13 610037 2 5.53i1.42 1161.5 t 143.6 1.07 <2.66 2.9
( 610037 3 <3.97 8 %.4 2 110.95 0.98 <3.93 4.5 610057 4 <4.16 789.75 2 97.86 1.01 <4.08 6.1 610057 5 <3.05 638.95 i 81.74 0.97 <3.58 7.7 610057 6 <4.22 820.15 1 101.65 1.02 <3.75 93 610057 7 < 1.70 648.051 8033 1.03 <3.04 10.9 620058 1 45.09 1 5.55 345.1 1 43.1 1.49 < MDA 13 i t 620058 2 11.68 2 2.10 100.5 1 13.0 1.03 < MDA 2.9 620058 3 <2.09 23.82 1 3.51 1.04 < MDA 4.5 620058 4 <0.48 3.1510.54 1.04 < MDA 6.1 620058 5 <0.46 1.431033 . 0.92 < MDA 7.7 620058 6 <033 1.1410.28 1.09 < MDA 93 620059 I 14.88 1 1.84 5038 6.41 1.68 <0.70 13 620059 2 <0.48 2.0020.41 1.02 <032 2.9 620059 3 <0.48 0.7810.25 1.08 <0.28 4.5 j I 620059 4 <0.48 1.0310.28 1.77 <032 6.1 \ _ L 620059 5 <033 <0.43 N/A <0.28 7.7 620059 6 (036 <0.4 5 N/A <0.27 93 i . 620059 7 <034 134 031 0.69 <0.27 10.9 ) 620060 1 <0.79 148512.01 1.60 <0.40 13 620060 2 < ! .09 3.0110.80 1.25 <0.5 2 2.9 620060 3 c.3.41 1.1520.29 135 <0.2 5 4.5 4-115 l l
~
Saxton Nucle:r Experimental C:rp::r:tian (SNEC) Site Characterization Report 7 / ) b Sample Core Average Of T/B Average Of T/B Top to Bottom Average oft /B Depth Number Sisce Counting Resmits Counting Results Ratio (Cs-137) Counting Results (cm) (Note 6) Co.60 pCVg Cs-137 pC1/g (Note 1) Eu-152 pCl/g (Note 4) 620060 4 (036 0.8810.26 0.84 <030 6.1 620060 5 <0.40 1.0120.28 0.83 <0.41 7.7 620060 6 <0.29 0.5810.21 1.00 <0.26 93 620060 7 <0.24 0.4710.18 0.82 <0.28 10.9 620061 1 <0.67 438i 1.03 1.74 <039 1.3 620061 2 <037 0.6210.21 0.91 <0.29 2.9 620061 3 <0.3 5 <0.42 N/A <031 4.5 620061 4 <0.27 <0.33 N/A <0.26 6.1 620062 1 23.88 1 3.65 260.65 1 32.84 1.68 <3.64 13 620062 2 1.20 i 0.36 21.73 1 2.86 0.99 7.27 i 1.01 2.9 620062 3 1.8910.48 24.05 1 3.15 0 94 8.2811.14 4.5 620062 4 1.141038 36.97 1 4.76 1.06 9.0211.24 6.1 620062 5 0.9610.34 25.501333 1.00 6.8310.98 7.7 l l 620062 6 1.2110.40 38.15 1 4.90 1.00 7.22 i 1.05 93 l 620062 7 0.971031 35.86 t 4.59 1.01 6.4510.94 10.9 620063 1 31.811632 430.85 1 53.70 1.46 <3.96 13 620063 2 <0.78 1.181035 130 5.8310.85 2.9 4 / h s 620063 3 1.2710.29 <034 N/A 6.4310.83 4.5 l ' <0.44 0.93 6.7610.93 6.1 620063 4 <0.91 620063 5 1.121034 <0.41 0.59 6.8010.93 7.7 l 620064 1 <l.94 334.60 1 41.80 1.20 <2 61 13
<032 223522.92 1.15 <0.41 2.9 620064 2 l 620064 3 <0.41 3.2610.57 1.12 <0 31 4.5 620064 4 <0.40 1.0410.28 0.93 <0.29 6.1 620064 5 <035 0.8910.25 1.I1 <0 29 7.7 620064 6 <038 0.5820.23 0.98 <0 33 93 620064 7 (0.41 <0.65 N/A <030 10.9 620065 i <2.24 74.85 1 9.86 1.43 <132 1.3 620065 2 <032 23210.45 1.19 <031 2.9 l 620065 3 <039 0.9910.26 0.73 < MDA 4.5 620065 4 <038 1.462032 1.24 < MDA 6.1 620065 5 <034 1.04 0.28 0 91 < MDA 7.7 620065 6 <038 0.6010.23 0.89 < MDA 9.3 610066 1 29.83 1 4.05 984.90 121.80 136 < MDA 13 610066 2 <038 213212.80 136 <0.50 2.9 610066 3 (035 2.15 ! 0.43 1.07 <030 4.5 610066 4 <033 1.48 0.34 1.06 <MDA 6.1 610066 5 <034 1.351 033 0.86 < MDA 7.7 610066 6 <032 1.0610.29 1.01 < MDA 9.3 ,/ N t 1 1.83 20.321339 13 620119 203013.05 568.40 1 70.51
( ) 1 1.51 25.55 3.99 2.9 620119 2 3.1010.57 2.7810.58 l 620119 3 3.6510.68 <0.67 N/A 30.841360 45 4-116 l i
Sixt::n Nuclx.r Experimental Carpmti n (SNEC) l Site Characterization Report l~ Q sameple Coee Average ort /B Average Of T/B Top to Bottom Average Of T/B Depth Namnber Slice Coundes Resuks Counting Resuks Ratio (Cs-137) Counting Reseks (Nese 6) Co4e pCl/s On-137 pCl/g (cm) ( (Note 1) Eu-152 pCl/g (Note 4) 620119 4 4.3220.75 0.7220.39 0.78 35.4824.I2 6. I 620119 5 4.2110.74 <0.74 N/A 37.951439 7.7 620119 6 533925.38 <0.83 N/A 9.3 25.89 1 3.02 620119 7 4.5610.79 (0.70 N/A 35.9814.I8 10.9 620119 8 4.4610.76 <0.68 N/A 33.58 2 3.90 12.5 620119 9 4.0210.69 <0.69 N/A 37371 4.32 14.1 620119 10 2.8410.64 <0.63 N/A 293113.43 15.7 I 620119 11 3.I420.60 <0.67 N/A 26.I813.09 173 620119 12 23010.32 <0.55 N/A 23.2I i 2.75 18.9 620!!9 13 2.1210.51 <0.56 N/A 22.55 2 2.69 20.5 l 410120 1 <0.75 20.09 2 3.06 1.45 < MDA 1.3 l 410120 2 <0.83 1.2420.54 0.84 < MDA 2.9 ) l 410120 3 <0.23 0.6020.16 1.31 < MDA 4.5 410120 4 <0.20 03420.13 0.81 < MDA 6.1 410120 5 <0.43 <0.45 N/A <0.46 7.7 410120 6 <0.27 <0.36 N/A <0.27 93 410120 ti <0.28 0.5410.15 1.18 <036 119.5 eg 410I20 13D <0.28 0.4120.13 1.22 1.1710.23 123.9 f r
\ 410120 13B <0.35 0.5010.11 See Note 5 2.4220.37 130.9 410120 13A <038 0.3520.15 0.75 2.7120.41 132.5 410120 13 <0.41 0.3520.15 1.19 2.88 0.43 134.1
} 410120 14 <0.51 0.5610.I8 0.84 3.5410.50 135.7 f 410120 15 0.4910.I8 0.5220.I8 0.89 3.7620.52 137.3 410121 1 <0.66 48.58 1 6.60 1.47 < MDA 1.3 410121 2 <0.78 11.75 2 2.04 1.04 <MDA 2.9 410121 3 <0.2 5 4.3310.64 1.15 < MDA 4.5 410121 4 <0.18 1.1620.24 1.04 < MDA 6.1 410121 5 <0.33 <0.46 N/A <0.29 7.7 410121 6 <031 <0.35 N/A <0.25 9.3 410121 13 <0.25 <0.20 N/A <0.29 103.6 ! 410121 14E <0.2 5 <0.18 N/A <031 111.4 410121 14D <0.27 <0.20 N/A *MDA 113 410121 14B <034 <0.20 N/A 1.2720.25 117.4 410121 15 <0.33 0.1910.14 0.68 2.4020.95 134.3 410122 1 <l.04 14.41 1 2.37 2.22 < MDA 13 410122 2 <0.28 <0.5 7 N/A < MDA 2.9 410122 3 <0.35 <0.59 N/A <032 4.5 410122 4 <0.36 0.9110.27 0.95 <0.31 , 6.1 410122 5 <0.31 0.57 t 0.24 1.53 <031 7.7 '
- 410122 6 <0.35 <0.44 N/A <0.27 9.3 g 410122 15 <0.24 <0.23 N/A <0.31 116.7 410122 16D <0.31 <0.2 5 N/A 1.1010.23 122.1 4 117
t Srxt:n NucleIr Experimental C:rp:rati:n (SNEC) l l Site Characterization Report l (U) Semple Core Average Of T/B Average ort /B Top to Bottom Average Of T/B Depth Number Slice Counting Results Counting Results Ratio (Cs.137) Counting Results (cm) (Note 6) Co.60 pCl/g Cs-137 pCl/g (Note 1) Es.152 pC1/g (Note 4) 410122 16B <033 <0.22 N/A 1.5910.27 126.5 410122 16A <0.32 <0.22 N/A 1.891031 128.1 410122 17 <038 <0.23 N/A 2.471038 130.5 410122 18 <0.39 <0.25 N/A 2.8020.42 132.1 410122 19 2.032038 <030 N/A 3.5120.50 133.7 410122 20 <0.46 <0.24 N/A 3.5510.51 1333 j 410123 1 <0.72 14.81 1 2.38 1.7) < MDA 13 l 410123 2 <0.85 2.8910.80 1.01 < MDA 2.9 410123 3 <0.88 2.7120.76 1.07 < MDA 4.5 410123 4 <0.28 3.2810.58 1.12 < MDA 6.1 410123 5 <035 2.0310.40 0.83 < MDA 7.7 410123 6 <0.25 2.5210.48 1.10 <MDA 9.3 410123 7 <039 23310.46 1.17 < MDA 10.9 I 410123 8 <031 2.4210.46 1.04 < MDA 12.5 ( 410123 9 <0.31 2.2820.44 1.08 < MDA 14.1 l 410123 10 <0.27 1.8310.40 1.00 < MDA 15.7 l 410123 10B <0.18 0.7110.44 0.92 <MDA 25.2 l l N 410123 10D <0.20 1.6110.29 0.99 <0.19 42.3 410123 12 <0.29 <0.48 N/A <0.43 105.2 410123 14 <0.47 <0.49 N/A <0.59 122.4 410123 15 <0.25 <0.22 N/A <0 32 124 l , 410124 1 <0.67 41.91 1 5.77 1.73 < MDA 1.3 410124 2 <0.75 4.1420.97 0.99 < MDA 2.9 ! 410124 3 <0.37 2.2110.44 1.00 < MDA 4.5 l 410124 4 <030 1.741038- 0.95 < MDA 6. I l i 410124 5 <0.36 2.5410.48 0.97 < MDA 7.7 410124 6 <0.33 2.6010.49 0.89 < MDA 93 l 410124 7 <039 5.1910.86 0.97 < MDA 10.9 410124 8 <0.41 1.2010.31 0.91 < MDA 12.5 410124 9 <032 1.2110.31 0.99 < MDA 14.1 410124 10 <0.28 0.8710.23 0.84 < MDA 15.7 410124 10B <0.26 <0.42 N/A <0.25 32.8 I 410124 12 <0.35 <031 N/A <034 107.2 410124 13 8 <030 <0.2 5 N/A <030 122.8 410124 14 <030 <030 N/A 0.7410.19 129.8 410124 15 <0.3 4 <0.27 N/A 0.9710.23 131.4 410125 1 <0.75 12.10 1 2.05 1.85 < MDA 13 410125 2 <0.26 0.6210.20 1.20 < MDA 2.9 410125 3 <032 <0.29 N/A <MDA 4.5
; 410l25 4 <032 <032 N/A <0.27 6.1
/ 410125 5 <0.23 0.2310.10 0.92 <0.26 107.8 410125 6B <0.28 <0.15 N/A <0.47 112.2 4 118 l
i
.- . - .- .. - - - - -- .- . - - -- =. -
- l 1
i Saxt:n Nucizr Experimental C:rp:rati:n (SNEC) Site Characterization Report i 1 hapie Core Aserage Of T/B Average oft /B Top to Bottom Average oft /B Depth I Namnber Stice Counting Results Counting Results Ratio (Cs-137) Counting Resuks (cm) 1 (Note 6) Co40 pC1/g Cs-137 pC1/g (Note 1) Eu-152 pC1/g (Note 4) , 410125 7 <0.29 0.2310.!! 0.67 1.9510.78 119.2 410125 8 <0.30 <0.27 N/A 1.5810.78 120.8 410125 9 <0.33 0.2010.12 0.39 2.2210.80 122.4 l 410125 11 0.5110.19 <0.24 N/A 4.9C 11.27 131.9 410125 13 5.6410.73 <0.30 N/A 4.6310.62 149.2 410125 14 <0.68 <0.39 N/A 6.0110.87 150.8 410125 15 1.4010.32 3.4310.55 1.62 5.9110.78 152.4 i 1 Average pCVg(Note 3) Average pCyg(Note 3) Average pCvg(Note 3) l 4 1.72i1.22 259 72 i 43.64 3.6711.29 ]. NOTE 1 - Ratio of TOP Activity to BOTTOM Activity (Cs-137 ONLY). Surface deposition indicated as ratio approaches 2.5. NOTE 2 - Average MDA values for 1000 sec. counts on " blank" concrete slice is: Co-60 = 0.28 pCi/g., Cs-137 = 0.30 pCi/g., Eu-152 = 0.27 pCi/g. MDA values for CM0 & Eu-152 increase, as higher positive values for Cs-137 are encountered. NOTE 3 - Average activity calculated using actual value from gamma scan or (highest) MDA value for isotopes not positively identified. i i NOTE 4 - Depth listed is maxunum (farthest distance from surface-normally bottom of slice). ! NOTE 5 - Due to very low radioactivity levels in this slice (near MDA), the TOP / BOTTOM counting result ratio for this slice is not representative (3.71). NOTE 6 - Prefix (SX8) and year (95), have been removed from sample numbers to conserve space (SX8-xx xxxx). 6 D i ( 4-119
Saxt n Nuclear Experimental Corporation (SNEC) Site Characterization Report l 5.0 HEALTH PHYSICS SURVEYS SUPPORTING CHARACTERIZATION 5.1 DOCUMENTATION (General) Proper documentation of every aspect of the program included an accurate mapping of the survey data, with its relationship to the site. Ir. addition, instrument measurements and analytical results were documented to include the following items as applicable: , ' 1. Location of surveys or samples taken (sketch, diagram, map etc.) l 2. Date or dates of surveys or samples collected
- 3. The measured concentration of the nuclides in bulk samples as applicable
- 4. Measurements of contamination reported as removable alpha, beta / gamma 2
l activity levels (dpm/100cm , uCi/ unit area, etc.)
- 5. The analytical error reported for error estimates are at the 95% confidence level
- 6. Name of surveyor (s), sampler and analyst, if applicable 1 7. Analysis date
; 8. Instrument Manufacturer, Model No. and reference to calibration data l 9. Reviewer's name t
5.2 HEALTH PHYSICS SURVEY RESULTS Survey data is presented such that: (1) the radiological condition of the site is reasonably represented; (2) the radiological condition of the site can be confirmed ! without further manipulation of the data; and (3) locations of conditions exceeding j release guidelines are easily ascertained. Radiologically contaminated areas are portrayed by using survey maps showing measurement locations together with tables of data taken at those locations. More complete information conceming contamination and radiation levels are given in tables so that intermediate data compilation and final conclusions can be checked. Original data is available and filed so as to be readily accessible. Facility general area (GA) radiological information is shown in Figures 5-1 to 5-5 at the end of this section. i l b 5-1 l
1 Sixt:n Nuclerr Experimental Ccrp:rati:n (SNEC) Site Characterization Report J 5.3 ALARA CONSIDERATIONS i During the SNEC facility Characterization effort, all work evolutions were carried out in accordance with SNEC procedure 6575-PLN-4542.01, "The SNEC Radiation Protection Plan * (Reference 37), and other applicable SNEC Radiological Controls procedures. All work instructions were written and carried out incorporating the concepts of the "As Low As Reasonably Achievable" (ALARA) philosophy. Radiological Controls personnel were consulted and provided valuable input during the preparation of relevant SNEC Work Instructions (SWl's). After preparation, work instructions were reviewed to ensure sound ALARA practices had been incorporated. SNEC facility Radiation Safety Officer concurrence was required for SWI approval. ! 5.3.1 Worker Exposure Control To control worker exposures and to minimize radioactive waste generation, specific ]
\ Radiation Work Permits (RWP's) were written for characterization activities. Worker exposure was controlled using RWP's which were written to provide specific radiological control guidance while working in radiologically controlled areas within the containment vessel. RWP's were required for all entries into contaminated areas, radiation areas, and during the handling of radioactive sample materials. Prior to each day's work activities and before all major work evolutions, the involved work force was briefed. Pre-job briefings were given by the task supervisor and the Group Radiological Controls Supervisor (GRCS). Briefing topics include ALARA, industrial safety, security and self-checking. Performance of this briefing was documented on a pre-job briefing check-off list.
5.3.2 Assessing Risks l A risk assessment of characterization activities that could potentially cause a release of radioactive materials from the CV was performed. The assessment reviewed all scheduled characterization tasks having the potential to generate radioactive i d 5-2 l I l
. - - . . - - - . . - . - . ~ - . - - -.-. .- . - . - . . . - - - . _ - . - _ - - - - - .
8:xt:n Nuclear Experimental Ccrparatisn (SNEC) -I Site Characterization Report materials inside the CV, and which therefore might also have potential to allow a release of radioactive materials from the CV. Three generic activities were identified which met this criteria: Contaminated system pipe cutting operations 1 Breaching of other highly contaminated systems Core bore sampling in contaminated environments The assessment addressed each of these activities and provided recommendations to mitigate the spread of contamination. This assessment was under the guidance of GPU Nuclear Memo number TMI-16575-95-167, dated April 19,1995, " Risk Assessment of Characterization Activities At SNEC Which Could Cause A Release Of Radioactive Materials From Containment". , 5.3.3 Special Considerations For Airborne Radioactive Material Generation Activities A continuous air monitor (CAM), was operated during all characterization evolutions as well as during other minor tasks which might have the potential to generate i l airt>ome contamination. The air monitor was positioned to monitor ambient airbome l radioactivity levels on the observation deck adjacent to the reactor cavity (El. 818' CV, AREA 4). This unit provided an alarm capability to ensure that no unexpected
; airborne radioa-tive materials would go unnoticed in the CV and verified that no release to the outside areas occurred during characterization activities. In addition, HEPA air filtration was used during entries into the Storage Well which is a high contamination and high radiation area as defined by the USNRC. This provided negative airflow into the reactor cavity to help ensure that if airbome radioactive materials were generated, they would be confined to the Storage Well (AREA 6).
The suction side of a HEPA filtration apparatus was also placed at the Steam ] Generator platform during manway removal, and near areas of piping cuts and/or ) highly contaminated system breaching. HEPA filtered vacuum cleaners were also used to help contain localized high contamination work. No airbome radioactive 5-3 l
- . . -- . . ~ - _ _ _ _ _ _ _ _ . . __ _ - . . _ _ _ . _ - . ___
Saxt:n Nuclerr Experimental C rporatinn (SNEC) Site Characterization Report /^\ U contamination level increases were noted on the observation deck (812' El.) CAM during the entire characterization project. Only a few breathing zone air (BZA) samples evaluated after work evolutions retained above background counting ! results, demonstrating that applied radiological work practices and precautions were satisfactory. 5.3.4 Air Sampling Results And Procedures . l l Breathing zone air sampling was routinely performed during characterization activities. Due to two unanticipated high long-lived alpha air sample results from the reactor cavity, a comprehensive airbome contamination evaluation was performed. While the air samples in question were being evaluated, an extensive test program was implemented which required taking a number of sample ~s from locations throughout the CV. These samples were' evaluated for the presence of both short and long-lived gross alpha and beta contamination. As appropriate, the SNEC facility j Gamma Spectroscopy system was used to identify the isotopes present in the samples. In each location specified by the SNEC facility RSO, five air samples were taken. Sample volumes were adjusted to ensure sufficiently low LLD values were obtained during counting operations. Results of this comprehensive air sampling evaluation indicate that there is no long-lived alpha activity present in the CV air during static conditions. Evaluation of two samples in question were performed by GPU Nuclear Radiological Engineering personnel. It was determined that the samples had been cross-contaminated and were not an indication of an actual uptake or ingestion by SNEC facility personnel. Approximately 225 air samples from the SNEC facility have been assayed to date with almost no assigned CEDE values. The following SNEC facility procedures governed air sampling and air sample analysis during characterization work activities: (1) 6575-ADM-4500.13. " Air Sampling" (reference 25) (2) 6575-ADM-4500.16, " Air Sample Analysis" (reference 27) 5-4
I i l Sixt:n Nuclair Experimental Corporation (SNEC) i y Site Characterization Report 5.3.5 Radon in The SNEC CV l Radon daughter products are routinely detected in the CV during air particulate sample analysis. During sample collection this short-lived activity (radon and it's daughters), is collected on the particulate filters along with the long-lived isotopes of interest. During sample counting, the more prodigious short-lived isotopes can mask l l the long-lived isotopes. Air samples must be decayed and re-counted to allow for analysis of long-lived radionuclides, thereby some delay is incurred before obtaining final results. Earlier studies have shown that typical radon daughter product concentrations in the CV average about 6.9 pCi/ml. An evaluation of the hazards associated with exposure to radon at the SNEC facility was completed March of 1994 (Reference 20). 5.3.6 Routine Radiological Surveys Routine and job specific radiological surveys were performed to evaluate radiological l O conditions at the work sites and to support RWP preparation. Surveys included U airbome radioactivity measurements, removable surface contamination surveys and contact General Area (GA) radiation surveys. Performance of routine surveys using standard survey methods is covered in the following SNEC Radiological Controls Procedures. ! + 6575-ADM-4500.11, " Radiation Surveys" (reference 23) l
- 6575-ADM-4500.10 " Surface Contamination Surveys" (reference 22)
The above procedures out':le a methodology for performing routine surveys. Surveillance surveys were performed at specified frequencies. All health physics related surveys were documented as per Section 5.1. These surveys were performed by qualified Radiological Controls technicians and reviewed by a qualified l Group Radiological Controls Supervisor (GRCS). Radiological Controls personnel are qualified according to GPU Nuclear Corporation, Radiological Controls , Procedures. i ( ( 5-5 i
)
Saxt:n Nuclerr Experimental Ccrp:rati:n (SNEC) Site Characterization Report 1 O Radiological surveys were documented in accordance with SNEC procedure 6575-ADM-4500.12, " Radiological Surveys: Requirements and Documentation" l (Reference 24). As specified in this procedure, each survey was given a unique survey number for tracking and identification purposes. Final characterization and ! routine survey data was obtained from 117 documented radiological survey maps of l the CV and surrounding facilities. System and component curie calculations specified by the SCP, were performed using data from these surveys. ! 5.4 CHARACTERIZATION SURVEY METHODS in order to provide radiation and contamination survey results specified in the SCP, l standard and non-standard survey methods were employed. Survey methodology and other guidance needed for proper completion of required surveys were specified in SNEC procedures and the following SNEC Work Instructions. gs (1) SWl-94-001, " Remove Core Bore Samples From SNEC CV structures" !( (Reference 56) ! (2) SWl-94-005, "Special Surface Contamination Measurements" (Reference 60) (3) SWl-94-007, " Unbiased surveys" (Reference 62) (4) SWl-94-009, " Estimating Loose Surface contamination" (Reference 63) (5) SWi-94-010, " Probe The Reactor Vessel intemals To Obtain Samples, Place TLD's, & Verify Path For Video Inspection" (Reference 64) . l l
- V 5-6
I l Srxtan NuclIrr Experimental Corporation (SNEC) Site Characterization Report 5.5 INSTRUMENTATION, USE AND PERFORMANCE Typical portable instrumentation used at the SNEC facility include those shown in the following table: l TABLE 5-1, Typical Health Physics instrumentation l INSTRUMENTATION RADIATION TYPE RANGE / SCALE APPLICATION l Eberline Instrument Beta / Gamma 0 to 5 R/h & Exposure Rate Surveys Corporation 0-50 R/h (EIC) RO-2 & RO-2A ASP-1 Multi-Purpose, Various Portable Survey Meter Dependent on Probe Used j ElC E-520 Gamma 0 to 2 R/h Exposure Rate Surveys l ElC E-530N Gamma 0 to 20 R/h Exposure Rate Surveys Ludlum Model 19 Gamma 0 to 5 mR/h Low-Level (uR/h) i (o to 5000 uR/h) Exposure Rate Surveys i ElC140N Beta / Gamma 0 to 50.000 CPM Personnel & Area w/HP-210T Probe Contamination Monitoring ElC RM-14 Beta / Gamma 0 to 50,000 CPM Personnel & Area w/HP-210T Probe Contamination Monitoring i Ludlum 2000 with Beta / Gamma Counts Per Unit Count Room ' \ HP-210 Probe Time Instrumentation , (scater) Air Samplers Particulate N/A Airborne Sampling l (All Types) l ElC SAC 4 Alpha Counts Per Unit Count Room Time Instrumentation (Scaler) Teletector Gamma 0 to 1000 R/h - Exposure Rate Surveys The above instrumentation listing includes most, but not all of the instrumentation used at the SNEC facility. Response of similar instrumentation to that shown above, ! is presented in Table 2, Appendix C and Table 1 Appendix D of NUREG/CR 5849 (Reference 13). All radiation monitoring equipment used in monitoring personnel exposures were calibrated and maintained by qualified GPU Nuclear personnel in accordance with GPU Nuclear policies and procedures.
- 5.6 HOT SPOT IDENTIFICATION PROGRAM
. A " Hot Spot" identification program was implemented for dose reduction purposes. Since exposure rates inside the CV, but outside the Storage Well are generally low, a 5-7
Saxt:n Nuclerr Experimental Ccrper:ti:n (SNEC) 1 Site Characterization Report i O a hot spot identification program was implemented as a way to keep exposures l j ALARA. Hot spot surveys were performed to augment other radiation surveys I required by the SNEC SCP. Hot spots outside the Storage Well were logged into a
" Hot Spot" log, identified and clearly labeled within the CV using hotspot stickers.
This log tracks components or areas outside the reactor cavity which present the t most significant radiation exposure hazards at the facility. Hot spot identification l criteria was liberally applied to any component outside the Storage Well found to l equal or exceed 50 mR/h on contact. Nothing within the Storage Well (a posted High ! Radiation Area), has been specifically labeled as part of the hot spot identification l program. ALARA precautions associated with entry into this area are covered by the l appropriate RWP and ALARA Review. Thirteen hot spots were identified within the CV, outside the Reactor cavity. Most of the hot spots identified are located within the primary compartment (AREA 2). The following table lists the hot spots identified in the SNEC hot spot log. Table 5-2, SNEC Hot Spot identification Log l NO. HOT SPOT LOCATION LOG mR/h ! ! Shutdown Cooling Heat Exchanger Flange,775' El. 100 l 2 Bottom Of Drain Tank,765' 8 El. 380 3 Letdown Recire. Vahe V-318,781' El. 100 4 Regenerative Heat Exchanger Bottom, 781' El. 2800 5 Non-Regenerative Heat Exchanger Flange,781' El. 280 l 6 Main Coolant Cold Leg @ RCP Inlet,795' El. 170 I l 7 RCP Discharge,795* El. 150 8 Main Coolant Out Of Steam Generator,791' El. 80 9 Bottom Of Hot Leg Piping @ S/G Inlet,791' El. 80 10 Bottom Of S/G @ Hot Leg Manway,793' El. 800 l 11 S/G Cold Leg Manway Flange,791' El. 100 12 S/G Cold Leg Manway Cover,789' El. 100 13 Pressurizer Relief Discharge Line,779' El. 50 5.7 PERSONNEL RADIATION EXPOSURES The total radiation exposure estimate for site characterization at the SNEC facility was projected to be about 2 person-rem. The final measured (TLD) dose to SNEC and contractor personnel is reported to be 1.96 Rem. Of this dose, ~73% was l h 5-8
( Srxt:n Nuclerr Experimental Ccrptration (SNEC) l Site Characterization Report received by Radiological Controls personnel who obtained surveys and samples l required by the characterization process, and who performed job coverage during l concrete core bore sample extraction work. Approximately 17% of the total exposure l l was received by contractor workers who performed concrete core boring operations. The remaining 10% was received by GPU Nuclear Project Management and contractor personnel who performed supervisory and specialty tasks such as rigging, asbestos removal and crane repairs. I i l l I ( i t l l . I I i l i 4
)
5-9
Caxt:n Nuclear Experimental Ctrp*r-tirn (SNEC) Site Characterization Report ("%
\
Average Contamination Levels and General Area Exposure Rates AREA 7 Outside _ CV ~
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Srxt n Nuclear Experimental C:rp: ration (SNEC) Site Characterization Report (V ) SNEC FACILITY AREA LOCATIONS Average Contamination Levels and General Area Exposure Rates " ggg8sWNMo4{il p,,, g$ QN Ek fljhh!jj
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l l 5:xt n Nucirr Experimental C:rp:r:ti:n (SNEC) Site Characterization Report lO SNEC FACILITY AREA LOCATIONS l Average Contamination Levels and General Area Exposure Rates
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- y_OSCALG l
! 5 5-14 l l
Sixten Nuclarr Experimintil Ccrporation (SNEC) Site Characterization Report o 6.0 QA/QC REQUIREMENTS 6.1 DATA COLLECTION The SNEC facility Quality Assurance Program (Reference 39), appropriate GPU l Nuclear and SNEC procedures, and SNEC Work instructions, were used to direct , [ the performance of all SNEC Site Characterization activities. In addition, laboratory Quality Control (OC) was maintained by using appropriate procedures or Quality Assurance (QA) approved outside laboratories. Data was examined in accordance with existing requirements to ensure a reasonable interpretation of results. Samples and measurements at all locations were collected using accepted procedures or SNEC Work Instructions to ensure well documented and consistent collection techniques. l On-site sampling and measuremeat instruments were subject to operational checks and periodic calibrations, in accordance with Saxton Nuclear Experimental
/ Corporation facility procedures. This was done to ensure both accuracy and i \
precision. Records and calculations were checked for error and proper use of appropriate recording and calculational techniques. Nuclear Safety Assurance (NSA) provided monitoring in support of site characterization. Laboratories were required to verify selected data analysis results. Chain-of-custody procedures were implemented to maintain the integrity of samples and corresponding analytical results. This program ensured that characterization data collected may be used to accurately evaluate site conditions. 6.2 MEASUREMENT QUALITY CONTROL To ensure measurement quality, the following controls were implemented: Five percent of soil samples were re-analyzed or re-taken and then re-analyzed. Five percent of swipes (system & structure) were re-analyzed or re-taken and then re-analyzed. 6-1 1
Saxt3n Nucitar Experim:ntal Ctrp:rati n (SNEC) Site Characterization Report ?^ f
\
- Five percent of direct measurements of total surface contamination were rotaken.
l Results from repeat measurements and sampling activities were reviewed and compared with previous results from other similiar activities. 6.3 SAMPLE RETENTION l Material samples will be retained by the SNEC facility radiological controls staff in accordance with the following schedule. i Sample Tvoe ' Retention Period Core Bore Samples Completion of Decontamination Activities
- l Smears Acceptance of Decommissioning Pian
- Bulk Samples Completion of Decontamination Activities
- Soll Samples Completion of Decontamination /Remediation*
Not all samples shall be retained. Representative samples for typical l locations throughout the facility (as determined by SNEC) will be retained. Samples that were analyzed prior to this retention requirement may not be j l available. l Completion of Decontamination Activities is defined as satisfactory l completion of NRC Confirmatory Survey Report which addresses SNEC facility structures. Completion of Decontamination /Remediation is defined as completion of NRC Confirmatory Survey Report which addresses SNEC facility soil. d-Acceptance of Decommissioning Plan is defined as USNRC authorization to commence Decontamination activities. !O 6-2 I
.-r
Saxt:n Nuclxr ExperimIntil Ccrp: ration (SNEC)
,. Site Characterization Report U
7.0 REFERENCES
- 1. Saxton Site Characterization Plan, Procedure No. 6575-PLN-4520.06.
- 2. Saxton Nuclear Experimental Corporation Facility Decommissioning Plan, February 1996.
- 3. Decommissionina Handbook. U.S. Department of Energy, DOE /EM-0142P, March 1994, Washington D.C.
- 4. "Saxton Nuclear Experimental Facility Reactor Vessel, Intemals, Ex-Vessel Lead, Structural Steel and Reactor Compartment Concrete Shield Wall Radionuclide Inventory" (Proprietary), Document No.
G01-1192-003, Rev. O, December 1995, prepared by TLG Services, Inc.
- 5. " Decommissioned Status of the Saxton Reactor Facility" forwarded to the United States Nuclear Regulatory Commission (U.S. Nuclear Regulatory Commission) on February 20,1975.
S 6. " Final Release Survey Report of the Reactor Support Buildings," GPU
) Nuclear Report, Rev. 3, dated March 1992.
s
- 7. "SNEC Reactor Support Buildings Demolition Report", GPU Nuclear Report, dated March 1994.
- 8. RESRAD, "A Computer Code for Evaluating Radioactively Contaminated Sites," Argonne National Laboratory.
- 9. "1994 Saxton Soil Remediation Project Report, GPU Nuclear," dated May 11,1995.
l 10. NUREG-1500, " Working Draft Regulatory Guide on Release Criteria f for Decommissioning: NRC Staff's Draft for Comment," U.S. Nuclear Regulatory Commission, August 1994. I
- 11. U.S. Nuclear Regulatory Commission IE Circular No. 81-07, " Control of Radioactively Contaminated Material," May 14,1981.
- 12. U.S. Nuclear Regulatory Commission IE Information Notice No. 85-92,
" Surveys of Wastes Before Disposal from Nuclear Reactor Facilities," l December 2,1985.
U l ! 7-1 ; i
. Saxt:n Nucle *r Experim:ntal Ctrpcrati:n (SNEC) l Site Characterization Report i 1
- 13. NUREG/CR-5849, " Manual for Conducting Radiological Surveys in ;
Support of License Termination," draft U.S. Nuclear Regulatory Commission, June 1992.
- 14. NUREG-1507, " Minimum Detectable Concentrations with Typical l Radiation Survey Instruments for Various Contaminants and Field Conditions," draft U.S. Nuclear Regulatory Commission, August 1995.
- 15. CCC-371, "ORIGEN2 - Isotope Generation and Oepletion Code - 4 Matrix Exponention Method," Oak Ridge National Laboratory Radiation Shielding Information Center, May 1991.
- 16. CCC-254, "ANISN-ORNL - One Dimensional Discrete Ordinates Transport Code System with Anisotropic Scattering," Oak Ridge l National Laboratory Radiation Shielding information Center, April 1991.
- 17. U.S. Environmental Protection Agency, Code of Federal Regulations,
" Protection of Environment", Title 40 Part 261,1994.
- 18. EPRI Report EA-5321, " Chemical Characterization of Fossil Fuel
[ Combustion Wastes",1987. I l ( l 19. EPRI Report CS-5355, " Evaluation of the Toxicity Characteristic ' Leaching Procedure (TCLP) on Utility Wastes",1987.
- 20. GPUNC Memorandum No. SNEC-94-0005, From G.E. Tomlinson, to R.G. Rolph, March 1,1994, " Radon Report".
SNEC Procedures
- 21. SNEC Procedure 6575-ADM-4500.04, " Records Retention" r ov.O.
- 22. SNEC Procedure 6575-ADM-4500.10, " Surface Contamination Surveys" Rev. O.
I
- 23. SNEC Procedure 6575-ADM-4500.11, " Radiation Surveys" Rev.1.
- 24. SNEC Procedure 6575-ADM-4500.12, " Radiological Surveys:
i Requirements & Documentation. [
- 25. SNEC Procedure 6575-ADM-4500.13, " Air Sampling" Rev. O.
(% 26. SNEC Procedure 6575-ADM-4500.15," Dosimetry Issue and Handling" Rev. O. 7-2 l
8::xton Nucl:ar Experim:ntal Ccrptrati:n (SNEC) Site Characterization Report
- 27. SNEC Procedure 6575-ADM-4500.16, " Air Sample Analysis" Rev.1
- 28. SNEC Procedure 6575-ADM-4500.19, " Release Surveys" Rev. 0
- 29. SNEC Procedure 6575-ADM-4500.20, " Radiation Work Permits" Rev.1.
- 30. SNEC Procedure 6575-ADM-4500.21, " Control and issue of Respiratory Protection Equipment" Rev. 0
- 31. SNEC Procedure 6575-ADM-4500.22, " Environmental Monitoring" Rev. 3.
- 32. SNEC Procedure 6575-ADM-4500.29," Radioactive Material identification and Handling" Rev. O.
- 33. SNEC Procedure 6575-ADM-4500.39, " Chain of Custody of Samp;es" Rev. O.
- 34. SNEC Procedure 6575-ADM-4500.41, " Work Instructions" Rev.1.
O Q 35. SNEC Procedure 6575-ADM-4500.44, "Saxton Technical Evaluations Procedure" Rev. O. 36 SNEC Procedure 6575-OPS-4524.33," Acquisition & Data Analysis of High Resolution Gamma Spectra Using the Inspector Portable Gamma Spectroscopy System" Rev. O.
- 37. SNEC Procedure 6575-PLN-4542.01, "SNEC Radiation Protection Plan" Rev. O.
- 38. SNEC Procedure 6575-PLN-4742.05, "SNEC Soil Characterization Plan" Rev. O.
- 39. SNEC Procedure 6575-QAP-4220.01," Quality Assurance Program for RadiologicalInstruments" Rev. 0
- 40. SNEC Procedure 6575-QAP-7200.01,"SNEC Quality Assurance Program" Rev 0.
- 41. SNEC Procedure 9400-ADM-4500.25, " Final Release Survey of Structures at the Saxton Site" Rev.1.
/
- 42. SNEC Procedure 9400-OPS-4524.20," Operation of the EIC
" RASCAL" PRS-1/2 For Alpha Detection" Rev.1.
7-3
S:xt:n Nucliar Experimsntal Ccrpsr; tion (SNEC) Site Characterization Report
- 43. SNEC Procedure 9400-OPS-4524.21," Operation of the EIC
" RASCAL" PRS 1/2 For Beta-Gamma Detection" Rev.1.
- 44. SNEC Procedure 9400-OPS-4524.25, " Operation of the ASP-1 Survey instrument" Rev. O.
l
- 45. SNEC Procedure 9400-OPS-4524.26," Operation of the EIC Count Rate Meter Model PRM-6" Rev. O.
- 46. SNEC Procedure 9400-OPS-4524.27," Operation of the Eberline Smart Portable Survey Meter, Model ESP-1" Rev. O.
- 47. SNEC Procedure 9400-OPS-4524.28," Operation of the Eberline Smart Portable Survey Meter, Model ESP-2" Rev. O.
- 48. SNEC Procedure 9400-OPS-4524.31," Operation, Handling and Servicing of HEPA-Filtered Controlled Vacuum Cleaners and Blowers" Rev.1.
- 49. SNEC Procedure 9400-PLN-4542.03, "A Survey Plan for SNEC to
" Show Compliance With Decommissioning Termination Survey Criteria" i Rev.1.
- 50. SNEC Procedure 6575-PLN-4542.07, "SNEC Soil Disposal Plan" Rev0.
Standing Order Memorandums
- 51. Saxton Nuclear Experimental Corporation (SNEC) Standing Order Memorandum (SOM) 6675-91-02, " Determining Average Residual Activity in Soil Around the Support Structures" Rev.1.
- 52. SNEC SOM 6675-91-03,"Use of TMl Radioactive Material Receipt, Disposal and Shipping Procedures at the Saxton Nuclear Experimental Corporation Site" Rev.1.
- 53. SNEC SOM 6675-93-01, " Implementation of 6675-PLN-4742.05 SNEC Soil Characterization Plan" Rev.1.
- 54. SNEC SOM 6575-94-01, "SNEC Survey Recording" Rev. O.
- 55. SNEC SOM 6575-95-03," Control of High Radiation Areas at Saxton" Rev. O.
7-4
i SIxt:n Nuclear Experimintal Ccrp:r: tion (SNEC) Site Characterization Report h o SNEC Work Instructions t
- 56. SNEC Work Instruction (SWI) 94-001, " Remove Core Bore Sample l from Saxton Containment Vessel Bldg. Structures" Rev. 2. ;
- 57. SNEC Work Instruction (SWI) 94-002, " Bulk Sample Collection from SNEC Site Facilities in Preparation for Offsite Analysis."
- 58. SNEC Work Instrudion (SWI) 94-003, " Systems Sampling at SNEC Facilities."
- 59. SNEC Work InsYuction (SWI) 94-004, " Measurement of Radiation Exposure Rates kom Saxton Reactor Vessel."
l
- 60. SNEC Work Instruction (6vVi) 94-006, -Special Surface Contamination l Measurements."
l
- 61. SNEC Work Instruction (SWI) 94-006, " Core 5 ample Analysis."
- 62. SNEC Work Instruction (SWI) 94-007, " Unbiased Surveys."
p)
\
- 63. SNEC Work Instruction (SWI) 95-009, "Et,timating Loose Surface Contamination."
- 64. SNEC Work Instruction (SWI) 95-010 " Probe the Reactor Vesrel Intemals to Obtain Samples, Place TLDs, and Verify a Clear Plath for a Video inspection."
l 65. SNEC Work Instruction (SWI) 96-012, " Video inspection of Reactor l Vessellnternals Via the N-3 Upper Head Penetration." SNEC Technical Evaluations ! 66. SNEC Technical Evaluation Calculation 5340-95-010, " Pressurizer," 10/19/95.
- 67. SNEC Technical Evaluation Calculation 5340-95-011, "RV Intemal Wetted Surface Area," 10/20/95.
- 68. SNEC Technical Evaluation Calculation 5830-95-012, " Spent Fuel Rack," 11/07/95.
- 69. SNEC Technical Evaluation Calcul6 tion 5340-95-013, " Pressurizer Discharge Tank," 11/17/95.
I 7-5 j l l
- -. . . - - - - - - . - . . - . - . - . . - -= .- __ .-- ._.
S:xt:n Nucle:r ExperimIntr.1 Carp:r; tion (SNEC) Site Characterization Report
\
- 70. SNEC Technical Evaluation Calculation 5340-95-014, " Boric Acid Demin," 11/17/95.
- 71. SNEC Technical Evaluation Calculation 5340-95-015, "SW Demin,"
11/17/95.
- 72. SNEC Technical Evaluation Calculation 5340-95-016, " Purification Demin," 11/17/95.
- 73. SNEC Technical Evaluation Calculation 5340-95-017, " Steam Generator," 11/17/95.
- 74. SNEC Technical Evaluation Calculation 5340-95-018, "Re-Gen /NonRe-Gen Heat Exchanger," 11/20/95.
- 75. SNEC Technical Evaluation Calculation 5340-95-019, " Shutdown Cooling Heat Exchanger," 11/20/95.
- 76. SNEC Technical Evaluation Calculation 5340-95-020, " Storage Well Heat Exchanger," 11/20/95.
O 77. SNEC Technical Evaluation Calculation 5340-95-021, " Main Coolant id Pump," 11/20/95.
- 78. SNEC Technical Evaluation Calculation 5340-95-022, "Purif & CW Filter," 11/20/95.
i
- 79. SNEC Technical Evaluation Calculation 5830-95-023, " Curie Estimate.
Saxton Piping System," 11/20/95.
- 80. SNEC Technical Evaluation Calculation 5830-95-024, " Storage Well Filters," 11/20/95.
- 81. SNEC Technical Evaluation Calculation 6575-95-032, " Concrete Calculation," 12/06/95.
- 82. SNEC Technical Evaluation Calculation 5340-95-033, " Auxiliary Air Handler," 12/22/95.
- 83. SNEC Technical Evaluation Calculation 5340-96-001, " Rod Room Air Handler," 01/02/96.
- 84. SNEC Technical Evaluation Calculation 5340-96-003, " Discharge Tank Pumps," 01/02/96.
7-6
Sext::n Nuclear Experimentil Ccrpsr; tion (SNEC) , Site Characterization Report () i l ! V 85. SNEC Technical Evaluation Calculation 5340-96-004, " Primary I i Coolant Instrument Rack," 01/02/96. l l
- 86. SNEC Technical Evaluation Calculation 5340-96-005, " Instrument
- Racks," 01/02/96. i I
l 87. SNEC Technical Evaluation Calculation 5340-96-006, "Superheated j Test Loop Tank," 01/02/96. l f
- 88. SNEC Technical Evaluation Calculation 5340-96-007, " Component l
! Cooling Surge Tank," 01/02/96. ]
- 89. SNEC Technical Evaluation Calculation 5340-96-008, " Shutdown Cooling Pumps," 01/02/96.
- 90. SNEC Technical Evaluation Calculation 5340-96-009, " Containment Vessel Sump," 01/02/96.
- 91. SNEC Technical Evaluation Calculation 5340-96-010, " Pump Drain Tank," 01/02/96.
l A 92. SNEC Technical Evaluation Calculation 5340-96-011, "Containmnent 1 () Vessel Sump Pumps," 01/02/96. )
- 93. SNEC Technical Evaluation Calculation 5340-96-012, " Storage Well
! Pumps," 01/02/96. l l 94. SNEC Technical Evaluation Calculation 5340-96-013, " Component Cooling Pumps," 01/02/96. ( 95. SNEC Technical Evaluation Calculation 5340-96-014, " Component Cooling Heat Exchanger" 01/02/96.
- 96. SNEC Technical Evaluation Calculation 5340-96-015, " Condensate l Tank," 01/02/96.
l 97. SNEC Technical Evaluation Calculation 5340-96-016, " Contaminated Pumps," 01/02/96.
- 98. SNEC Technical Evaluation Calculation 5310-96-017, "CV Primary Equipment Air Handlers," 04/08/96.
- 99. SNEC Technical Evaluation Calculation 5310-96-018, "CV Ops Floor Air Handler," 04/08/96.
f~ V) . 7-7 ' l
. _ - .- - -. _.. = - - . - . . - . -
Srxt:n NucleIr Experirnantal Carp:riti:n (SNEC) Site Characten?.ation Report 100. SNEC Technical Evaluation Calculation 6575-95-001, "Saxton Lead Inventory",09/11/95. SNEC Radiological Surveys l l l- 101. SNEC Radiological Survey 849-95-046, " Moveable Refuel Bridge," 04/26/95. l 102. SNEC Radiological Survey 845-95-047, "812'-818' Concrete Walls," 04/26/95. 103. SNEC Radiological Survey 846-95-052, "SNEC Containment Vessel 818' el. Steel Platform," 04/27/95. 104. SNEC Radiological Survey 843-95-053, "SNEC Containment Vessel l 818' el. Concrete Floor," 04/27/95. 105. SNEC Radiological Survey 840-95-060, "SNEC Containment Vessel Escape Access Hatch," 05/02/95. ( 106. SNEC Radiological Survey 210-95-087, "781' Containment Vessel
\
Regen Heat Exchanger," 05/09/95. 1 107. SNEC Radiological Survey 842-95-088, "843' Dome," 05/09/95. l l 108. SNEC Radiological Survey 874-95-089, "814' Concrete Shield l Horizontal Wall," 05/10/95. l l SNEC Radiological Survey 871-95-090,"811*-817' Outside
- 109.
l Containment Vessel Wall," 05/10/95. 110. SNEC Radiological Survey 875-95-091,"804'-814'(811-814) Actual ! Outside Shield Wall Concrete," 05/10/95. 111. SNEC Radiological Survey 831-95-094, " Area 3 - 810' el. Ceiling," 05/12/95. l 112. SNEC Radiological Survey 826-95-095, " Area 2 - 807' el. Steel Platform Steam Generator Pressurizer," 05/12/95. 113. SNEC Radiological Survey 822-95-096, " Area 2 - 814' el. Ceiling Steam Generator Pressurizer Area," 05/12/95. ! 114. SNEC Radiological Survey 814-95-097, " Area 1 - 781' el. Ceiling," 05/12/95. 7-8 i
1
. Saxtrn Nuclerr Experimsntal Cerporatirn (SNEC)
! Site Characterization Repon F O l0 115. SNEC Radiological Survey 814-95-098, " Area 1 - 775' el. Ceiling of Basement," 05/12/95. 116. SNEC Radiological Survey 813-95-099, " Area 1 - 765' el. Sump," 05/12/95. 117.- SNEC Radiological Survey 830-95-0105, " Containment Vessel 795' Piping Hot Spot," 05/11/95. 118. SNEC Radiological Survey 830-95-0106, " Containment Vessel 781' Piping Hot Spot," 05/11/95. 119. SNEC Radiological Survey 812-95-107, " Containment Vessel 781' Top of Filter Cubicle Piping Hot Spot," 05/11/95. 120. SNEC Radiological Survey 811-95-0111, "765' Containment Vessel Piping Hot Spot," 05/11/95. L 121. SNEC Radiological Survey 811-95-0112, "765' Containment Vessel l Piping Hot Spot," 05/11/95. O t 122. SNEC Radiological Survey 834-95-0116, "795' el. Component Cooling Containment Vessel," 05/16/95. 123. SNEC Radiological Survey 812-95-0117, "765' Containment Vessel Basement Piping," 05/17/95. 124. SNEC Radiological Survey 812-95-0119, "765' Containment Vessel l Basement Discharge Line," 05/17/95. 125. SNEC Radiological Survey 818-95-0120, " Containment Vessel 765' , SW194-002; 2,3 Sample Location," 05/17/95. 126. SNEC Radiological Survey 216-95-0152, "765' el. Containment Vessel," 05/30/95. 127. SNEC Radiological Survey 871-95-0153, "1M2 Containment Vessel Wall," 05/31/95. 128. SNEC Radiological Survey 232-95-0157, " Containment Vessel 781' Fuel Pool Through 4' Core Bore," 06/01/95. 129. SNEC Radiological Survey 844-95-0164, "818' el. - 812' el. Concrete Walls / Floor," 06/02/95. 1 EJ 7-9 l
l l Saxt n Nucle r ExperimIntal Ccrptrati:In (SNEC) Site Characterization Report i V ! 130. SNEC Radiological Survey 841-95-0165, " Polar Crane Ladder," 06/02/95. 131. SNEC Radiological Survey 846-95-0166, "840' el. Steel Platform," l 06/02/95. 132. SNEC Radiological Survey 898-95-0168, " Containment Vessel 781' Air Handler Unit & Filter," 06/05/95. l 133. SNEC Radiological Survey 898-95-0169, " Containment Vessel 781' Primary Air Handler Unit & Filter," 06/05/95. 134. SNEC Radiological Survey 225-95-0188, "781'-795' Steam Generator Pipe Cut item #3," 06/06/95. l 135. SNEC Radiological Survey 223-95-0189, "781' Steam Generator Pipe Cut SX-8-25-95-0040, item #6," 06/06/95. 136. SNEC Radiological Survey 225-95-0191, "795' Steam Generator Pipe ! Cut SX-8-25-0043, item #2," 06/07/95. l ' 137. SNEC Radiological Survey 227-95-0192, "781' Steam Generator Pipe y) Cut SX-8-25-0042, item #6," 06/07/95. 138. SNEC Radiological Survey 225-95-0193, "795' Steam Generator Pipe
- Cut SX-8-25-95-0044, item 4," 06/07/95.
l 139. SNEC Radiological Survey 225-95-0196, "765' Pipe Cut #9," 06/08/95. l 140. SNEC Radiological Survey 225-95-0197, "765' Pipe Cut #13," 06/08/95. 141. SNEC Radiological Survey 898-95-0204, " Miscellaneous System Piping Smears," 06/13/95. 142. SNEC Radiological Survey 89B-95-0208, " Dose Rates on Miscellaneous Bulk Sampling," 06/13/95. i 143. SNEC Radiological Survey 89B-95-0210, " Dose Rate 780' Primary Air i Handler Sample," 06/14/95. 144. SNEC Radiological Survey 867-95-0228, "779' Containment Vessel i Storage Well, Reactor Exterior," 06/20/95. i A
~
145. SNEC Radiological Survey 861-95-0229, " Containment Vessel Spent i Fuel Pool 765' Floor / Walls /Demin," 06/20/95. 7-10
l Saxt:n Nucle r Experimentil Ccrpsraticn (SNEC) i Site Characterization Report
/~N l L 146. SNEC Radiological Survey 821-95-0235, " Area 2, Containment Vessel i l 779'," 06/16/95.
l l 147. SNEC Radiological Survey 824-95-0236, " Area 2, Containment Vessel l 795',"06/16/95. l 148. SNEC Radiological Survey 893-95-0242, " Pipe Cut SX-8-14-95-0054," 06/22/95. i 149. SNEC Radiological Survey 810-95-0262, " Discharge Tank," 06/27/95. 150. SNEC Radiological Survey 89A-95-0273, " Steam Generator Hand l Hole Secondary Side," 07/05/95. ] 151. SNEC Radiological Survey 202-95-0292, "818' el. Floor Plug #4," 07/11/95. i l 152. SNEC Radiological Survey 204-95-0293, "812' el. Containment Vessel l Floor Plugs #1,2,5,6,7," 07/11/95. l 153. SNEC Radiological Survey 205-95-0294, "795' el. Containment Vessel Primary Side Core Bore," 07/11/95. l 154. SNEC Radiological Survey 210-95-0295, "781' el. Primary Side," 07/11/95. l 155. SNEC Radiological Survey 216-95-0296, "765' el. Core Bore Location j #0030," 07/11/95. 156. SNEC Radiologic 61 Survey 873-95-0303, " Containment Vessel Dome Outside 845' to 870'," 07/11/95. 157. SNEC Radiological Survey 872-95-0304, " Containment Vessel Outside 817' to 845'," 07/11/95. l 158. SNEC Radiological Survey 872-95-0305, " Containment Vessel i Outside 817' to 845'," 07/11/95. l 159. SNEC Radiological Survey 898-95-0306, " Lead Inventory j Containment Vessel 812' el," 07/12/95. I ! 160. SNEC Radiological Survey 898-95-0307, "Pb inventory 765'-8" Top of Cubicle," 07/12/95. ] 7 11 l l
Saxtin Nucle r Experimental Ccrp rati:n (SNEC) Site Characterization Report ! 161. SNEC Radiological Survey 898-95-0308, "Pb Inventory Containment Vessel 795'," 07/12/95. l 162. SNEC Radiological Survey 89B-95-0309, "Pb inventory Containment Vessel 781'," 07/12/95. ; ! 163. SNEC Radiological Survey 821-95-0320, " Containment Vessel 781' l Primary (Area 2)," 07/18/95. 164. SNEC Radiological Survey 822-95-0321, " Containment Vessel 795' Primary (Area 2)," 07/18/95. 165. SNEC Radiological Survey 811-95-0322, " Containment Vessel 765' l (Aree 1)," 07/18/95. 166. SNEC Radiological Survey 834-95-0325, " Containment Vessel 795' Fixed Point Characterization," 07/19/95. 167. SNEC Radiological Survey 833-95-0326, " Containment Vessel 781' Fixed Point Characterization," 07/19/95. [ 168. SNEC Radiological Survey 817-95-0327, "Contcinment Vessel 765' l l Q] Fixed Point Characterization," 07/19/95. ! 169. SNEC Radiological Survey 821-95-0328 " Containment Vessel 781' Primary Fixed Point Characterization," 07/19/95. 170. SNEC Radiological Survey 824-95-0329, " Containment Vessel 795' Primary Fixed Point Characterization," 07/19/95. 171. SNEC Radiological Survey 843-95-0330, " Containment Vessel 812' Fixed Point Characterization, "07/19/95. 172. SNEC Radiological Survey 891-95-0343, " Steam Generator Manway Cover Removal," 07/26/95. l 173. SNEC Radiological Survey 891-95-0344, " Steam Generator Manway l Cover Removal," 07/26/95. 174. SNEC Radiological Survey 89A-95-0355, " Containment Vessel 795' Survey of S/G Blowdown and River Water Pipe Cuts," 07/28/95. 175. SNEC Radiological Survey 898-95-0356, " Containment Vessel Air l HandlerNent/ Mixing Fan," 07/28/95. L 7 12 c
Saxtrn Nucle:r Experiment:1 Ccrp:r;ti:n (SNEC) Site Characterization Report O 176. SNEC Radiological Survey 898-95-0365, "795' Containment Vessel Vent Housing," 07/31/95. 177. SNEC Radiological Survey 895-95-0366, "765' S/D Cooling," 08/01/95. l ! 178. SNEC Radiological Survey 895-95-0371, "765' S/D Cooling," 08/01/95. 179. SNEC Radiological Survey 895-95-0372, "779' S/D Cooling," 08/01/95. 180. SNEC Radiciogical Survey 893-95-0373, "Demin Sluice Line Pipe Cut," 08/02/95. 181. SNEC Radiological Survey 899-95-0376, " Containment Vessel 765' Vent & Drain Header Discharge Line," 08/02/95. l 182. SNEC Radiological Survey 899-95-0377, " Containment Vessel 765' Piping Between Sump Pump & CV Wall," 08/02/95. p 183. SNEC Radiological Survey 899-95-0378, " Containment Vessel 781'- Q 795' Primary Comp. Piping Between Sump Pumps & CV Wall," 08/02/95. 184. SNEC Radiological Survey 899-95-0379, " Containment Vessel 795' Vent & Drain Header Discharge Line," 08/02/95. 185. SNEC Radiological Survey 899 95-0380, " Containment Vessel 781' Primary Comp. Vent & Drain Header Discharge Line," 08/02/95. 186. SNEC Radiological Survey 892-95-0381, "795' Pretsurizer," 08/03/95. 187. SNEC Radiological Survey 892-95-0382, "795' Pressurizer Spray Line," 08/03/95. 188. SNEC Radiological Survey 892-95-0383, "795' Pressurizer Relief Valve Discharge Line," 08/03/95. 189. SNEC Radiological Survey 892-95-0384, "779'-795' Pressurizer Relief Valve Discharge Linel' 08/03/95. 190. SNEC Radiological Survey 892-95-0385, "765' Pressurizer Relief j Valve Discharge Line," 08/03/95. 191. SNEC Radiological Survey 88H-95-0388, " Weir," 08/04/95. 7 13
l Sixt:n Nucle:r ExperimIntal Corp;r;ti:n (SNEC) Site Characterization Report 192. SNEC Radiological Survey 600-95-0395, " Core Bore At Saxton IGA & High School," 08/07/95. 193. SNEC Radiological Survey 891-95-0396, "795' Primary Recire Pump (MCP) & Piping," 08/04/95. 194. SNEC Radiological Survey 891-95-0397, "795' Primary S/G," l 08/04/95. 195. SNEC Radiological Survey 891-95-0398, "795' Primary Cold Leg S/G to MCP," 08/04/95. i 196. SNEC Radiological Survey 307-95-0408, "Penelec Yard - Core Bores," 08/07/95. 197. SNEC Radiological Survey 891-95-0426, " Containment Vessel , Reactor Vessel Hot & Cold Leg," 08/16/95. 198. SNEC Radiological Survey 867-95-0427, " Containment Vessel Reactor Cavity Super Heat System," 08/16/95.
/
l ( 199. SNEC Radiological Survey 867-95-0428, " Containment Vessel Reactor Vessel," 08/16/95. l 200. SNEC Radiological Survey 869-95-0429, " Containment Vessel l Storage Well Boric Acid, SW, Pur. Demin," 08/16/95. 201. SNEC Radiological Survey 896-95-0432, " Containment Vessel l Reactor Cavity Seal Injection Piping," 08/17/95. I 202. SNEC Radiological Survey 867-95-0434, " Containment Ve.ssel Reactor Vessel," 08/17/95. j 203. SNEC Radiological Survey 897-95-0445, "781' Containment Vessel Storage Well System," 08/21/95. 204. SNEC Radiological Survey 842-95-0450, " Containment Vessel inner l Dome 6' Above Polar Crane," 08/22/95. 205. SNEC Radiological Survey 852-95-0452, " Containment Vessel Tunnel Area," 08/22/95. 206. SNEC Radiological Survey 893-95-0455, " Containment Vessel 779' O Primary Non-Regen Heat Exchanger," 08/22/95. s.) 7-14
Stxten Nucl;cr Experimental C:rpsrati:n (SNEC) l Site Characterization Report b 1
\)
207. SNEC Radiological Survey 893-95-0456, " Containment Vessel 781' l Primary Regen Heat Exchanger," 08/22/95. 208. SNEC Radiological Survey 893-95-0457, " Containment Vessel 779' Primary Letdown Line Between Cold Leg & Regen Heat Exchanger," 08/22/95. 209. SNEC Radiological Survey 892-95-0458, " Containment Vessel 795' Primary Surge Line Pressurizer to Cold Leg," 08/22/95. 210. SNEC Radiological Survey 836-95-0480, "781' Fixed / Loose - Wall," 08/29/95. 211. SNEC Radiological Survey 811-95-0481, "765' Fixed / Loose - Floor," 08/29/95. t 212. SNEC Radiological Survey 836-95-0482, "795' Fixed / Loose Survey - Wall," 08/29/95. 213. SNEC Radiological Survey 864-95-0571, " Containment Vesse! S18' Under Shield Blocks," 09/20/95. t O l Q 214. SNEC Radiological Survey 884-95-0591, " Septic Tank Pump Well Manways," 09/27/95. 215. SNEC Radiological Survey 871-95-0593, " Fixed Contamination Outer
' Containment Vessel Wall & Concrete," 09/28/95.
216. SNEC Radiological Survey 89A-95-0595, "S/G Secondary Handhole Cover," 09/29/95. 217. SNEC Radiological Survey 89A-95-0597, " Fixed Contamination Component Cooling River Water, Surface Bottom Blowdown," 09/29/95. i 218. SNEC Radiological Survey 85Z-95-0629, " Containment Vessel l Tunnelc Septic Tanks, Weir," 10/10/95. I l 219. SNEC Radiological Survey 862-95-0638, " Reactor Cavity & Fuel Pool," 10/12/95. ! 220. SNEC Radiological Survey 214-95-0648, " Containment Vessel 765' Discharge Tank Pumps & Filter / Cubicle," 10/16/95. 221. SNEC Radiological Survey 898-95-0651, " Containment Vessel 818'
'g Borated Water Pumps," 10/17/95.
t 7-15 l
l l
, Srxt n Nucl:ar ExperimIntrl Ccrporation (SNEC)
Site Characterization Report ' t 222. SNEC Radiological Survey 898-95-0652, " Containment Vessel 765' Condensate Tank & Pump," 10/19/95. 223. SNEC Radiological Survey 84C-95-0665, " Personnel Airlock," 10/23/95. 224. l SNEC Radiological Survey 847-95-0666, " Fixed Point inner Walls 4 812'-820',"10/23/95. 225. SNEC Radiological Survey 84Z-95-0667, " Fuel Transfer Trolley," 10/24/95. 226. SNEC Radiological Survey 83Z-95-0668, "795' Component Cooling i Tank," 10/24/95. l 227. SNEC Radiological Survey 894-95-0670, "795' Containment Vessel l Comp. Cooling Heat Exchanger," 10/24/95. 228. SNEC Radiological Survey 81Z-95-0671, "765' Containment Vessel," 10/24/95. o tg 229. SNEC Radiological Survey 82Z-95-0672,781' Containment Vessel Primary instrument Rack," 10/24/95. 230. SNEC Radiological Survey 502-95-0675, "SX-8-56-95-0165, SX-8 95-0107,"10/25/95. L 231. SNEC Radiological Survey 818-95-0742, " Containment Vessel 765' Instrument Racks," 12/5/95. L I 232.' SNEC Radiological Survey 893-95-0755, " Containment Vessel 765' _ Storage Well Discharge Filter," 12/5/95. ) l 233. SNEC Radiological Survey d9B-95-0756," Containment Vessel Superheated Sample Drain Tank," 12/06/95. 234. SNEC Radiological Survey 898-95-0757," Containment Vessel Ventilation Units," 12/07/95. l 4 7-16 , i , , , - _
, I S:xt:n Nuclear Experimental C rp:rati:n (SNEC) Site Characterization Report 7._ '( 'l V List of Appendices
- 1) Appendix A..... .. ...SNEC Site Characterization Plan, Procedure Number 6575-PLN-4520.06, Rev. O,3/24/95, Pages 1 to E4-3
- 2) Appendix B.. ....... ....SNEC Facility Photographic Section, Pages 1 to 32
- 3) Appendix C... .. .... . Regulatory Guide 1.86, " Termination of Operating Licenses for Nuclear Reactors", June 1974,5 Pages
- 4) Appendix D.. . ... . ..Off-Site Radiological Analysis of SNEC Facility Samples.
a) B&W Report No. 9507002, September 14,1995, includes , Supplemental Information in Report No. 9507002S, January 10, 1996,36 Pages. Samples include Pipe Sections, Scrapings, Filter Solids & Sludge Materials. l l I , (, ,) Attached to reports are the individual sample percentages of each v' isotope for each sample. Systems are identified, TRU & U Content is estimated for each sample analyzed. Ratios between various nuclides are calculated. Assumes "Less Than" values are positive results. b) B&W Report No. 9509065, November 2,1995, includes Supplemental Information-TCLP Metals Analysis, Bulk Asbestos Results and Total Lead Analysis Results,28 Pages. Samples include Pipe Sections, Tank Sediment, Paint Chips, SW Resin, S/G Samples and analysis of Reactor Cavity Steel. Attached to reports are the individual sample percentages of each isotopo for each sample. Systems are identified, TRU & U Content is estimated for each sample analyzed. Ratios between various nuclides are calculated. Assumes "Less Than" values are positive results. c) B&W Report No. 9510024, December 6,1995, Concrete Core Bore Results,17 Pages. Attached to report is the individual sample percentages of each [ ) isotope for the Hi-n-Flux (cetivated), core bore sample and a (_/ composite of other core bore sample results from non-activated i l l
Scxten NucleIr Experimental Ccrporttisn (SNEC) Site Characterization Report areas (No-n-Flux). Also included is the sample number key relating the original core bore sample numbers with the samples sent to B&W.
- d) B&W Report No. 9511048, December 22,1995, CV Tunnel, RV Sample, Rebar Sample & Condensate Tank Sediment Results,10 Pages.
Attached to reports are the individual sample percentages of each isotope for each sample. Systems are identified, TRU & U Content is estimated for each sample analyzed. Ratios between various nuclides are calculated. Assumes "Less Than" values are positive results, e) Teledyne Brown Engineering Environmental Services Reports on Composite Smears From CV AREA's 1,2,3/4 & 6. Dated 1/15/95 and 6/26/95,7 Pages. Normalized results from these reports are presented in section 4.1.
- 5) Appendix E..... .. .... ..Results of SNEC Facility Piping System Walk-p Down, Spring 1995,2 Pages.
- 6) Appendix F................SNEC Facility Chain of Custody Log for SNEC Samples, Pages 1 to 11.
i i
. . , . . . , . . . - _ - . ~ .
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b;- .\ g 812' elevation looking l east over the shield f blocks. The " key block" k - l is visible laying down , p" p' g
- while the temporary .
nm~. - covers are visible in the
. g3.
foreground. - Ap . ._ g. g y g]. 1
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812' elevation fuel ,.
, .. m handling bridge ladder E ,' ' "I ' " - ,.,:
. ,o
! 3 4
l to storage well. _ 4 4 i 6 i 4.-- _-- _ _ _ _ _ _ - _ _ . _ _ __ ,_ . . _ _ _ _ __ _,
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+ nw View from 812' elevation open il,'
'$y^ W
- hatch looking down and to ?[-
- 27 l l the south. Primary compart- IN,(;g : ,[" ..d$F ment is to the right. ,' 'T -
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Looking down from s - t . - is. .' 812' elevation to 765' b -
- U '
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>e llI elevation. Primary .
7r~ -- e . . . compartrnent outer -
\' :
~
south wall is visible at ,
-If n -l~* _;-_ f_
7, the top of the picture. %. ; ...i , nif i=== _. .R. 7
f__________ i i i 1 Uk. , , " __ r t
, A i
Auxiliary compartment 795' elevation component closed cooling / river water i I heat exchangers. ( y l l I I i l .
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'! 1 _$ -
compartment looking ?. . east, northeast. Visible is - :-
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l the closed cooling water . -; . _, _ t . . i {p system piping above the ' -
'{ / 'jp ~ ! - .
two pumps. . !. . : :- _ f 8 i
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i (- .;i f ql- - .?. e .. 795' elevation Auxiliary ;. L . ~? - :/~
, s. u compartment east wall, . ,n7 ,ft~. ..,.
y river water check valve. @['J g g;'- ... ..A: ! i
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j Auxiliary compartment " '~ ' ~ ^ ~ ' ' 795' elevation #1 com-i ponent cooling pump. - f 9gg l } 9
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l i.. :,;. . . . l 1 [.y}.2[ ' . I l g i l 795' elevation Auxiliary - j ! compartment air handling - ! unit in the ceiling. -
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_ m mme mny w y, i.f l
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1 t j 781' elevation Auxiliary compartment i- , l view of access into the Primary com- Ih,ki j partment looking west. ti j ,. [ SM 10 i'
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j Storage well cooling . g
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pump #2 (1 of 2). 781' i <- t elevation Auxiliary .. j compartment. ,
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l > j r*, t 781' elevation Auxiliary i Y. .
. I 1
4
!g -
i compartment, storage 's ' ! well heat exchanger. - A I 11 I
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x ' 781' elevation Auxiliary compartment looking . l northeast. Storage 6 well cooling suction l
- iI piping and strainer are visible. 6V1785 ,- . - -
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m& 781' elevation Auxiliary hE" , compartment looking l:f)W south at 1LRC-21V. Visible to the left is an ; E?A Af incore drive assembly. __ P p, J..; I
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1 : : : :7 l : -: -- - 4.. e j 781' elevation Auxiliary ,
.:: : [: ,,,
l compartment air han- . . s Jy;l.f.gi,; i ' .' ' l dling unit (#2) looking . east. Open hatch at Lf.$1 - ,:? right.
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1, l 765' elevation Auxiliary compartment :. ] southeast side. Condensate return w
~
s i ~ .. t[ d. ..*jI unit. th. - i 13 1
- 1. . . .. . . __ _ , _
1 l 7-
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e l. i - 1 l 765' elevation Auxiliary l compartment looking I north at the south end , of the drain tank. _ i l 1 f 1 g - i i i i e i l I i = 1 e i l I ( j 765' elevation Auxiliary i compartment south --
%4%
l west side. Shutdown
! cooling pump #2 (lof -
qa
~~ "
- 2) on the 768'-3" ledge. "-
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i B. f;Y i < .. f 1 765' elevation Auxiliary 2 .
- (( ' '.
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compartment looking b - 1 north. Area above the 7gy, - b ,
'y
). I drain tank. Visible above the tank is the shutdown *
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cooling heat exchanger. - 1
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l i f 765' elevation Auxiliary compart-I ment looking east at the instrument } cabinent on the outer west wall of i the filter room. I 15 i i
. _ _ _ _ _ _ _ . _ _ . _ - _ _ - . _ - ._~._ _ _ .__.__.__ _ __._..__._.. __.____ _
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New fuel storage rack, SH w 765' elevation. jn! g . i n .. n I
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I 765' elevation filter room facing 9 d south west corner. L .. . ' . [. a 16 A u _ _ _ _ _____ __ .__ _ - . - . --
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Primary compartment looking "a. L. down on the 795' elevation [, "N, p? % M -
+
y' - - j platform. Visible is the main y "
!O '
coolant cold leg from the 4.g
- g4 l discharge of the main coolant .
l pump which is on the right. .- . r ! The piping above the cold leg n ' is the pressurizer spray line.
~
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h,$hh,f.w!ffff . g ..
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! 795' elevation Primary compartment 4?f . .-
. 351% 1d:J main coolant pump looking north. 'A _
i 17
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g l ,, i ' 1 3 . 795' elevation Primary compartment looking .
? ,. .
g : north at the main , , j coolant pump discharge .. . line (cold leg) as it
.p 4 w passes through the l '
? g .t i south wall of the j .) .
- storage well. The pipe ' '
4iB . : ~- N l lagging and " bricks" . A , L l visible are asbestos. 4 M2 . h#!j?. l I i 1 I i ,. L.u. 1 . .
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_e i s-795' elevation Primary _ _ . .
~t ' .
7 s compartment looking
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l y 779'-8" elevation Primary compartment j
'f. p l
- looking west. East end of the regen heat '
- f. :g
- exchanger.
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i j i i 7' l 770'-8" Primary compartment looking
} west at the west end of the regen heat exchanger.
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779'-8" elevation Primary compartment ! ; 4 l1 valve & piping arrangement in front of , f - 5 f,) i the non-regen heat exchanger, the north _ f -
-f
! end of which is visible in the back- .A 'U.Q l ground. l';Jg ' . [ ] i } i i s l . k, ! e ! 4 l .
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779'-8" elevation Pri- . sk , ~ 4 .j i j mary compartment 3mW w j j looking up at the hot leg side Steam Genera-ggg g ' f
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- tor manway. !
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[ f, a 779'-8" elevation Primary Nh , compartment looking up J
.( p, s[}ift s
! ,} " l at the suction of the 7 .. gg i Main Coolant Pump. u ~ T.s.aSm ds ww? ,
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j 779'-8" elevation Primary
~
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s.'h j compartment area behind the instrument cabinet
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and above the non-regen g Sf {7 p's - i l 7: j heat exchanger looking ^ F3 ; .
,_ W ' N.. TY f,,;
I ! north. , . 1 1 1 21 1
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1 I j . g t....' 779'-8" elevation Pri-
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g- .g~.* 1 f . l g j mary compartment ** ' 3 ,y l valve & piping arrange- - j ment above regen heat ,,'. '.f. '
--'g .
?;
{ exchanger looking v. _ .~ y 8 ' i l north. i 4MQ - [ f .L l g $ ;;, d
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L I g^ g T "~' J'?t.g;,'ih. i I ti' ~ r e.w: l y[; e,O i diiiik l pg, ! I. j - 779'-8" elevation Primary compart- j% l I( i ment looking east. South half of the fij _ y ? e i f
~
l B R R:.c.'s ! instrument rack. ' w-- : : ; -- . 22 E
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s
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., ^ J. - M i 779'-8" elevation Primary compartment , ,
l air handling unit (#3) looking west. . { m% _. - _- . . j -[$8&; . 2j .' f f.sLm ! e W<'I d-! } s b'8.h;l , : 1 m l } Mp e Q... t ,: 'e -l= ? .,( .
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j 765' elevation looking north into the ,jM i
; Rod Room. CDRMs and lower reactor l instrument penetration are visible.
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. .a 765' elevation Rod Room r j looking north. .
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, "th j y0 e<
j 765' elevation under the 4 , j Reactor Vessel looking i' up into the Storage Well ; , Penetration Rod Room. , , e, . 4 i 24
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765' elevation Rod ,. , Room " Sump" Core Bore [- -
^'
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Location. , , , , .
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812' elevation view from j shield blocks down to
), .
j the 793' elevation . j platform of the Reactor .,., f, , Head. Note the steel
~
-r f shield segments sur- . id. . / -
rounding the vessel on . I: i
; .I l
the grating. 'N k AO , .
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4 Y-5 25
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4)x 1 Superheated steam test . t i project components. [ ., y -l i Reactor Storage Well. ., .a- . , , , , ,4 800' elevation platform 7 N swg . (- wer., gr. .")N$ki t .e < h-
. 4
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l l l >i 6 9t 8 j Tank, part of the A
- es j superheated steam ]' y.
i test project. Reac- .
} tor Storage Well, se*
l west end 800' ' elevation. 'g , i i 26 1 1
k 1 i 4 4 1 l ~. , e. . 3 4 , A. . y ,.u e Ar.r 1 ., e 3:?pL; n .:n.awf. ,.p*' ^ 5&
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a. e, a 793' elevation Storage " i * '
?.M'W
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PI ' GI Coat- ! WeIlI ex in8 repatr"(painted over duct taper), fSfkmg;;2;w.,.. .[W. s.
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l $v + 5 *I +", ] 1 I fg'f ' c t f;; '; 1 795' elevatto ? *P e of extensive cracktnE ( x,y"ff'fly'*yfy*y;A[F W fzf,QQQ h wn9 v X
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i I s < [ , I' I 793' elevation Reactor ' Storage Well. Reactor head penetrations. h ~
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! 3 \ ~ fA ! Ti 779'-8" elevation looking
- up at Reactor outlet (hot , W l leg), ..I d. i 28 i;
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_l 779' elevation Reactor ~I'. U [ ' ~ ' '# Storage Well. Example
'!' - -/ -
S of Ni mounting assem- ;g .. j - bly (1 of 8). % e a'
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. . $ $3II
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] Storage Well safety injec- fp ) I I i j tion inlet check valve - 792' g elevation, north wall. No- .- !-
~
tice the cracking and delamination of the wall [!, g' l l lining. Valve 3-V253. -- 2 29 1 i ,
l J J i i s 6 l a enwm . - 3
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1 , i j 793' elevation Reactor i Storage Well. Reactor [
- j Vessel internal instru- p b Md .,
ment assembly lifting
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.. 1 Spent Fuel Pool south side, close-up of the % , % . * / 7'~
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. i.I <l i a j
d[f y - l J$$~ Spent Fuel Pool south j
- side taken from 779'-8"
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,)/,.
1 4 1 32 i
p . . l l l 4 APPENDIX C , i ! l } 1 t } 4 i ) i i I t 4 / 1 1 1 4 4 a i i e 1 s d s a f 1 4
l June 1974 [#" #% U.S. ATOMIC ENERGY COMMISSION o o 4 s,, , ,.f r REG'ULATORY GUIDE DIRECTORATE OF REGULATORY STANDARDS REGULATORY GUIDE 1.86 TERMINATION OF OPERATING LICENSES FOR NUCLEAR REACTORS A. INTRODUCTION A licensee having a possessionenly license must retain, with the Part 50 license, authorization for special Section 50.51, " Duration of license, renewal," of 10 nuclear material (10 CFR Part 70, "Special Nuclear
- CFR Part 50, " Licensing of Production and Utihzation Material"), byproduct material (10 CFR Part 30," Rules Facilities," requires that each license to operate a of General Applicability to Licensing of Byproduct production and utilization facility be issued for a Material"), and source material (10 CFR Part 40, specified duration. Upon expiration of the specified " Licensing of Source Material"), until the fuel, radio-period, the license may be either renewed or terminated active components, and sources are removed from the by the Commission. Section 50.82, " Applications for facility. Appropriate administrative controls and facility termination of licenses," specifies the requirements that requirements are imposed by the Part 50 license and the must be satisfied to terminate an operating license, technical specifications to assure that proper surveillance including the requirement that the dismantlement of the is performed u.d that the reactor facility is maintained facility and disposal of the component parts not be in a safe condition and not operated.
inimical to the common defense and security or to the health and safety of the public. This guide describes A possession.only license pennits various options and methods and procedures considered acceptable by the procedures for decommissioning, such as mothballing, p Regulatory staff for the termination of operating entombment, or dismantling. The requirements imposed ( licenses for nuclear reactors. The Advisory Committee Q)' on Reactor Safeguards has been consulted concerning this guide and has concurred in the regulatory position. depend on the option selected. Section 50.82 provides that the licensee may dis. l mantle and dispose of the component parts of a nuclear B. DISCUSSION reactor in accoidance with existing regulations. For research reactors and critical facilities, this has usually When a licensee decides to ternunate his nuclear meant the disusembly of a reactor and its shipment reactor operating license, he may, as a first step in the offsite, sometimes to another appropriately licensed process, request that his operating license be amended to organization for further use. The site from which a restrict him to possess but not operate the facility. The reactor has been removed must be decontaminated, as advantage to the licensee of converting to such a necessary, and inspected by the Commission to deter. possessiononly license is reduced surveillance require- mine whether unrestricted access can be approved. In ments in that periodic surveillance of equipment im. the case of nuclear power reactors, dismantling has portant to the safety of reactor opezation is no longer usually been accomplished by shipping fuel offsite, required. Once this possession.only license is issued, making the reactor inoperable, and disposing of some of reactor operation is not permitted. Other activities the radioactive. components. related to cessation of operations such as unloading fuel from the reactor and placing it in storage (either onsite Radioactive components may be either shipped off. of offsite) may be continued. site for burial at an authorized burial ground or secured USAEC RECULATORY GUIDES ces o, .u ,r., otn c. dive "U'f ,,
"*"*J'I ' '."".#" .;,d G.. '"a' '"E"C"'.E
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s .uq M.thoth .nd ut d if , .,n out m Th. Ou.dn .ro i u.d 6n th. vollo ,n. t.n tiro.d d,n.,on.- Et!,."."J .'",f 'f.*f. ,'.%'";t'.', 'T,,"Oca'02. '"'"*" " i. - .
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. .o . s in (m) lf,,%"J. .*d O ,%?" """',O"*.'l".%" **o". *"*"' " *"*"""*" i U.",*.".~,"'fd."#,".'1,on d ^2".'T """
on the site. Those radioactive materials remaining on the fluids and waste should be removed from the site. l site must be isolated from the public by physical barriers Adequate raliation monitoring, environmental sutveil-i / lance, and appropriate security procedures should be l (\ j]. or other means to prevent public access to hazardous
!crels of radiation. Surveillance is necessary to assure the established under a possession only license to ensure that the health and safety of the public is not endangered.
long term integrity of the barriers. The amount of wr.eilbrice required depends upon (1) the potential lard to the healdi and safety of the public from b. In-Place Entombment. In place entombment con-radioactive rnaterial remaining on the site and (2) the sists of sealing all the remaining highly radioactive or integrity of the physical barners. Before areas may be contaminated components (e.g., the pressure vessel and released for unrestricted use, they must have been reactor intemals) within a structure integral with the decontaminated or the radioactivity murt have decayed biological shield after having all fuel assemblies, radio-to less than presenbed limits (Table 1). active fluids and wastes, and certain selected com-ponents shipped offsite. The structure should provide The hazard associated with the retired facility is integrity over the period of time in which significant evaluated by considering the amount and type of quantities (greater than Table I levels) of radioactivity remaining contamination, the degree of confinement of remain with the material in the entombment. An the remaining radioactive materials, the physical security appropriate and continuing surveillance program should provided by the confinement, the susceptibility to be established under a possession only license. ! release of radiation as a result of natural phenomena, and the duration of required surveillance. c. Removal of Radioactive Components and Dir,- mantling. All fuel assemblies, radioactive fluids and C. REGULATORY POSITION waste, and other materials having activities above ac-cepted unrestricted activity levels (Table I) should be
- 1. APPLICATION FOR A LICENSE TO POSSESS BUT removed from the site. The facility owner may then have NOT OPERATE (POSSESSION-ONLY LICENSE) unrestricted use of the site with no requirement for a license. If the facility owner so destres, the remainder of .
the reactoi facility may be dismantled and all vestiges I A request to amend an operating license to a possessionenly license should be made to the Director removed and disposed of, of Licensing, U.S. Atomic Energy Commission, Washing-ton, D.C. 20545. The request should include the d. Conversion to a New Nuclear System or a Fossil
) following information: Fuel System. This alternative, which applies caly to
( nuclear power plants, utilizes the existing turbine system (v j
- a. A description of the current status of the facility- with a new steam supply system. The original nuclear steam supply system should be separated from the
- b. A description of measures that will be taken to electric generating system and disposed ofin accordance prevent criticality or reactivity changes and to minimize with one of the previous three retirement alternatives.
releases of radioactivity from the facility.
- 3. SURVEILLANCE AND SECURITY FOR THE RE.
- c. Any proposed changes to the technical specifica- TIREMENT ALTERNATIVES WHOSE FINAL tions that reflect the possession only facility status and STATUS REQUIRES A POSSESSION-ONLY the necessary disassembly / retirement activities to be LICENSE performed.
A facility which has been licensed under a posses-
- d. A safety analysis of bo'th the activities to be sion-only license may contain a significant amount of accomplished and the proposed changes to the technical radioactivity in the form of activated and contaminated specifications. hardware and structural materials. Surveillance and commensurate security should be provided to assure that
- e. An inventory of activated materials and their the public health and safety are not endangered.
' location in the facility.
- a. Physical security to prevent inadvertent exposure
- 2. ALTERNATIVES FOR REACTOR RETIREMENT of personnel should be provided by multiple locked barriers. The presence of these barriers should make it Four alternatives for retirement of nuclear reactor extremely difficult for an unauthorized person to gain facilities are considered acceptable by the Regulatory access to areas where radiation or contamination levels staff. These are: exceed those specified in Regulatory Position C.4. To prevent inadvertent exposure, radiation areas above 5 1
- a. Mothballing. Mothballing of a nuclear reactor mR/hr, such as near the activated primary system of a l facility consists of puttmg the facility in a state of power plant, should be appropriately marked and should protective storage. In general, the facility may be left not be accessible except by cutting of welded closures or i
O) intact except that all fuel assemblies and the radioactive the disassembly and removal of substantial structures ( /
" 1.86 2
and/or shielding material. Means such as a remote- (1) Environmentalsurveys, readout intrusion alarm system should be provided to indicate to designated personnel when a physicalbarrier (2) Facility radiation surveys, is penetrated. Security personnel that provide access control to the facility may be used instead of the (3) Inspections of the physical barriers, and
. physical barriers and the intrusion alarm systems.
(4) Abnormaloccurrences.
- b. The physical barriers to unauthorized entrance into the facility, e.g., fences, buildings, welded doors, and access openings, should be inspected at least 4. DECONTAMINATION FOR RELEASE FOR UN.
quarterly to assure that these barriers have not deterior. REKIRICTED USE . ated and that locks and locking apparatus are intact. If it is desired to terminate a license and to eliminate
. . c. A facility radiation survey should be performed at any further surveillance requirements, the facility should least quarterly to verify that no radioactive materialis be sufficiently decontaminated to prevent risk to the ;
escaping or being transported through the containment public health and safety. After the decontamination is 1 I barriers in the facility. Sampling should be done along satisfactorily accomplished and the site inspected by the most probable path by which radioactive material the Commission, the Commission may authorize the such as that stored in the inner containment regions license to be terminated and the facility abandoned or could be transported to the outer regions of the facility released for unrestricted use. The licensee should per. , and ultimately to the environs. form the decontamination using the following guide. lines:
- d. An environmental radiation survey should be performed at least semiannually to verify that no a. The licensee should make a reasonable effort to signficant amounts of radiation have been released to the eliminate residual contamination.
environment from the facility. Samples such as soil, vegetation, and water should be taken at locations for b. Ido covering should be applied to radioactive which statistical data has been established during reactor surfac'es of equipment or structures by paint, plating,or operations. other covering material untilit is known that contamina. tion levels (determined by a survey and documented) are
- e. A site representative should be designated to be below the limits specified in Table I. In additica, a responsible for controlling authorized access into and reasonable effort should be made (and documented) to movement within the facility. further minimize contamination prior to any such (b covering.
- f. Administrative procedures should be established for the notification and reporting of abnormal occur. c. The radioactivity of the interior surfaces of pipes, rences such as (1) the entrance of an unauthorized drain lines, or ductwork should be determined by person or persons into the facility and (2) a significant making measurements at all traps and other appr,opriate change in the radiation or contamination levels in the access points, provided contamination at these locations facility or the offsite environment. is likely to be representative of contamination on the interior of the pipes, drain lines, or ductwork. Surfaces
- g. The following reports should be made: of premises, equipment, or scrap which are likely to be contaminated but are of such size, construction, or (1) An annual report to the Director of Licensing, location as to make the surface inaccessible for purposes U.S. Atomic Energy Commission, Washington,' D.C. of measurement should be assumed to be contaminated 3 20545,' describing the results of the environmental and in excess of the permissable radiation limits.
facility radiation surveys, the status of the facility, and an evaluation of the performance of security and d. Upon request, the Commission may authorize a surveillance measures. licensee to relinquish possession or control of premises, equipment, or scrap having surfaces contaminated in (2) An abnormal occurrence report to the Regula- excess of the limits specified. This may include, but is tory Operations Regional Office by telephone within 24 not limited to, special circumstances such as the transfer hours of discovery of an abnormal occurrence. The of premises to another licensed organization that will abnormal occurrence will also be reported in the annual continue to work with radioactive materials. Requests report described in the preceding item. for such authorization should provide:
'h. Records or logs relative to the following items (1) Detailed, specific information describing the should be kept and retained until the license is termi. premises, equipment, scrap, and radioactive contami-nated, after which they may be stored with other plant nants and the nature, extent, and degree of residual records: surface contamination.
k I'.86 3 i
-. . - - - - - - . . .- . =
(2) A detailed health and safety analysis indi- or a change in the technical specifications should be cating that the residual amounts of materials on surface reviewed and approved in accordance with the require- [
\
areas, together with other considerations such as the prospective use of the premises, equipment,or scrap, are ments of 10 CFR 950.59. N unlikely to result in an unreasonabic risk to the health if major structural changes to radioactive components and safety of the public. of the facility are planned, such as removal of the pressure vessel or major components of the primary
- e. Prior to release of the premises for unrestricted system, a dismantlement plan including theinformation
- use, the licensee should make a comprehensive radiation required by 950.82 should be submitted to the Commis survey establishing that contamination is within the sion. A dismantlement plan should be submitted for all limits specified in Table 1. A survey report should be the alternatives of Regulatory Position C.2 except filed with the Director of Licensing, U.S. Atomic Energy mothballing. However, minor disassembly activities may Commission, Washington, D.C. 20545, with a copy to still be performed in the absence of such a plan, the Director of the Regulatory Operations Regional provided they are permitted by existing operating and Office havingjunsdiction. The report should be filed at maintenana procedures. A dismantlement plan should least 30 days prior to the planned date of abandonment. include the following:
The survey report should:
- a. A description of the ultimate status of the facility (1) Identify the premises;
- b. A description of the dismantling activities and the (2) Show that reasonable effort has been made to precautions to be taken.
reduce residual contamination to as low as practicable levels; c. A safety analysis' of the dismantling activities including any effluents which may be released. I (3) Describe the scope of the survey and the I general procedures followed;and d. A safety analysis of the facility in its ultimate status. (4) State the finding of the survey in units l specified in Table 1. Upon satisfactory review and approval of the dis. mantling plan, a dismantling order is issued by the After review of the report, the Commission may Commission in accordance with 950.82. When dis-mantling is completed and the Commission has been
- kj) inspect the facilities to confirm the survey prior to granting approval for abandonment. notified by letter, the appropriate Regulatory Opera-tions Regional Office inspects the facility and verifies
- 5. REACTOR RETIREMENT PROCEDURES completion in accordance with the dismantlement plan.
If residual radiation levels do not exceed the values in As indicated in Regulatory Position C.2, several Table I, the Commission may terminate the license. If alternatives are acceptable for reactor facility retirement. these levels are exceeded, the licensee retains the If minor disassembly or "mothballing" is planned, this possession only license under which the dismantling could be done by the existing operating and mainte- activities have been conducted or, as an alternative,may nance procedures under the license in effect. Any make application to the State (if an Agreement State) planned actions involving an unreviewed safety question for a byproduct materials license. l l s 2 1.86 4 I
, _ . _ _ . _ .m _ . _ _ . _ _ . _ _ . . ._ _ .._ _ _ _ __.____ _ _. _ . _ _ _ _ _ _ . __ ___ _ , TABLEI ! 't ACCEPTABLE SURFACE CONTAMINATION LEVELS NUCLIDEa AVERAGEb c MAMMUMbd REMOVABLEb e - U-nat, U-235, U 238, and 5,000 dpm a/100 cm2 15,000 dpm a/100 cm2 1,000 dpm a/100 cm2 associated decay products 1 l Transuranics, Ra 226, Ra 228, 100 dpm/100 cm2 300 dpm/100 cui2 20 dpm/100 cm2 Th-230.Th-228,Pa 231, Ac-227,1 125,1129 Th-nat, Th 232, Sr-90, 1000 dpm/100 cm2 3000 dpm/100 cm2- 200 dpm/100 cm2 Ra 223 Ra-224,U 232, 1126,1131,1133 Beta-gamma emitters (nuclides 5000 dpm M/100 cm2 15,000 dpm h/100 cm2 1000 dpm h/100 cm2 with decay modes other than alpha ' emission or spontaneous fission) except Sr 90 and others noted above.
*When surface contamination by both alpha and betagammaemitting nuclides exists, the lisaits established for alpha- and ber - " ting nuclides should apply ' ' , ^tly.
i bas used in this table, dyn (didatagrations per annute) means the ate of enaission by radioactive matedal as detonnined by correcting the counts per minuts observed by an appoopriate detector for background, of5ciency, and secanstric factors associated with the instrumentation. - l cMeasurements of average contandnant should not be averaged over more than 1 squase meter. For objects ofless surface area, the average should be derived for each such object. 2 dThe mamunum con *====*w level applies to an ama of not more than 100 cm ,
'The amount of removsble radioactive material per 100 cm of 2 surface anoa should be detsnained by wiping that area with dry fDter or soft absorbent paper, applying moderate pressuse, and asesemos the amount of radioactsre materiet on the wipe with an appropriate l
instrument of known efficiency. When sonnovable contaraination on objects of less surface area is determined, the pertinent levels should be reduced propornonally and the entire surface should be wiped. l l l i e i L l 86-5 { i
APPENDIX D 1 O I l O . l l
l
- B&W NUCLEAR ENVIRONMENTAL SERVICES, INCORPORATED i
NUCiFAR ENVIRONMENTAL LABORATORIES ? - 1 j l DATA REPORT NUMBER 9507002 i 10CFR61 ANALYSIS RESULTS 1 t i Submitted to ) GPU Nuclear Corporation i Saxton Site - Decommissioning Project l 1 g September 14,1995 i l Prepared by: # David S. Kohl, Supervisor Radioanalytical Chemistry Laboratory Released by: Nb- A N[k James L. Clark, Project Manager O Nuclear Environmental Laboratories .
= . - - _
f. i, . . 4 k GPU-Saxton Data Report September 14,1995 j DATA REPORT
SUMMARY
l Laboratory Project Summary i 1 e Plant Site: Saxton Decommissioning Project 4 e Shipment Number: 95 003 i e Receipt Date: July 10,1995 i , e NEL COC Number: SN-009274,009275,009276 l l e NEL SDG Number: 9507002 1 i e NEL Proposal: NEL5-1258 r e Contract Number: 0397355 k
- Analysis Service: Routine i
, e Laboratory QC: Routine - Batch j 4 I h l j i i 4
O-GPU Saxton Data Report September 14,1995 l 9 l l Sample Delivery Group Summary l Customer ID NEL Sample ID Sample Type Description SX815950010 9507002-01-AA Pipe Section item 12 SX825950011 9507002-02-AA Pipe Section Demin Water - SX825950040 9507002-03-AA Pipe Section . Item 6 SX825950041 9507002-04-AA Pipe Section item 3 SX825950042- 9507002-05-AA Pipe Section item 7 SX825950043' 9507002-06-AA Pipe Section item 2 l- .SX825950044' 9507002-07-AA Pipe Section Item 4 l- SX825950045 9507002-08-AA Pipe Section Test Cut SX825950046 9507002-09-AA Pipe Section item 9 i SX825950047 9507002-10-AA Pipe Section item 13 SX815950006 950700211-AA Scrapings Instrument Rack 1 SX815950007 9507002-12 AA Scrapings Instrument Rack 2 SX825950038 9507002-13-AA Filter / Solids item 11 SX835950039 9507002-14-AA Filter / Solids Aux Air Handler SX815950050 9507002-15-AA Filter / Solids Control Rod Room SX815950005 9507002-16-AA Sludge CV Sump SX885950048 9507002-17-AA Sludge Septic Tank A SX885950049 9507002-18 AA Sludge Septic Tank B O l ii
1 l0 ? GPU-Saxton Data Report September 14,1995 a l Sample Physical Characteristics Summary l { NEL Sample ID Matrix Leach Volume (mL) Pipe Mass' Leached (g) } { 9507002-01-AA Pipe Leach 3755 163 9507002-02-AA Pipe Leach 513 10 9507002-03-AA Pipe Leach 258 10 ! 9507002-04-AA Pipe teach 220 10 { 9507002-05-AA Pipe Leach - 648 11 9507002-06-AA Pipe Leach 1039 53 9507002-07 AA Pipe Leach 2919 2 9507002 08-AA Pipe Leach 1858 0 9507002-09-AA Pipe Leach 1988 1 9507002-10-AA Pipe Leach 3676 69 NEL Sample ID. Sample Matrix Mass / Volume Density i 9507002-11-AA Mixed solids 14.65 g 1.33 g/cc 9507002-12-AA Mixed Solids 12.32 g 1.54 g/cc
- 9507002-13-AA Mixed Solids 15.46 g 1.10 g/cc 9507002-14-AA Mixed Solids 3.74 g 0.93 g/cc 9507002-15-AA Mixed Solids 2.34 g 2.34 g/cc 9507002-16-AB Sludge Liquid 43 mL 1.40 g/mL 9507002-16-AC Sludge Solids 67 g 0.99 g/cc 9507002-17-AB Sludge Liquid 14 mL 2.16 g/mL 9507002-17-AC Sludge Solids 83 g 1.28 g/cc 9507002-18-AB Sludge Liquid 71 mL 1.33 g/mL 9507002-18-AC Sludge Solids 60 g 1.39 g/cc iii
l l l I GPU Saxton' Data Report September 14; 1995 1 i l Pipe Section Leach / Decon Summary l l NEL Sample ID % Activity Removed ("Co) % Activity Removed (*Cs) 9507002-01-AA 77
- 2 77
- 3 9507002-02-AA <0 65
- 2 9507002-03-AA <1 .O
{ 9507002-04-AA' 17
- 1 Not Detected 9507002-05-AA 21
- 2 64
- 1 l 9507002-06-AA 8.7
- 0.7 Not Detected 9507002-07-AA 43
- 6 32 t 1 950,002-08-AA 0 30 6 9507002-09-AA <1 61
- 2 9507002-10-AA 97
- 1 92
- 1 l.
4 IV I
1 l !O l GPU Saxton Data Report
~
~
September 14,1995 1
~
l Sample Analysis Summary
- l l Customer ID NEL Sample ID Sample Matrix Notes i
! SX815950010 9507002-01-AA Acid Leachate (1,2,3) SX825950011 9507002-02-AA Acid Leachate (1,2) l SX825950040 9507002-03-AA Acid Leachate (1,2) { !- SX825950041 9507002-04-AA Acid Leachate (1,2) SX825950042 9507002-05-AA Acid Leachate (1,2) SX825950043 9507002-06-AA Acid Leachate (1,2) SX825950044 9507002-07-AA Acid Leachate (1,2) SX825950045 0507002-08-AA Acid Leachate (1,2) SX825950046 9507002-09-AA Acid Leachate (1,2) SX825950047 9507002-10-AA Acid Leachate (1,2) } SX815950006 9507002-11-AA Mixed Solids --- ! SX815950007 9507002-19-AA Mixed Solids --- 1 SX825950038 9507002-13-AA Mixed Solids (4) SX835950039 9507002-14-AA Mixed Solids (4)
- SX815950050 9507002-15-AA Mixed Solids (4) il SX815950005 9507002-16-AB Sludge Liquid (5).
[ 1 SX815950005 9507002-16-AC Sludge Solids (5) j SX885950048 9507002-17-AB Sludge Liquid (5) 1 j SX885950048 9507002-17-AC Sludge Solids (5) j SX885950049 9507002-18-AB Studge Liquid (5) SX885950049 9507002-18-AC Sludge Solids (5) }
~
1O i V 4 i _ - _
I O GPU Saxton Data Report September 14,1995 l Sample Analysis Summary' Notes l Note-1 Acid leaching with 30% HNO3 / 5% hcl for 3-4 hour period. Note-2 ' Leach solution subsequently analyzed includes small volume rinses. Note-3 Leach solution combined with significant loose solid deposits. Note-4 Loose solids combined with surface scrapings from filter media. Note-5 Solid / Liquid phases separated by filtration. Solid analyzed on wet weight basis. - l Activity Results l l l See Attached Report Sheets I vi l l
,~
w v ' b Sexton fwutts Summary 9507002 (Revreed 09/29/95)
/ o fdl ID 9507002 4t 9507002-02 9507002 43 2SUNC 9507002 44 e507002 45 ACTMTY. ACitVITY 2SUNC ACTMTY 2SUNC ACTMTV tuCWs Removed 2SUNC ACTMTV 23UNC lucueRemoved guCug Removed tuCUg Removed luCVg Removed H-3 < 054E44 < 1.12E43 '-
< 7.33E44 -
< e.24E44 <
C.14 < t .59E44 - < 1.5SE 44 < 2.23E44 l.87E43 - Fe 65
< 1.6eE44 - . < 3 05E44 3.41E44 7.14E45 < s.e4E46 -
5.37E43 8.72E44 96-5 9 4.95E43 4.55E43 7.39E44 2.55E42 4.14E43 7.23644 < 9.59E46 - 8 69E44 991E45 5.30E43 7.92E 44 E83 1.47E45 2.15E42 2.19E44 2.01E43 3.01 E-04 1.58E45 4.seE42 7.12E43 3.34E41 4.eeE42 ~ 1.24E45 se e9 < 3.30E44 - < 4.62E46 - < 3.10E46 < 1_e0E 42 ss.co 8 70E43 3.01E46 - < e.01E46 - 1.37E44 2.81 E45 4.47E48 2 99E42 1.53E 44 1.41E42 Nti94 < 1.03E44 3.97E44 2.57E45 2.04E45 - < 2.0eE46 - < s.75E45 < 2.46E44 Te99 2.97E44 5.99E45
< - 7.20E45 -
< e 79E45 - < 5.90E45 - < 5.87E45 -
< 1.22E43 g.129 < 3.e2E 45 - < 2.17E45 - < 4.05E45 < 1.53E45 <
3.09E45 - No 237 < 3.99E47 -
< 4.26E47 -
< t .99E43 < 1.79E43 <
3.93E43 - h 238 4.2GE45 2.39E4e 1.03E45 1.2 t E46 2.73E42 1.96E43 4.24E43 5.e5E44 4.10E42 3.54E43 N239/240 S.96E45 4.70E 46 4.10E45 2.42E46 7.40E42 4.37E43 9.14E43 9.2eE44 a.4eE42 4.90E43
- N241 f.74E43 1.71 E44 8.64E44 8.52E45 1.16E + 00 1.19E41 1.09E41 2.17E42 1.29E + 00 1.3eE41 N 242 < 2.07E4e - < 3.74E40 - < s.26E45 - < e.13E45 - < 2.36E44 -
Am-241 2.87E45 2.t SE46 3.73E45 3.39E46 S.33E42 4.51E43 0.20E43 4.06E44 3.53E43 5.7eE44 Am.243 < 1.23E47 - < 2.50E47 - < 4 52E45 - < 3.14E45 -
< 9.95E45 -
Cm.24 2 < 3.50E47 - < 1.30E48 - < 4.2SE44 - < 9.72E45 -
< 3.00E44 -
Cm-24 3/244 < 7.10E47 - < 7 93E47 - 1.99E43 2 pee 44 < 2.42E44 - < 4.34E44 - U.234 <. 9.24E4e - < l .83E47 - 1.00E 45 9.5 t E 47 9 00E47 1.57E47 t.34E4e 2.85E47 U-235 < 5.34Eco - < t .05E47 - 9.24E47 1.58E47 < 9.65E48 - < l.eGE47 - U-23e < t .07E47 - < 6.07E48 - 7.22E4e 6.45E47 4.40E47 9 93E40 9.18E47 2.14E47 Co.60 2.05E43 2.53E44 8.44E46 2.32E46 9 90E42 2.16E 44 1.23E41 1.72E44 1.14E42 1.02E43 Ru.106 < 4.t?E44 - < 2.72E46 - < 5.37E44 - < 5.19E44 - < 5.03E44 - As-10em < 4.0SE45 - < 3 39E46 - < 4.96E45 - < 4.75E45 - < 6 03E 45 - 6b.125 < 1.85E44 - < l.50E45 - 4.70E 45 1.51 E45 < 1.21 E44 - < 1 OeE44 - Co-134 < 2.20E45 - < 2.34E46 - < e.49E45 - '< 0.4eE45 - < 8.00E45 - Co-137 7.37E42 9.91E43 3 79E43 5.01E44 5 99E43 7.99E45 l.94E43 5.5sE45 4.29E42 4.59E43 Co-l44 < t .97E44 - < 2.40E45 - < 2.21E44 - < 1.9tE44 - < 2.5 t E44 - Eu.15 2 < 2.09E44 - < 1.59E45 - 3.94E43 8.37E44 2.20E43 6.4SE44 4.73E44 3.82E44 Eu.154 < 4.09E45 - < 6.10E48 - 7.82E43 1.72E44 1.00E43 1.17E44 3.21E44 1.14E44 . Eu.165 < S.54E45 - < t 05E45 - 206E43 1.20E44 2.4eE44 8.40E45 < 6.06E45 - Am441 < t.10E44 - < l.34E45 - S.72E a* 3.04E44 1.99E42 4.18E44, 1.00E42 1.04E43 __.__m.___ _ _ _ . . _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ . _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ . _ . _ . _ _ _ . _ _ _ . _ _ _ _ _ _ _ _ _ _ _ _ _ _ . _ _ _ _ . _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ . _ _ _ _ _ _ _ _ . _ . _ _ _ _ _ _ _ _ . _ _ _ _ _ _ _ _ _ _ _ . _ _ _ _ _ _ _ _
m..__m.. -- _ ._ _ - - - _ - _ - - _ - - _ - - - . _ . - - - _ - - - - - _ - - _ - _ _ - _ - _ - - - _ - - - _ _ - - - _ - - - - - - - - - - - - _ - - - - - - _ - - - - - _ - - - . _ . - - - - _ - - - - - _ _ - - _ - - - - . _ - - - _ - - - _ - _ - _ - - - - - - - _ . - - - _ - - - - _ _ _ - _ - - - _ - - - - _ - _ - _ . - _ . - - _ . en k N 5enton Resuhe C - y 9607002 (Revised 09/29/966 NELIO # 95G7002 48 9607002 47 9507002 48 9507002 49 ACTMTV 2SUNC ACTMTY 2SUNC 9507002-10 ACTMTV 23UNC ACTMTV 2S UNC ACTMTV 2SUNC tuCdeDemewed _ lucue memoved lucu sMem*wedl* H.3 < 5.57E44 < tuCue nomewee suCue namoved 4.14E 42 -
< 5.29E42 - < 5.05E42 <
C.14 < e.51E45 < t .51E43 - 9.3 t E43 -
< e.stE43 < 0.39E43 < 3.27E44 Fe 55 4.90E43 8.etE44 281E43- 4.4 t E44 <
1.SSE 44 - 1.94E42 3.24E43 1.50E43 2.45E44 4 59 1.2 t E42 SJeE43 7.24E42 1.09E 42 < t .09E43 - < 7.99E44 - 1.93E42 2 83E43
, 96-6 3 8,25E41 1.30E45 3.59E+00 5.25E 41 1.30E42 1.90E43 3.82E42 5.41E43 1.33E + 00 1.95E41 se-99 < 2.84E4e - < 2.13E 44 -
< 1.7?E44 <
5e-90 l.70E4 4 - < l.40E45 - 1.00E43 2.70E45 4.18E43 4.23E44 4.50E44 8.47E45 1.83E43 2.29E44 E94 < t.47E45 < 3.40E42 3.39E44 e.44E44 -
< e.41E48 -
2.7eE43 2.84E44 < e eeE44 - To99 < 0.07E45 -
< 4.eeE43 < 3.05E43 < e.22E43 1129 <
- - - < 2.30E44 -
t .74E45 - < 7.52E44 - < l .72E43 < 4.26E43 < 3.4eE45 - Np-237 < 1.52E43 < 3 9tE45 < 5.09E45 < t .00E44 -
< 5.29E43 -
h23e 1.47E43 3.18E44 7.94E44 5.eeE45 1.4eE43 1.00E44 4.87E43 2.47E44 < 1.83E43 - N239/240 3.22E43 4.7eE44 1.81E43 1.00E44 3.59E43 2.20E44 S.95E43 5.09E44 < 1.75E43 N241 0.37E42 7,84E43 3.14E42 3.12E42 0.54E42 0.54E43 1.89E41 1.eeE42 < 9.22E42 - 4 242 < 7.39E45 - < 2.3eE46 - < 4.74E4e - < 3.70E46 -
< 2.SSE44 -
Am-241 3.43E43 3.17E44 7.11E44 S.44E45 2.06E43 2.27E44 S.40E43 3 82E44 5.43E43 2.09E44 Am 243 < 2.43E45 - < 9 52E4e - < t.20E45 - 1.23E44 2.89E45 < 1.37E44 - Cm-242 < t .2 t E 44 - < 3. tee 45 - < e.40E45 - 1.23E44 2.09E45 < 3.00E44 - Cm 243/244 < 1.73E44 - < 3.93E45 -
< 7.3eE45 -
2.7eE44 4.49E45 < 7.79E44 - U-234 6.35E4e 3.7eE4e < 5.43E4e - < e.oeE46 - < 9.11E4e - < t. elf de - U-235 < 3.06E4e - < 1.18E46 - < 3.10E4e - < 5.27E46 - < e 23E47 - tA236 < 5. lee 40 -
< 2.58E46 - < 4.4eE4e - < 3.40E46 - < 0.79E47 -
Co-se 0.53E43 7 90E44 2.10E41 8 95E43 4.25E44 4.70E45 2.49E42 2.50E43 3.30E41 2.07E42 Ru.100 < t .50E44 - < 5.40E42 - < t.33E44 - < 5.43E44 - < 5.91E43 - As.100m < 1.10E45 - < 5.44E43 - < 1.32E45 - 2.23E43 2.87E44 < 5.04E44 - St.125, < 3.2eE45 - < 2.74E42 - < 5.00E45 - 1.9tE44 1.5eE44 < 2.27E43 - Co 134 < 2.33E45 - e.73E 43 2.00E43 < 1.29E 45 - 0.03E 45 3.4eE45 < 8 93E44 - Co-137 1.13E44 2.41E45 1.77E + 03 3.05E + 00 1.00E42 1.14E43 2.17E41 2.47E42 e.74E45 e.SeE42 Co-144 < 4.7eE45 - < 3.12E42 - < e.seE45 - < 3.SeE44 - < 4.10E43 - Eu 152 9.99E44 2.92E45 < t.95E42 - < t.00E44 - 7.24E44 3.3 t E44 < 5.99E43 - Eu 154 1.02E44 4.?5E45 <- 3 9eE43 - < 3 47E45 - e 67E44 1.31E44 < e.05E44 - Eu.155 2.19E45 1.81E45 < 1.30E42 - < 4.00E45 - 2.2 t E44 1.17E44 < 1.20E43 -
- Am.24 9 2.59E 43 2.76E 44 < 2.10E42 - 1.02E44 4.00E45 8.24E43 7.59E44 1.05E42 2.47E43
fr "'] Senten Resulte *- _ry 9507002 (Revised 09/29/953 NEL ID 07002 -11
/ li / '
9507002-12 07002-93 9507002-14 ACTMTV 2SUNC ACTMTY 9607002 15 2S UNC ACTIVITV 2SUNC ACTMTV 2SUNC tuCV96 ACTIVITY 2SUNC H-3 luCWel luCve l quCue l luCVol
~
< 9.93E46 - < l .17E44 -< 7.0 t E46 C.14
- < 0.4eE45 - < e 05E44
< 7.82E46 - < 0 70E46 -
< 1.0eE44 < 2.25E44 i
Fe 65 1.t eE44
- < 2.12E44 -
2.82E46 1.50E44 2.64 E 46 7.4 t E46 1.20E46 2 66E44 5.10E45 i M 69 1.27E44 2.90E45 9.24E46 1.90E46 1.19E44 1.08E46 1.11E44 1.75E45 e14E46 1.30E45 < N 63 0.66E43 2.4eE45 - 1.2SE43 8.84E43 1.29E43 6.61E43 0.22E44 8.31E43 . 9.24E 04 Se-89 < 3.43E46 5.50E46 e.07E48 - < l .30E46
- < 6.41 E4e < t.80E 05 <
Se-90 2.SeE46 - 4.4GE44 3.SeE46 e.43E46 1.59E46 6.22E46 860E46 3 94E46 e.4eE46 < t.64E45 - hS4 < t.49E46 - < 2 36E46 -
< 3.11E46 < t.49E 06 To-99
< 9.1 t E46 - *
< t .93E46 - < 3.90E44 - < 5.5eE44 - < 1.30E43 - < e.10E45 -
1129 < 3.87E46 - < S.00E46 < 2.19E46 < t .53E45
< 3.79E46 -
NP-237 < t.47E47 - < 6.26E47 - < t.07E47 - < 6 91E47 -
< t .40E46 - !
N238 2.00E46 1.12E46 9 34E46 e.6 2E47 0.00E47 1.23E47 1.25E46 S.30E47 < e.84E47 - Pu.239/240 6.06E46 2.e4E46 2.64E46 1.60E46 2.13E46 2.10E47 2.90E45 1.6eE46 < 6.64E47 - N241 1.61E46 9.72E46 3.10E44 3.18E46 3.11E46 S.22E46 3.67E44 3.86E45 < 6.18E46 - Pu.24 2 < 9.31E49 - < 2.64E4e - .< 1.75E4e - < 2.7eE4e - < t.18F47 - Am241 3.19E46 1.76E48 3.03E46 6.3eE46 2.22E46 1.63E46 1.57E45 1.40E46 3.07E46 7.04E47 Am.243 < 2.2SE4e - < 1.02E46 - < 6.93E48 - < 9 90E40 - < t.79E 07 - l C=242 < 1.91E47 - < 3.43E46 -
< 2.69E47 - < 4.29E47 - < 6.11E47 -
C=243/244 7.92E47 8.30E47 < t.7eE46 - < 4.67E47 - < 6.86E47 - < 3.SeE47 - U 234 < 1 e8E47 - < S.SSE4e - < 8 0eE47 - < t .03E46 -
- 1. tee 46 2.02E46 u-236 <
< 9.40E4e - 0.66E4e - < 2.e9E47 - < 6.05E 47 -
< 1.12E46 -
U.2 30 < t.79E47 - < 8 66E4e - < 4.56E47 - < 9.seE47 - 9.49E46 1.7eE46 Co-60 2.02E43 1.79E44 2.14E43 1.94E44 9.86E43 4.74E.06 2.60E43 3.46E46 3.94E45 9.69E46 Ru.106 < 6.42E44 - < 8.7eE44 - < 1.33E44 - < t .83E44 - < 7.76E46 - As-100m < 6.34E46 - < - S.70E46 - < t .32E46 - < t .42E45 - < I .20E46 - 1 St-126 < 2.SSE44 - < 3.64E44 - < 6.30E46 - < 7.89E45 -
< 2.30E45 -
Co-134 4.70E46 1.30E46 < 2.65E46 -
< 0.05E46 - < t .64E45 - < 9.00E46 -
Co-137 2.90E41 2.90E42 1.74E41 1.73E42 4.70E42 4.3eE43 2.96E42 0.2SE46 3.95E44 1.51E45 Co-144 < 4.SlE44 - < 0.06E44 - < 7.2SE46 - < 9.74E45 - < 4.20E46 - Eu 152 < t.30E44 - < 2.20E44 - < 7.84E46 - < 1.39E44 - < e.10E46 -
- Eu-164 < 2.96E46 - < 4.73E46 - < t .42E46 - < 2.66E45 -- < 2.84E46 -
Eu-155 * < l.11E 44 - < t .82E44 - < 3.22E46 - < 44eE45 - < 2 05E46 - Am-241 < 2.0SE 44 - < 2.79E44 - < 4 94E46 - < 7.50E45 - < 3.75E45 - l
R
- p. e s
( \ v~ Seston Resulte Summery 9507002 tRevised 09/29/958
/ r NELID. 9507002 -leas 9507002 -t 7As 9507002 -leas j9507002 leAC y /
ACTMTV 2SUNC ACTMTV 2S UNC 9507002 -17AC 9507002 -leAC AC1MTV 2S UNC ACTMTV 2S UNC ACTMTV 2SUNC luCWmLI tuCWmL) ACTIVITY 28UNC tuCWmLl tuCWgt fuCUgl H.3 5.43E44 1.58E44 < l .01E44 < luCuel - l .01E44 - 6.e7E44 1.80E44 < l .00E44 < C-14 < 1.33E45 - < 1.4 2E45 -
< 5.10E46 <
9.33E45 - Fe55 < 3.5eE47 1.02E45 -
< 9 82E4e - < t.83E45
- < 2 07E46 -
< 7.4 7E47 2.57E44 Ni-59 < 3.84E47 <
1.04E44 < 4 92E46 - < 2.2tE45 - 2.13E46 - < l.2 t E46 - 9.77E 45 1.e3E45 < 1.27E45 Ni.e3 6.SeE4e 1.0SE46
- < 9.02E4e -
1.79E45 2.7eE46 8.83E45 1.30E45 7.58E43 1.llE43 e.49E45 1.33E45 Se.09 < l .07E4e < 2.40E46 < 5.84E45 7.90E4e 0.49E47 - < 4.4eE4e - < 4.37E4e < 4.12E4e Sr.90 3.20E45 3.38E4e < 1.75E4e 1.00E4e 1.82E47 5.83E47 2 52E4e < 3.19E46 - < 3.00E4e Nb-94 < 3.13E47 -
< t.00E47 < t.35E47 <
To.99 < 9.95E46 - < .* 04E47 - < 1.01E47 7.24E46 -
< 7.SeE46 - < 4.32E46 - < t .14E45 < 2.53E46 I-129 <
- - < 3.3eE46 -
1.17E4e - < 5.24E4e - < 1.04E48 < 1.37E46 < 3.02E46
- - - < t.97E4e -
Np-237 < 2.70E4e -
< 2.25E47 < 3.05E48 <
4.55E47 - < 1.85E47 - < I .e4E47 - N230 < 0.27E49 -
< 9.00E4e -
< e.39E49 -
5.32E45 2.seE46 < 4.30E4e - 5.5eE4e 3.52E47 A239/240 < t.74E4e -
< 7.05E4e -
< 9.04E49 - 1.22E44 S.3eE4e < 4.70E4e - < 9.etE4e - '
Pu-241 < 7.74E47 - < e.4 7E46 - < l.17E46 - 1.e5E43 . t .61 E44 < 3.esE4e - < 4 5eE4e - Pu.24 2 <- 2.Jt E49 - < 2.03E4e - < 9.04E48 - < t .95E48 - < 0.39E49 - < t .09E4e - Am.241 . 3.75E4e 0.00E49 < e.59E4e - < t.etE46 - 2.85E44 1.33E45 < 4.etE4e - < e.85E4e - Am-243 < l.34E49 - < e.99E49 -
< 2.79E49 - < 1.97E 4e - < 7.32E49 - < l .2t E4e -
Cm 242 < 2.59E49 - < 3.12E46 - < t.03E4e - 1.45E46 1.72E47 < 3.20E4e - < 2.70E4e - Cm.243/244 < l .02E 4e -
< 3.44E4e - < l.50E4e -
4.44E4e 3.52E-0 7 < 2.74E4e - - < 3.e4E4e - U-234 < t .25E47 - < 4.13E46 - < 2.40E4e - < 5.05E47 - < 5.84E47 - < 4.50E47 - U 235 < 9.10E4e - < 2.07E46 -
< 2.90E4e - < t .e5E47 - < 2 02E47 - < 2.01E47 -
U 238 < 4.30E4e - < 2.07E46 - < t.10E4e - < 2 75E47 - < 4.39E47 - < 4.15E47 - Co-80 < 4etE47 - < t.95E47 - < l.75 E47 - 3.21E43 2 82E44 1.07E46 35eE47 1.23E4e 1.09E47 h-100 < 2.43E45 - < 1.SeE46 - < t.23E4e -
< 2.7t E44 - < l.95E4e -
< 9.94E47 -
Ag-100m < 2 55E4e - < 1.03E47 - < 1.e t E.07 - < 2.75E45 - < 1.97E47 - < t.0eE47 -
$4-125 < 1.20E45 - < 4.50E47 - < 3.08E 47 - < l .18E44 - < 5.58E47 - < 3.3eE47 -
Co-134 3.42E 46 5.50E 47 < 2. ole 47 - < t.48E47 - 5.eeE45 1.13E45 < 2.20E 47 - < 1.12E47 - Co-137 1.00E42 S.4eE44 < 2.05E47 - < t.4 t E47 - 1.04E41 1.79E42 2.tlE45 2.02E4e 3.20E45 3.44E46 Co 144 < t .82E45 - < 9.5 3 E47 - < 0.13E47 - < 1.82E44 - < 5.etE47 - < 5.53E47 - Eu-152 < 3.07E46 - < 1.49E46 - < l.15 E46 - < 9.30E45 - < t.76E46 ,- < e.e8E47 - Eu.154 < 1.00E4e - < 5.70E47 - < 4.52E47 - < t .88E 45 - < e 85E47 - < 3.20E47 - Eurl55 < 7.29E4e - < 4.56E47 - < 2.e4E47 -
< t .2 t E44 - < 2.50E47 - < 2.03E47 -
Am.241 < 1.03E45 - < t .54E47 - < 5.94E47 - 1.87t44 8.54E45 < e.77E47 ' - < t.40E47 -
i
- l 4 i j .
j B&W NUCLEAR ENVIRONMENTAL SERVICES, INCORPORATED NUCLEAR ENVIRONMENTAL LABORATORIES 4 } J } DATA REPORT NUMBER 9507002S 4 SUPPLEMENTAL INFORMATION ! l e 1 l 4 1
- Submitted to f GPU Nuclear Corporation 3 Sar_ ton Site - Decommissioning Project i
1 1 Januaqr 10,1996 1 j i. 1 i i l I 1 } l Released by: dfEJ L . ///oNG James L. Clark, Project Manage'r Nuclear Environmental Laboratories
I i DATA REPORT
SUMMARY
l o) -
\
Laboratory Project Summary? ! e Plant Site: Saxton Decommissioning Project i e Shipnient Number: 95-003 l e Report Date: September 14,1995 ( i i e NEL COC Number: SN-009274,009275,009276 I e NEL SDG Number: 9507002- _ ! e NEL Proposal: NEL5-1258 i i I e Contract Number: 0397355 i e Analysis Service: Routine ! l e Laboratory QC: Routine - Batch l 1 l t i l l : l O i
9507002S - Supplemental January 10,1996
- sSAmple Delivery Group' Summary) -
Custo'mer?ID' LNEL Sa'mple'IDE eSample Type I ; Description' SX815950010 9507002-01-AA Pipe Section item 12 SX825950011 9507002-02-AA Pipe Section Demin Water SX825950040 9507002-03-AA Pipe Section item 6 SX825950041 9507002-04-AA Pipe Section item 3 SX825950042 9507002-05-AA Pipe Section item 7 SX825950043 9507002-06-AA Pipe Section item 2 SX825950044 9507002-07-AA Pip'e Section item 4 SX825950045 9507002-08-AA Pipe Section Test Cut SX825950046 9507002-09 AA Pipe Section item 9 SX825950047 9507002-10-AA Pipe Section item 13 1 l Pipe Sections / Physical Characteristics Summary NEL Sample ID Le'ngth (in) OD (in) ID (in) ; 9507002-01-AA 12 5/8 2.422 2.222 1 9507002-02-AA 12 1/4 1.910 1.793 9507002-03-AA 12 1.325 1.039 9507002-04-AA 5 1/2 1.321 1.055 9507002-05-AA 13 2.380 2.030 9507002-06-AA 13 1/2 3.480 2.963 9507002-07-AA 12 1/4 4.560 4.313 9507002-08-AA 13 3.560 3.322 9507002-09-AA - 14 1/2 3.500 3.373 l 9507002-10-AA 9 7/8 2.420 2.260 il
. . - ..- ..- --- -____--... -~...._ - --___--___ .
l 9507002S - Supplemental January 10,1996
~
~
l' Sample Aliquot Summary 6 i
- l. NEL'i@ ' ...' Matrix ! Leach Volume ~ (mL)' f Pihe ~ Mass /R' e m'oved (g) i 9507002-01-AA Acid Leach 3755 2068 /163 j 9507002-02-AA Acid Leach 513 1047 /10 l 9507002-03-AA Acid Leach 258 1338 /10 9507002-04-AA Acid Leach 220 654 /10 i 9507002-05-AA Acid Leach 648 3737 /11 9507002-06-AA Acid Leach 1039 6880 / 53 9507002-07-AA Acid Leach 2919 4964 / 2 9507002-08-AA Acid Leach 1858 3600 / O 9507002-09-AA Acid Leach 1988 2418 / 1 9507002-10-AA Acid Leach 3676 1416 / 69 o -
iii
B & W RESULTS 9/14/95 - Sexton Samples . M SYSTEM = = = > Ve*ns & Drains - Insumel Surfince ofpiping between CVsump pumps & CVwan(1.5*to 2" WL152R). Samfon il B& 4f I 4 1.89941 SammAe Numbert Samode humber I Total TRU Content (nCi/g) l j SX815950010 W 9507002-01-AA I i 0.00031 Total Uranium Content (nci/g) i Sample Type il Pipe Secten l l 1 Description ,1 Item 12 i Results Normalized l Nuclide Notes uCi/a to 100 % Nuclides RATIOS
- 11 H-3 i LESS THAN l 6.54E-04 i 0.2728% //####/////////#//###//#/Au\%%%%%%\%\h\\1 21 C-14 ! ' WSS THAN i 1.59E-04 0.0663% P////#/#/##/####/#/#//#A\%%%%%%%%%%%1 l 31 Fe-55 I f 3.41E-04 0.1423W Fe-56/Co-80 0.128681 4i Ni-59 i j 4.95E-03 2.0651 %:r/####/#/#/#/#/#####/ amu %%%%%%%%%\1 5P ki-os !
l .1.47E-01 61.3274 %PN##/##//#/#/#/#####/A\\\\%%%%%\u\%%\1 6h St-89 I WSS THAN ! 3.36E-04 0.1402 %p/#/#//#######/#/##/#A%%%i%%%%%%%W 7! St-90 ! 6.70E-03 2.7952%l St-90/Cs-137 0.00091 Si Nb-94 I WSS THAN 2.04E-05 0.0085 %!f#/#####///#/##/##/////An\\n\n\\u\wi\u\W 91 Tc-99 - WSS THAN 7.28E-05 0.0304 %U//#/##/##///##/#/#/#/NA\%%%%%%\n%%%\1 106 l-f29 ! WSS THAN 3.62E-05 0.0151 % F//#!##/////#/###///###/ Ann \%%%%\%%\%%j til No-237 ! WSS THAN 3.99E-07 ! 0.0002% l f/#/##//#/#//##//####//A\%%%\\\%\u\\%%%1 121 Pu-238 li I 4.26E-05 l 0.0178 % Pu-238/Am-241 1.59551 l l 13: Pu-239/240 il 8.85E-051 0.0369 % f/#//####/#/#/##/#/#///A%\%%%%%%%%%%) l 14i Pu-24 f 6 1.74E-03 i 0.7259 % Pu-241/Am-241 65.16854 l 15 Pu-242 i WSS THAN 2.67E-06 0.0000 % fl#////#/#//#/##//#//#/#/A%wh\wunnukuw)
\ 16i Am-24 7 2.67E-05 0.0111 % Am-241/Pu-230
(") 0.626761 i
- 17l Am-248 i WSS THAN 1.23E-07 0.0001 % f#//#/#/////#/#///#//##/#A\%%\\\wn%%\\\wn)
! 18) Cm-242 i WSS THAN 3.50E-07 0.0001 % f###//########///////#Auw\wwwwhuguw) l 195 Cm-243/244 i WSS THAN 7.10E-07 0.0003 % f///#####/#/##/#/////////Au\WWWWWu\u\uut 20W U-234 WSS THAN 9.24E-06 0.0000 % fn#///###/##/#/#//#/##Anu\%%%%\%%%%\1 l 21i U-235 WSS THAN I 5.34E-06 0.0000 % f////#/##//////#/##//#///#Au\\wu%%%%%%\nf I 221 U-238 i GSS THAN 1.07E-07 0.0000 % U-238/Am-241 i c.004011 23! Co-60 l 2.65E-03 1.1056 % Am-241/U-238 i 249.53271i 24i Ru- 106 ! WSS THAN 4.17E-04 0.1740 % f//##//#///#/#/#//##//////Au%%%%%%%%%%\1 25i Aa- f 00m i WSS THAN 4.06E-05 0.0170 % f///#///#///#//#///////##/#A%%\Wu\\%%%\%%\1 l 26! Sb- f 25 i WSS THAN 1.65E-04 0.0688 % f//#/#/##/#/#/////##//////Au\%%%%%%%%BW 27: Cs- f 34 WSS THAN 2.26E-05 0.0095 % Cs-134/Cs-137 1 0.00031 28I Cs-f37 7.37E-02 30.7472 % Co-80/Cs-137 i 0.03596I i 291 Ce- 144 WSS THAN 1.97E-04 0.0622 % Co-137/Co-60 27.81132 l 30; Eu- f 52 WSS THAN 2.09E-04 0.0672 % f//#///#/#//#////////###///Au%%%%%%%%%\W 31; Eu- f 54 WSS THAN 4.09E-051 0.0171 % M%%%%%%%%%%%\WW%%%%%%%%%%%%%%\1 32l Eu- f 55 WSS THAN 8.54E-051 0.0356 % '//#/##/#/###/##/###//A%\%%%%%%%%%%1 li Totais= = > il 2.40E-011 100.00m I t l l f O
O B & W RESULTS 9/14/95 - Saxton Samples -
]
\ SYSTEM = = =i> Dominerahzed Water System i Sarton BAW I l 0.9642H Samo4r Number Samsde Number Total.TRU Content (nCilg) SX825950011 W 9507002-02-AA ll - 0.00031 > Total Uranium Content -(nCi/g) i i Sample Type ll Pipe Section 11 i i Descnotion 4 Domin Water Pipe ll l Results Normalized Nuclide Notes uCi/a ,to 100 % Nuclides RAT 108 ili H-3 i LESS THAN 1.12E-03 ! 17.2119% F//////#//#////#/##////##A%%%\%%\%%%%%\1 2i C-14 i WSS THAN 1.58E-04 i 2.4281%9//#/###//#/#/##//### Ambu %%%%%%%%1 , ; 3i Fe-55 I WSS THAN 8.842-06' O.1359% Fe 55/Co-60 1.04739! l 4 ii Ni-59 W ESS THAN 9.59E-06 0.1474% '###/##/#/#///#/#//#//A%%%%%%%%%%%\1 l Si Ni-63 e 2.19E-04 3.3655% P/##/###/##//#/#/#/#/Aw%%%%%%%%%m1 6 l- St-89 W WSS THAN 4.62E-061 0.0710 % "#/#/####/#/#/#/#//#/Au\%%M\\%%\u\u\1 l 70 Sr-90 i 2.81 E-051 0.4318% Sr-90/Cs-137 0.00741, l 8!! Nb-94 i GSS THAN 2.06E-06 0.0320 % '//##/###//#/###/##//Au%%%%%%%%%%) 9 11 Tc-99 it ESS THAN 6.79E-05 1.0435 % '/#//#/#//#/#//#///#///#Am%%%%%%%%%1 1011 /-f29 li WSS THAN 2.17E-05 0.3335 %F/######//#//##/#/##AMW%%%%%%%%W 1111 ha-237 4 WSS THAN 4.26E-07 ! 0.0065 %F/##//##//N#//####/#/Ahm%%%%%%%%W 121! Pu-236 1 1.83E-05 ! 0.2812%I Pu-238/Am-241 0.49062 13:1 Pu-239/240 4.18E-051 0.6424 %P#////#/###//###//#/#/Auwuwmuhuwww 14ll Pu-24 f 8.64E-04 ! 13 2777 %I Pu-241/Am-241 1 23.16354 i p 15 Pu-242 ! ESS THAN 3.74E-08 i 0.0006% '////###///##//#/#/###A%m%%%%%%mW l / 16i Am-24f i' 3.73E-05 0.5732% Am-241/Pu-2381 2.03825 ' ( 17}___Am - 243 18I Cm-242 ESS THAN WSS THAN 2.50E-07 1.30E-06 0.0038 % '//#/##//##//###/#/#//Aum%%%\w\%%%W 0.0200 % '/#/////#////###/#//#//#Am%%%%%%%%%Wi { i 19i Cm-243/244 H ESS THAN 7.93E-07 0.0122%J//##//##//#//////##///#Au\hWWWuw\%%%\1 20 q,_ U-234 I GSS THAN 1.63E-07 - 0.0025 %P///##/##////#/#//#/##/Au\W%%%%%%%%W 211 U-235 i GSS THAN 1.05E-07 - 0.0016 %F/##///#/#////##//#//##A%\wu\\\W\%%\um 22t U-238 t WSS THAN 6.87E-06 0.0011 %! U-23s/Am-241 0.00184 23: Co-60 l 8.44E-06 0.1297 %! Am-241/U-238 542.94032 241 Bu- fot ESS THAN 2.72E-05 0.4180 %F////#//#////#/#//#//##//Amunwnh\\\Wu\W 25- Aa- f oSm ESS THAN 3.38E-06 0.0519%F/#/#/#//#/#/#///////#///AWWWu\\Wuhum\\' 261 Sb-f25 ESS THAN 1.50E-05 0.2305 % '/#/#/##//#/#//#/////////AuguuhWunwum 27 Cs- f 34 WSS THAN 2.34E-06 0.0360 % Cs-134/Cs-137 0.00062 2811 Cs- f 37 3.79E-03 58.2438 %! Co-60/Co-137 0.00223 l 29ll Ce-144 i GSS THAN 2.40E-05 0.3688 %I Cs-137/Co-60 l 449.05213l 301: Eu-f52 i WSS THAN 1.58E-05 0.2428%F/##//#####/##///#//#ABum%%%%%%%%t 31 ll Eu- f 54 1 ESS THAN 6.10E-06 0.0937%F##///##//##//////#/##/Auw%%%%%%%\%%) 32i Eu-f55 i GSS THAN 1.05E-05 0.1614 %F/###/##////#//##/#/#/Amuuhunw\wwww l Totals = = > W 6.51E-03!I 100.00% W i (Aw) I l 1 l l i
!h O B & W RESULTS 9/14/95 - Sexton Samples . . SYSTEM = = = > Purncebon - Intemal Surtface ofpoing between M.C. pipeg & inlet to Pur. Domin. (fPU250f R). f' .I Samfon ! 8AW isemple Number Samole Number i l ll 1349.0358 W l Total TRU Content (nCilg) LSX825950040 1 9507002 AA i 11 0.018'Til l Total Uranium Content (nCi/g)
? Sample Type 4 Pipe Secten il
.! Description j item 6 il 1
Results Normalized Nuclide Notes uCi/c to 100 % Nuclides RATIOS l 11 H-3 ! LESS THAN 7.33E-04 i 0.0470% if ////////////////////////////////A%W\\w\W\\\%\\\\\\\f l 21__C s. 14 l LESS THAN 2.23E-041 0.0143% f////////////////////////////////A%%\%\w\wh\W\\%\1 l 3Il Fe-55 I 5.37E-031 0.3442% I Fe-55/Co-80 0.05480i 44 Ni-59 i 6.68E-03 0.4281% /////////////////////////////////AWW\%%%\%\W\W\\\i Sil Ni-63 4.88E-02 3.1276% /////////////////////////////////AWW\\n\n%%%\mni 6 11 St-69 LESS THAN 3.10E-06 0.0002% f////////////////////////////////A%%%%%%%%%%%1 7il St-90 l 2.99E-02 1.9163%l Sr-90/Cs-137 4.99165 l l 8 11 Nb-94 LESS THAN I 8.75E-051 0.0056% /////////////////////////////////A\\w\\wh\%\%%%\\\1 96 Tc-99 LESS THAN I 5.88E-051 0.0038% lf////////////////////////////////(%%%%\%%%%%%W 10i /-129 ' LESS THAN I 4.05E-05 0.0026% y////////////////////////////////Am%%%%%%%%%W , ill No-237 ! LESS THAN I 1.89E-03 . i 0.1211% f////////////////////////////////Au%%%%%%WWWW l 121 Pu-236 il i 2.73E-021 1.7497% l Pu-238/Am-241 l 0.32773l 13! Pu-23ar240 l i 7.40E-02I 4.7427% f////////////////////////////////(%%\wm%%\w\\\my 14t Pu-24f 5 l 1.16E+001 74.3447% i Pu-241/Am-241 13.92557! l A 1 511 Pu-242 ! LESS THAN I 8.26E-05! 0.0053% f////////////////////////////////A%%%m%m%mww I l 3 16i Am-24 f I I 8.33E-02I 5.3387%ll Am-241/Pu-238 3.05128 C/ 17: Am-243 ! LESS THAN ! 4.52E-05 0.0029% lf////////////////////////////////A%%%%%%%%%%m1 i l 18j Cm-242 i LESS THAN I 4.28E-04 0.0274% /////////////////////////////////A\%%%%%%%%%%f l 19 !- Cm-244/244 i i 1.99E-03 ' O.1275% 1////////////////////////////////Au%%%%%%%%%%\f 20; U-234 I i 1.00E-05 0.0006% /////////////////////////////////AMuunuu%%%%%f l 21: U-235 I 9.24E-07 0.0001 % !////////////////////////////////A%\%%mWuuumf 22t U-238 i t 7.22E-06 0.0005% U-238/Am-241 0.00009i l 23r Co-60 i i 9.80E-02 6.2808 % Am-241/U-236' 11537.39612i 24l Hu-106 i LESS THAN I 5.37E-04 0.0344 % f////////////////////////////////u%%%%%%%%%%f 25e As-f00m i LESS THAN I 4.96E-05! 0.0032% f////////////////////////////////A%%%%%%%%%%%V 26 Sb-f25 I i 4.70E-05l 0.0030 % '////////////////////////////////h%%%%%%%mWWWV 27, Cs-f34 ! LESS THAN l 8.49E-05i 0.0054% Cs-134/Cs-137 0.01417! ! 28: Cs-f37 i I 5.99E-031 0.3839% Co-80/Cs-137 16.36060l l 29i Co-f44 I LESS THAN I 2.21E-04i 0.0142 % Cs-137/Co-Go 0.06112! 30t Eu-f52 l ! 3.94E-03 0.2525 % '////////////////////////////////un%%%%%%%%%%i 31 IE'u- f 54 l ! 7.62E-03 0.4884 % '////////////////////////////////n%\%%%%%%%%%f 32! Eu-f55 l ! 2.86E-03 l 0.1833% f////////////////////////////////un%%%%%%%4%%f l Totals = = > ll 1.56E+00_JI 100 00%{ i l i n ( v)
(N ) i B & W RESULTS 9/14/95 - Saxton Samples , , l Ll SYSTEM = = = > Pressure Control & Rehet - /ntemal Surface of piping between M.C. piping Tokt Leg'& inlet to Pressunzer Spray Une (f MC250f R). Sauron 4 84 W l ll 212.8119!!
. Sampde NumberLSample Number ! Total TRU Content (nCi/g)
*SX825950041 4 9507002-04-AA l ll 0.0015 ll Total Uranium Content (nCl/g) 1_ Sample Type i Pipe Section W i Descnotson i item 3 ll Results Normalized Nuclide Notes uC4/a to 100 % Nuclides RATIOS 1 I ~H-3 t ESS THAN 6 6.24E-04 0.0890% f//#/#/#//#//#####/##A\%%%%%%%%\WW\'
2 FC- f 4 i ESS THAN ! 1.68E-04 0.0240 %?/#/#/#######/##//#/A%%%%%%%%%%\W 34 Fe-55 i i 4.55E-03 0.6489 %i Fe-56/Co-80 i 0.03699 44 Ni-59 I ! 5.30E-03 0.75 S%F/#//##////#/###/#### Ann \h\\\\\WW\www Sil Ni-63 I I 3.34E-01 47.6306 %//#/#/#/#/#/##/####//A\%%%%\\W\%%%%\' 6d St-89 ESS THAN i 3.01 E-06 0.0004% f/####/#/#//###//#//#A\%%%%%%%%%BW 71 Sr-90 l 1.41E-02 2.0100 %! St-90/Cs-137 7.26804l 81 Nb-94 LESS THAN i 2.46E-04 0.0351 %P/##/###/#/////#/##/#hnWB%%%%%%%%W,
- 91. Tc-99 ESS THAN l 5.87E-05 0.0064 %p#/###/#//####///##/hn%%%%%%%%%%W ,
10I l-129 ESS THAN i 1.53E-05 0.0022 %3/#/#####///#//###//#Auu%%\%%%%%%W= i 11' Np-237 ! ESS THAN l 1.78E-03 0.2538 %F#//####/####/#/##//A%%%\whuunuwn) 121 Pu-238 8 4.24E-03 0.6047 %I Pu-236/Am-241 0.51707 13; Pu-239/240 I 9.14E-03 1.3034 %P/#/#//####/##/#/#/#hn%%%%%%%%%%W , 14il Pu-247 ! 1.89E-01 26.9527 %! Pu-241/Am-241 23.04878 /O I 154 Pu-242 W- ESS THAN 8.13E-05 0.0116 %p/###/#//#//#//##/#//#AhuhWW\%%%%\\W ( 164 Am-24f li 8.20E-03 1.1694 %) Am-241/Pu-238 1.93396l 1711 Am-243 il ESS THAN 3.14E-05 0.0045 % ?##/#/#/#/##/#//##/#hn%%%\BhW\B%huj 18i Cm-242 ._jI~ LESS THAN 9.72E-05 0.0139 %p//#/#/##////#/#/####hWhWW\%%%%%%\%1 19: Cm-243/244-~li LESS THAN 2.42E-04 0.0345 %J/#//#/##//##/#/#/#//#Au\WWW\wn%%%\%y 20; U-234 1 9 80E-07 0.0001 %J/######/#/#///////#//#nh\W%%%%%%%%\1 21 I U-235 i LESS THAN 9.65E-06 0.0000%F#///##/#/#//#/##////#hn%%%%\%%%%%%\1 22: U-238 1 4.40E-07 0.0001 %! U-236/Am-241 0.00005! 23' Co-do 1.23E-01 17.5406 %i Am-241/U-238 18636.36364! 24i Ru-f06 ESS THAN 5.19E-04 0.0740 %F/#////###///#//##/#/#/A%%%%%%%%%%\%1 25! Aa- foem LESS THAN 4.75E-05 0.0068 %F//#/#/#/#///#//#/////////A\\%%%%%%%\%%W , 261 Sb- 125 LESS THAN 1.21 E-04 0.0173 %P/#/////#///#/#////#/#/#/AWWWu\WW\%%%%% 271-~ 6- f 34 i LESS THAN 8.48E-05 0.0121 %! Ca-134/Cs-137 0.043711 26N Cs- 137 -I 1.94E-03 0.2767 %l Co-60/Cs-137 63.402061 29 ll Ce- f 44 LESS 1HAN 1.91 E-04 0.0272% Cs-137/Co-60 l 0.01577: 30a Eu-f52 2.20E-03 0.3137%J///////#//#/##//#/##/#/Au\WWWWMWWWWW) 311 Eu- f 54 1.00E-03 I 0.1426 %F/#//#///#//#/#/#/####AWh\%%%%%\%%%4 32'f~ Eu-f55 ! 2.48E-04 ! 0.0354 %F#####/#/#/#/#/#/##/Au\WWWWW\WWWWW i Totals = = > ll 7.01 E-01 ll 100.00%ll 1
\
i,
\
v)
! l I l l
/m\ B & W RESULTS 9/14/95 - Saxton Samples - \~j SYSTEM = = = > Shutdown Cooling - Intemel Surface of piping between the connection with the main coolant piping and the connecuan mtn the Safetyinjecten piping (2SC250fR).
j Sarton i BAW ! Li _1404 3375ll pmate Numbed Sanpie Number _i Total TRU Content (hcl /g) , l SX825950042 19507002-05-AA j h 0.0024il Total Uranium Con:ent (nCi/g) uSam_ple Type ! Pipe Section !I J Description # item 7 ll Results Normalized Nuclide Notes uCl/a to 100 % Nuclides RATIOS 1d H-3 LESS THAN I 1.67E-03 0.1033% F#l//#//##/////#/#/#///#/A\ww\\%\w\\\\\%\m4 2b C- f 4 r LESS THAN i 3.05E-04 0.0189% 9///#/#////####///####/Au\%%%%%%%%%%9 j 34 Fe-55 ! 2.55E-02 1.5780% Fe-55/Co-60 I 2.23684 dil Ni-59 i l 2.01 E-03 0 1244% 'l#/##/#////////#//////#/NAgu\m\Wmungu\t 5i Ni-63 , i 1.24E-01 7.6737% /#####//#////####/#///A%%%%%%\\%\%\mi 6! St-89 LESS THAN I 8.01E-06 0.0005% F###/###/###/####/#Aum%%%%%%%%%) 7l Sr-90 1 3.97E-04 0.0246 %l St-90/Cs- 137 1 0.00925. i Si Nb-94 l 2.97E-04 0.0184 %F#//##/#/#/######/////A%%%%%%%%%%mt l 9i Tc-99 LESS THAN r 1.22E-03 0.0755%F/#/#/##/#//#/#/##//##Amunnumnugu\f l 10! l-f29 LESS THAN l 3.89E-05 0.0024% f/#/##/#/#/#/##//////##Ak\%%%%%%%%%%) til No-237 LESS THAN l 3.93E-03 0.2432% 1/##/##//#/////##/////##An\W\W\wh\\w\w\%) 12j Pu-238 i 4.10E-02 2.5373%Il Pu-236/Am-241 11.61473 13[ Pu-23S/240 6.48E-02 4.0101 % '#//#/#####/#//#//#////n%%m\%%%%%%%) 14t Pu-24f 1.29E+ 00 79.8307% Pu-241/Am-241 365.439091 (, j 15ll Pu-242 LESS THAN 2.36E-04 0.0146 % '/#//###//#i///##//#/////nmuuhumn%%%1 , V 16i Am-24 f i 3.53E-03 0.2185 % Am-241/Pu-238 0.06610. 17! Am-243 I LESS THAN 9.95E-05 0.0062 %F////##//#////#////#/#/#//n\Mun%%%%%%%f 18g Cm-242 i LESS THAN 3.06E-04 0.0191%F#///##/#/##/#///#/##//n\\B\%\W\\%%\uMT 19i Cm-24d/244 l LESS THAN I 4.34E-04 0.0269%i'/#///###///#////#/#//#//n\%%\%%%%%%%%1 204 U-234 i i 1.34E-06 0.0001 % '#/####//##/#/#/////#// n%%%%%%%%%%%f 21i U-235 i LESS THAN } 1.68E-07 0.0000% '/#////#//###//##//#/#//n%%%%%%%%%%%T 221 U-236 9.18E-07 0.0001 % U-238/Am-241 0.00026II 23 Co-60 1.14E-02 0.7055 % Am-241/U-238 3845.31590) 24; Bu-fof LESS THAN 5.03E-04 0.0311 % //#/#/////##//////#/#////// n%%%%\m%%%%%f 25i As-108m LESS THAN 5.03E-05 0.0031 %l/#/#//////#/##/#//////////An%%%%%%%%\wnf 26; Sb- f 25 LESS THAN 1.08E-04 0.0067% '##/#///##///#///#/#///#A\%%%%%%%\WWmf 27if-~ Cs- f 34 I LESS THAN 6.00E-05 0.0037% Cs-184/Cs-137 0.00140l 28;l Cs-f37 4.29E-02 2.6546 % Co-60/Cs-137 0.265731 291 Ce-f44 LESS THAN 2.51 E-04 0.0155 % Cs-137/Co-60 3.76316l 304 Eu- f 52 4.73E-041 0.0293 % '//////#////////###//////////n%%%%%%%%%%%i 31.i Eu-f54 3.21 E-04 0.0199% '#//#//#///#///#/#####/n%%%%%%%%%%%Y 32i Eu-f55 LESS THAN 6.66E-05 0.0041 % '/////###/####///#/##//n%%%%\%%%%%mt I- Totals = = > il 1.62E+00W 100.00%i l i f) 1 \ l v
l O B & W RESULTS 9/14/95 - Saxton Samples - SYSTEM = = = > Pressure Control & Relet - IntemalScJace of piping between M.C. piping ' Hot Leg"& inlet to the Pressunzer (Surge Line), (3MC250fR). 1 Sarton ll 8& W il l 73.73221 LSampde Numbers Sampde Number 4 Total TRU ConteTn (nCi/g) I i SX825950043 3 9507002-08-AA 11 l 0.00021 Total Uranium Content (nci/g) J Sample Type H Pipe Section l i l Description li stem 2 ll l I Results Normalized - Nuclide - Notes uCi/c to 100 % Nuclides RATIOS 11 H-3 ! LESS THAN 5.57E-041 0.0601 % (#/#/#/#////#//#/////###AiWWW\\n\\%%%\Wh 2l C-f 4 LESS THAN i 8.51E-051 0.0092% l/##/#/#/##/##//#///##A%%%%\n%%%%%%\ 3! Fe-55 1 4.90E-03 0.5287% Fe-55/Co-80 l 0.574441 l All_ N/-59 i 1.21E-02 1.3056% '#/#/##//#/#/#######A%%%%%%%%%\%%V ' Sli Ni-83 ! 8.25E-01 89.0192% 'l#///#//#l#/####/#//##A%%%%%%%%%%\%V ! 6i Sr-89 i LESS THAN 2.84E-061 0.0003% f///////##/#/###//#/#/#/An%%%%%%%%\WMV 7 i Sr-90 1 1.06E-03 0.1144% 1 Sr-90/Cs-137 l 9.38053i Sf Nb-94 LESS THAN i 1.47E-05 0.001e% f/#/#/######/#//#/#/#A\%%%%%%%%%%%N l 9j Tc-se I LESS THAN I 6.67E-05 0.0072% f#/##/###/#//#/#///##/An%%%%%\%%\%%% l 101 /-129 i LESS THAN i 1.74E-05 0.0019% f/#/#/////#/#####/#//#/Au\\%%\\W%%%%%% 11! No-237 i. LESS THAN I 1.52E-03 0.1640% f/#####/#//##/#/#/##/An%%%%%%%%%%% l 12i Pu-238 6 i 1.47E-03 0.1586% I Pu-2Se/Am-241 1 0.42657
- 13) Pu-23WE40 l i 3.22E-03 0.3474% '/#//##/##/#/#//#/##/#An\n\W\u\W\n\%%%
l [ 14! Pu-24 f i i 6.37E-021 6.8734% Pu-241/Am-241 i 18.57143 l (' 15f Pu-242 i LESS THAN l 7.39E-05) 0.0080% jf#/#/###/####/##/#/#A%%%%%%%%%\%%N 16i Am-24 f i I 3.43E-031 0.3701%II Am-241/Pu-238 i 2.333331 17! Am-243 I LESS THAN i 2.43E-05 0.0026% /#/#//##//#//##//#//#///A\%%\%%%%%%%%% 181 Cm-242 i LESS THAN I 1.21E-04 0.0131 % if//###//##/#####/###Au%%%%%%%%%%% 19l_ Cm-24W244 i LESS THAN 1.73E-04 0.0187% 1#//#/#//#///#/#/##/#/#An%%%%%%%%\W\% 204 U-234 6.35E-06 0.0000% lf#//##/#////////##/#/#///An%%%%%%%%\w\% l 21 ll U-235 LESS THAN 3.65E-06 0.0000% f##/////##/#/###/####Anu\Mukun%\WWW i 22s U-238 ' LESS THAN 5.16E-06 0.0000% i U-238/Am-241 1 0.00002! ! 23! Co-80 i 8.53E-03 0.9204% i Am-241/U-2Se 106472.868221 24i Ru- f as LESS THAN 1.50E-04 0.0162% f/##///#/#/##//#////#/#/An%%%%%\%%%%\% 251- Aa- f 00m LESS THAN 1.10E-05 0.0012% ir#/#////#/##//##/#/###Au%%%%%\n\%%%% 26 Sb- 125 LESS THAN 3.26E-05 0.0035% lf/#/#####/#/#/#/#/##/A%%%%%%%%%%%\N 27 Cs- f 34 l LESS THAN 2.33E-05 0.0025% D Cs-134/Cs-137 0.20619: 281 Cs-f37 i 1.13E-04 0.0122%) Co-60/Co-137 75.46673 l 29: Co-i44 LESS THAN 4.76E-05 0.0051%)l Cs-137/Co-60 0.01325-
.30; Eu- f 52 1.99E-04 0.0215% jf#//#//#/##/#/##///#/#/A\%%%%%%\W\%%%V 31 Eu- f 54 1.02E-04 0.0110% !f///#//#/#//##///#/////#//A%%%%%%%%%%%%8 l 32; Eu- f SB 2.19E-05 0.0024% if//#/###/#/#Ull//###//A%%%%%\%%%M 1
i Totais= = > ll 9.27E-01 L) 100.00% k I r i 1 .% J
rv B & W RESULTS 9/14/95 - Saxton Samples - - SYSTEM = = = > Pressure Control & Relet - /ntemel Surface of poing between Pressureer relief valves
& theDnscharge Tank (4PR302@
l i Samfon !I BAW il 34 6371 ll _Sampde Numbee Sampde Number Total,TRU Content (nCi/g) l SX825950044 ll 9507002-07-A_A_ ll 0.0092 l1 Total Uranium Content (nci/g) Sample Type il Pipe Section il I Desenption il item 4 11 l i Results NormCized l Nuclide . Notes uCi/a to 100 % Nuclides RATIOS ) 1 ll H-3 i LESS THAN i 4.14E-02 0.1896% F//////////////##////#/////#A%%\Wh\WWWW\%%) l
' 21 C- f 4 I LESS THAN I 9.31E-03 0.0426 % "###//#//#/#/#//#////#/AuwhB%%%%%\WW 1 34 Fe-55 1 2.61 E-03 0.0120 % Fe-55/Co-60 0.01243 4: Ni-59 i 7.24E-02 0.3316 % '##/////##/####/#/#/#A\%%%%%%%\%%%Y , ,
5 Ni-63 i i 3.59E+00 16.4413%J#///#/#////#/###///##/Auw%%%%\WWWWW! ! 6- St-89 LESS THAN i 2.13E-04 0.0010 %4#//###/#/#/#/#/#/#///AWW%%%%%%%%%Y 7 Sr-90 f 4.16E-03 0.0191 % Sr-90/Cs-137 1 0.00024 81 Nb-94 LESS THAN I 8.44E-04 - 0.0039%//##///#////#/#/#////#/#A%%%%%%%%%\WM 91 Tc-99 LESS THAN I 4.86E-03 0.0223%F#////#/#####//##//##A%W\%%%%%\%%%Y i 10! /-129 LESS THAN l 7.52E-04 0.0034 %F##//#/###//##/#//#/#Au%%%%%%%%%%Y; j 11i Np-237 LESS THAN l 3.91E-05 0.0002%F#/#////###/##/#/#///#AWW%%%%%%%\Wut l 12! Pu-239 1 7.94E-04 0.0036 %! Pu-238/Am-241 1.11674 13 Pu-239/240 1 1.61 E-03 0.0074%PN#/#//#//###//##//#//A\\%%\\\\\\\\\\\%\\\Y h 14i Pu-24 f 6 3.14E-02 0.1438%l Pu-241/Am-241 44.16315 g 15s Pu-242 LESS THAN I 2.36E-06 0.0000% F///#/##////#/#//#///#///A%%%%%\%\%%%%\1
% 161 Am-24 f I i 7.11E-04 0.0033%l Am-241/Pu-238 0.89547i t 1761 Am-243 i LESS WAN i 9.52E-06 0.0000 % '/###//#///#/###/##///A%%\%\%%%%%%%\1 l
18il Cm-242 I LESS THAN I 3.18E-05 0.0001 % '/#//////#/////////#///////#A%%%%%%%%%%%\1 19! Cm-244/244 LESS THAN I 3.93E-05 0.0002%F#/#/##/#/###///////////AWWWW%%%%\%%%) 20 U-234 LESS THAN I 5.43E-06 0.0000%P#/##////////##//#/##///AWW\%%%%%%%%%1 l 211 U-235 LESS THAN 1.16E-06 0.0000 % ///#/////#///##/#/#//#//A\%%%\WW%%%%%Y l 22: U-238 LESS IHAN 2.58E-06 0.0000 % U-238/Am-241 0.00863 23: Co-60 i 2.10E-01 0.9617 % Am-241/U-238 275.58140 l 24e Ru-fos ' LESS THAN i 5.40E-02 0.2473% '#////#//#/#/#///#////////A%%\%\%%%%%%%Y 25: An- f oem LESS THAN i 5.44E-03 0.0249% '/#/#/////###/#/#/##/#A\\\\\\\\\\\\\\\\\\\\\\Y l 26i Sb- 125 ! LESS THAN I 2.74E-02 0.1255 % '#///##//#/##///##//////Au%%%%%%%%%%Y 27i Cs-734 I 8.73E-03 0.0400% Cs-134/Cs-137 0.00049 28i Cs- f 37 l 1.77E+01 81.0616 % Co-60/Cs-137 0.011861 291i Ce-f44 LESS THAN ! 3.12E-02 0.1429% Cs-137/Co-60 84.285711 30s Eu- f 52 ' LESS THAN I 1.95E-02 0.0893% F///#/###/#/###//#//#/A%%%%\n%%%%%%) 31: Eu-154 LESS THAN I 3.96E-03 0.0182% '#/i#///#/#/#//#//###//A%%%%%%%\W%%%1l 32: Eu- 185 i LESS THAN ! 1.38E-021 0.0632% '/#/#//##/#/#///#/#/#//AWWW\%%%%%%%%) 11 Totals = = > ll 2.18E+01il 100.00% jl [ l l f (v /
B & W RESULTS 9/14/95 - Saxton Samples -
- SYSTEM = = = > Caa &Orr&rit n Cooling Water System, 3* Diameter P&e S=*vs (Up-Streem of CC Pump).
Down Straem of CC Hx. Samfort ! SAW i ll 73.5370ll
\_Samoie Number Samoie Number ! Total TRU Content (nCi/g) i SX825050045 I 9507002-06-AA i l 0.013716 Total Uranium Content (nci/g) 2 Samole Type 1 Pice Section !!
il Descriotion i Test Cut li Results Normalized
'Nuclide Notes uCi/c to 100 % Nuclides RATIOS 1l H-3 ! LESS THAN 5.28E-02 32.1317%f#####/#//#/##///////#/Auwuww%%%%%%y 2! C-14 i LESS THAN 6.61 E-03 4.0226% F##//#/#/#/##/#/#!/##/a\hW\whWWW\ hun ,
31 Fe-55 I LESS THAN 1.66E-04 0.1010%) Fe-55/Co-60 di 0.39059 W Ni-59 i LESS THAN I 1.09E-03 0.6633%1//#//##//##/#//#####/A%%%%\%%%%%\u\t 5l Ni-83 I i 1.30E-02' 6! 7.9112% ///#/////N#//#///////####A\%%%\\\u\u\n\uut Sr-89 i LESS THAN 1 1.77E-04 0.1077% '//######///#/#///##///A%%%\%%%\n%%%\f 7i Sr-90 e i 4.50E-04 . 0.2739% Sr-90/Cs-137 0.04500I
' 8! Nb-94 LESS THAN I 6.41E-06 0.0039% '##/###/#/#/##/##//#/A\%%%%%%%%%%%t 9h Tc-99 LESS THAN I 3.85E-03 2.3429% F/##/##//###/###//##/A%%%\W%%%%%%%f 1 01 l-f29 i LESS THAN ! 1.72E-03
#p-237 1.0467% U//###/#/#######/#/#Anun\%%%%%%%W 11l LESS THAN 5.09E-05 0.0310%F/##//#//#///###/#//##/Aun\WWWWW\WWuv 12! Pu-238 1.48E-03 0.9007%I Pu-238/Am-241 0.51748 13[ Pu-23G&40 3.59E-03 2.1847% '#////#/##/###/##/#//#(WWu\WWWWuwww O 14[ Pu-24f 6 54E-02 39.7995 % Pu-241/Am-241 22.86713
( 15l Pu-242 LESS THAN 4.74E-06 16; 0.0029%F##//#/#/#######/#/#(%%%\%%%%%%%%1 Am-247 2.86E-03 1.7405% l Am-241/Pu-238 1.93243! 17 Am-243 LESS THAN 1.28E-05 0.0078% J//###/##/#///#//#/###A%%%%%%%%%\\%\t 18 _ Cm-242 i LESS THAN 6.48E-05 0.0394% #/#//##/#//#######/#A%%%%%%%%%\%%t 19l Cm-24W244 ' LESS THAN 7.38E-05 0.0449% f##/###/##/#//##//##/A\%%%%%%%%%\\Wj 20i U-234 i LESS THAN ' 6.06E-061 0.0037%F//#//#/##////#/###/##/A\%%%%%%%%%%%0 211 U-235 I LESS THAN 3.16E-061 0.0019%F#/##//####/#/#/##/#/A%uuiW%%%%%%W 221 U-238 LESS THAN 4.46E-06 ' O.0027%l U-238/Am-241 0.00156 23i Co-80 4.25E-04 0.2586%I Am-241/U-236 641.25561 24i Ru-f06 LESS THAN 1.33E-04 0.0009% F/##/#/#/#/#/#/#/#/#/#A%%%%%\n%%%%W 25! Aa-108m LESS THAN 1.32E-05 0.0080%F#/#////###/###/#/#//#An\%%%%%%%%%W 26; Sb- f25 LESS THAN 5.00E-05 0.0304 %F####/##///N#####/#/AWWuw\%%%%%%W 276 Cs-f34 LESS THAN 1.29E-05 0.0079% Cs-134/Cs-137 0.00129 28; Cs-137 1.00E-02 ' 6.0856 % Co-60/Cs-137 0.04250 291 Co-f44 LESS THAN 8.86E-05 0.0539% Cs-137/Co-60 23.529411 301 Eu-f52 LESS THAN 1.06E-04 31; 0.0645% F#/#/#/#/####//##/#///A%%%%%\W\%%%Wt Eu-f54 LESS THAN 3.47E-05 0.0211% F####/###/##/#//##/NA%%%\W%%%%%%WM
- 32) Eu-fM LESS THAN 4.00E-05 0.0243% ///###/#/////###//##////A%%%%%%%\%%%%f Totale= = > H 1.64E-0111 100.00%)I v -
t
l B & W RESULTS 9/14/95 - Saxton Samples , ,
% SYSTEbe = = = > Stornae Wen - /ntemal Suthece d piping between itMet side d the Storage WeX Pumps to the
- l. decharge point d the Storage WeH (3SWf 5f & 4SWf 5fJ.
l l l l Sauron ll SAW d 189.9277ll iSemode Numbert Samode Number Total TRU Content (nCi/g) JX825950046 4 9507002-09-AA il 0.0178 ll l Total Uranium Content (nCi/g)
' Sample Type ll Pipe Section ll l l Descriptum il item 9 ll 1
l Results Normaltzed ! j Nuclide Notes uCi/a to 100 % Nuclides RATIOS l 1i H-3 ! ESS THAN 5.65E-02 9.8499% P/#/#/#/#//###///////##A\\%\W\\WB\\\\\Wh\t i 20 C-f4 i ESS THAN 8.39E-03 1.4627%P/#/##/#/#/////////#/#/#Au\wu\\%%%\\\\%\M 31! F9-55 1.94E-02 3.3821 % Fe-55/Co-60 l 0.77912: l 41 $li-59 LESS THAN 7.99E-04 0.1393 % '/#####//#####//#////AW%%%%%%%%%%\1 I Sit Ni-63 3.62E-02 6.3109% p//#/###//////#///##/////A%%%%%%%%%%%\1 j 6; Sr-69 WSS THAN 1.70E-04 0.0296% F#///##/###//#/#//##//AW\%%%%%%%%%%1 1 7t St-90 1.83E-03 0.3190 %! St-90/Cs- 137 0.00643 t ar Nb-94 2.78E-03 0.4847 %P///////##//#///#/////##//M\\\%\Wu\%%%\%%) l 9: Tc-99 GSS THAN 6.22E-03 1.0844% Pu##/##//#!/#####/#/An%%%%%\wu\nh\1 ! 10! /- f 29 WSS THAN 4.25E-03 0.7409 %J/#//##////#//////#//#////An\%%%%%%%%\W\1 tile #p-237 WSS THAN 1.00E-04 0.0174 %F////##///##///#////###/AW\\\\%\%%\\%\\\%\~ 1 21' Pu-238 4.87E-03 0.8490 %! Pu-238/Am-241 i 0.75154!
- 1311 Pu-239/240 ( i 8.95E-03 1.5603% f///#;#/#/####//#///#//A%%%%%%%%%%\%1 141 Pu-24f I I 1.69E-01 29.4626%i Pu-241/Am-241 1 26.000251
) 15t Pu-242 WSS THAN I 3.70E-06 0.0006 % ?///#;/#/#/####///#/////A%%%%%\w%%%%%1 l Am-241/Pu-2381
/ 16: Am-24 f , 6 6.48E-03 1.1297%l 1.330601 174 Am-243 i 1.23E-04 0.0214 % ?/#/###/#//#//#/#//#///A%%%%%%\%%%%%1 18t Cm-242 i 1.23E-04 0.0214 % ?//#/##//##//////#/#//#/A%\%%%%%%%%%%)
19L Cm-244/244 2.78E-04 0.0485% P//#/#//###//#/#///#/#/A\W%%%%%%%%%%1 20 U-234 LESS THAN 9.11 E-06 0.0016 % ?////#///#//////////#////////A\W%%%%%%%%%%1 21! U-235 WSS THAN 5.27E-06 0.0009 % ///#####/#/#///////#////A%%%\%%%%%%%%1 22; U-238 LESS THAN 3.40E-06 0.0006 %i U-238/Am-241 0.000521 ! 23 ! Co-80 2.49E-02 4.3409 %i Am-241/U-238 1905.88235i 246 Au- f os i LESS THAN 5.43E-04 0.0947 %J/////#/#///#/#/##/#/////Au%%%%%%\%%%W 25y As-foem 2.23E-03 0.3888 %J//#####///##//##/##/M%%%%%%%%%%%) l 26i Sb- f 25 1.91 E-04 0.0333 %J//#//##////////#//#/#/#lA%%%%%\%%%\W\W1 27i Cs-f34 8.03E-05 0.0140 %i Cs-134/Cs-137 0.0003711 28a Cs-137 2.17E-01 37.8307%I Co-60/Co-137 0.11475il i 29il Ce- f 44 LESS THAN 3.68E-04 - 0.0642%I Cs-137/Co-60 8.71486ll ! 3011 Eu-f52 7.24E-04 0.1262%////u/////#/#/#///#///#/##/#////#/#///#/#//// i 31.'f Eu-f54 8.67E-04 0.1511 %F/###//###/////#////##/##/////////#/###//// 321 Eu-f55 2.21 E-04 i 0.0385 %?!/#/#///##////#//#///////#/#///#///#///#///#/ ! Totais = = > 11 5.74E-0_1JI 100.00% I I l t k t (m./) i l
,Q B & W RESULTS 9/14/95 - Saxton Samples .
SYSTEM = = = > Vents & Drains - /ntemn/ Surkrce ofpiping on vent and drain hender discharging to the Discharge Tank (2DLO76R). 1 Saxton i BAW ! ll 107.2610ll Semote Numbert SampJe Number '
~
Total TRU Content _(nCi/g) _5_X825050047 i 9507002-10-AA ) ll 0003111 Total Uranium Content (nci/g)
! Sample Type i Pipe Section I i Description i item 13 4 Results Normalized Nuclide Notes uCi/c to 100 % Nuclicles RATIO 8 1!! H-3 il LESS THAN : 1.51 E-03 0.0555% v###MM/m//M//M#/Ma%\WWWW%%%%%%Y 2! C- f 4 i LESS THAN i 3.27E-04 0.0120 % '#M#///MMMMN/M////A\WWW%%%%%%%W 3! Fe-55 1 i 1.50E-03 0.0551 % Fe-55/Co-60 0.00455 4! Ni-59 4 i 1.93E-02 0.7094% F#//M//#///#////#//////#//A\\\\\%\w\hW\\WW\V Si N/-63 i
! 1.33E+00 48.8869% WMMNM#MNM/M#M#A\WWuu%%%\%%%\h 6l Sr-se i LESS THAN I 1.40E-05 0.0005% pMN//##M##/////#/##//m\%%%%%%\%%%\Y 7! Sr-90 t 1 3.40E-02 1.2497%i Sr-90/Cs-137 0.03890 8! Nb-94 i LESS THAN 6.66E-04 0.0245% V/M//////M/////##/#/###A\\\n%%%\%%%%%\t 9i Tc-99 LESS THAN 2.38E-04 0.0067%FMM#M#/M//#MN/##/A\%%%%%%%%%\W\j 10! /- f 29 LESS THAN 3.49E-05 0.0013% !#MM/#MM/#/##/#M/m%%%%%%%%%%\t 11! #p-237 LESS THAN 5.29E-03 0.1944 % '#M////#///M##M#$#/A%\%%%%%%%%%W '
12!) Pu-236 LESS THAN 1.03E-03 0.0379 % Pu-238/Am-241 0.18969 13t Pu-238V240 ! LESS THAN 1.75E-03 0.0643 % '/M//M/##/M///##MM/a\n%%%%%%\nwny 14! Pu-24f LESS THAN 9.22E-02 3.3890%l Pu-241/Am-241 16.97974 15t Pu-242 LESS THAN 2.66E-04 0.0090% F/##/####/#/#//#M#Ma\Wh\%%%%%%%%V
\
v 16 Am-24f 5 43E-03 0.1996% l Am-241/Pu-2Se 5.27184! 174 Am-243 : LESS THAN 1.37E-04 0.0050%7MNMM/##MM/M#M/Mu\%%%%%%%%! 18t Cm-242 _J LESS THAN 3.80E-04 0.0140%2M/MM##/#M/#MN//#a\%%%%\%%%%%%g 199 Cm-244/244 i LESS THAN 7.78E-04 0.0286%F##M##M///#MMM//#A%%%%\%%%%%\\W 20!! U-234 l LESS THAN 1.61 E-06 0.0001 %P/MMM#M#M#//MM#aw%%%%%%mn%W 216 U-235 i LESS THAN 6.23E-07 0.0000 %FM/M//#M//###NMM/Aw\wh\mn%%%\%\1 224 U-238 LESS THAN 8.79E-07 0.0000 %l U-238/Am-241 0.00016 234 Co-60 3.30E-01 12.1290%l Am-241/U-236 6177.47440 241 Ru- f os i LESS THAN 5.91 E-03 0.2172 %P//MM#MMM//#//M/#/L\Wh%%%%%%%%W 25i As- f oem LESS THAN 5.84E-04 0.0215% fMM/##//M/M###M///MWh\%%%%%%%W 261 Sb-125 LESS THAN 2.27E-03 0.0634 % 'M#M##//MM/M//MN/MWMwwwwhwW\W 273 Cs-134 i LESS THAN 6.93E-04 0.0255 % Cs-134/Cs-137 0.00079 261 Cs-f37 I 8.74E-01 32.1257% Co-80/Cs-137 0.37757 29s Ce- f 44 ! LESS THAN 4.16E-03 0.1529 % Cs-137/Co-60 2.64848 30; Eu-f52 i LESS THAN 5.99E-03 0.2202%J/MM#M#M///M#/M///n%%%%%%%mWWV 31: Eu-f54 I LESS THAN 9.05E-04 0.0333%P//M/#M/#//W///M//#///MMW%%%%%%%%Y4 32 Eu-f55 I LESS THAN 1.20E-03 0.0441% jMMMM//MMM#//MU/,Mmuww\Wuw%Mt l Totals = = > h 2.72E+00}i 100.00%Il
ff O ~ B & W RESULTS 9/14/95 - Saxton Samples . I SYSTEM = = = > Core /niet TransmitterRack - AREA f, west of sump. Santon i sa w 1 il O.1193J LSerrpie Number Sample Number I! Total TRU Content (nCi/g) __S_ XE15950006 i 9507002-11-AA i! il 0.0005 il Total Uranium Content (nCi/g) Sample Type i Scrapinos t CDescription n instrument Rack III Results Normalized Nuclide Notes uCi/o to 100 % Nuclides RATIOS 1s H-3 .i LESS THAN , 9 93E-05 t 0.0327% f//////////////////////////////A\\\w\\\ww\\S\@ 21 . C- 14 1 ESS THAN i 7.62E-05: 0.0251 % J//////////////////////////////Aw\wwwwwwwung 3i Fe-55 _j t 1.18E-04l 0.0380%t Fe-55/Co-60 i 0.05842i 41 Ni-59 i
! 1.27E-041 0.0419 %)//////////////////////////////Au\www\wwwwwn)
Si Ni-63 e i 8.55E-03 i 2.8191 % J//////////////////////////////Awwww\w\wwwwwi ! 61 Sr-89 ESS THAN i 8.67E-061 0.0029% J//////////////////////////////Awwwwwwwwwww) l 7i Sr-90 4 i 4.48E-041 0.1477 %i Sr-90/Co-137 i 0.00154i ! 84 Nb-94 !! ESS THAN t 1.49E-05i 0.0049 % f//////////////////////////////Awwwwwwwwww\w l 9 'I Tc-99 J E8S THAN ! 1.93E-051 0.0064 %J//////////////////////////////Awwwwwwwwwwwt 10 I- 129 , ESS THAN i 3.67E-06 6 0.0012 %J//////////////////////////////Awwwwww\wwwwu 111 AD-237 i ESS THAN i 1.47E-07 6 0.0000D/////////////////////////////Awwww\wwwwwwg p 12'i Pu-238 13i Pu-239V240 4
,
- 2.06E-051 5.05E-051 0.0068 %I Pu-238/Am-241 0.645774 0.0167 %J//////////////////////////////Awwwww\wwwwwg
(' 14:1 Pu-241 i 1.51E-051 0.0050 %; Pu-241/Am-241 1 0.473351 154 Pu-74TE ! ESSTHAN < 9.31E-09 4 0.0000 %J//////////////////////////////Awwwwww\wwww\i 16a Am-241 *
' 3.19E-051 0.0105%y Am-241/Pu-238 i 1.548544 17t Am-243 ESSTHAN ! 2.25E-08) 0.0000% ///////////////////////////////Au\wwwwwwwww\)
181 Cm-242 1 ESS THAN 1.91E-07 ! 0.0001 % f//////////////////////////////Aww\wwwwwww\w) 19a Cm-243/244 .: t 7.92E-07 i 0.0003% J//////////////////////////////A\ww\wwww\ww\\W 20i U-234 4 ESSTHAN i 1.88E-07 i 0.0001% J//////////////////////////////A\\ww\\\wwwwwww 21 4 U-235 ( ESS THAN i 9.40E-081 0.0000% J//////////////////////////////Awwwwwwwwwww) 22i U-238 i ESS THAN I 1.76E-071 0.0001 % i U-238/Am-241 1 0.00552 ii 23: Co-60 1 2.02E-03 6 0.6660 %I Am-241/U-238 l 181.250001 i 24; Ru- 106 ESS THAN ! 5.42E-04; 0.1787% ///////////////////////////////A\\\wwwwwwwwwg j 25i Aa- 108m ESS THAN 5.34E-05 i 0.0176% ///////////////////////////////A\\w\ww\www\\\wg 261 Sb- 125 i ESS THAN 1 2.86E-041 0.0943% ///////////////////////////////Awww\www\wnuw) 271 Cs- 134 i 4.78E-051 0.0157%ll Cs-134/Cs-137 i 0.00016: 28: Cs-137 f 2.90E-018 95.6173%IC7-60/Cs-137 ; 0.00607 294 Co- 144 ESS THAN 6 4 91E-04i 0.1619%O-137/Co-60 143.56436 30: Eu-152 ' ESS THAN I 1.30E-041 0.04293~[//////////////////////////////Awwwww\www\n\n I 31 Eu- 154 GSS THAN i 2.66E-05i 0.0088 %J//////////////////////////////Anuwww\wwwwww 32 Eu- 15lF ESSTHAN i 1.11 E--04 i 0.0386%J//////////////////////////////Awwwwww\qhggg l Totals = = > l 3 03E-011 100.00% i l 4 Y g (v/ l
I l
- l l I l
f% V)
^ \
l B & W RESULTS 9/14/95 - Saxton Samples l SYSTEM a = = > AuxiliaryCompartmentInstrumentRack - AREA 1, eastofsump. i _ _ Sarton d B& W 4 n 0.381811
. Sample Number __ Sample Numberj Total TRU content (nCi/g) l 5 SX81_950007_u9507002-12-AA 4 i 0.0197 ll Total Uranium Content (nC4/g) l ~-Sample Type i Scrapings 9 l Description ilnstrument Rack 2ll l
Results Normalized Nuclide Notes uCi/a to 100 % Nuclides RATIOS l 1: H-J LESS THAN i 1.17E-04 i 0.0620%'f////////////////Tllll/T((T////lWWW\\\\\\\\\\\\\\
- 2i C- 14 ESSTHAN 8.70E-05 8 0.0461 %'f//////////////////////////////A\\\\\\\\\\\\\\\\\\\\\\\
3! Fe-55 ! 1.50E-041 0.0796%ll Fe-55/Co-60 ! 0.070091 4! Ni-59 i 1.19E-04 ' O.0631 % 1//////////////////////////////A\\\\\\\\\\\\\\\\\\\\\\j 5I Ni-63 8.84E-03 i 4.6882% f//////////////////////////////A\\\\\\\\\\\\\\\\\\\\\\1 61 Sr-89 ESS THAN 1.38E-051 0.0073% f//////////////////////////////A\\\\\\\\\\\\\\\\\\\\\\j 7J Sr-90 1 i 8.43E-05 i 0.0447%l Sr-90/Cs-137 t 0.000484 81 Nb-94 : ESS THAN i 2.36E-05 0.0125% f/////////////////////////// ///A\\\\\\\\\\\\\\\\\\\\\\) 9l Tc-99 1 ESS THAN i 3.96E-04i 0.2100% i///////////////////////////////A\\\\\\\\\\\\\\\\\\\\\\1 10): 1-129 J ESS THAN ; 9.00E-06 i 0.0048% f////////////////////////// ///A\\\\\\\\\\\\\\\\\\\\\\1 11-i /Wp-237 i G$~S THAN i 5.25E-07 ! 0.0003% P////////////////////////://///A\\\\\\\\\\\\\\\\\\\\\\1 [N 12'i Pu-238 13! Pu-23W240 1 i : 9.34E-06 ! 2.54E-05 i 0.0050 %i Pu-238/Am-241 1 0.30825 W 0.0135% ///////////////////////////////A\\\\\\\\\\\\\\\\\\\\\\j i (
\ 141 Pu-241 i i 3.10E-04 I 0.1644%I Pu-241/Am-241 1 10.23102]
15i Pu-242 1 ESS THAN i 2.54E-081 0.0000%f//////////////////////////////A\\\\\\\\\\\\\\\\\\\\\\1 16:! Am-241 ! 6 3.03E-051 0.0161%I Am-241/Pu-238 i 3.24411 il 171 Am-243 1 ESS THAN i 1.02E-061 0.0005% J//////////////////////////////A\\\\\\\\\\\\\\\\\\\\\\j ! 18 Cm-242 1 ESS THAN > 3.43E-061 19a-Cm-243/244 i ESS THAN i 1.76E-061 1 6.55E-06, 0.0018% J//////////////////////////////Ai\\\\\\\\\\ 0.0009% f//////////////////////////////A\\\\\\\\\\\\\\\\\\\\\\ 20 U-234 4 ESS THAN 0.0035% f//////////////////////////////A\\\\\\\\\\\\\\\\\\\\\\j 21- U-235 ESS THAN ! 6.55E-06i 0.0035% J//////////////////////////////A\\\\\\\\\\\\\\\\\\\\\\1 22 U-238 ESS THAN i 6.55E-06i 0.0035%i U-238/Am-241 0.21617: l 23; Co-60 i 2.14E-03 l 1.1349%9 Am-241/U-238 4.625051 l 24 Ru-106 : ESS THAN i 6.76E-04; 0.3585% ///////////////////////////////A\\\\\\\\\\\\\\\\\\\\\\) l 25 Aa- 108m ESS THAN 6.70E-05 : 0.0355% ///////////////////////////////A\\\\\\\\\\\u\\\\\\\\j 26 Sb- f 25 ESS THAN i 3.54E-041 0.1877% J//////////////////////////////A\\\\\\\\\\\E\\\\\\\j 27! Cs-134 , ESS THAN i 2.55E-05 i 0.0135 % Cs-134/Cs-137 : 0.00015. 28 Cs-137 ' i 1.74E-01 1 92.2793%; Co-60/Cs-137 i O.01230; 29i Ce-144 ESS THAN i 6.05E-04 i 0.3209% Cs-137/Co-60 i 81.30841i 30 Eu-152 ! ESS THAN ! 2.20E-04: 0.1167 %J//////////////////////////////A\\\\\\\\W \\\\\\\\\\\\ i 31 Eu-154 4 ESS 1HAN ' 4.73E-05 i 0.0251 %?//////////////////////////////A\\\\\\\\\\\\\\\\\\\\\\\ 32 1 Eu-155 1 ESS THAN 1.82E-04 i 0.0965% ///////////////////////////////A\\\\\\\\\\\\\\\MM i Totals- = > d 189E-01 ti 100 00%ii l l t [ p 4
Q . Q l B & W RESULTS 9/14/95 - Saxton Samples SYSTEM = = = > Cooling, headng and Vend /adon - Pnmary Compartnent Air Handler, ARF.A 2 t_Saxton 4 64 W i ti 0.0573!I LSample Numbe_r.l Sample Numbre Total TRU Content _(nCi/g) sx825950038 (9507002-13-AA i b 0.0015ll' Total Uranium content (nci/g) Sample Type 4 Filter / solids i
! Description 1i Item 11 q Results Normalized Nuclide Notes uCi/3 to 100 % Nuclides RATIOS 1i H-3 ! LESS THAN 7.01 E-051 0.1253% 3/####/###/#/####/Au\\uu\\ugu\n\u) 21 c- 14 :l ESS THAN 1.08E-041 0.1931 %f#/#/#///##/##/#////#/A\uu\u\uuu\unut 3 L Fe-55 4 i 7.41 E-05 i 0.1325%! Fe-55/Co-60 6 0.040053 4!! Ni-59 3 i 1.11E-04! 0.1964%/###/#/##/#/#////###Anu\wunu\uunu) 5l Ni-63 4 i 5.61E-031 10.0289% /###//#/###/##/#///#Auguuuuuuuuu) 64 Sr-ap 4 ESS THAN 5.41 E-06 i 0.0097 %N/####/#####/#/#/#Anunuunuunuu) 71 Sr-so 1 i 5.22E-05 6 0.0933 %; Sr-90/Cs-137 i 0.00111i 84 Mb-94 i ESS THAN ! 3.11E-061 0.0056 %3/##/#/##/#/##/##/#Auguu\ununu\wt j 9i Tc-99 . ESS THAN ! 5.58E-04) 0.9975 %W///##//#///#/#//#/#/NAuuuuuuuuuuuj 104 l- 129 ESS THAN ! 2.18E-051 0.0390 %/#//#/#####/#/#////#/Annuuuuuuuuuj 11 L__Ap-237 ESS THAN 1.87E-07 i 0.0003%?//##/#/#//#/##/#/#/#A\u\uuuuuuuuuj O 121 Pu-238 -
8.80E-07 - 0.0016 %i Pu-238/Am-241 1 0.03964 ;l (v 131 Pu-23W240 1 2.13E-06 0.003868#//#/#//#//#/#/#/NAnununu\uuuu\\ 14( Pu-241 '
- 3.11 E-051 0.0556 % Pu-241/Am-241 1 1.40090 d 153 Pu-f42 i ESS THAN i 1.75E-08 0.0000% ?#/#//#////##/#//#//##A\\uunuu\n\uu\\t 16 FAm-241 '
i 2.22E-05 0.0397%I Am-241/Pu-238 i ) 25.227271 175 Am-243 i ESS THAN i 5.93E-081 0.0001 % /#/#/##///##//#///#//#A\ ,uunu\uunu\w) 189 Cm-242 i ESS THAN i 2.59E-07 0.0005% j#//##//#//#/#///#//////A\\w\n\uuuuuuuj ! 19 Chi-f43/244 1 ESS THAN i 4.67E-07 0.0008% f#/#//#//#//#/#/#//#///A\\n\n\nu\unu\ut l 20" U-234 , ESS THAN ! 8.08E-07 0.0014% /#/#//////##///#//#////#Anu\\n\\uu\uu\\\1 ' 21a U-235 i ESS THAN i 2.69E-07 0.0005% 3/#///##//////////#///##Anu\uguuuuuu\j l 22il U-2M 4 ESS THAN i 4.56E-07 0.0008 % U-238/Am-241 i 0.020549 234 Co-60 i i 1.85E-03 3.3072 % Am-241/U-238 i 48.684214 244 Ru- 106 i ESS THAN 1.33E-04 0.2378% //#/////#/###/////#/#/#Au\wuuuuuuuu% 25; Aa-10ern il ESS THAN 1.32E-05 , 0.0236% /#/#///##////#///#//#///Anununuu\unun l 26:! Sb- 125 i' ESS THAN ! 6.30E-05 i 0.1126% J##///////////#/#/#//#/NAuuununu\u\Wuh 1 2711 Cs- 134 , ESS THAN ! 8.85E-06i 0.0158%! Cs-134/Cs-137 I 0.00019i ! Cs- 137 84.0213%! Co-60/Cs-137 i 283 i i 4.70E-021 0.03936i 29i Ce- 144 i ESS THAN 1 7.28E-05 i 0.1301 %i Cs-137/Co-60 25.40541-30i Eu-152 i ESS THAN 1 7.84E-05 I 0.1402 %////#///###////#/#/////NAu\uuuuuuuuu% l 31i Eu-154 i ESS THAN i 1.42E-05 6 0.0254%#//#//#/#///###//#/////Anunuuunu\Wu% 32; Eu-155 i ESS THAN I 3.22E-05! 0.0576% 1////#/#/////###/#///////Auguuu@unu\u
;l Totals = = > I! 5.59E-02k 1TO.00%l V
i (.A)
^
v B & W RESULTS 9/14/95 - Saxton Samples SYSTEM = = = > Cooling, Headng and VenWation - Auvi/iary Compartment Air Handler, AREA 3 i Sexton BA V il ll 0.4161N LSample _._ Number Samode Aumber il Total TRU Content (nCi/g) SX835950039_,,i_9507002-14-AA 0 ji 0.0025(
~5 ample Type il Filter / solids 4
! Description L Aux. Air Handler ll Results Normalized Nuclide Notes uCi/a to 100 % Nuclides RATIOS l
1i H-3 ESS THAN 8.48E-05 : 0.2046% N/####/####/#####A\\\\\\\\\\\\\\\\\\\\\\\ ! 21 C- 14 - ESS THAN I 2.25E-04i 0.5428%Ju##/####/####//##A\\m\\mmmmmg l 3i Fe-55 , i 2.66E-04i 0.6418 % Fe-55/Co-80 1 41 Ni-59 0.10640i l ! 8.14E-05i 0.1964 %P####/##/#####///#/A\mmmmmmmi i 5" Ni-63 - i 6.31E-03: l 6i Sr-89 ESS THAN 15.2236 %##//####/##//##//##Am\\m\\m\mm\\) I 1.60E-05! 0.0386 %?##/###/###/#/#//##Am\mmmm\mm l 7i Sr-SD 4 i 3.94E-051 0.0951 0 Sr-90/Cs-137 0.001331 8 st Nb-94 i ESS THAN i 1.49E-05 i 0.0359 %//####/###////##/#/#A\mmmmmmm) 93 Tc-99 4 ESS THAN ! 1.30E-031 3.1364% !'/###//####/##/##/NAmmmmmmm\1
- l. 101 /-129 si ESS THAN i 1.53E-05; 0.0369 %s#////#/##########/A\\m\m\mmmmg i 111 A(p-237 i ESS THAN I 6.91E-071 0.0017 %J#####/##///######Amm\mm\\mmij
- (K 121 Pu-238 i i 1.25E-05 I 0.0302% ! Pu-238/Am-241 ' O.79618 j l
( )j
! 13:1 Pu-235V240 144 Pu-241 i t 2.90E-051
! 3.57E-04 p 0.0700 %PN/#//###/####/#/#//Ammmm\mmm) 0.8613 %! Pu-241/Am-241 1 22.73885 J 15L Pu-242 :1 ESS THAN t 2.78E-081 16! Am-241 i 0.0001 %P##//##/###/#//#/#/#Amm\mmmmm) i 1.57E-05i 0.0379 0 Am-241/Pu-238 l 1.258005 l 17" Am_-243 !! ESS THAN i 9.90E-08!
18,i 0.0002 %?////#/##/#//###//##//Ammmmmm\m) , Cm-242 1 ESS THAN i 4.29E-071 0.0010 %P//#/####///#/#/#//#//A\\m\\\\mm\\mm l 19.i cm-243/244 _ ri ESS THAN i 6.86E-07 6 0.0017% //#/#///////##//#/////#//A\mm\\\\\mmm\) ' 204 U-234 4 GSS THAN i 1.03E-061 0.0025% //###//#/#/##//ul##//Ammmm\mmm) 21, U-235 i ESS THAN } 5.05E-071 0.0012% J//#/#////////#/////////////Ammmmm\mmi 221 U- 2Mf~'-'~i GSS THAN 9.8BE-071 0.0024 % U-238/Am-241 i 0.062931 23.6 Co-60 i 2.50E-03 ! 6.0315 %I Am-241/U-238 15.89069 j i 241 Ru- 106 i WSS THAN 1.83E-04 I 0.3933 %/##/#/#//#/#//#/#//##A\mmmmmmm) 25i Aa- foam i GSS THAN 1.42E-05 ; 0.0343% /#///#/##/#//#////#/////A\mmmmmmm) 26j Sb- 125 GTS THAN I 7.89E-05i 0.1855% /##/////////#/////###/#/Ammmm\mmm) 271 Cs- 134 t WSS THAN I 1.54E-05 t 0.0372 % Cs-134/Cs-137 i 0.00052: 28r Cs- 137 i -
! 2.96E-021 71.4135 % Co-60/Co-137 i 0.084463
, 29: Ce- 144 ' WSS THAN 1 9.74E-051 0.2350 % i Cs-137/Co-60 ' 11.84000 l 30: Eu-75P ' WSS THAN i 1.30E-04 ; 0.3354 %/##/#/##//#//#//##/#/A\mmmm\mmm 31__Ju- F50 i GSS THAN ! 2.86E-05i 0.0642%//#//###/#####/#/#//A\\mmmmmmm 32! Eu-75U WSS THAN I 4.48E-05: 0.1081 % #/#/#/#/##/#/#/###//Ammmm\\m\mu ~~ ~ ' i Totals = = > 14.14E-0TI 100.00WI
\
b l
~
1 1 i n - ( . ( l B & W RESULTS 9/14/95 - Saxton Samples SYSTEM = ==> Cooling, Headngand Vend /adon - ControlRodRoom AirHandler, AREA f
. Saxton i B& W ll i 0.0585li LSamp/e_ Number Sample Number _ji Total TRU Content (nCl/g) l SX815950050 3507002-15-AA il 1 0.0224 h
! Total Uranium Content (nCi/g) Sample Type >i Filter / solids N i Descnotion d A/H in Rod Room W j Results, Normalized l Nuclide Notes uCi/c to 100 % Nuclides RATIOS l 1; H-3 - LESS THAN ' 6.05E-041 3Z6291 % ///////////////////////////////A\\m\\\\\\\\\\\\\m\\ 2! c- 14 ESS THAN i 2.12E-04 11.4337% ///////////////////////////////A\mmmmmmm) 3? Fe-55 ' i 2.90E-05 1.5640 %I Fe-55/Co-60 1 0.73604'! 4i Ni-59 ii ESS THAN ' 2.48E-05 i 1.3375% ///////////////////////////////A\\\\\\\\\\\\m\\\\\\\1 5.i Ni-63 i I 3.43E-051 1.8499% f//////////////////////////////A\\\M\\\\\\\\\mm\\ 6 F- Sr-ap i ESS THAN : 2.68E-05i 1.4454% ;f//////////////////////////////A\\mmmmmm\\t 71 St-90 ' ESSTHAN I 1.64E-05 6 0.8845 % Sr-90/Cs-137 i 0.042601
.8i Mb-94 ESS THAN i 9.11E-06! 0.4913 % ;f//////////////////////////////Ammmmmmmg 91 Tc-99 WSS THAN ! 6.10E-051 3.2899 % f//////////////////////////////A\\mm\M\\m\\\\\\
104 /- 129 WSSTHAN i 3.79E-051 2.0440% ' f//////////////////////////////Am\mmmmmm) 11:1 Np-237 ' WSS THAN t 1.40E-061 0.0755 % f//////////////////////////////Au\mm\m\\\\\\\\\) [A
\,._,-
129 14J Pu-238 131 Pu-239/240 i Pu-241 3 4 ESSTHAN ESS THAN ESS THAN i 6.64E-071 i 5.54E-071 I 5.16E-05I 0.0358 % Pu-238/Am-241 i 0.216291 0.0299 % f//////////////////////////////AmummmhWWM 2.7829 %. Pu-241/Am-241 1 16.807829 154 Pu-742 1 ESS THAN I 1.16E-071 0.0063 % ;'//////////////////////////////Ahmmmuuuuun 16a Am-241 a
! 3.07E-06I 0.1656%l Am-241/Pu-238 I 4.623 74 17: Am-243 4 ESSTHAN I 1.79E-071 0.0097% f////////////////////////////// Anum %%%%%\wm 189 Cm-242 i ESSTHAN i 5.11E-071 0.0276% 1//////////////////////////////Au%%%%%%%%\Wn 19I Cm-243/244 ; WSSTHAN i 3.58E-071 0.0193% f//////////////////////////////An\%%%\%%%BW4 20i U-234 i i 1.18E-051 0.6364% (//////////////////////////////Aumuuuumumn 21 i U-235 r ESSTHAN 1.12E-06 0.0604% f//////////////////////////////Aww%mmuguuw) 221 U-238 i 9.49E-06 0.5118 % U-238/Am-241 l 3.09121 q
- 23. Co-do ;
I 3.94E-05l 2.1249 % Am-241/U-238 I 0.323504 24: Ru- 106 g ESS THAN ! 7.75E-05! 4.1796% f//////////////////////////////AWW%%%%%%%%%) 25 Aa- 108rn i ESS THAN ! 1.20E-051 0.6472% ///////////////////////////////A%%%mwhunung 26' Sb- 125 ESS THAN i 2.30E-05! 1.2404% ///////////////////////////////AumummWmuuj 27: Cs-134 ESS THAN 9.60E-06 i 0.5178%" Cs-134/Cs-137 i 0.02494, 284 Gs- TJ7 3.85E-04 i 20.76e.0% Co-60/Cs-137 i 0.10234-29 ; Ge- 144 WSS THAN 4.26E-05 l 2.2975 % Cs-137/Co-60 9.77157' 30- Eu- 152 , ESS THAN : 8.10E-05 4 4.3685 %f//////////////////////////////ABWmBMWmmun 31: Eu- 154 . ESS THAN ! 2.64E-051 1.4238 % ;f//////////////////////////////AmWWum\mmm) 32- Eu- 155 WSS THAN ! 2.05E-05: 1 1056% 1/////////////////////////////lhhuR\\nhWBMW Totals = = > ! 1.85E-0Tf-- iOO.00%! l l j , j . I ' \'w/ /
i l i n B & W RESULTS 9/14/95 - Saxton Samples [V ) - l l
. SYSTEM = = = > Containment Vesse/ Sump S/udge (Uquid Ptuse), AREA f il Samfon ll BAW I il 0.000911 1Samode Numbesi Samsde Number I Total TRU Content (nCl/ml)
SX815960006 il 9507002-16AB l 9 0.0003il Total Uransum Content (nci/ml) i Sample Type i Sludoe-LP il 1 Descnot on 't CV Sump 4 l Results Normalized , Nuclide Notes uCi/ml to 100 % Nuclides RATIOS 1!I H-3 , i 5.43E-04 i 4.7305 % /##/////////##//////####AWuw\hw\w\gugut 2ll C-14 i WSS THAN 1.33E-05 l 0.1159%J/#/#//#///#//##/#//#///Au\h\\\w\\\\\ww\\\\t , ; 34 Fe-55 WSS THAN 3.56E-07 0.0031 %I Fe-55/Co-60 I 0.772231 l 4 li Ni-59 WSS THAN I 3.84E-07 0.0033% F//#/#//////#/##//#//#/#An%%\wn\w\wwww 5( Ni-63 1 6.88E-06 0.0599 %;'//##/////##///////#///##AW\WuhW\\%%\\\\W 64 Sr-89 WSS THAN 1.07E-06 0.0093 %;'/#/#////#/##//#//##/#/AWM\uhunghWh%) 7!i Sr-90 i 3.26E-05 0.2640 %i St-90/Cs-137 l 0.003021 l 84 Nb-94 ! WSS THAN 3.13E-07 0.0027 %!'#/#///#/#/#//###/##//Ang\\w\\n\w\\\\uul l 9F Tc-99 ' WSS THAN i 7.24E-06 0.0631 %;'/#/////##/##//#/##/#//Mw\nuguggggggt j 10! /-129 i WSS THAN I 1.17E-06 0.0102%i'##/#////#/##/##/###/Anng\wngungun" l til No-237 i WSS THAN I 2.70E-06 0.0002%?###/#/#////#/#//#//#//Anuw\wwwwwwww I 121 Pu-238 i WSS THAN I 8.27E-09 0.0001 %I Pu-238/Am-241 0.22053 13i Pu-239/2#n i WSS THAN I 1.74E-06 0.0002%i'////##//#/###/#/###///nwngwungungw 14] Pu-24 f W WSS THAN i 7.74E-07 0.0067 %i Pu-241/Am-241 20.64000i fq 15:1 Pu-242 i WSS THAN I 2.01E-09 0.0000% '/#/#/#/###/#/#/#//#//h%%%\%%%%%%%n) 16lL._Am-24 f I ! 3.75E-06 0.0003 % Am-241/Pu-238 ' 4.5344611 ( )
\d 17ll Am-243 WSS THAN i 1.34E-09 0.0000 % '####//##//##/#//#//# Ann \nnunununni 18:! Cm-242 WSS THAN 2.59E-09 0.0000 %,'/##/////#////#/#//#//#//A\\\n\\\\\\\\\n\\n\uy 19 FCm-243/244 WSS THAN 1.02E-06 0.0001 %,'////#//##//###///##/#/A\wn%%\wngu\wnt 20'l U-234 WSS THAN i 1.25E-07 0.0011 % p/#/#/////#/##/#//////////ABuw\nn\uw\n\\nj 21 it U-235 # WSS THAN 9.10E-06 0.0006 % '////#/####/#/###///#//nuwwwwwwwwwnt 22P U-238 # WSS THAN 4.30E-06 0.0004 % U-2SS/Am-241 '1.14667!
23; Co-80 11 WSS THAN 4.61 E-07 0.0040 % Am-241/U-238 0.872091 24i Bu- fos li WSS THAN 2.43E-05 0.2117% J#//#/////#///#/#/#/#/#//nwwwwwwwwwwW, 25; Aa- 108m v LESS THAN 2.55E-06 0.0222 %F###//#///////##//#/##/hwww\%%%wungw; 264 Sb- f 25 !! WSS THAN 1.28E-05 0.1115%F#/###////#/#//#/#/##//nnwwwwwwwwwWi 2711 Cs- f 34 il 3.42E-06 0.0296 % Cs-184/Cs-137 0.000324 2e F- Cs- f 37 ! 1.06E-02 94.06e2% Co-60/Co-137 0.00004 j 291 Ce- f 44 i WSS THAN 1.62E-05 0.1411 % Cs-187/Co-60 23427.33109] 30 ti Eu-f52 ! WSS THAN 3.0'7E-06 0.0267 % ;/#//##///#/#/##///#////hW%\%%%%\\\\\wh\1 31i Eu- f 54 ! WSS THAN 1.06E-06 0.0092%F//##///##//#//////##/////n\WWW%%%%%%%\\ 321 Eu- f 55 t WSS THAN 7.28E-06 ! 0.0634% "//#/#//////#//#////#///#//nuunnwwwwwwM il Totals = = > p 1.15E-02ij 100 00 %)I t d
l l . . l l p
- B & W RESULTS 9/14/95 - Saxton Samples - .
~
SYSTEM ===> Containment VesselSump S/udge (SolniPhase). AREA f l I
~
Santon il SAW $ 2.09661 ISamWe Number! Samde Nu_mbe_r_ Total TRU Content (nCL/g) SX815950006 il 9507002-16AC ll 0.00091 Total Uranium Content (nCi/g) I l 1 Sample Type 4 Studeo-SP 11 i Descnotum n CV Sump il Results Normalized Nuclide Notes uC#a to 100 % Nuclides RATIOS iil H-3 t 6.67E-04 6 0.3355% p///#//////////////////////////Aw\\W\\\\%\W\\WWu] 2ll C- f 4 i WSS THAN 1.02E-05 i 0.0051% F///////////////////////////////A\%%%%%%%%%%%1 3ll_ Fe-55 i 2.57E-04 0.1293%l Fe-55/Co-60 0.00006 l 40 Ni-59 1 9.77E-05 0.0491 %F///////////////////////////////A\\W\\\\%\\\\\\\h\\%j Sl Ni-63 i 7.58E-03 3 8126%F////////////////////////////////W%%%\%%%%%%\\\j 61 Sr-89 I WSS THAN 4.46E-06 0.0022%F///////////////////////////////Au%%%%%%%\w%%1 7 41 St-90 h 5.63E-07 0.0003%l St-90/Co-137 0.00000l 8f Nb-94 f ESS THAN 9.95E-06 0.0050 %F///////////////////////////////Q%%%%%%%%\%%%1 9i Tc-se WSS THAN 1.14E-05 0.0057 %P///////////////////////////////A\\\wu\\\\\\\\\\\w\\\f 10t /- f 29 i WSS THAN 1.37E-06 0.0007% F///////////////////////////////A\\\w\\\\\\\\\\\\\\u\\1 l 111 No-237 i WSS THAN 4.55E-07 0.0002% F///////////////////////////////A%\u\\\W\\w\n\ww\1 i 12i Pu-238 I 5.32E-05 i 0.0268 %I Pu-238/Am-241 0.20075i 13y Pu-23s/240J 1.22E-04 0.0614% P///////////////////////////#//A%%%%%%%%%%\%1 144 Pu-24 f i 1.65E-03 0.8299%l Pu-241/Am-241 6.22642i 15il Pu-242 WSS THAN 1.95E-06 0.0000% F////////////////////////#/////A%%%%\ww\%%%%Y 0.1333 %) Am-241/Pu-236 g 161 Am-24f 2.65E-04 4.98120 N 17!! Am-243 WSS THAN 1.97E-06 0.0000% J#//##/#/#/////###//////Awwww%%%%%%%Y 18 FCm-242 l 1.45E-06 0.0007% J////#/###//####///#///Aw\\\\w\w\WWWWuY ( 190 Cm-243/244 4.44E-06 0.0022% F#//##//#/#/#/#/#////#/A\WW\WWWWWWwhyi 20; U-234 WSG THAN 5.05E-071 0.0003 %F###//###/#####//////A\wwWWWW\uw\\uY , 21: 0 - 235 WSS THAN 1.65E-071 0.0001 %P/#//###/##/#///##/////Awww%u\%%%%%W ! 22! U-238 ! WSS THAN 2.75E-07 - 0.0001 %I U-238/Am-241 0.00104 23t Co-60 3.21 E-03 1.6146 % Am-241/U-238 l 963.63636 24i Bu-106 i WSS THAN 2.71 E-04 0.1363 % '/#//###/#//######/#A%%%%%%%%%%%Y 25i As- 108m j WSS THAN 2.75E-051 0.0138 % '###/########/##/#A%%%%\WW\\WW\WY 26! Sb- 125 i GSS THAN 1.16E-041 0.0583 % '####//##//N##//#/#//A\wwwwwwww\WWY 27 Cs-f34 1 5.8eE-051 0.0296 % Cs-134/Cs-137 I 0.00032 28i Cs-f37 ! 1.64E-01 i 92.5492% Co-60/Cs-1371 0.01745 294 Co- f 44 I WSS THAN 1.62E-041 0.0615 % Cs-137/Co-60 i 57.32007i 30'l Eu-152 i WSS THAN 9.30E-051 0.0468 % W##//#/##/#/#////#/#/Au%%%\\WW\n%%%1l 314 Eu- f 54 I WSS THAN 1.66E-051 0.0083 %J/#/######/#/#####/Aun%%%%%%%%\%1 321 Eu- f55 !
.ESS THAN 1.21 E-04 i 0.0609%Un#/##/#/#########AuwwhWWWWWWWW -
! "otais = = > W 1.99E-0_1JI 100.00 %)I l
l t 4 'i
\
l
l I l Q B & W RESULTS 9/14/95 - Saxton Samples . . SYSTEM = = = > Septc Tank A. S/udge (L4uid Phase) AREA 8 l il Samfort il Ba W i l 0.0090 ll ilsamsde Numbed Samode Number ! Total TRU Content (nCl/ml) LSmama 11 9507002-17AB I ll 0.0063R Total Urgnium Content (nCl/ml) i Sample Type !I Studoe-LP W I Descnoten il Septic Tank A il Results ' Normalized Nuclide Notes uC/mi to 100 % Nuclides RATIOS 1f~ H-3 > WSS THAN i 1.011 -04 56.5852%P&//M#M#MW/MM#/A%\%%%\%%%%%\Wi' 21 C- f 4 i GSS THAN I 1.42E-05 7.9555%1MMMMM#/M//#/MNAn%%%%%%%\uwW l 31 Fe-56 WSS THAN : 2.07E-06 1.1597 % Fe-55/Co-60 10.61538, l 4) Ni-59 i GSS THAN I 2.13E-061 1.1933 % //MM///M/#M/#M/M#AWW%%%%%%%%\W I 5i Ni-83 ! i 1.79E-05 10.0265%lfM/M#M/MM/MMMNAh%%%%\%%%%%%y ) 61 Sr-89 i WSS THAN I 2.40E-06 1.3446 %PM/#/#M//MM/MMM/A%\%%%%%%%%%%f 7 11 Sr-90 GSS THAN 1.75E-06 0.9604% j St-90/Cs- 137 8.53659i 8! Nb-94 WSS THAN 1.80E-07 0.1006%i'/#MMW/#MMMW/#/A%%\\\WW\\\n\\%%\t 9N Tc-99 WSS THAN l 7.96E-06 4.4596 %)M/#MMM//M/M/MMAu\%%%WWW\%%%\t 10 1 t- f 29 i GSS THAN 5.24E-06 2.9357 %P//M#/MM////MMMM/A\WWWWW\%%\B%%j 11 I No-237 WSS THAN 2.25E-07 0.1261 % fMMMu/#/MM/M#//NAu\M\\\\n\\\\%\\\\w) 12 Pu-238 WSS THAN i 9.06E-06 0.050e% Pu-230/Am-241 1.37481 131 Pu-PSGM40 WSS THAN I 7.85E-06 0.0440 % 'MMMM//M/#MMM#A\%%%%%%%%%%W
- 14) Pu-24 f WSS THAN I 8.47E-06 4.7453 % Pu-241/Am-241 126.52807 15l Pu-242 WSS THAN 2.03E-OS 0.0114 % '//#M#MM#M/M/MMA%\%%%%%\%%%%Y
.O 16! Am-24 f WSS THAN 6.59E-oe 0.0369 % Am-241/Pu-238 0.727371 I
I ( 171 18! Am-243 Cm-242 WSS THAN WSS THAN 8.88E-00 f 3.12E-06 0.0050 % f/#//##MMM#//W#M/A%%%\%%%%%%%%1 0.0175% '//##MM////#/MM#/#/AWWWWWWW\%%%%1
)
19! Cm-244/@ WSS THAN I 3.44E-06 0.0193 % '/MN/#NMW/#$M/#/AWWW\%%%%%%%W, 20lL U-234 h GSS THAN ! 4.13E-06 2.3136% 'l//M##M//##//M##M/A\WWWWWu%%%%Wj-211! U-235 !! GSS THAN 2.07E-06 1.1597% 'MM//#l/M/#//#M/# MAW %%\WWWWWWWW i ! 22: U-238 li WSS THAN 2.07E-06 1.1597 % U-238/Am-241 31.41123-23t Co-do il GSS THAN 1.95E-07 i 0.1092% Am-241/U-238 0.031841 24! Mu- 108 i WSS THAN 1.e6E-06 I 0.9300 %1f/MMMMM////MMW//A%%\%%%%%%%%W 25t As-100m i GSS THAN 1.83E-07 ; 0.1025% fMMMW//#/#MMM#!Ahuu\wwwnhWWW i 26! Sb- f 25 i WSS THAN 4.56E-07 0.2555 % 'WMMW//##/MMM///A%%%%%\\\WuWWW 27I Cs- f S4 WSS THAN 2.01 E-07 0.1126 % Co-134/Cs-137 0.90049 l 260 Cs-f37 WSS THAN 2.05E-07 0.1149% Co-60/Cs-137 0.95122
- 29! Co- f 44 WSS THAN 9.51 E-07 0.5328% Cs-137/Co-60 1.05128 l 30) Eu-f52 ! WSS THAN 1.49E-06 0.8348 %"MMM//M/#MM//M#/A%%%%%%\wwwww
! 316 Eu-f54 i WSS THAN 5.70E-07 0.3193 % 'M//M/M/M/M//#/////#A%%%%%%%%%\WW 32t Eu-f55 i
.ESS THAN 4.56E-07 0.2555% 1/#NM##M##/M/#M/An%%\WWWWW\%%Y l "otais = = > R 1.78E-04 tl 100.00% X l
l i 1
\ (
z
\
l e 1 I
i [\ / B & W RESULTS 9/14/95 - Saxton Samples . . SYSTEM = = = > Septic Tank A, Sludge (Solid Phase), AREA 8 1 Santon 4 8& W d 0.0042l J Sampde Numboe SamWe Number Total TRU Content _(nCi/g) ISX8ese50048 4 9507002-17AC l 0.00130 - Total Uranium Content (nCi/g)
! Sample Type 4 Studge-SP ll l Descriptum :l Septic Tank A ll Results Normalized Nuclide Notes uCi/a to 100 % Nuclides RATIOS 1h H-3 i GSS THAN i 1.00E-04 38.4563% f///////////////////////////////A\WWWWWWW\W\Wut 2f C-f4 ! WSS THAN I 9.82E-06 3.7764% '///////////////////////////////Aguwinuw%%%%%1 3lI Fe-55 WSS THAN I 4.92E-06 1.8920 % Fe-56/Co-60 4.59613!
4il Ni-59, LESS THAN I 1.27E-05 4.8839% J///////////////////////////////An%%\n%%%%%%%1 5! Ni-63 I I 8.49E-05 32.6494 % ?///////////////////////////////AWW%%%%\%\%%\%) 6! St-89 i WSS THAN i 4.37E-06 1.6805 % ?///////////////////////////////AWWWWWWWuhWWY 7! Sr-90 i GSS THAN 3.18E-06 1.2229 %! Sr-90/Cs-137 i O.15071 Si Nb-94 i LESS THAN 2.04E-07 0.0785% J///////////////////////////////AW\\\\\%\hbu\Wu\Y , 9I Tc-99 i WSS THAN 2.53E-06 0.9729% J///////////////////////////////A\\\\\\\\\\\\\\\\h\\\\\] l 106 /-f29 ! LESS THAN 3.62E-06 1.3921% J///////////////////////////////AWW\\WWWW\Whuhl til No-237 i LESS THAN 1.65E-07 0.0635% F///////////////////////////////A\\\\\%\uw\\n\\\hwi 12t Pu-238 i GSS THAN 4.30E-08 0.0165 %I Pu-238/Am-241 0.91684 l 13r Pu-239/240 ! LESS THAN 4.70E-08 0.0181% U///////////////////////////////nWWWWuhunghhy 1411 Pu-24 f H WSS THAN 3.86E-06 1.4844 %I Pu-241/Am-241 82.30277 3 1 511 Pu-242 li GSS THAN 9.39E-09 0.0036 % '///////////////////////////////h\h\\%\\MW\%%%%Y r \ 16ll Am-24 f 4 ESS THAN 4.69E-08 0.0180 % Am-241/Pu-238 1.09070 (/ 17 cam-243 181 Cm-242 ll il LESS THAN LESS THAN 7.32E-09 3.26E-06 0.0028 %J///////////////////////////////AWW%%%%%%%%\W) 0.0125 % 19( Cm-243/244 W LESS THAN 2.74E-08 0.0105 % F///////////////////////////////AWuhw\whwuuhuj) F///////////////////////////////A%%%\%%%%%%%% 20: U-234 W LESS THAN I 5.64E-07 0.2169 %J///////////////////////////////AW\\\M\W\\\%%\\\\\\j 21i U-235 i WSS THAN I 2.92E-07 0.1123 %"///////////////////////////////AW\n\WM\\%\%\\\\%) 22! U-236 i LESS THAN i 4.39E-07 0.1688 % U-236/Am-241 9.36034l 234 Co-60 I i 1.07E-06 0.4115 %. Am-241/U-238 0.10683! 24!i Ru- 106 i LESS THAN 1.85E-06 0.7114 % '///////////////////////////////Au\M\WWWWuhuwY 2SJ Ag-106m I LESS THAN 1.97E-07 0.0758 % '///////////////////////////////AWW\%\\\%%%%%\\Y 261 Sb-f25 ' LESS THAN 5.58E-07 0.2146 % '///////////////////////////////A%\\\\W\n%%%%%%) 27i Cs-134 l LESS THAN 2.20E-07 0.0846 % Cs-134/Cs-137 I 0.01043! 28! Cs-f37 l 2.11 E-05 8.1143 % Co-60/Cs-137 I 0.05071! 29! Ce- f 44 LESS THAN 5.88E-07 0.2261 %! Cs-137/Co-60 l 19.71963! 30t Eu- f 52 i GSS THAN 1.76E-06 0.6768 % '///////////////////////////////AWh\\n\WWWWWWBt 31 ll Eu-f54 ! LESS THAN 6.65E -07 0.2557 % '///////////////////////////////AWWW\\\ hun %%%%% : 32ll Eu-f55 _J LESS THAN 2.50E-07 I 0.0961 % '///////////////////////////////A\MuuMWWWMuhut LTotals = = > W 2.60E-04 :1 100.00 %) l i
\v) !
l l l l
i l . B & W RESULTS 9/14/95 - Sexton Samples SYSTEM = = = > Septsc Tsnk B. Sludge (Liqund Phase) ARE4 6 l Seaton il BA W D 0.00131 iSamode Numbed Samode Aumber Total TRU Content (nCi/mQ
.SN e9 W 9507002-18AB h 0.00011 Total Uranium Content (nCi/mQ i Sample Type il Studge-LP !l
, I i
il Descnotion !! Sectic Tank B ll Results Normalized Nuclide Notse uC4/mi to 100 % Nuclides RATIOS 1 ll H-3 LESS THAN 1.01E-04 6 47.9669 % f//////#////###/##/#/////Ah\M %\%%%%\uw 24 C- f 4 WSS THAN 5.10E-061 2.4231 % '//##///#/#//#/##/##///A\%%%%g%%%%%% 3 11 Fe-55 WSS THAN 7.47E-071 0.3549 % Fe-55/Co-60 4.26857 [ 44 Ni-59 i ESS THAN ' 1.21E-061 0.5749 % '###/##////#/#/##///#/A%%%\\%%%%%\WW1 ! SQ Ni-63 1 6.63E-05 i 41.9529 %P/#/#//###////#//####/A%%\%%%%%%%%%1 6ll St-89 WSS THAN 6.49E-071 0.4034 %F##/#///#///#/##/#///#/A\\\%\\\W\%%%\\\h% 7' Sr-90 1.00E-06 I 0.4751 %I Sr-90/Co-137 7.09220 6t Nb-94 WSS THAN 1.35E-07 i 0.0641 %P/##/###///#/#///////#//A\Mi\n\\WWW\M\\h% 9 !! Tc-99 4 WSS THAN 4.32E-061 2.0525%F/##/####//##/####//AW\%%%%\%%%%%y 10} l-129 i GSS THAN 1.64E-06 I 0.6742% P/##/#//#///#/##/////////A\\h\n\n\\%%%%%g 11! No-237 ESS THAN 3.05E-06 i 0.0145%,'/####/####///#//###n\%%%%%\%%%%\Wj 12} Pu-238 ESS THAN 6.39E-09 i 0.0030 %I Pu-2SS/Am-241 0.39689 l 13i Pu-PSS/240 WSS THAN 9.04E-09 i 0.0043 %F/#/##/####//##/###/n%%\WWWh%%%%W l 14il Pu-24f ESS THAN 1.17E-06 I 0.5559%l Pu-241/Am-241 72.67061 l j] j 15li Pu-242 WSS THAN 9.04E-091 0.0043%P/#//##/##########/A%%%%M MughM l l (U 16l 171' Am-Nr Am-243 i WSS THAN WSS THAN 1.61 E-06 ! 2.79E-09 ! 0.0076 %I Am-241/PU-238 0.0013%P##///###///##/#/##/#Ah\MhWWM\%%\Wt 2.519561 l 18t Cm-242 i ESS THAN 1.03E-06 l 0.0049 %F###/#/##/#/#//#/##//Aw\%%%%%%%%%$ j 19! Cm-243/244 i WSS THAN 1.50E-06 I 0.0071 %PN///##//###//#/###/#n\WMMW\WMW\\\%j l 20! U-234 i WSS THAN 2.40E-06 l 0.0114 %J###/#//##///####/#//nWMWWW\WMW) . 211 U-235 _ WSS THAN 2.90E-061 0.0136 %P/##/###//////###//##/A\WWM%%%%%%%1 I 22 !. U-2Ss WSS THAN 1.10E-06 I 0.0052%l U-238/Am-241 0.683231
- 23) Co-op WSS THAN 1.75E-07 ~ 0.0631 %I Am-241/U-238 1.46364 24, Ru-f06 WSS THAN 1.23E-06 l
0.5844 %J////#####//#/#///////#/AM\%%%MMWW 25 Aa- 10pm WSS THAN 1.61 E-07 0.0765 %F////##//#/#//##////###A\\\M\\n\Mww\\Y ! 26; Sb- 125 WSS THAN 3.osE-07 0.1463 %1/###/##!/////#//##//#/A% M W M M WW\W i 27! Cs-f$4 WSS THAN 1.46E-07 0.0694 % Cs-184/Ce-137 1.03546 l ! 26t Os-137 LESS THAN 1.41 E-07 0.0670 % Co-60/Cs-137 1 1.24113 29! Co-f44 LESS THAN 6.13E-07 0.2912% Cs-187/Co-80 1 0.80571 i 30t Ju-152 WSS THAN 1.15E-06 0.5464 %F/#/#//#//#/#####////#A%%%%%\%%%BMW ! 31r Ju-f54 WSSTHAN 4.52E-07 0.2146%F/###//####////#//#/#/AW%%\%%%%%%%%11 32!L Ju- f85 .ESS THAN 2.64E-07 0.1254% U#/////#/##/#/////##/#//A%%%%%%%%%%%%I l l "otals = = > il 2.10E-04ll 100.00%j l i t t i v i
l- . . i l l 1
/'m\
t 1 B & W RESULTS 9/14/95 - Saxton Samples - l
- SYSTEM = = = > Septic Tank B, S/udge (Sohd Phase). AREA 8 l
1 Samfon 11 S A ter i ll 0.0106 ll l
@imode NumbetSamsde Number l Total TRU Content (nCilg) iOX8650500de (_9507002-18AC i 0.0012H Total Uranium Content (nCyg)
.1 Sample Type 9 Studge-SP #
!i Descnotion 4 Septic Tank B i i
Results Normalized Nuclide Notes UCVa to 100 % Nuclides RATIOS 1 H-3 WSS THAN 9.33E-05 2 C- f 4 37.7797%P//////////////////////////////Th\\\\n\%%%%%%\\\\1 l i WSS THAN I 1.03E-05 4.1707% '///////////////////////////////AWWW\WWWWWWWut j 3 Fe-55 i WSS THAN I 2.21E-05 8.9489 % N -55/Co-60 i 17.96748I 4i Ni-59 WSS THAN I 9.82E-06 3.9764 % '//////////ini/////////////////A\%\\\%\n\W\\\W\\%) Si Ni-63 1 1 5.04E-05 6t 20.4063%////////////////////////////////AWWW%%%%%%\Why Sr-69 i WSS THAN i 4.12E-06 7! 1.6683%F///////////////////////////////A%%%\%%%%%%%%1 St-90 1 WSS THAN ! 3.00E-06 1.2148 %I Sr-90/Cs-137 I 8i Nb-94 O.09375i i WSS THAN ' 1.01E-07 9! Tc-99 0.0409 %F///////////////////////////////A\\\%\%%%%%\\u\\\\ i WSS THAN ' 3.38E-06 1.3687%F///////////////////////////////Ahu%%%%%%%\WWj 101 /- ff9 l WSS THAN 6 1.97E-06 0.7977 %////////////////////////////////Aw\%\%\\%%\%\W\%" 111 No-237 WSS THAN : 1.64E-07 0.0664 % f////////////////////////////////nw\%%\\whWW\n\Y i
- 12) Pu-238 i 5.56E-06 2.2514 %! Pu-238/Am-241 62.82486 l
13ll Pu-239/240 WSS THAN i 9.61E-06 0.0389%F///////////////////////////////Au\%%%\%\WW\%%Y l 1411 Pu-24 f WSS THAN i 4.59E-06 1.8586%l Pu-241/Am-241 51.864411 15ll Pu-242 WSS THAN I 1.09E-06
./\ m 16i Am-24 f MSS THAN i 8.85E-06 0.0044% p///////////////////////////////Ahw%%\%%%\n%%\1 0.0358 %I Am-241/Pu-238 0.01592l 17h Am-243 WSS THAN I 1.21E-06 0.0049% F///////////////////////////////A\huhWWWWWWWW1
- 18) Cm-242 WSS THAN i 2.70E-06 /
0.0100%i ////////////////////////////////n%%\\%%%%%%%%1, 19! Cm-243/244 WSS THAN i 3.64E-06 0.0147% F////////////////////////////////n%%\\\WW\\\W\\\\Wj 20t U-234 WSS THAN i 4.58E-07 0.1655 %F////////////////////////////////n%%%%%%%%\\WW 21t U-235 i GSS THAN i 2.81E-07 0.1138 % '////////////////////////////////n%%%%%%%%\uuY 22: U-238 i WSS THAN i 4.15E-07 0.1680 % U-236/Am-241 4.68927
- 23) Co-60 1.23E-06 0.4961 % Am-241/U-238 0.21325l 24} Au-fos ESS THAN I 9.94E-07 0.4025 %F///////////////////////////////A%%%%%%%%%%%%
25! An-708m WSS THAN I 1.06E-07 0.0429 %P///////////////////////////////A%%%%%%%%%%%% 26; Sb- f 25 WSS THAN i 3.36E-07 0.1361 % '////////////////////////////////n%%%%%%ku%%%1 27t Cs- f 34 WSS THAN 1 1.12E-07 0.0454 % Co-184/Co-137 0.00350l 28! Cs- 137 __ I 3.20E-05 12.9577 % Co-60/Cs-137 0.03644 l 291 Ce- f44 i WSS THAN i5.53E-07 0.2239 % Cs-137/Co-60 26.01626I 30s Eu-152 i WSS THAN I 8.68E-07 0.3515 % '///////////////////////////////Anw\%%%\%%%%V'\ 31i Ju-154 WSS THAN ! 3.26E-07 0.1320 % '///////////////////////////////Au%%%%%%\u%%%\ 32'l Ju-f55 WSS THAN I 2.03E-07 : 0.0622% '///////////////////////////////A%%%%%%%%\W%%1 1 Totale = = > il 2.47E-04 ji 100.00 %1 3 V
B&W NUCLEAR ENVIRONMENTAL SERVICES, INCORPORATED NUCLEAR ENVIRONMENTAL LABORATORIES l DATA REPORT NUMBER 9509065 l 10CFR'61/ CHEMICAL ANALYSIS RESULTS i l l Submitted to GPU Nuclear Corporation Saxton Site - Decommissioning Project November 2,1995 Released by: 6(E b IA / //2.[9f James L. Clark, Project Manager Nuclear Environmental Laboratories O .
1 l GPU Saxton Data Report November 2,1995 DATA REPORT
SUMMARY
Laboratory Project Summary e Plant Site: Saxton Decommissioning Project
- Shipment Number: 95-006 ,
o Receipt Date: September 14,1995 i e NEL COC Number: SN-12306, SN-12307 l e NEL SDG Number: 9509065, 9509066, 9509067
- NEL Proposal: NEL5-1258 e Contract Number: 0397355 e Analysis Service: Routine
- Laboratory QC: Routine - Batch O -
i
O GPU-Saxton Data Report - November 2,1995 Sample Delivery Group Summary l Customer ID NEL Sample ID Sample Type . Description SX865950053 9509065-01 S-INJ Pipe Section item-1 (Leach) SX865950053 9509065-01-AA S-INJ Pipe Section item-1 (Metal) SX865950069 9509065-02 S-INJ Pipe / Table B-6 Item-3 (Leach) SX865950069 9509065-02-AA S-INJ Piperrable B-6 Item 3 (Metal) SX835950085 9509065-03 CC River Water Pipe item-7. (Leach) SX825950086 9509065-04 SGBD Line (Bonom) ltem-8 (Leach) SX825950087 9509065-05 SGBD Line (Surface) Item 9 (Leach) SX866950116 9509065-06-AA RX Cavity Steel item-13 (Metal) SX865950118 9509065-07 Make-Up Demin Pipe item-14 (Leach) SX815950068 9509066-01/01 AA Tank Sediment item-2 SX8C6950083 9509066-02AA Paint Chip Comp Item-5 SX8C6950084 9509066-03AA Paint Chips - SW l' tem-6 SX866950115 9509066-04/04AA Storage Well Resin item-12 (Lig) SX866950115 9509066-04AB Storage Well Resin item-12 (Solid) SX835950071 9509067-01 SG Scrape item-4 SX825950110 9509067-02 SG Manway Scrape item-10 (NA) SX825950111 9509067-03 SG Bowl Smear item-11 Note that "NA" in the preceding Table indicates that sample was not analyzed. This was due to sample being collected on Duct-Tape and being of inadequate total mass for removal and analysis. O ii
. _ _ _ . _ _ _ _ _ . _ _ _ . _ _ _ . _ . . _ _ _ _ _ . _ . . _ . _ _ _ _ . _ . - . . . _ . ~ . . ..___.._-_.m_.. __ __ _ . _ _ _ _ _ . .
- GPU.Saxton Data Report November 2,1995 Pipe Sections / Physical Characteristics Summary l NEL Sample ID Length (in) OD (in) ID (in) 9509065-01 9 3/4 3.519 2.614 9509065-02 13 3.512 2.610 9509065-03 10 1/8 4.576 4.005 9509065-04 11 7/8 2.380 1.906 9509065-05 11 1/8 1.925 ~1.502 9509065-07 8 5/8 1.910 1.660 l Pipe Section Leach / Decon Summary l NEL Sample ID % Activity Removed ('*Co) % Activity Removed (*Cs) 9509065 01 86 i8 65 6
; 9509065-02 0 0 .
9509065-03 0 17
- 7 9509065-04 52 i 59 18 4 9509065 05 0 52
- 3 9509065-07 0 0 O
iil
~
O GPU Saxton Data Report November 2,1995 l: _ se
.: Sample"ANguot Sdmmary" '~
- l NEL Sample ID. iMatrix! i$Volumo#l' Density '
IPipe Mass / Removed'(g) l 9509065-01 Acid Leach 1109 mL /1.08 g/mL 5492 / 0 9509065-02 Acid Leach 1158 mL /1.15 g/mL. 7328 / 0 9509065-03 Acid Leach 4095 mL /1.08 g/mL 4236 /134 9509065-04 Acid Leach 953 mL /1.14 g/mL 2167 / 58 9509065-05 Acid Leach 965 mL /1.08 g/mL 1479 / 34
'9509065-07 Acid Leach 503 mL /1.08 g/mL 891 / 2 NEL Sample ID Sample' Matrix ' Mass / Vslume ' Density 9505065-01 AA Metal Solids NR NR 9509065-02AA Metal Solids .NR NR 9509065 06AA Metal Solids NR NR 95090CS-01/01 AA Wet Sediment 3.27 g 0.93 g/cc 9509066-02AA Paint Chips 15.02 g 1.67 g/cc 9509066-03AA Paint Chips 15.11 g 1.51 g/cc 9509066-04/04AA Resin Liquid 62 mL 1.01 g/mL 9509066-04AB Resin Wet Solids 3.98 g 0.99 g/mL 9509067-01 Scrape Solids 0.504 g 0.84 g/cc 9509067-02 Scrape Solids NA NA 9509067-03 Smear 0.922 g 0.46 g/cc Note triat "NA" in the preceding Table indicates that sample was not analyzed. This O was due to sample being collected on Duct Tape and being of inadequate total mass for removal and analysis. "NR" indicates.that the listed analysis was not requested.
iv
j . ] GPU-Saxton Data Report November 2,1995 i Sample Analysis Summary l j Customer ID NEL Sample ID Sample Matrix Notes ! SX865950053 9509065-01 Acid Leachate (1) { SX865950053 9509065-01-AA Metal Shavings (2) l SX865950069 9509065-02 Acid Leachate (1) SX865950069 9509065-02-AA Metal Shavings (2) l SX835950085 9509065-03 Acid Leachate (3) 1 j SX825950086 9509065 04 Acid Leachate (1) ; i SX825950087 9509065-05 Acid Leachats - (1) SX866950116 9509065 06-AA Metal Shavings (2) j SX865950118 , 9509065-07 Acid Leachate (1) l SX815950068 9509066-01/01 AA Wet Sediment Solids --- SX8C6950083 9509066-02AA Paint Chips (4)
- SX8C6950084 9509066-03AA Paint Chips (4)
SX866950115 9509066-04/04AA Resin Liquids l (5) l SX866950115 9509066-04A8 Wet Resin (5) 2 SX835950071 9509067-01 Scrape Solids --- SX825950110 9509067-02 Tape Adhered Solids (6) SX825950111 9509067-03 Smear Solids --- O - V
OPU Saxton Data Report November 2,1995 Sample ~ Analysis Summary Notes-l Note-1 Acid leaching with 30% HNO 3 / 5% hcl for 8-24 hour period. Leach solution subsequently analyzed includes small volume rinses. Note-2 Metal Shavings collected from multiple locations to yield representative sample with minimal surface contamination. - Note-3 Leach solution combined with significant loose solid deposits. Note-4 Asbestos Analysis performed by B&W NESI Approved Sub-Contract' Laboratory. Note-5 For 10CFR61 Analysis: Solid / Liquid phases separated by filtration. Solid analyzed on wet weight basis. For TCLP Analysis: Sample analyzed as mixed Solid / Liquid Matrix. Note-6 Sample not analyzed due to inability to remove from tape adhesive upon which it was collected. [ Analytical Results See Attached Report Sheets vi
-- . . ~ , . . . . - _ _ - - - _ - - _ _ _ _ _ _ _ - - - _ - - - - - - _ - _ - _ - - - _ - _ - - - - - - - - - - _ - - - - - _ - _ - _ - _ _ _ _ _ _ _ _ _ - - - - - _ - _ _ _ _ - - - - - - _ - - - _ - _ - - - - _ - - _ - - - - - _ _ _ _ _
'\
U) Saxton Results Summary NEL ID 9509065 -01 9509065 -02 9509065 -03 ACTIVITY 2S UNC ACTMTY (uCi/g Removed) 2SUNC ACTMTY 2S UNC H-3 < (uCi/g Removed) (uCi/g Removed) 8.37E-03 --
< 8.74E-03 C-14 <
- < 2.31E-04 1.19E-02 --
< 7.86E-03 Fo-55
- < 7.36E-04 5.27E+00 8.56E-01 5.58E-03 Ni-59 1.07E-03 < 1.83E-05 1.14E+00 1.68E-01 < 2.05E-04 Ni-63
- < 6.17E-06 -
9.12E+01 1.33E+01 1.44E-02 Sr-90 2.14E-03 4.10E-05 6.48E-06 5.91E-01 7.23E-03 < 1.39E-04 Nb-94 9.84E-06 226E-06 1.65E-02 1.14E-03 < 2.81 E-03 < Tc-99 <. 1.39E-05 - 5.86E-04 - < 2.91E-04 < l-129 < 8.95E-06 - 1.41E-03 - < 1.57E-03 < 127E-05 - Np-237 < 3.80E-01 - < 4.31E-05 < 1.61E-06 - Pu-238 227E+00 2.44E-01 6.32E-05 129E-05 < 4.00E-07 - Pu-239/240 4.99E+00 4.08E-01 123E-04 1.85E-05 < 6.81E-07 - Pu-241 5.72E+01 6.57E+00 2.61 E-03 322E-04 < 4.03E-05 - Pu-242 < 2.35E-02 - < 2.86E-06 < 1.15E-07 - Am-241 7.75E+00 5.03E-01 1.96E-04 1.58E-05 1.02E-06 5.06E-08 Am-243 < 1.52E-02 - < 9.41E-07 - < 4.77E-08 - Cm-242 < 8.11E-02 - < 1.02E-06 - e 1.73E-07 - Cm-243/244 < 9.62E-02 <- 4.98E-06 < 1.65E-07 - U-234 < 2.36E-02 - < 4.96E-06 - < 9.11E-07 - U-235 < 6.54E-03 - < 7.01E-06 - < 2.57E-07 - U-238 < 1.13E-02 - < 3.51 E-06 - < 4.19E-07 - Co-60 3.58E+01 6.54E-02 3.12E-03 3.01E-04 8.71 E-06 9.05E-07 Ru-106 < 1.34E-01 - < 8.14E-04 - < 1.63E-06 - Cs-134 <. 2.33E-02 - 6.62E-05 2.78E-05 < 2.17E-07 - Cs-137 8.30E-01 1.85E-02 3.16E-01 4.47E-02 4.80E-06 6.97E-07 C3-144 < 4.34E-02 - < 6.31E-04 - < 8.34E-07 - Am-241 1.05E+00 3.68E-02 3.37E-04 2.52E-04 1.06E .06 9.29E-07 Ag-108m < 1.45E-02 - < 8.71E-05 - < 1.64E-07 - Eu-152 < 1.07E-01 - < 2.45E-04 - < 1.89E-06 - Eu 154 6.65E-02 3.30E-02 < 5.04 E-05 -- < 5.07E-07 - Eu 155 1.70E-02 1.36E-02 < 3.06E-04 - < 4.22E-07 -
< 3.17E-02 < 4.33E-04 < 4.31E-07 Str 125 --- - -
3
O Saxton Results Summary O O NEL ID 9509065 -04 9509065 -05 9509065 -07 ACTMTY 2S UNC ACTMTY 2S UNC ACTMTY 2S UNC (uCi/g Removed) (uCilg Removed) (uCL/g Removed) 4 H-3 < 1.24E-04 - < 2.14E-04 -- < 1.90E-03 - C-14 < 1.36E-04 - < 3.01E-04 - < 2.33E-03 - Fo-55 < 3.99E-06 - < 9.37E-06 - 7.39E-04 1.31E-04 Ni-59 < 1.32E-05 - 3.77E-06 5.52E-07 1.77E-04 4.91 E-05 Ni-63 1.41E-04' 2.15E-05 2.21 E-04 3.34E-05 1.65E-02 2.42E-03 Sr-90 5.03E-06 1.21E-06 < 4.40E-06 -- 1.00E-03 1.12E-04 Nb-94 < 4.67E-06 - < 4.29E-06 - < 4.78E-04 - Tc-99 < 4.35E-06 - < 7.12E-06 - < 6.61E-05 -
< 1.53E-05 < 1.74E-05 < 3.62E-04 I-129 - -
< 1.13E-06 < 1.98E-06 < 7.82E-05 -
Np-237 - - Pu-238 2.37E-06 4.07E-07 < 1.13E-07 - 5.98E-04 5.58E-05 3.96E-06 5.46E-07 < 3.92E-07 - 1.72E-03 1.18E-04 Pu-239/240 4.60E-05 5.78E-06 < 4.09E-05 - 2.42E-02 2.58E-03 Pu-241
< 1.13E-07 < 4.17E-Gd -
Pu-242 < 7.46E-08 - 3.M E-06 3.11E-07 1.59E-05 1.02E-06 1.91 E-03 9.45E-05 Am-241
< 2.98E-08 - < 2.29E-06 -
Am-243 < 2.6 ? E-08 -
< 9.17E-08 - < 1.41E-05 -
Cm-242 < 7.31 E-08 -
< 1.60E-07 - < 1.71E-05 -
Cm-243/244 < 1.07E-07 -
< 3.12E-07 < 1.77E-05 -
U-234 2.73E-06 3.20E-07 -
< < 1.04E-07 - < 4.57E-06 -
U-235 1.36E-07 -
< 4.77E-07 - < 4.57E-06 -
U-238 - 2.33E-06 2.88E-07 3.35E-05 3.29E-06 2.99E-03 5.43E-05 Co-60 3.98E-05 4.14E-06
< < 1.46E-04 -
< 6.52E-06 - 1.33E-06 ,
Ru-106 < - < 2.19E-05 - Cs-134 5.59E-07 3.55E-07 - 523E-03 5.88E-05 1.51E-03 1.73E-04 5.70E-05 7.55E-06 Cs-137 < 8.09E-05 -
< 1.39E-06 - ,
Co-144 <' 5.06E-06 - 1.69E-04
< 1.31E-06 - 2.61E-03 Am-241 2.92E-06 1.90E-06
< 1.54E-05 -
< 6.93E-07 - < 3 01E-07 -
Ag-108m < 1.97E-04 -
< 3.35E-06 - < 3.52E-06 -
Eu.152 < 6.02E-07 - 1.13E-04 3.90E-05 Eu-154 < 6.90E-07 -
< 2.92E-05 -
< 6.93E-07 -
Eu-155 < 2.38E-06 -
< 5.66E-05 -
< 3.30E-06 -- < 9.28E-07 -
Sb-125
V Saxton Results Summary ' NELID 9509065 -01AA 9509065 -02AA 9509065 -06AA ACTMTY . 2S UNC ACTMTY 2S UNC ACTMTY 2S UNC (uCi/9) (uCi/g) (uCi/g) H-3 < 6.32E-05 < 6.06E-05 -- < 2.63E-05 C-14 < 4.08E-04 < 2.39E-04 - < 1.23E-04 - Fe-55 < 4.08E-08 - < 3.68E-06 4.98E-06 2.08E-06 Ni-5.9 1.62E-05 - 4.84E-06 4.00E-06 1.66E-05 3.63E-06 Ni43 3.33E-04 5.00E-05 4.47E-04 6.68E-05 5.25E-04 7.78E-05 Sr-90 < 3.78E-06 - 2.86E-05 5.06E-06 < 1.56E-06 - Nb-94 < 1.20E-06 - < 9.75E-07 < 5.61E-07 - Tc-99 < 4.44E-06 - < 4.11E-06 < 2.05E-06 - I-129 < 4.39E-06 - < 6.92E-06 - < 6.22E-06 - l Np-237 < 1.25E-06 - < 8.52E-07 -
.< 7.27E-07 Pu-238 <
)
3.52E-07 - < 2.95E-07 - < 3.79E-07 - 1 Pu-239/240 < 6.96E-07 - < 3.97E-07 - < 4.78E-07 - Pu-241 < 3.82E-05 - < 3.20E-05 - < 2.13E-05 - Pu-242 < 1.06E-07 - < 8.88E-08 - < 5.07E-08 - Am-241 < 5.64E-07 - < 1.98E-07 - 9.85E-07 628E-08 Am-243 < 6.90E-08 - < 4.42E-08 - < 1.14E-08 - Cm-242 < 7.90E-08 - < 1.52E-07 - < 5.37E-08 - l Cm-243/244 < 1.95E-07 - < 1.17E-07 - < 6.63E-08 - l U-234 < 1.55E-07 - < 8.31E-07 - < 4.73E-08 - I U-235 < 6.95E-08 - < 5.76E-07 - < 1.50E-08 - l U-238 < 6.95E-08 - < 4.97E-07
- < 2.12E-08 -
Co-60 1.43E-04 1.32E-05 1.82E-04 5.30E-06 6.08E-04 5.38E-05 - l Ru-106 < 1.11E-05 - < 1.74E-05 - < 1.11 E-05 - I Cs-134 < 1.51 E-06 - < 2.40E-06 - < ' 1.60E-06 - Cs-137 3.91 E-06 1.51 E-06 7.19E-06 1.90E-06 1.94E-06 9.96E-07 Co-144 < 4.92E-06 - < 7.57E-06 - < 4.62E-06 - Am-241 < 1.11 E-06 - < 6.40E-06 - < 2.20E-06 - Ag-108m < 1.06E-06 - < 1.75E-06 - <- 1.10E-06 -
< 1.32E-05 < 2.30E-05 - < 1.48E-05 -
Eu-152 -
< 3.38E-06 < 4.54E-06 - < 2.59E-06 -
Eu-154 -
< 1.96E-06 < 3.61E-06 - < 1.99E-06 -
Eu.155 -
< 2.61 E-06 < 4.02E-06 - < 2.53E-06 -
Sb-125 + -
Saxton Results Summary
- NELID 9509066 -01 9509066 -04AA 9509066 -04AB ACTMTY 2S UNC ACTMTY 2S UNC ACTMTY 7S UNC (uCi/g) (uCi/mL) (uCi/g)
H-3 7.76E-04 120E-04 2.90E-04 E4 AE-05 3.37E-04 ti.48E-05 C-14 < 1.02E43 - < 1.16E-04 - < 1.96E-03 - Fe-55 1.19E43 2.28E-04 - 1.23E-05 2.84E-06 228E-03 4.36E-04 Ni-59 1.47E42 2.16E-03 3.52E-05 520E-06 1.32E-03 1.95E-04 Ni-63 9.93E41 1.45E-01 2.81E-03 4.10E-04 - 1.32E-03 1.95E-04 St-90 1.44E-02 3.30E-04 2.92E-04 3.74E-06 1.01E-01 1.47E-02 Nb-94 < 1.47E-04 - 1.05E-06 - 7.32E-03 < 5.66E-06 - Tc-99 < 3.42E-05 - < 3.38E-06 - < 2.55E-05 - l-129 - < 4.86E-06 - < 1.51E-06 - < 8.54E-06 - Np-237 < 1.20E-04 - < 1.02E-07 - < 7.34E-06 -- Pu-238 9.19E-04 7.22E-05 9.64E-07 7.58E-08 1.23E-04 9.70E-06 i Pu-239/240 2.09E-03 1.35E-04 1.74E,06 1.17E-07 2.73E-04 1.7/E-05 i
' Pu-241 3.83E-02 3.98E-03 2.88E-05 3.04E-06 4.02E-03 #.24E-04 l Pu-242 < 3.86E-06 < 4.06E-09 - < 528E-07 -
l Am-241 1.87E-03 924E-05 1.16E-06 5.76E-08 3.36E-04 1.98E-05
~
< 2.40E-09 < 3.94E.07 [
Am-243 < 2.81E-06 - - - I
< 1.82E-05 < 1.08E-08 < 1.87E-06 -
Cm-242 - -
< 2.63E-05 < 1.70E-08 < 3.65E-06 - f Cm-243/244 - -
U-234 < 1.87E-05 - < 3.51E-07 - < 1.52E-07 - [
< 1.03E-07 < 8.80E-08 -
U-235 < 7.48E-06 - -
< 7.31 E-08 < 7.18E-08 -
U-238 < 6.48E-06 - - 4.92E-04 5.60E-05 7.11 E-03 3.56E-04 l Co-60 - 2.32E-01 322E-03
< 6.62E-03 l Ru-106 < 2.56E-02 - < 4.39E-03 - -
1.47E-03 < 1.88E-04 - 6.91E-04 2.04E-04 Ca-134 4.51E-03 1.85E+00 2.60E-03 3.75E+00 5.62E-03 i Cs-137 1.49E+01 2.02E-02 ,
< 3.34E-03 - < 4.49E-03 -
Co-144 < 2.43E-02 -
- I
< 2.15E-03 - < 2.13E-03 -
Am-241 < 1.33E-02 -
< 4.35E-04 - < 6.60E-04 -
Ag-108m < 1.99E-03 -
< 8.87E-04 - < 1.49E-03 -
f u 152 < 1.19E-02 -
< 2.45E-04 - < 3.70E-04 -
l'u 154 < 1.99E-03 - i
< 1.49E-03 - < 2.12E-03 -
Eu155 < 1.15E 02 --
< 2.38E-03 - < 4.23E-03 -
stit?5 < 1.66E-02 - i
- ~ . _ - _ _ _ . . . _ . _ _ _ . . _ . . _ - _ _ _ - - - - . _ _ . - . - - _ _ _ _ _ _ - _ _ _ _ _ _ _ - _ _ _ _ _ _ _ _ _ _ - _ _ . ~. -
m a ' Saxton Results Summary NEL ID 9509067 -01 9509067 -03 ACTMTY 2S UNC ACTMTY 2S UNC (uCi/g) (uCi/ Sample) H-3 < 1.48E-03 - < 7.50E-04 - C-14 < 4.67E-03 - < 2.58E-03 - Fe-55 < 3.25E-05 -- 8.46E-02 1.38E-02 Ni-59 < 7.81E-05 - 1.57E-02 2.30E-03 Nif,3 < 7.81E-05 - 1.43E+00 2.10E-01 Sr-90 1.30E-04 . 2.98E-05 2.64E-01 5.14E-03 Nb-94 < 8.50E-04 - 5.17E-04 3.60E-05 Tc-99 < 5.14E-05 - < 2.52E-04 - l-129 < 8.87E-05 - < 6.37E-05 - Np-237 < 2.43E-05 - < 1.80E-04 - Pu-238 1.26E-05 2.78E 6.88E-04 7.96E-05 . Pu-239/240 < 9.11 E-06 - 1.45E-03 1.23E-04 Pu-241 < 2.73E-04 - 1.99E-02 2.30E-03 Pu-242 < 6.77E-07 - < ri.32E-06 - Am-241 3.63E-05 4.44E-06 1.73E-03 1.08E-04 Am-243 < 5.53E-07 - < 2.91 E-06 - Cm-242 - < 2.10E-06 - < 1.37E-05 - Cm-243f244 < 2.47E-06 - < 2.38E-05 - U-234 , 2.80E-05 4.44E-06 < 1.38E-04 - U-235 < 2.01E-06 - < 6.52E-05 - U-238 2.61 E-05 4.26E-06 < 4.61E-05 - Cs-60 1.52E-04 3.64E-05 7.60E-01 1.80E-02 Ru-106 < 2.10E-04 - < 1.24E-02 - Cs-134 < 2.30E-05 - < 1.74E-03 -
' Cs-137 1.54E-03 2.14E-04 9.58E-01 8.90E-02 Co-144 < 1.07E-04 - < 4.67E-03 -
Am-241 < 6.84E-05 - 2.04E-01 1.14E-02 Ag-108m < 2.15E-05 - < 1.23E-03 - I u 152 < 1.91E-04 - < 1.22E-02 - t . . I ',4 < 6 22E-05 - 7.05E-03 2.26E-03 t u 155 < 4 21E-05 -- < 1.22E-03 -- 1;ti 125 < 6 32E-05 -- < 3.68E-03 --
A eN A L Y S I S O R E P O R L n T DATE: 10/31/95 CONTRACT #: 431-1258-00 CLIENT: GPU-SAXTON CLIENT REPRESENTATIVE- B. H. Brosey NEL SAMPLE #: 9600006 SAMPLE DESCRIPTION: Miscellaneous Matrices NUMBER OF SAMPLES: 4 SAMPLE RECEIPT DATE: 9/15/95 Customer ID: SX815950068 SX8C6950083 SX8C6950084 SX866950115 NEL #: 9509066-01AA 9509066-02AA 9509066-03AA 9509066-04AA ANALYTE METHOD UMII UNITS TCLP EXTRACTION / METALS ANALYSIS Arsenic SW8461311/6010 <0.5 <0. 5 <0.5 <0. 5 5.0 mg/L SW8461311/6010 <10.0 <10.0 <10 0 <10.0 100.0 mg/L Barium SWR 461311/6010 0.40 <0.1 < 0.1 0.12 1.0 mg/L Cadmium SW8461311/6010 <0. 5 <0.5 <0.5 <0.5 5.0 mg/L Chromium mg/L SW8461311/6010 <0.5 <0.5 <0.5 7.2 5.0 Lead 0.2 mg/L SW8461311/7470 <0.021 0.055 <0.021 <0.021 Mercury mg/L
<0.16 <0.16 <0.16 <0.16 1.0 Selenium SW8461311/6010
<0.5 <0.5 <0.5 <0.5 5.0 mg/L Silver SW8461311/6010 The sample aliquot to extraction fluid ratio required by TCLP was maintained. but the sample volume was reduced due to actMty.
Extracts were diluted for analysis without exceeding the reporting limits for TCLP.
- e Reviewed By: MfJ b. Date: 1Of38 [W -
3 l0;llhI . DLte: Approved By
/
P. O. Den 11165 paw %.. e se senwice. Inc Roues 726 A.. e e.. .. . i .e*"** Lynehbene. VA24608 9ISE .re .e . ss.'
y 0e - Air Monitoring
- Testing Laboratory
- Project Management
- Surveys REPORT OF BULK SAMPLE ANALYSIS Report To: Babcock & Wilcox N.E.S.I. Lab No: 9510073 RD #1, B Client Code: BWA Vandergn.ox ft, 355PA 15690 Client No: 431-1258-00 Sampled by: Client Attention: Ms. Joan Maritto Date: October 20,1995 Projact: Saxton lSampleI.D. l84158 l84159 l l client I.D. 9509066-02-AA 9509066-03-AA Sample
Description:
0101 35.9730g 0101 19.94g Is It Homoaeneous? No No Does It Contain Lavers? Yes Yes Is the Samole' Fibrous? Yes No Samole Colort Blue /Grav/ Tan Black / White / Tan
'Does the sample contain / Asbestos Fibers? No No Asbestos Tp Presents (Type and Percent)
None None lTotalPercent Asbestos l 0% l0% l l other Fibrous Materials (Type and Percent) Cellulose Cellulose - 1-2% <1% Nonfibrous Constituents Not Analyzed Not Analyzed Sample analysed according to EFA Medied EFA/606/R-93/116. Results are reported as estimaam of pertest area. subject to vanabihty and are speci6c for masenal analysed. Reports cammet be d=phr=8ad or altered webaut expressed wntten consamt of AGX. lac. Ipro B / 4//.4 Analyzed By: [M AGX, Inc. Daniel Winkle AGX, Inc. Daniel Winkle Laboratory Manager Industrial Hygienist i v 50 Progress Avenue, Cranberry Township, PA 16066, (412) 776-1905, FAX (412) 776 5714
0/ 610 tm N NUCLEAR ENVIRONMENTAL LABORATORIES P. O. BOX 11165 LYNCHBURG, VIRGINIA 2450G-1165 l LEVEL 1 INORGANICS DATA REPORT MULTIPLE SAMPLES OR MULTIPLE ANALYTES Report Date: November 17,1995 Report No.: 9509066S NEL Contract No.: 431-1258-00 Customer SDG No.: N/A Customer: GPU Nuclear Corporation Customer Site: Saxton Decommissioning Project Customer Representative: B. H. Brosey Customer Authorization No.: 0397355 S:mple
Description:
Saxton Paint Chips / Spent Resin No. of Samples: 3 Sample Receipt Date: 09/14/95 Analysis Date: 11/16/95 Customer's NEL's Analytical Sample ID Sample No. Method Analyte Results Units . SX8C6950083 9509066-02AA SW846 6010 Total Lead 34,200 mg/Kg SX8C6950084 9509066-03AA SW846 6010 Total Lead 16.4 mg/Kg SX866950115 9509066-04AA (Liquid) SW846 6010 Total Lead 29.7 mg/L SX866950115 9509066-04AB (Solid) SW846 6010 Total Lead 1880 mg/Kg 4 bk Date: / / / f ~7 /T5~ l Data Released Hv: M [
g B & W RESULTS 11/3/95 - Saxton Samples i }\ \ SYSTEM = = = > Safety /niecdon Piping - Pipe Sec6on Between RV and Check Valve. ' i Sarfon I BAW l 72806.00001 iSample Numbest samole Number Total TRU Content (nCi/g) l 8X865950053 I 9509065- 01 l 41.4400ll Total Uranium Content (nCl/g) 1 Sample Type l Pipe Section I i Description ll SCP, item 8. Int. N Results Normalized Nuclide Notes uCi/a to 100 % Nuclides RATIOS tl H-3 i LESS THAN 8.37E-03 0.0040% '#//####/#/#!#///#/#/#///a%%%%%%%%%%%%' 21 C-f 4
)_ LESS THAN 1.19E-02 0.0057 % '######/#/##/###/##/#A%%%%%%%%%%\ny 3 Fe-55 s 5.27E+00 2.5318 % Fe-55/Co-80 0.14721 . ;
4 N/-59 1.14 E + 00 0.5477% '###//######//##////##/A%%%\\%%%%%%%\Y 5 Ni-63 9.12 E + 01 43.8136% '#//##############///a%%%%%%%%%%%W , 6 Sr-89 LESS THAN N/R 0.0000% '#########/##/#/#/##/a%%%%%%%%%%%W 7 Sr-90 5.91 E-01 0.2839 % S t- 90/Cs- 137 0.71205 8 Nb-94 1.65E-02 0.0079 % '#######/#/#//##//###/a%%%%%%%%%%%%) 9 Tc-99 LESS THAN 5.86E-04 0.0003% '##########//##/#/##/a%%%%%%%%%%%%t 10 /-129 LESS THAN 1.41 E- 03 0.0007 % '####//##//##//#////###/a\\\\%\wn\\\\\\\\\w\\j 11 No-237 LESS THAN 3.80E-01 0.1826 % '/#/#####//#/#///###/#//a%%%%%%%%%%%Wj 12 Pu-238 2.27E+ 00 1.0905% Pu-238/Am-241 0.29290 13 Pu-239/240 4.99E+00 2.3973% '/#/#######////#//##//#/A%%%%\W%\%%%%W 14 Pu-247 5.72 E+01 27.4796 % Pu-241/Am -241 7.38065 15 Pu-242 LESS THAN 2.35E-02 0.0113 % '//###//#/u///////###//##/A\nW%%\Wh%%%%g1 16 Am-24 f 7.75 E+00 3.7232 % Am-241/Pu-238 p% 3.41410! t \ v) 17 18 19 Am -243 Cm -242 Cm-243/244 l 1 LESS THAN LESS THAN LESS THAN 1.52E-02 8.11 E-02 9.62 E-02 0.0073 % '//#///#/#####//#/#/#/#//a%%%\\%%%%\%\ uni 0.0390% '###/#/####//#######/A%%%%%%WWWWWW1 0.0462 % '#####//#//########/#A%%%%%%%%%%%%1 j 20 U-234 LESS THAN 2.36E-02 0.0113% '######//##/#/#////###/A%%%%%%%%%%%%) 1 21 U-235 LESS THAN 6.54E-03 0.0031 % '####/#/#//#//######///A%%%%%%%%%%%%1 j 22 U-238 LESS THAN 1.13E-02 0.0054 % U- 238/ Am-241 0.001461 j 23 Co-60 3.58E+ 01 17.1968% Am-241/U-238 685.84071) j 24! Ru- 106 LESS THAN 1.34 E-01 0.0644 % '#########//####/#/#/AWWWWhhuwwwwh 1 25! Aa- 106m LESS THAN 1.45E-02 0.0070% '#//##//#//##//##////###/a%%%%%%%% l 26 Sb-f25 LESS THAN 3.17E-02 0.0152 % '######//####/##/###/a%%%%%%%%%\%%\J 27 Cs- f 34 LESS THAN 2.33E-02 0.0112 % Cs-134/Cs-137 0.02607i 28 Cs-f37 8.30E-01 0.3957 % Co-60/Cs-137 43.13253i 29 Ce-144 LESS THAN 4.34E-02 0.0208% Cs-137/Co-60 0.023181 30 Eu-152 LESS THAN 1.07E-01 0.0514 % '#//#//#//#//####//#////#/a%%%%%%%%%%%%1 31 Eu- f 54 ! 6.65E-02 0.0319% M%%%%%%%\Wu\%%%%%%hh%%%%%%%%%%%) 32 Eu-f55 l 1.70E-02 0.0082 %l'#/###/#/#/##//#////###/a%%%%%%%%%%%%t l Totals = = > ll 2.08E+0211 100.00% ll N/R - Not Reported g
\v) ,
j l l
!~ 1 B & W RESULTS 11/3/95 - Saxton Samples . SYSTEM = = = > Sanety /niection Piping - Pope Secnon Between RVand Check Vahe. 1 Sexton 4 SA F ! il 0.04151 i Samode Numbert Samode Aumber l Total TRU Content (nCi/g)
~TX86""? 4 9508065-01-AA l , ll 0.00031 j
3 Total Uranium Content (nCi/g) Sample Type 4 Pipe Section i i a Description ~ SCP. Item 8. Metal Results Normalized Nuclide Notes uCl/a to 100 % Nuclides RATIOS 1,5 H-3 LESS THAN
- 6.32E-05 i 5.9246% f/#//#/#/##//##/#//#/#//Augggggmn\Mn) 21 C- f 4 4 LESS THAN i 4.06E-04 I 38.2472% F/#/#/#/#/#//#//#///////#/A\\w\\n\uu\\nggut 31 Fe-55 '
LESS THAN i 4.06E-061 0.3825%l Fe-55/Co-60 i 0.026534 l 4: Ni-59 > LESS THAN i 1.62E-05 l ' 1.5186% f/###/##/###//#/#//##/Auw\wun%%wwwuj 5, Ni-83 , 1 3 33E-04 i 31.2165% F##/#/#####/###///##/Annuununngmn) 6 Sr-89 LESS THAN ! N/R I 0.0000% F/#/#####//####//##/#Agggugmuggggj 7 St-90 LESS THAN I 3.78E-061 0.3543 %l St-90/Cs-137 i 0.966751 l 8: Nb-44 LESS THAN : 130E-061 0.1125 %F###/###/#//###/####Aumnununnuguj l 9- Tc-99 LESS THAN i 4.44E-061 0.4162% , fun //#//#///#///###///////A\w\\\\\\ng\w\gggj 10: /-f29 i LESS THAN i 4.39E-06 I 0.4115% yu/#//#/#/#//##///#/#/#/ Ann \unnununnut 11i No-237 i LESS THAN i 1.2SE-06 I 0.1172% fl##/#//#/###/###//#/#Aqung\n\n\ugggt 12.' Pu-238 i LESS THAN i 3.52E-07 r 0.0330%! Pu-238/Am-241 1 0.624111 131 Pu-239/240 i LESS THAN t 6.96E-07 i 0.0652% g//#////##////#/#//#///#/A\\\unnu\wunnu\t j , '14i Pu-247 s LESS THAN i 3.82E-051 3.5810%I Pu-241/Am-241 1 67.73050? I 154 Pu-242 LESS THAN I 1.06E-07 0.0099 %If/#/#/#/#//##/#///#/////#A%WWW\\n%%%%%M [3 ( 16 tl 174 Am-24 f Am-243 i i LESS THAN LESS THAN i 5.64E-07 1 6.90E-08 i 0.0529 %! Am-241/Pu-238 l 1.60227 N 0.0065 %F/#/###/##/##//##/////#Annuu\%www\\w\\t 184 Cm-242 ' LESS THAN i 7.90E-os 0.0074 %if/##////#/#//###/#####Anwhnw\\\\\\\\\w\\g l 19L Cm-243/244 LESS THAN 1.95E-07 0.0163% if#////#//##//#/#/#//#///#A\\\\\\uww\w\\nn\\j l 20$ U-234 i LESS THAN 1.55E-07 ' O.0145% P/#///#////#/#//##///##///Auwungungwww\) l 214 U-235 4 LESS THAN 6.95E-061 0.0065 %F#/#//#/#/####//##///#/Aggggwununum! ! 22; U-23J i LESS THAN . 6.95E-06 i 0.0065 %! U-238/Am-241 i 0.12323 231 Co-do i i 1.43E-04 ! 13.4053%l Am-241/U-238 i 8.11511; 249 Ru-f06 i LESS THAN , 1.11 E-051 1.0405% y#/#////#///##/#/##/////#Awwwwwwuwmuut
- 25' Aa-f00m LESS THAN ! 1.06E-06 i 0.0994% f#//##////////#////#/#///#/A%%%%%%%%%%mi 26: Sb- f 25 LESS THAN i 2.61E-06 i 0.2447% ////#/##//###/#///##/////Aumnunwunn\n\1 27: Cs- 134 LESS THAN I 1.51E-06 ! 0.1416%4 Cs-134/Ca-137 1 0.38619 j 28 il Cs-137 i i 3.91E-061 0.3665 %I Co-60/Cs-137 i 36.57289l 294 Co- f 44 i LESS THAN I 4.92E-061 0.4612 %1 Cs-187/Co-60 t 0.027341 301 Eu-152 LESS THAN I 1.32E-05 l 1.2374% l'/////////////#/#//##///////#A%%%%%%%\\%%%)
31J Eu- f 54 LESS THAN ! 3.38E-06 0.3169 %h%%%\\B%%\B\\\\%%%M%%\%%%B%%%%%%%t 32$ Eu- f 55 LESS THAN i 1.96 E-06 0.1837% U/#/#/#/#/##/#/#/#//##/AWWW%%BMMuuhl 1 Totals = = > 1 1.07 E-03 il 100 00%l l N/R - Not Reported t 1 6 l V l l
i l . l
/',m) B & W RESULTS I U3/95 - Saxton Samples -
)
SYSTEM = = => SaW.y/njection Piping - Pipe Section Between Check VsNe and CV Wall. i l 1 Santon il BAW l ll 3.04511-l LSamode Numbed Samode Number i Total TRU Content (nCi/g) [S.X_865950069 !! 9500065 -02 1 ILoy5] Total Uranium Content (nCi/g)
' Sample Type H Pipe Section i j Description ' SCP Table B-6 (
Results Normalized Nuclide Notes uCl/o to 100 % Nuclides RATIOS ! 14 H-3 1 ESS THAN i 8.74E-03 i 2.3853% f/////////////////////////////////Awwwumbunggg) 2 il C-14 ESS THAN i 7.86E-03 i 2.1451 % ! f/////////////////////////////////A%ugmungwngwj 31 Fe-55 i 5 58E-03; 1.5229 %i Fe-55/Co-60 . 1.786461 4 Ni-59 ESS THAN I 2.05E-04 ! 0.0559%P/////////////////////////////////Anghmwnnngwnj 5t Ni-63 i ' 1.44E-02 ! 3.9300 %if/////////////////////////////////A\\m\guwmugugl 6: Sr-89 1 ESS THAN N/R . 0.0000% jf/////////////////////////////////Anu%%uumhuum 7 Sr-90 !I ESS THAN ,1.39E-04 i 0.0379%l St-90/Cs-137 i O.00044 84 Nb-94 s ESS THAN ' 2.81 E-03 ' O.7669%if/////////////////////////////////A\%\%hu%%%%%%) 91 Tc-99 .I ESS THAN < 2 91E -04 l O.0794 %P/////////////////////////////////A%%%%mug\%nMt 104 /- f,29 ll ESS THAN 1.57E-03 0.4285% V/////////////////////////////////Annw%muununut 11!1 No-237 il ESS THAN 4 31E-05, 0.0118% lr/////////////////////////////////Annu%uwwwwnm) 124 Fu-238 4 6 32E-051 0.0172%! Pu-238/Am-241 1 0.322451 134 Pu-239/240 1 i 1.23E-04 i 0.0336% ! f/////////////////////////////////Anunnunnunumt 144 Pu-24f 4 2.61 E-03 ; 0.7123% l Pu-241/Am-241 1 13.316331 l fm 154 Pu-242 9 ESS THAN 2.86E-061 0.0008 % f/////////////////////////////////Ammmmmm\wuj [ \ 16i Am-24 f ii 1.96E-041 0.0535 % Am-241/Pu-238 1 3.10127 j () 17d 18 ll Am-243 Cm-242 4 i ESS THAN ESS THAN 9.41 E-07 r
' 1.02E-06 6 0.0003 %i.'/////////////////////////////////Aumwwmwwwumt 0.0003 %i'/////////////////////////////////Amwwwm\muum) 194 Cm-243/244 ESS THAN 4.98E-061 0.0014 %if///////////////////////////////i/Anunmmwwwwm1 20; U-234 ESS THAN . 4.96E-06 i 0.0014 % f/////////////////////////////////Auwww\wunnunut i
21i U-235 LESS THAN : 7.01 E-06 l 0.0019 %P/////////////////////////////////Am%%mmum%m) 22I U-238 ESS THAN 3.51 E-06 i 0.0010 %1 U-238/Am-241 1 0.01791 a , 23 i Co-60 i e 3.12E-031 0.8515 %i Am-241/U-238 ! 55.84046i l 24- Ru-106 4 ESS THAN i 8.14E-04 ! 0.2222% P/////////////////////////////////Anunum%%mmut 25' Aa- 708m i ESS THAN i 8.71E-051 0.0238 %f/////////////////////////////////AWWMm%%%%umt 26i Sb-f25 i ESS THAN i 4.33E-04 ' O.1182% J/////////////////////////////////Au%umumuunm) 27' Cs-ts4 i ; 6.62E-05 : 0.0181 %I Cs-134/Cs-137 i 0.00021 281 Cs-137 i 3.16E-011 86.2426 %I Co-60/Cs-137 ! 0.00967i 291: Ce- f 44 - .I ESS THAN i 6.31 E-041 0.1722 % i Cs-137/Co-60 1 101.282051 30h Eu- f 52 i GSS THAN 1 2.45E-04 i 0.0669% y/////////////////////////////////A%%%%%%%%n%%\1 31# Eu-154 ' ESS THAN ! 5.04E-05 i 0.0138%%%mmmmu%%%%mmmMummm\%MmW 32" Eu-f55 i ESS THAN 3 06E-04 ' O .0835 % )/////////////////////////////////A%%%%%%%%%%%4 s Totais = = > I 3 66E-01 ii 100.00 % s N/R - Not Reported l l l
i
-[3 \
B & W RESULTS 11/3/95 - Saxton Samples -
)
SYSTEM = = = > Safety /niection Piping - Pipe Sectort Betwen Check Valve and CV Wa/I. Sarfon i BAW i ii 0.0341l
!Sampde Nwnbe(Jampde Number i TotalSU__C_ontent (nCi/g) l l SX865950069 !l 9509065 AA l L_ O.00191 l
) i_ Sample Type 4 Pipe Section _ Total Uranium Content (nC1/g) l J
. 1 1 Desenption 4 SCP TB-6, Metall I l l Results Normalized Nuclide Notes uCi/g_ to 100 % Nuclides RATIOS 1( H-3 i WSS THAN i 6.06E-05 5.5840 %J/#/////#////#///#//#/#/##A\uguwwigg\%wny 24 C- 74 4 GSS THAN 2.39E-04 22.0225 %P/#/#///##/#/#####iu///Augnuguggnggq 39 Fe-55 a WSS THAN : 3.68E-06 0.3391 %I Fe-56/Co-60 ! 0.02022 41 Ni-59 a 4.84E-061 0.4460 %U##/#/###///##/###/##Au\wunggggn\gy 5i Ni-63 '.
4.47E-041 41.1886 %F#///#/#/#/#///#/#/#/#/#Anununn\ngung1 61 St-89 GSS THAN' i N/R i 0.0000 %f///##/#/#####/#/####Ag\gngunnggwul 71 St-90 ' t 2.86E-05 i 2.6353 %i Sr-90/Cs-137 ! 3.97775 l 8(__Nb-94 y WSS THAN < 9.75E-071 0.0898 %?///#////##/##//#/##//##A%\W%%%%ggung) l 94 Tc-99 4 ESS THAN 4.11 E -06 0.3787% J//##/##///##/#/////#/#//An\\nuu\ngunng) l 101 /- f 29 LESS THAN i 6.92E-06 0.6376% U//#///#//#//#/////##/##//A\\\\\\\\n\w\\\\\\\\wt l 119 No-237 i GSS THAN I 8.52E-07 0.0785 %?/##/////////#//#///#//##//A\w\wwwww\nw\\w) 121 Pu-238 1 WSS THAN ! 2.95E-071 0.0272 %! Pu-236/Am-241 ! 1.489901 134 Pu-239/240 l GSS THAN ! 3.97E-07 0.0366 % f##////#////////#//////#/#//A\\n\www\\wwwwn) i 14s Pu-241 WSS THAN i 3.20E-05 2.9486 %! Pu-241/Am-241 i 161.61616I , 154 Pu-242 ESS THAN 1 8.88E-06 0.0082% P//#//##/#/#//####!#/#/Anu\ngunwwwwuj l \ 164 Am-24 f ESS THAN i 1.98E-07 0.0182%I Am-241/Pu-236 ' [/ 174 Am-243 Cm-242 WSS THAN i 4.42E-08 6 O.671194 0.0041 %if//#////#//##/#/#/#///#///Aww Juw\wunung 184 GSS THAN ! 1.52E-07 i 0.0140%'f#/#//#///#////#/#////#/#/Aw\\w\wwwwunnut 194 Cm-243/244 ESS THAN i 1.17E-071 0.0108 % f/////#/#/#//##//#///////#!Aw\w\w\wni\unuw) 201 U-234 4 WSS THAN ! 8.31E-07 ! 0.0766 %if#//#/#////////////#//////#//An\\w\ww\www\wn) l 211 U-235 1 WSS THAN t 5.76E-071 0.0531% U///#/////#///////////####//A%\WW%%%%BWWW\ 22i U-238 ' LESS THAN ' 4.97E-07 I 0.0458 % U-23s/Am-241 ! 2.51010; 23' Co-60 ' 1.82E-04 ; 16.7703 % Am-241/U-238 I 0.39639! , 244 Ru- 106 1 WSS THAN ' 1.74E-05 i 1.6033 % r/##//#/#//#/#/###///#//A\nw%%%%%%%%%) 4 251 Aa- foem 4 WSS THAN i 1.75E-06 0.1613% r//#///##/#/#//##//#///#/Auww\wwwununu) l 264 Sb- f 25 4 ESS THAN ! 4.02E-06 0.3704 % r////#///#////////#/#//N#/#A\n\%\%%%%n\\wut I 275 Cs- f 34 l ESS THAN i 2.40E-06 0.2211 % Cs-134/Cs-137 1 0.33380 I 28i Cs- f 37 ii ! 7.19E-06i 0.6c25%i Co-60/Cs-137 t 25.31293i
- 29) Ce-144 Ni WSS THAN 6 7.57E-06 i 0.6975 %I Cs- 137/Co-60 ! 0.03951i 30i Eu- 152 :l ESS THAN I 2.30E-051 2.1193 %F///#//##///###//#/#/#///A%%%%%%%%Bu%\l l 314 Eu- 154 l WSS THAN I 4.54E-061 0.4183 %I i%%%\%%%%%%%%\u%%%\%%%%%%Bu\uuBM) l 32i Eu-f55 i WSS THAN ' 3 61E-06 t 0.3326% J//////////##/####!/##//#Annun%%%%Butu l il Totals = = > 1 1.09E-03 9 100.00% l l
N/R - Not Reported l 2
,/
v l
l . L ( l [m)
\ /
B & W RESULTS 11/3/95 - Saxton Samples -
~
l SYSTEM = = => Riwr Water' Pipe Section l l d Sarfon J BAW d il 0.0445X iSampde Numbesit Samode Number l Total TRU Content (nCi/g) 1 SX835950085 N 9500065-03 1 !) 0.00168 Total Uranium Content (nCi/g)
! Sample Type 4 . Pipe Section i I Description ! SCP Table B-4 II l Results Normalized Nuclide Notes uCi/a to 100 % Nuclides RATIOS 11 H-3 LESS THAN 2.31 E-04 i 20.2000% if/////###/##/#//#/#/#/#/Anugu\wwww\w\wj 2 11 C-f4 .
LESS THAN i 7.36E-04 i 3# Fe-55 WSS THAN 64.3602% PN/N#//##/#/////##/###/Annuw\wngwnngj i 1.83E-05 i 1.6003 %I Fe-56/Co-60 1 2.101031
- 4) Ni-59 WSS THAN i 6.17E-061 0.5395% lf#//#///#/////##/#/#/#/#/A\%%Buwwwwwwwj 5!I Ni-63 !
i 4.10E-05 i 3.5853% '#////#/##/##/////####//Annun\nungungt 6t St-89 i WSS THAN i N/R i 0.0000 % r/##/#/##//#/#////#/#////Anunnug\unngut l 7! Sr-90 .I 9.84E-061 0.8605 % St-90/Cs-137 1 2.050001 St Nb-94 -1 LESS THAN i 1.39E-051 1.2155% f#/###/##/#/##/####//Annu\%%uungwuj 9: Tc-99 4 WSS THAN 8.95E-06 ; 0.7826% f//###/##//#/#/#////###A\\un\\\wwwnnggj l 10; l-129 4 WSS THAN 1.27E-05 i 1.1106% /##/####/#//##//##/##A\%%wungwunung 11i No-237 .I ESS THAN 1.61 E-061 0.1406% fu#///##/##//#//###/##Au\unnununu\n\t l 12! Pu-238 1 WSS THAN 4 00E-071 0.0350%I Pu-238/Am-241 1 0.39216 l 13: Pu-239/240 WSS THAN 6.81 E-07 i 0.0596% 1/##/#/#/#/#////#/#/### Ann \nnwnnnwun) 144 Pu-24f 4 WSS THAN 4.03E-05 i 3.5241%I Pu-241/Am-241 ! 39.50960 j n 154 Pu-242 1 WSS THAN 1.15E-07 ' O.0101 % P///#/#/###/#/#////##/#/A\w\wun\\\www\wy f ) 16il Am-24f , 1.02E -06 0.0692 %I Am-241/Pu-238 2.55000 il (/ 17s Am-243 i 1.ESS THAN i 4.77E-061 0.0042% F//###////#//#/###/###/Mwwwun%%%%%B1 l 186 Cm-242 WSS THAN I 1.73E-071 0.0151 % lr##/##/##/####//##/#/Mwwwwwwwngwn) 196
~
Cm-243/244 WSS THAN I 1.65E-07 I 0.0144 %Ir/#///#/////#//#//#//#//////Auw\n%%%wn\\nu) 206'-~U-234 __ WSS THAN t 9.11E-071 0.0797 %9#/###/#/////#///#///#////Au\\\hhhhhu\W\B) 21 U-235 WSS THAN I 2.57E-07 ; 0.0225 % 22 U-238 fu////////////////##/////#////ABWWuw%u\unwu) i WSS THAN ! 4.19E-071 0.0366 % U-238/Am-241 1 0.410784 23: Co-60 .i . 8.71E-061 0.7617 % Am-241/U-238 l 2.43437! I 24! Ru-fos 't WSS THAN I 1.63E-06 i 0.1425 % fu//#/##//#/#####/#/#/Auwnuunu%%\unj i 25i Aa-fosm .I WSS THAN I 1.64E-071 0.0143 % r/#/##////#//##////#/###AukWWWWWWh\%%) l 26' Sb-125 4 WSS THAN } 4.31E-07 i 0.0377 % r/#///////#////#///#/#/////#A\wwwununnuuut l 27- Cs- f 34 4 WSS THAN i 2.17E-07i 0.0190 % Cs-134/Cs-137 1 0.04521j l 28 Cs- f 37 'l i 4.80E-06 i 0.4197 %. Co-60/Cs-137 6 1.814584 291 Ce-f44 I WSS THAN i 8.34E-07 i 0.0729 %I Cs-137/Co-60 1 0.55109' 30: Eu-152 ' WSS THAN i 1.89E-06 i 0.1653% jr#//##/#//##/////#/#/#/#A\uwWuMWBunuu$ ! 31. ! Eu-f54 ! WSS THAN i 5.07E-07! 0.0443% h%%%\n%%%%%%%%%%%%%%%%%%%%%B%%% 321 Eu- f 55 i WSS THAN ' 4 22E-07 ' O .0369% F####/////#////#////##//#AWMMMWBWWBB%R l g iatals = = > 4 1.14E-03 il 100.00 % 1 N/R - Not Reported l l
, i Gi -
?
l - l B & W RESULTS 11/3/95 - Sexton Samples . i SYSTEM = = = > Steam Generator - Bottom Blow Down Line 1 I Samrorp ll 84 W ll 0.0566 il Samode Nannbeg Samsde Number Total TRU Content (nCi/g) 1 l SX625960066 il 9509065-04 ll 0.00521 Total Uranium Content (nCi/g) Sample Type .1 Pipe Section 9 Descnotson i SCP, Table B-10 '4 l Results Normalized Nuclide Notes UCi/n to 100 % Nuclides RATIOS l 14 H-3 .i LESS THAN - i 1.24E-04 i 5.9561 %ir///#/////###//#/#/####/Au\W\%%%ggugggg 2; C- f 4 4 LESS THAN ' 1.36E-04 ! 6.5325 % r/##////#/#////#//##/##//A\guggggggggggt i 3l Fe-55 <l WSS THAN 6 3.99E-06 ! 0.1917% Fe-55/Co-60 1 0.10025I l l 4: Ni-59 4 WSS THAN i 1.32E-051 0.6340% if##/####/###/#///###/Au%%w%%%%%%%q
- Si Ni-63 i i 1.41E-04 i 6.7727% F#//#/######//#/##///#A\gn%%wwww\wn\\
6: St-89 WSS THAN i N/R I 0.0000% F##/##/##/######/#///Annuungugwguq i 7i Sr-90 : 5.03E-06 ? 0.2416% l Sr-90/Cs-137 ' O.00333 i l ! 81 Mb-94 t WSS THAN i 4.67E-061 0.2243% F/##/#/##//#/#/###/##/Auwwunwwww%uw l 9P Tc-99 WSS THAN . 4.35E-06 6 0.2069% F/##/#///##/#/##/#/#/#/AuwhuhunW%%%g 10i /-129 WSS THAN 1.53E-05 I 0.7349% F#/#//#/##//##/###/#/#A\\h\huu\%\\w\\ug 114 Np-237 WSS THAN i 1.13E-06 i 0.0543% Ir/#/##//#/##/#//##///##Anun%%%%%%ggg 121 Pu-238 4 i 2.37E-061 0.1136 %! Pu-238/Am-241 1 0.77199! 131 Pu-239/240 ' ! 3.96E-06 i 0.1902% '/#/#####/##//#/#///////A%%%%%%%%gugg 141 Pu-24 7 t 4.60E-05 6 2.2095 % Pu-241/Am-241 1 14.983711 154 Pu-242 i LESS THAN i 7.46E-061 0.0036% f//#///##/#/////#/##///////Ahnhwinug\\\\ng) h, 161 Am-24f I ! 3.07E-06 i 0.1475 %I Am-241/Pu-238 i 1.29536: (j
- 17[ Am-243 16a Cm-242 LESS THAN WSS THAN
' 2.61 E-061 i 7.31E-061 0.0013 %if//#/#/##///##/#/#/##!#A%%%%%%%%\%uut 0.0035 %ir//#/#//###//#/#//###///Agu\n%%%%%wwn) 194 Cm-24W244 'l LESS THAN i 1.07E-07 ! 0.0051 %;r/#f##f///##/##//#///#f#Aggggggggu\ggi)
- 20) U-234 4 1 2.73E-06 6 0.1311 %ir////##////#///###//#///#/A%%%%%%%%%%%w 21 ; U-235 1 LESS THAN i 1.36E-07 i 0.0065% lr/#/#/////#/##/##//#/#///A\%%\uguwwwwww 221 U-238 H i 2.33E-06 i 0.1119%I U-236/Am-241 1 0.75896' 23 <; Co-60 ll ' 3.96E-05 ! 1 9117 %i Am-241/U-236 l 1.31760!
24l Ru-106 LESS THAN t 6.52E-061 0.3132 %;r//#####/#//////#///##/#AuguwnuwwuM\wt 251 Aa- foem ! LESS THAN i 6.93E-071 0.0333 %'r//#////#####/#///#//##/AWWWugunnuung 264 SD-f25 5 LESS THAN t 3.30E-06 i 0.1565 % r///##////#/////N#////#/#//Auguwnuwwwwww l 274 Cs-134 Il I 5.59E-071 0.0269 % Cs-134/Cs-137 - 0.00037 281 Cs-f37 !! ! 1.51E-031 72 5299% Co-60/Cs-137 . 0.02636 29! Ce-f44 LESS THAN l 5.06E-06 i 0.2430% l- Cs-137/Co-60 6 37.93970 30!~~~~Eu- f 52 , WSS THAN I 3.35E-06 0.1609 %ir/#///#//#//#######/////A%%%%\%%%%BW14 31t Eu- 154 i LESS THAN i 6.90E-07 0.0331% 4W%%\%\%%\%%%%%%%%%%%%%%%%%%%%B4 l 321 du-f55 i LESS THAN i 2.36E-06 01143% #/#//#/#/#/#/#/#/#//##//A%%%%B%%%%w04 ll Totals = = > i 2.06E-03i! 100 00% I I N/R - Not Reported l l l l
\
L l
l ? l i g B & W RESULTS 11/3/95 - Saxton Samples . SYSTEM = = = > Sanam Generator - Sunince Blow Down Proing l l 1 l Santon ' B&W *
? O0597!
! LSamsde Number, Samsde Number l Total TRU Content (nci/g)
' SX825950067 4 9509065 -05 3 d 0 00093 Total Uranium Content (nCi/g) i i Sample Type 4 Pipe Section 1 Desenption : SCP, Tabee B-10 8 Results Normalized Nuclide Notes uCi/a to 100 % Nuclides RATIOS 11 H-3 1 GSS THAN i 2.14E-04 i 22.7131 % F###//###//##//#//##///Aiguwi\%%\\%%%%)
2 fl C- f 4 GSS THAN I 3.01E-041 31.9470% f#/#/#///////#//##////#//#Au\uuu%%%%\\%%) 3! Fe-55 LESS THAN : 9.37E-06 i O.9945%l Fe-55/Co-60 i 0.279701 4l Ni-59 : 3.77E-061 0.4001 % F//#//#//#####//##/#/#/A\n%%%%%%%%%%1 5, Ni-83 l 2.21E-04 i 23.4561 % f/##/#/#/#/####//#///#/Anu\n%%%%\%\%%) 61 St-89 GSS THAN i N/R I 0.0000% fun ##//#/####/##/#/#/A%%%%%%%%%\%%j 74 Sr-90 1 LESS THAN i 4.40E-06 i 0.4670% l Sr-90/Cs-137 } 0.077191 8: Nb-94 y ESS THAN : 4.29E-061 0.4553% f/##//##/////#/##//#/##/A\n%%%%%%%%%%t , 9' Tc-99 ' GSS THAN i 7.12E-06 i ' 0.7557% f///#///#/##///#/#////#////A%%%%\%%%%\\%%t 10! /-129 GSS THAN i 1.74E-05 l 1.8468% ifu/////####/#/###//#/#/A%%%%%%%%%%%\t 11i Np-237 LESS THAN 1.96E-06 ! 12 ;Tu-238 0.2101 %1f#//////##///#//##//#///#/A\W\n%%%%\%%%%) GSS THAN ' 1.13E-07 i 0.0120 %I Pu-238/Am-241 1 0.007111 13n Pu-239/240 LESS THAN : 3.92E-07i 0.0416 %P///#/////#####/#//###//$%%%%%\wunwhW) 14( Fu-24 f GSS THAN ! 4.09E-05 i 4.3410 %1 Pu-241/Am-241 i 2.572331 154 Pu-242 LESS THAN 1.13E-07 i 0.0120 % r//#/#/#///#//##/##//#/#A%%%%%\WWuBBM1 l [] ( 16i 171 Am-241 Am-243 4 LESS THAN
! 1.59E-051
' 2.98E-08 6 1.6876 % Am-241/Pu-238 i 140.707961 0.0032%'r/##////##/#/#///#/#//##A%\%%%%%%\%%%W 18s Cm-242 i LESS THAN 9.17E-08 i 0.0097 % f###/#/////#/#//////#/#///A\\%%%%%\%\%%%\1 l
l 19 lcm-243/244 ! LESS THAN i 1.60E-07 I 0.0170% i r#///////#//####//##////#Ag\%\W%%%\%%%%) 209 U-234 i LESS THAN i 3.12E-07 i 0.0331 %ifu#////##//#/##/#//##/#Aww%%%%\W%%%%1 21 ; U-235 :i LESS THAN 1.04E-07 i 0.0110% '#///////#////#//#/#/////////Aw\%%%%%%\%%%\1 22I U-238 1 GSS THAN 1 4.77E-07 ; 0.0506 %! U-238/Am-241 i 0.03000 231 Co-80 j 3.35E-05 l 3.5556 %I Am-241/U-238 I 33.33333 24 [ Bu- fp6 i LESS THAN 1.33E-06 ; 0.1412%P//#/#/#/#/#/##////#/#/#Aw%%%%%%%\%%%) 25J As- f08m 4 LESS THAN t 3.01E-071 0.0319 %FI///###///#//#/###/////#A%hW\MW%%%%%%) 261 Sb- 125 4 LESS THAN 1 9.28E-07 I 0.0985 %b/////#///#/##//////#////////A%WWWW\%%%%%%t 274 Cs- 734 i LESS THAN ! N/R . 0.0000 %1 Cs-134/Cs-137 1 0.000004 28i Cs-f37 . 1 5.70E-05 6 6.0498 % Co-60/Cs-137 I 0.587721 29J Ce- f 44 ! LESS THAN i 1.39E-061 0.1475 %i Cs-137/Co-60 i 1.701491 304 Eu- f 52 LESS THAN i 3.52E-061 0.3736 %F/#/#///###/#/#//###//#A%%%%%%%%%%\W\ 31i Eu-f54 i LESS THAN : 6.02E-071 0.0639 %4%%%%%%%%%%%%%%%%%%\%%%%%%%%%%t 32i Eu-155 ! LESS THAN I 6 93E-07 ! 0.0736 %J//#/#///####//#/#/#/#//A%%%%%%%%%%%W Totals = = > l 9 42E-04 4 100.00 % N/R - Not Reported i I l l (3 - y l l \
l l l ! B & W RESULTS 11/3/95 - Saxton Samples O)
\
G SYSTEM = = = > Reactor Cavrty 3800/ Support Structure - NNE Quadrant ( i Saarron 9 8AW ! ji 0.02416 iSample Numbed Samode Numberj Total TRU Content (nCi/g) ! l LSX866950116 il 9500065-06-AA I !I 0.00014 J l Total Uranium Content (nCi/g) l
- Sample Type i Steel Support 4 l l i Descnota i Support of ECIS (
Results Normalized l Nuclide Notes uCi/c ' to 100 % Nuclides RATIOS t il H-3 'l W5FTHAN , 2.63E-05 i 1.9049% f#/###/#////#//#/#//###Au\WWW\%%%%%\nj 28 C- f 4 i LESS THAN i 1.23E-04 6 8.9057 %:'##/#/#//#/#/##//#/##//A\%%%%%%%%%%%q 3: Fe-55 -
~ 4.98E-06 ! 0.3607 %i Fe-55/Co-60 1 0.006191 4i Ni-59 1.66E-051 1.2023 %l'#/#/#//#/##/#/#/####/A%u\WWunwwwwuj 5l Ni-63
- 5.25E-04 i 38.0249 %P/###//#/###/#####/#/A\%%%%\Wuw%%%\t 6J Sr-89 ! LESS THAN : N/R i 0.0000 %F/#/#/######//##/##/#A\%%%%uu%wunut 72 Sr-90 i LESS THAN I 1.56E-061 0.1130% l St-90/Cs- 137 I 0.804121 84 Nb-94 i LESS THAN i 5.61E-07 ! 0.0406 %:r//###/#/#///////#//##/#/Auguwwwwww\wnt 94 Tc-99 { LESS THAN 2.05E-061 0.1485% fu/####//##/#/#//n#/#/AbhW\%%%%%%%%j 104 /-129 9 LESS THAN !
6.22E-06 0.4505% f#//##/////##/#/###/#/#A\%%%%uu\%%%ng 114 Np-237 LESS THAN i 7.27E-07 0.0527% f/##/#/#/#//####//###/A\\\\w\hnww\%%wn 123 Pu-238 LESS THAN i 3.79E-07 ! 0.0275%s Pu-238/Am-241 i 0.384771 l 130 Pu-239/240 1 LESS THAN - 4.78E-07 : 0.0346% //#////#/##//##//#///#/#/A\%\\%\n\\%\\W\\\\\1 l 1411 Pu-24 f a LESS THAN ' 2.13E-051 1.5427%I Pu-241/Am-241 1 21.624371
' 5.07E-081 153 Pu-242 i LESS THAN 0.0037% //#/#/#//////////##//#/#///A%%%%%%%%%%%W ,O g 16i Am-24f >
i 9.85E-071 0.0713%l Am-241/Pu-238 l 2.596944 (j 17i 18:1 Am-243 Cm-242 i LESS THAN LESS THAN 1.14E-08 I 5.37E-08 i 0.0008% //##/#/###/#////##/#////A%%%%%%%%%%%\j 0.0039% W/##//##//#///#/#/###/#Auwuw%%wu%%%g l 19t _Cm-243/244 LESS THAN i 6.63E-08 ! 0.0048% //#/#///##///#//#/###////An%%%%%%%%%%g 204 U-234 s LESS THAN i 4.73E-081 0 0034% ' /#//#////#/#/###/###////A%wwwwwww\%%%j 21i U-235 ' LESS THAN ' 1.50E-06 i 0.001 1 % 1/#/#/#/###//#/#/####/A%%%%%%%%\%%%1 i 221 U-238 1 LESS THAN i 2.12E-081 0.0015 % U-238/Am-241 1 -0.021521 ) 23; Co-do i 6.06E-04 l 44.0364 % Am-241/U-236 ' 46.46226i ! 24d Ru-106 i LESS THAN ' 1.11E-05 ! 0.8040 %'f/#//#//#//#/#//#/##/#/#A%%\Wuwwwwww%j 25J Aa-106m n LESS THAN i 1.10E-061 0.0797 % f####///##/###/#/##///Auguh\%%%%%%%j j 26i Sb-f25 o LESS THAN ! 2.53E-06 ! 0.1832% f#/#/#//#/#/#///#//#/##/A%%%%%%%\%Muk) l 274 Cs- f 34 4 LESS THAN 1.60E-06 i 0.1159 % Cs-134/Cs-137 l 0.82474l 281 Cs- 137 4 7 1.94E-06 i 0.1405 % Co-60/Cs-137 6 313.40206j 29i Ce- f 44 .I LESS THAN I 4 62E-061 0.3346 % Cs-137/Co-60 1 0.00319l 301 Eu-152 LESS THAN f 1.48E-051 1.0719 % f//#//#//#///////#//#/##/#A%h\u\hW\unnhuk 31i Eu- f 54 ; LESS THAN , 2.59E-061 0.1876 % 6%%%%%%%%\n\u%%%%%%%%%%%%%%%%%W 321 Eu-155 ! LESS THAN ' 199E-06 ! 0.1441 % r#//#//##////##///##/#///A%%%%%%%%Buun j Totals = = > 1 138E-031 1.00.00 % l N/R - Not Reported i
\
d l l
i 1 v% s B & W RESULTS 11/3/95 - Saxton Samples I SYSTEM = = = > Resin Sluice From Domineralizers
._ Sarfon a BAW jt 28.5439i
_Sampie Number' Samde Number Total TRU Content (nCL/g) ITX866950118 I 9509065-07 d ) 0.0268 k Total Uranium Content (nCi/g) ESample Type 4 Pipe Section !! l l Description i SCP, item 10 Il Results Normalized Nuclide Notes uCi/c to 100 % Nuclides RATIOS 11 H-3 i ESS THAN
- 1.90E-031 3.1146% F/////////////////////////////////Au\uwwwWn%%%%)
24 C- f 4 i LESS THAN i 2.33E-03 ! 3.8195 %if/////////////////////////////////A\w\%\WuB%\www\g 3i Fe-55 i i 7.39E-04 i 1.2114% l Fe-55/Co-60 i 0.247161 4; Ni-59 1.77E-041 1 0.2902% 1 f/////////////////////////////////A%\%%%%%%%%%%) 5 Ni-63 4 i 1.65E-021 27.0479%if/////////////////////////////////Au\\%\%%%%\%%%\j 61 St-89 !i ESS THAN > N/R i 0.0000%if/////////////////////////////////Au\%\%%%%%%\\%% 7;l St-90 1.00E-03 ! 1.6393% Sr-90/Cs-137 i O.19120l 8; Nb-94 WSS THAN i 4.78E-04 ! 0.7836 % f/////////////////////////////////A%%%%%%%%%%%% 94 Tc-99 WSS THAN : 6.61E-05! O.1064 % r//////////////f//////////////////jgugggggWg%%gg 10:1 I-129 i ESS THAN : 3.62E-041 0.5934% f//////////////////////////////////(WW\%%%%%%%%%) 11i Np-237 ' LESS THAN ' 7.82E-051 0.1282% P//////////////////////////////////(W%\%%%%%%%%%).
, 121 Pu-238 i 5.98E-04 ) 0.9803% l Pu-236/Am-241 1 0.313094 134 Pu-239/240 y 1.72E-03 I 2.8195 %P/////////////////////////////////$%%%%%%\W%%%%)
148 Pu-241 il ! 2.42E-021 39.6703% l Pu-241/Am-241 1 12.670161 151 Pu-242 9 ESS THAN I 4.17E-061 0.0068%lf/////////////////////////////////A\%%%%%%%%%%%) 16s Am-24 f H s 1.91E-03 6 3.1310%i Am-241/Pu-238 I_ 3.19398? 171 Am-243 {_, WSS THAN t 2.29E-06 ! 0.0038 %if/////////////////////////////////A%%%%%%%%%%%4 18:1 Cm-242 j WSS THAN i 1.41E-05 : 0.0231 %F/////////////////////////////////A%%%%%%%%\%%%) 191 Cm-243/244 4 WSS THAN i 1.71E-051 0.0280 %P/////////////////////////////////AWWW\%%%%%%%%j l 20; U-234 ESS THAN 1.77E-05 i O.0290%i'/////////////////////////////////A%%%%%%%%%%%\1 211 U-235 ; WSS THAN 4.57E-061 0.0075 %F/////////////////////////////////A%%%%%%%%\%%%) 22d U-238 :l ESS THAN 4.57E-061 0.0075 % U-236/Am-241 ! O.002391 23: Co-80 s t 2.99E-031 4.9014 %i Am-241/U-238 ! 417.94311] 24' Ru- 106 i ESS THAN 1.46E-041 0.2393 %'f/////////////////////////////////AWWW%%\%%%%%\% 25t An- 108m I WSS THAN ' 1.54E-05 ! 0.0252% l'/////////////////////////////////A\%%%%%%%%%%%) 26i Sb-f25 WSS THAN i 5.66E-05 i 0.0928 %/////////////////////////////////AWWu%\%%%%\%%% 27d Cs- f 34 i ESS THAN t 2.19C-05 i 0.0359 %I Cs-134/Cs-137 1 0.004191 28 ( Cs- f 37 ' 5.23E-03 i 8.5734 %I Co-60/Cs-137 1 0.57170: 291 Ce-f44 ESS THAN ' 8.09E-05 : 0.1326 % Cs-137/Co-60 t 1'.74916i 30i Eu- f 52 ESS THAN
- 1.97E-04 i 0.3229 %J/////////////////////////////////AWWWW\%%%%%%\%
311 Eu-754 1 1.13E-04 i 0.1852% 4%%%%%%%%%%%%%\%%%%%%%\%%%%%%\W 321 Eu- fSE ESS THAN e 2.92E-051 0.0479% t/////////////////////////////////ABu%%%%%%%%%_4 Totals = = > l 6.10E-02!I 100.00 % i N/R - Not Reported l l 1 U
. l l
l l B & W RESULTS 11/3/95 - Saxton Samples . SYSTEM = = = > Dscharge Tank Sedenent Sarton N SA W I il 43.35021~ _ Semple Nam 6er Samode humber Total'TRU Content (nCi/g) l ! SX815950068 ll 9500068 - 01 r p 0.03271 ) i Total Uranium Content (nCl/g) CSample Type i Sediment 4 i Descnotion ! SCP, item 5 4 ! Results Normalized Nuclide Notes uCi/c to 100 % Nuclides RATIOS 11 H-3 i 7.76E-041 0.0048% f///////#//#///###/#//#///TA%%\WWWWW\wiwwj 24 C- f 4 LESS THAN 1.02E-03 i 0.0063% ///#//####/#//##/#//##/A%%%\W\W%%%%%4 3J Fe-55 4 ' 1.19E-03 I 0.0073 % I Fe-56/Co-60 I 0.005131 l 4f Ni-59 i 1.47E-021 0.0902% //#///#//#/##//##//#///#/A\W\%%%%%%\n%%1 l 54 N/-63 , I 9.93E-01 I 6.0924% //#/#/#/#/##/##//#N/#//AW%%%%%%%%%%Q 6i St-89 ESS THAN i N/R I 0.0000% //##//#///#/##//////##//#AM\\%\WWW%%%%%) l 71 Sr-90 4 1.44E-02 0.0883% $ Sr-90/Cs-137 i 0.000971 8i Nb-94 :i LESS THAN < 1.47E-04 0.0009% //#/#///#/#/##//###//////ABW%%%%%%%%%g , 91 Tc-99 LESS THAN I 3.42E-05 0.0002% f///#/###///###/#/#/##/A\%%%%%%%%\W%\\ l l
*101 /-f29 ,
LESS THAN i 4.86E-06 0.0000% //#/##//##//##/#/#/###Au\%%%%%%%%%%) l
" No-237 LESS THAN 1.20E-041 0.0007% y/##//##/#/#//N#/###///A\%%%\\\\whW\M\\g l 12. __
Pu-238 : 9.19E-04 ! 0.0056%q Pu-238/Am-241 0.491441 13s Pu-239/240 f 2.09E-031 0.0128% lf/#//#///#///#/##///#/#/#A\%%%\%%\%%%%%) 14e Pu-24 f '
! 3.83E-02 0.2350 %I Pu-241/Am-241 20.481281 151 Pu-242 i LESS THAN : 3.86E-06 0.0000 %if/##/#////##/#//#///////#/Au%\%%%%%%%%%)
b 164 Am-24 f ' 1.87E-03 0.0115 %i Am-241/Pu-238 2.034824 (* 171 Am-243 i LESS THAN i 2.81E-06I 0.0000% if//##///#////#//####/#///A%%\%%%%%%%%%i 184 Cm-242 il LESS THAN i 1.82E-05 : 0.0001 %i f/#//#//####//#/#///###Au%%%%%%%%%%\) 194 Cm-243/244 i LESS THAN l 2.63E-05 i 0.0002% jf//##/##//#/////##/#/##/Au%%%%%%%%%%g t 201 U-234 l LESS THAN i 1.87E-05 l 0.0001 %ifun###//#/////##/#/###A%%%%%%%%kWWW l 21J U-235 i LESS THAN i 7.48E-06 i 0.0000 %;f///////#//#/////##///#/##/A%%%%%%%%%%%g l 224 U-238 ' LESS THAN : 6.48E-061 0.0000 % I U-238/Am-241 1 0.00347 2 i 23 Co-60 . 2.32E-011 1.4234% l Am-241/U-238 i 288.580254 l 24i Ru-106 9 LESS THAN 2.56E-02 I 0.1571 % ; f/#///#///#///#/#///#/#////A%%%%%%%%\%WW) 25- Aa- foam ii LESS THAN 1.99E-03 i 0.0122% jf/##/#/////###///#////#/#Au\%%%%%%%%%%) 26' Sb- f25 i LESS THAN 1.66E-021 0.1018 %'f//##//#////#/#/#///#/#///Au%%%%%%%%%%g 27' Cs-f34 i 4 4.51E-03 0.0277 % Cs-134/Cs-137 1 0.00030-284 Cs-f37 ' 1.49E+01 91.4164 % Co-60/Cs-137 i 0.01557' 293 Ce-f44 GSS THAN i 2.43E-02 0.1491 % Cs-137/Co-60 I 64.22414-30i Eu- 152 i LESS THAN i 1.19E-02 0.0730 %'f/##/#/////##//###/##///A%%%%%\wu%%%%) 31! Eu-154 i LESS THAN ' 1.99E-03 0.0122% n%%\W%%%%%%%%\%%%%%%%%%%\%%%%%\n 321 Eu-f55 ; LESS THAN 1.15E-021 0.0706% F/##/////#/#///#//////#/#//Annuununu\WWM j' Totals = = > 1 163EMI 100.00 % i N/R - Not Reported o
l l O B & W RESULTS 11/3/95 - Saxton Samples - SYSTEM = = => storage we// Assin Sample l I Santon il SAW ! ll 0.03281 LSampde Numbed Samode Number i Total TRU Content (nCl/mi)
; SX88695011579500088-04/04AAl Il 0.00051 Sample Type il Total Uranium Content (nCi/ml)
Rosin (LP) i l l t Descnotum 1 SW Floor Resin l Results Normalized Nuclide Notes uCi/mi to 100 % Nuclides RATIOS 1! H-3 . 1 i 2.90E-041 0.0155% i f//#/###////#//#///#///////Anush\wn\\u\W\%\g l 21 C- 14 il LESS THAN ' 1.16E-04 i 3s Fe-55 i i 1.23E-051 0.0062%if/###/#/#/#/#///////#/////A%%%%%%%%%%%\1 0.0007% l Fe-56/Co-80 i 0.02500
)
4i Ni-59 i 6 3.52E-05i 0.0019%If##/#///#///#///##//###/A%%%%%%%%\%%gi l Si Ni-83 l s ' 2.81 E-031 0.1505 %g//#///#/#//##/#/#/#t#/#A%%%%%%%\%%\W\) 61 Sr-89 i l LESS THAN . N/R , 74 Sr-90 0.0000 %U/#//##/#/##////##//###A%%%%%%%%\%%%) 4 i 2.92E-041 0.0156 %i St-90/Cs-137 ! 0.00016i 84 Nb-94 4 1.05E-06 i 0.0001 %J/##/###////##///#/##/#A%%%%%%%%%%\ug 9: Tc-99 9 LESS THAN : 3.38E-06 I 0 0002%)f////##////##/#/#///###//Au%%%%%%\n%%%\ 10: /-f29 i LESS THAN 1.51 E-06 i 0.0001 %if/##/###/#///#////////#///A\\uww\\\WWWWugg 11l No-237 4 LESS THAN 1.02E-07 i 0.0000 % '/#/###/#/#///#####////A%%\%\%%%\%%%%1 12! Pu-238 :l : 9.64E-07I 0.0001 %i Pu-238/Am-241 1 0.83103' ! 13; Pu-239/240 ,i 1.74E-061 0.0001% j'##////////##/////##/#/##A%%%%\n%%%%%%t 14, Pu-24 f ? 2.88E-051 0.0015 %l Pu -241/Am-241 1 24.82759i 15! Pu-242 . LESS THAN ! 4.06E-09 ! 0.0000%p#####/#//##/###/###A%%%%\%%%%%%%(
) 16 Am-24 f f 1.16E-061 0.0001 %! Am-241/Pu-238 1 1.20332;
. \j l 17l Am -243 ___,,i LESS THAN i 2.40E-091 0.0000 %g#/#/##//#/##//####/#/A%%%%%%\%%%%%\ 18! Cm-242_. LESS THAN l 1.08E-08 0.0000 %g##/#//##/#/##/##/###A%%%%%%%%%%%g , 19 Cm-243/244 ; LESS THAN 1.70E-06 0.0000 %?##//#///////##/##/#//#//A%%%%%%\w%%%%\ } 204 U-234 LESS THAN 3.51 E-071 0.0000 %/##/##/#///##/#/#/#///#A%%%%%%%%%%\wi 21 L_,_U-235 LESS THAN 1.03E-071 0.0000 %v///#/#/#///#/#///#//##///Au%%%%\n%%%%%\ 225 U-238 41 LESS THAN I 7.31E-08 I 0.0000 %i U-238/Am-241 4 0.06302' ! 23r Co-80 f i 4.92E-04 0.0263 %i Am-241/U-238 15.86867: 24: Bu- fos LESS THAN ! 4.39E-03 0.2351 %P#####/#//#/#///#//#////AuguM\Wu%%%%%\ 254 Aa- f oSm LESS THAN 4.35E-04 0.0233 %J#/#/#/////#/////#/#/#///#A%%%%%%%%\%%%\ l 264 Sb- f 25 LESS THAN 2.38E-03 0.1274 %J##/#//##/#/#//#/#!##//A\%%%%%%%%%3\j, 271 Cs-134 LESS THAN 1.88E-04 0.0101 %i Cs-134/Cs-137 ! 0.00010 l 28," Cs- 137 ! 1.85E+00 99.0660 %; Co-80/Cs-137 1 0.000275 29 [ Ce- f 44 LESS THAN I 3.34E-03 0.1789 %i Cs-137/Co-80_ ! 3760.16260 301 Eu- f 52 _ LESS THAN ! 8.87E-04 0.0475 %P//###//#///#/#////#/#////Au%\%%%%%%%%%) Eu- 154 311 ' LESS THAN i 2.45E-041 0.0131 %i%%%%%%%%%%%%%%%%%%%%%%\%%%%%%W 32! Eu-f55 i LESS THAN 1.49E-03 ! o 0798%.f/#////#///##/#//#//////////Au%%%%%\WBWMt L Totais = = > 41.87E+00 f 100.00 % 1 N/R - Not Reported
\
k I l
f i I B & W RESULTS 11/3/95 - Saxton Samples {' . Y SYSTEM = = = > Storage We# Bes#1 SampAs Sexton .I SAW i l 4.7658 l Simode Number Samode Number ! Total TRU Content (nCi/g) LE X866960115 'l 9600066-04AB . $ 0.0003] Total Uranium Content (nCi/g) Sample Type a Rosin (SP) ij i Description ! SW Floor Rosin i Results Normalized Nuclide Notes uCi/a to 100 % Nuclides RATIOS 1i H-3 4 4 3.37E-04 i 0.0087% f/#/#///#//#///#///###/#/A%%%%%%%%%\wut 21 C-f4 Il LESS THAN i 1.96E-03 : 0.0504% ir/#/#/##/##/////#//#/##/A%%%%%%%%\%%%j 3ll Fe-55 4 : 2.26E-03 i 0.0586 %1 Fe-56/Co-60 I 0.32066 il 4 :i Ni-59 4 1.32E-03 i 0.0339 %if/#/##//#/##//#/#####/A%%%%%B\\\\W%\\\\ , 5I Ni-63 i 1.32E-03 : 0.0339 %;f/#//##/#/###/##/#/#/#A%%%%%%%%%%%% l 64 Sr-89 i LESS THAN N/R t 0.0000 %J##////#/#//####/##//#/A\%%%\%%%%%%%g 7t Sr-90 '! 8 1.01 E-01 : 2.5959 %; Sr-90/Cs-137 i D.02693i 8< Nb-94 i LESS THAN : 5.66E-06 ; 0.0001 %.,f#####/##/########/A%%%%%%%\%%%%t 9i Tc-99 d LESS THAN i 2.55E-05 i O.0007 %J/#######/######/##A\w\\%%%%\\%%%\n 10 ( , _ I- f 29 4 LESS THAN 8.54E-06 6 0.0002 %J#####/##/#/###/##/#AwWwwwunw\%%\n 11 , No-237 4 LESS THAN i 7.34E-06 i 0.0002 %J/####/#/#/#/####/#/#A%%%%%%%%\w%g 12s Pu-238 ll 1.23E-04 > 0.0032% Pu-236/Am-241 I 0.36607f 13i Pu-239/240 4 i 2.73E-04 > 0.0070% J/#############/#/#4\\\\%\wwuu\\%%%j
-143 Pu-24 f 4 i 4.02E-03 ! 0.1033% l Pu-241/Am-241 l 11.96429 l 15( Pu-242 'i LESS THAN i 5.28E-07 i 0.0000 %J#####/#/#/#/##/####A\\w\%%%%%%\\\\W) g#,h 16i Am-24 f ;
t 3.36E-04 ! 0.0066 % Am-241/Pu-236 1 2.73171i , '(j 17i 18d Am-243 Cm-242 LESS THAN LESS THAN i 3.94E-071 1.87E-06 i 0.0000 %JN#######/#/##/####A%Buuuu\%%%%%j l 0.0000% //####/#//##/#/#/####A%Bu%%%%\whWW) 19 Cm-243/244 - LESS THAN i 3.65E-06 I 0.0001 %//#/#########/#####A%%%%%\%%%%%%) 201 U-234 LESS THAN 1.52E-071 1 0.0000 %J###/###/#/#####/#/#A%%%%%%%%%%\\\\ 21i U-235 4 LESS THAN ! 8.80E-08 i 0.0000 %J###/#############Aw\%%%%%%%%%%1 22: U-238 1 LESS THAN ; 7.18E-061 0.0000 %: U-236/Am-241 ! 0.000211 232 Co-60 i , 7.11 E-03 ! 0.1827 %i Am-241/U-236 1 4679.665fM 241 Ru-106 l LESS THAN I 6.62E-031 0.1701% J' /###/####/#/##/#/#/#A%%\%%%%\\\\%%%) 25 il Aa- foSm .! LESS THAN i 6.60E-04 ! 0.0170% ! f##/##/###/#/##/###//An%%%\%%%%%%%) 26i Sb- f 25 a LESS THAN 4.23E-03 i 0.1067%J######/##/###//###/A\%%%%%%%%%%%) 274 Cs-134 4 l 6.91E-041 0.0178% j Cs-134/Cs-137 i 0.000181 28: Cs- f 37 1 3.75E+00 i 96.3611%l Co-60/Cs-137 1 0.00190i 29; Ce- 144 a LESS THAN ! 4.49E-03 i 0.1154 %! Cs-137/Co-60 1 527.426161 30.i Eu-152 il LESS THAN 1.49E-031 0.0363%lf##/#/##/######/#/##An%%%%%\%%%%%( 31 C_ Eu- f 54 LESS THAN i 3.70E-041 0.0095% hu\%%%%%\%%%%%%\%%%%%%%%\%%%%%%%) 32I Eu- f 55 ! LESS THAN 2.12E-031 0.0545% #######/##/####/###A\%%%%%%%%%%%i
;l Totais = = > ii 3.69E+00 a 100.00 % I N/R - Not Reported i
Nj
r l I B & W RESULTS 11/3/95 - Saxton Samples . SYSTEM = = = > Secondary Side of Steam Generator - Smn// Manway i Santon il B4W t !! O.36114
- Samode Numbeel Samode Number l Total TRU Content (nCi/g) r SNSM71 4 9509067-01 I ll 0.05611 Total Uranium Content (nCi/g) ;
Sample Type i S/G Scrape t ( i Descnotion SCP, Table B-10 i Results Normalized ! Nuclide Notes uCi/c to 100 % Nuclides RATIOS 11 H-3 d LESS THAN i 1.46E-034 14.3857% . f####//#////////###/#////A\ugung\wunu%g 2! C- f 4 LESS THAN i 4.67E-03 ' 3; 45.3926 %;f##/##//#/#///#/#/#/#/#Anh\%%%%\g\gghj ; Fe-55 LESS THAN : 3.25E-051 0.3159%I Fe-55/Co-60 i 4i Ni-59 0.21382i l i LESS THAN 7.81 E-051 0.7591 %if//#////#/##/##//#////#/#Au\\ngn%%gu\ ggt Si l Ni-63 . LESS THAN I 7.81 E-05 ! 0.7591 % , f/#//#/##//##/#/#####/Au%%%%%%%%%%\j 64 Sr-89 c LESS THAN ! N/R - 0.0000% i r/####//#/##/###/#/##Anhuw%%%%ugg\1 71 Sr-90 l 1.30E-04 ! 1.2636 %i Sr-90/Cs-137 1 0.06442i 8J Nb-94 ! LESS THAN ! 8.50E-04 I 8.2620%;f###////#//#/##//#####Agungungg\%wgj ! 91 Tc-99 LESS THAN i 5.14E-05 i O.4996% lf##/#####/#/#//#/##/#A%%%%%%%%%%%g 101 /-f29 1 LESS THAN ' 8.87E-05 : 0.8622% 7##/##/#/#/##//#/####Agnuunung\wwgt 113 No-237 I LESS THAN ! 2.43E-05 t 0.2362% f/#/#//##//#//###//#/##Aw\\ugungwngw\\ 124 Pu-238 e 1.26E-051 0.1225%l Pu-238/Am-241 1 0.347111 134 Pu-239/240 , - LESS THAN ! 9.11E-06 f 0.0885% f###/#/#/#/######///# Ann \%uwwwwwWht 141 Pu-24f I LESS THAN i 2.73E-04 2.6536%l Pu-241/Am-241 I 7.52006i 151 Pu-242 i LESS THAN i 6.77E-07 ' O.0066% l fl#/#///////#/#//#/////N//// Ann \w\\\n%nnwh\j
, \ 16i Am-24f i I 3.63E-05 t 0.3528 %i Am-241/Pu-238 1 2.88095 d I ) 17e Am-243 i LESS THAN t 5.53E-071 0.0054 %if/#####/#///##///#///#//A%w\%ww%%www\g 18a Cm-242 i LESS THAN i 2.10E-06 ! 0.0204 %ifl##//#//####/#//##!##AWuwwn%%%%%%g 19 lcm-244L24_4_, LESS THAN i 2.47E-06 i 0.0240 %if/#/#/##/###/##///##/#A%\w%%%%%%%%%)
20; U-234 i 1 2.80E-05 ; 0.2722 %;fl#########//#///#/##A\\\%%%%%%%%%%1 211 U-235 LESS THAN i 2.01E-06 i 0.0195 %;f/#/#//###/##/#/////##//AW%%%%%%Buhun 224 U-238 ' 2.61 E-05 i 0.2537 %I U-23s/Am-241 ' 0.71901 231 Co-do i 1.52E-04 i - 1.4774%l Am-241/U-238 ! 1.39060 4 24d Ru-106 i LESS THAN I 2.10E-041 2.0412%;f//#/##/####//#/#####A%\%%%%%%%uunt 25! Aa-f00m LESS THAN i 2.15E-05 : 0.2090 %;f/###///#//#//##////###/Ahu%%%%%B%%%\1 261 Sb- 125 i LESS THAN i 6.32E-05 271 Cs- f 34 LESS THAN i 2.30E-05 : 0.6143 %!r//##////##/#/#/#/#///////AnuwwWuhuuu\u3 0.2236 %! Cs-134/Cs-137 I 0.01494: 28l i Cs-137 ' 1.54E-03 i 14.9689% l Co-60/Cs-137 1 0.09670 j i 29i Ce- 144 J LESS THAN I 1.07E-04 l 1.0400 %i Cs-137/Co-60 1 10.13158d 301 Eu-152 LESS THAN l 1.91 E-04 ' 1.8565 %3#//##//###//##/#/#/////Au%%%%%%\%%%%\ 314 Eu- 154 ' LESS THAN ! 6.22E-05 ' O.6046 %huu\WunhuuhWWWWWu\W\uMWWuMWMWu\ut 321 Eu- f 55 ! LESS THAN ! 4 21E-05 0.4092% f/####//#//##//#/#/##//AWMWB\%%%%%%B1 a Totals = = > W 1.03E-02 4 100 00% i N/R - Not Reported s
- O f
t l l-l
/t B & W RESULTS 11/3/95 - Saxton Samples '
SYSTEM = = = > Steam Generator - Cold Leg Side. Bowl Sment i Saxton il BAW I li 23.99671 iSemode Number Semote Number Total TRU Content (nCl) l l_S_X825960111 1 9509067-03 i p 0.2493 il l Total Uranium Content (nCl) l Sample Type 4 S/G Smear i l l i Description ' SCP, item 1 1 Results Normalized I Nuclide Notes uCi to 100 % Nuclides 'iATIOS 1 H-3 ESS THAN i 7.50E-04 I 0.0209% ///##/##//##/#/#//#//////Awwwguggggggg 2 C-f4 ESS THAN ' 2.58E-03 i 0.0720% U/###/////#//##////#/#/#/AW\\n\ww\wngg\ng i 3: Fe-55 8 46E-021 2.3599 %I Fe-55/Co-60 ! 0.11132 j 41 Ni-59 1 1.57E-02 l 0.4379% l f#//#/#/###/#//#/#//#///A\\\\\%ggg\g\\\\\g)j 51 Ni-83 4 ! 1.43E+00 i 39 8896 %if/#/##/#/##/#///#i/###/AWWWggggggggg 6i St-89 i ESS THAN t N/R 0.0000% F///#/#//#//##//////#//#///Ag\\%%%%uwg\gg) 7' Sr-90 J ' 2.64E-01 ' 7.3642%1 St-90/Ca-137 ! 0.27557) Si Nb-94 i
' 5.17E-04 ' O.0144% if#//#/##///##/##/#/#/#/Angg\wngwnugug l
9i Tc-99 4 ESS THAN 2.52E-04 i 0.0070 %J/#////##//###///#//////#/Anw%g\wwwwwwn)
- 10) /-f29 i ESS THAN i 6.37E-05 : 0.0018 %J/////##/#/####/#/#//#//Anwwwwwww%nng 11J #p-237 LESS THAN 1.80E-04 ! 0.0050 %fu/###///#/#/##//#/#////Awwww\\wn\ng\wg l 12 :! Pu-238 ! 6.88E-041 0.0192 %i Pu-238/Am-241 1 0.39769l l l 131 Pu-239/240 i l 1.45E-031 0.0404 %Pu//#///##/#/#///#/#//////Awwwwgungunga l 141 Pu-24 7 i 1.99E-021 0.5551 %i Pu-241/Am-241 l 11.502891 l 3 15:1 Pu-242 i ESS THAN i 8.32E-061 0.0002 %P/###//#/#/#//#////#/##/A\ungwunnuwww l f i 16 t__Am-24 f a i 1.73E-03 ! 0.0483 %i Am-241/Pu-238 l 2.51453 l j
/ 174 Am-243 !! ESS THAN ' 2.91E-06 ! 0.0001 %U/######/#/#/#/#///#///A\ughnunnug\wn l l 16 L_ Cm- 242 j ESS THAN 1.37E-051 0.0004 %P//##//#/#///#/###/#/##Annuw\\un;\n\wn) l
( 191 Cm-243/244 i ESS THAN I 2.38E-051 0.0007% ifu//#/#///#////#//##//#/#Au\%%\u\\%%w\uut j l 20 L U-234 1 ESS THAN 1 1.38E-041 0.0038 %if#/#/####////###//#/#//A%%%%%uunn\uw) 1 21:1 U-235 i ESS THAN I 6.52E-051 0.0018 %!'//#/#/#/#/##//#///##//#Au%nnunuwwwuq l l 22d U-238 I ESS THAN ' 4.61 E-05 ! 0.0013 % U-238/Am-241 ! 0.026854 1 23 1 Co-80 3 7.60E-01 1 21.2000 %I Am-241/U-238 37.52711i
-244 Bu-106 'I~~ ESS THAN i 1.24E-02 ! 0.3459 %J##//#//#/#/##//#/#//////AWWWWWW\%\%\%%)
254 Aa-f00m i ESS THAN i 1.23E-03 i 0.0343 %if///##/##/#/#/#/####/#AWWWWW\n%%%\%% 261 Sb-f25 :( ESS THAN ! 3.68E-03 6 0.1027 %,f/#//#/#/##/#//#/#####A\\wwwwwwwwh\%) [ l 27'L Cs-f34 a ESS THAN l 1.74E-03 ! 0.0485 %i Cs-134/Cs-137 i 0.00182j l i 28:l Cs-f37 , I 9.58E-011 26.7232%: Co-60/Cs-137 i 0.79332i l 29 T Ce- 144 ESS THAN : 4.67E-03 ; 0.1303 % Cs-137/Co-60 : 1.260531 I 30 L,_ Eu- f52 a ESS THAN t 1.22E-02 ! 0.3403 %N/#///#/#/#/##///#/#/////Aku%%%%%\%%%%) j 311 Eu-f54 4 ESS THAN 7.05E-03 i 0.1967% hWUMW\%%%%%%\WWWW\\%%BuhuMuu%%%% l 32! Eu-f55 il GSS THAN 1.22E-03 ! 0.0340% )//##/#/////#//]//##/##i//A%%%Wunwuu\W\W f'~~~Totals = = > I 3.58E+00 h 100.00% l l N/R - Not Reported ('v) l l l l
1 I 1 -
\
B&W NUCLEAR ENVIRONMENTAL SERVICES, INCORPORATED 1 NUCLEAR ENVIRONMENTAL LABORATORIES i .i i
- DATA REPORT NUMBER 9510024 1 ,
RADIOCHEMICAL ANALYSIS RESULTS i l 1 i Submitted to I 1 GPU Nuclear Corporation l Saxton Site - Decommissioning Project t i i \ December 6,1995 J i 5 4 i 1 Released by: \ 6Mr b . /2/6/T5' James L. Clark, Project Manager Nuclear Environmental Laboratories
l GPU-Saxton Data Report December 6,1995 DATA REPORT
SUMMARY
l l$ntSbratory AdjecF$$mniafyjY0 #'
$h 4 x ",s ~
- Plant Site: Saxton Decommissioning Project
- Shipment Number: 95-007
- Receipt Date: October 5,1995 -
- NEL COC Number: SN-12537,12538,12539,12543
- NEL SDG Number: 9510024
- NEL Proposal: NEL5-1258
- Contract Number: 0397355
- Analysis Service: Routine
- Laboratory QC: Routine - Batch v
b i
n GPU Saxton Data Report December 6,1995 l l
; , M $ SamhleiD3ii0iiirh[NEoiipllIDd$rasisNsltidi50rNinary; - , ,
' ?CUh$$ierilDE $NEd$Arnhld!I65 ISarhhidNNA$[$lices? ETdialMiss'h) ^
SX801950127 9510024-01 Core Bore / 2 208.90 l l SX811950128 9510024-02 4 Core Bore / 4 405.13 ' l SX821950129 9510024-03 Core Bore / 4 439.34 l SX832950130 9510024-04 Core Bore / 2 190.07 SX832950131 9510024-05 Core Bore / 2 185.17 SX832950132 9510024-06 Core Bore /1 111.27 l l SX841950133 9510024-07 Core Bore / 6 595.08 i
,Q SX852950134 95100,24-08 Core Bore / 4 - 405.61 SX861950135 9510024-09 Core Bore / 4 382.37 i
SX861950136 9510024-10 Core Bore / 2 208.22 SX811950137 9510024-11 Core Bore /1 82.21 SX811950138 9510024-12 Core Bore /1 85.65 SX811950139 9510024-13 Core Bore /1 89.29 SX811950140 9510024-14 Core Bore /1 100.36 SX822950141 9510024-15 Core Bore / 2 208.48 SX822950142 9510024-16 Core Bore / 2 192.10 SX822950143 9510024-17 Core Bore /1 98.93 SX822950144 9510024-18 Core Bore /1 105.21 SX822950145 9510024-19 Core Bore / 2 208.37 I o i d ii i f l
$ ~ GPU-Saxton Data Report December 6,1995 Minplel dellUsfybi6sp'I Ch"sifa cieElstids IS0minAh $Chn tirid e d
!LCustumeflD e ENEOSnindlilDJ [Samhlhyhaf[Sdse'sj > <$jdtaI(Ma~ssl (g)r SX822950146 9510024-20 Core Bore / 2 191.29 SX822950147 9510024-21 Core Bore /1 101.58 SX822950148 9510024-22 Core Bore /1 103.87 SX841950149 9510024-23 Core Bore /1 80.62 SX841950150 9510024-24 Core Bore /1 99.24 SX841950151 9510024-25 Core Bore /1 106.20 SX841950152 9510024-26 Core Bore /1 111.28 SX861950153 9510024-27 Core Bore / 2 208.67 SX861950154 9510024-28 Core Bore /1 100.22 SX862950155 9510024-29 Core Bore / 2 187.05 SX862950156 9510024-30 Core Bore /1 105.41 j SX862950157 9510024-31 Core Bore / 2 192.60 I i
SX862950158 9510024-32 Core Bore /1 117.38 l SX862950159 9510024-33 Core Bore / 2 198.75 SX862950160 9510024-34 Core Bore /1 109.72 l SX862950161 9510024-35 Core Bore / 2 216.91 SX862950162 9510024-36 Core Bore /1 93.18 SX862950163 9510024-37 Core Bore /1 111.91 SX862950164 9510024-38 Core Bore /1 109.11 Note: Sample Number 9510024-36 contained a section of Rebar Metal which was 7 removed prior to analysis. This Rebar section weighed 57.74 g. The mass d given in this Table has been corrected for the Rebar Section weight. iii
_ _ . . ~_ ...._. _ _ ___ _ l l
. I GPU.Saxton Data Report December 6,1995 1
+ , ..,...o...x . ;.:..., . . . . I 13
, ,,;'4 BWAnd. .. .. . ; .:.. ...lytical Resultsk ', ':e- '; , , , " ~ l I
See Attached Report Sheets l e t 0 IV
~
O O ' O GPU Saxton Results Summary B&W NESI, Nuclear Environmental Laboratories 2/06/95 (Background) Area 1-Floor Area 2-Floor Area 3 -4 ' Wall Area 6 - 4 ' Wal SX801950127 SX811950128 SX821950129 SX832950130 SX832950131 NEL ID 9510024-01 9510024-02 9510024-03 9510024-04 9510024-05 NUCLIDE uCilg +/- 2S uCi/g +/- 2S uCilg +/- 2S uCi/g +/- 2S uCi/g +/- 2S Co-60 < 9.89E-09 - 4.97E-07 7.%E-08 6.37E-06 6.64E-06 2.58E-07 7.74E-08 1.32E-06 1.62E-07 Ru-106 < 1.12E-07 -
< 3.56E-06 -
< 5.03E-06 -
< 9.55E-07 -
< 5.26E-07 -
Cs-134 < 1.64E-08 - 3.81 E-07 7.06E-08 7.67E-07 1.43E-07 < 4.96E-08 .- < 5.05E-08 - Cs-137 1.16E-07 2.14 E-08 1.48E-03 2.02E-04 4.69E-03 8.02E-04 5.86E-05 8.02E-06 1.42E-05 2.44E-06
. Ce-144 < 5.58E-08 -
< 2.28E-06 -
< 3.57E-06 -
< 6.25E-07 -
< 3.37E-07 -
Am-241 < 5.16E-08 -
< 2.08E-06 -
< 2.68E-06 -
< 5.66E-07 -
< 2.55E-07 -
p.g-108m < 1.11 E-08 -
< 3.71 E-07 --
< 5.29E-07 -
< 1.01 E-07 -
< 5.90E-08 -
Eu-152 < 1.14E-07 -
< 3.54E-07 -
< 7.32E-07 - < 1.42E-07 -
< 4.42E-07 - ,
Eu-154 < 4.06E-08 -
< 9.29E-08 -
< 1.29E-07 -
< 1.46E-07 -
< 1.02E-07 -
Eu-155 < 1.07E-07 -
< 1.12E-06 -
< S.67E-07 -
< 3.03E-07 -
< 1.63E-07 -
Sb-125 < 2.97E-08 -
< 1.91 E-06 -
< 2.64E-06 - < 5.15E-07 -
< 2.27E-07 -
H-3 < 1.73E-05 -
< 1.69E-05 -
< 1.89E-03 -
<- 1.80E-05 -
< 1.71 E-05 -
C-14 < 2.13E-05 -
< 1.74E-05 -
< 2.12E-05 -
< 2.06E-05 -
< 2.37E-05 -
Ni-59 < 3.35E-06 --
< 4.11 E-06 -
< 4.36E-06 -
< 1.33E-05 -
< 4.24E-06 -
Ni-63 < 5.32E-06 -
< 5.13E-06 -
2.35E-05 3.66E-06 < 4.73E-06 -
< 5.08E-06 -
Nb-94 < 1.21 E-08 -
< 4.25E-08 -
< 8.38E-08 -
< 3.87E-08 -
< 4.96E-08 -
Tc-99 < 1.50E-06 - < 1.45E-06 -
< 1.52E-06 - < 1.64E-06 -
< 1.52E-06 -
I-129 < 3.39E-06 -
< 2.99E-06 -
< 8.94E-06 -
< 2.08E-06 -
< 1.89E-06 -
Gross Alpha < 3.90E-06 -
< 2.12E-06 -
< 106E-06 - < 1.99E-06 -
< 2.45E-06 -
Sr-90 NA NA NA NA < 5.08E-06 - Page 1 of 8
O O O GPU Saxton Results Summary B&W NEST Area 6 - Area 6 - Area 4 - 812 ', Nuclear 12/06/95 Area 6 - 4 ' Wall FloorEnvironmental Area 5-Tunnel Laboratori[s Wa 1 Floor Shallow Floor Deep GX832950132 SX841950133 SX852950134 SX861950135 SX861950136 NEL ID 9510024-06 9510024-07 9510024-08 9510024-09 9510024-10 NUCLIDE uCilg +/- 2S uCi/g +/- 2S uCi/g +/- 2S uCilg +/- 2S uCi/g +/- 2S Co-60 3.43E-08 2.06E-0E 9.81 E-08 2.36E-08 3.10E-08 1.66E-07 3.19E-05 3.40E-06 2.15E-06 2.64E-07 Ru-106 < 2.02E-07 -
< 2.66E-07 -
< 221E-07 -
< 1.53E-05 -
< 3.48E-06 -
Cs-134 < 2.60E-08 -
< 2.11 E-08 -
< 1.76E-08 -
1.87E-06 4.36E-07 < 9.12E-08 - Cs-137 1.24E-06 2.16E-07 1.35E-05 1.88E-06 3.06E-06 5.26E-06 1.38E-02 1.88E-03 9.45E-04 129E-04 Ce-144 < 1.21 E-07 -
< 1.42E-07 .
< 1.09E-07 -
< 9.66E-06 -
< 2.18E-06 -
' Am-241 < 9.57E-08 -
< 1.08E-07 -
< 8.63E-08 -
< 1.02E-06 -
< 2.01 E-06 -
Ag-108m < 2.02E-08 -
< 2.73E - < 1.E :.-08 -
< 9.16E-07 -
< 3.61 E-07 -
Eu-152 < 1.78E-07 -
< 1.43E-07 -
< 1.29E-07 -
< 4.49E-06 -
< 6.36E47 -
Eu-154 < 6.36E-08 -
< 5.36E-08 -
< 4.53E-08 -
< 5.14E-07 -
< 1.39E-07 -
Eu-155 < 8.33E-08 -
< 7.12E-08 -
< 3.46E-08 -
< 4.79E-06 -
< 4.11E-07 -
Sb-125 < 6.16E-08 -
< 1.07E-07 -
< 6.31 E-08 -
< 8.15E-06 -
< 1.86E-06 -
H-3 < 1.93E-05 -
< 222E-05 -
< 2.57E-05 -
< 2.35E-05 -
< 2.14E-05 -
C-14 < 2.62E-05 -
< 2.80E-05 -
< 3.49E-05 -
< 2.43E-05 -
< 2.16E-05 ---
Ni-59 < 5.37E-06 - < 5.66E-06 -
< 1.03E-05 -
< 620E-06 -
< 5.50E-06 -
Ni-6) < 4.34E-06 - 2.28E-05 3.58E-06 < 4.61E-06 - 4.41 E-05 6.78E-06 7.99E-06 1.28E-06 Nb-94 < 1.91 E-08 -
< 2.13E-08 -
< 2.98E-08 -
< 4.29E-07 -
< 5.47E-08 -
Tc-99 < 1.52E-06 - < 1.55E-06 -
< 1.49E-06 -
< 2.11E-06 -
< 2.17E-06 -
l-129 < 1.34E-06 -
< 1.34 E-06 -
< 1.85E-06 -
< 2.54E-06 -
< 2.30E-06 -
Gross Alpha < 2.52E-06 -
< 2.98E-06 -
< 2.39E-06 -
< 224E-06 -
< 2.66E-06 -
Sr-90 NA NA NA < 7.07E-06 - NA Page 2 of 8
'N
/]
GPU Saxton Resu!!s Summary B&W NESI Nuclear Environmental Laboratories Area 2 12/06/95 Area 1 Floor Area 1 Floor Area 1 Floor Area 1 Floor 5' Wall Surface SX811950137 SX811950138 SX811950139 SX811950140 SX822950141 NEL ID 9510024-11 9510024-12 9510024-13 9510024-14 9510024-15 NUCLIDE uCilg +/- 2S uCi/g +/- 2S uCi/g +/- 2S uCilg +/- 2S uCi/g +/- 2S Co-60 1.90E-06 2.46E-06 2.92E-07 6.64 E-08 2.12E-07 6.60E-08 5.27E-08 4.18E-08 1.60E-07 2.96E-08 Ru-106 < 1.00E-05 -
< 3.32E-06 -
< 7.17E-07 -
< 5.44E-07 -
< 7.06E-07 -
Cs-134 2.14E-06 3.20E-07 4.91 E-07 9.%E-08 < 620E-08 --
-< 5.79E-08 -
< 2.75E-08 -
Cs-137 7.18E-03 1.00E-03 1.87E-03 3.18E-04 3.14E-05 4.38E-06 1.15E-05 1.60E-06 1.27E-04 2.16E-05 Ce-144 < 4.74 E-06 -
< 2.24E-06 -
< 2.34E-07 -
< 2.99E-07 -
< 4.97E-07 -
Am-241 < 3.65E-06 -
< 1.77E-06 -
< 2.81 E-07 ---
< 2.96E-07 -
< 3.73E-07 -
Ag-108m < 1.02E-06 -
< 3.45E-07 -
< 7.61 E-08 -
< 5.29E-08 -
< 7.38E-08 -
Eu-152 2.16E-06 8.16E-07 2.61E-06 5.80E-07 1.18E-06 4.14E-07 < 4.80E-07 -
< 1.71E-07 -
Eu-154 < 2.34E-07 -
< 1.04 E-07 -
< 1.53E-07 -
< 1.49E-07 -
< 5.23E-08 -
Eu-155 < 2.60E-06 -
< 1.12E-06 -
< 1.78E-07 -
< 1.09E-07 -
< 2.43E-07 -
Sb-125 < 4.51 E-06 --
< 1.85E-06 -
< 2.85E-07 -
< 2.30E-07 -
< 3.68E-07 -
H-3 < 2.54E-05 -
< 2.46E-05 -
< 2.28E-05 -
< 2.25E-05 -
< 2.36E-05 -
, C-14 < 2.74E-05 -
< 2.56E-05 -
< 4.34E-05 -
< 2.58E-05 -
< 4.11E45 -
Ni-59 < 2.82E-06 -
< 3.88E-06 -
< 5.10E-06 -
< 9.31E-06 -
< 4.56E-06 -
Ni-63 < 4.09E-06 -
< 4.78E-06 -
< 5.36E-06 -
8.56E-06 1.37E-06 < 5.10E-06 - Nb-94 < 1.05E-07 -
< 4.97E-08 -
< 4.93E-08 -
< 5.63E-08 -
< 1.95E-08 -
Tc-99 < 2.29E-06 -
< 1.47E-06 -
< 2 39E-06 -
< 2.26E-06 -
< 1.75E-06 -
l-129 < 5.75E-06 -
< 1.61 E-06 -
< 1.46E-06 . < 1.18E-06 -
< 1.67E-06 -
Gross Alpha < 3.56E-06 -
< 1.91 E-06 -
< 2.54E-06 -
< 1.19E-06 -
< 2.16E-06 -
Sr-90 NA NA NA NA NA Page 3 of 8
o o~ l l L D GPU Saxton Results Summary Area 6 - B&W NESI, Nuclear Environmental Laboratories Above H2 O Level Above H2O Level im 5' Wall Surface Area 6 - 5 ' Wall Area 6 - 5' Wall . Area 2-Wall 5' Area 6 - 5' Wall SX822950142 SX822950143 SX822950144 SX822950145 SX822950146 NEL ID 9510024-16 9510024-17 9510024-18 9510024-19 9510024-20 NUCLIDE uCilg +/- 2S uCilg +/- 2S uCi/g +/- 2S uCilg +/- 2S uCVg +/- 2S Co-60 1.46E-05 1.54E-06 121E-07 2.70E 08 < 2.95E-08 -- 9.12E-08 1.99E-08 2.67E-06 2.90E-06 Ru-106 < 1.24E-06 - < 2.69E-07 -
< 2.71 E-07 --
< 5.61 E-07 -
< 4.55E-07 -
Cs-134 < 1.02E-07 -
< 2.76E-08 -
2.70E-08 1.75E-08 . 4.33E-08 1.74E-08 < 5.16E-08 - l Cs-137 2.46E-04 3.36E-05 1.38E-05 2.36E-06 8.98E-06 126E-06 1.32E-04 1.80E-05 2.78E-05 3.80E-06 Ce-144 < 7.15E-07 - < 1.76E-07 -
< 1.33E-07 -
< 3.55E-07 -
< 2.40E-07 -
Am-241 < 4.30E-07 - < 1.40E-07 - < 1.08E-07 -
< 327E-07 -
< 121E-07 -
Ag-108m < 1.30E-07 - < 3.11E-08 - < 2.75E-08 -
< 5.94E-08 -
< 4.74E-08 -
Eu-152 < 8.91 E-07 - < 1.84 E-07 -
< 1.77E-07 -
< 1.45E-07 --
9.31 E-06 1.07E-06 Eu-154 < 1.47E-07 - < 6.60E-08 -
< 6.36E-08 - < 4.83E-08 -
5.87E-07 1.47E-07
- Eu-155 < 1.16E-07 - < 7.55E-08 -
< 1.04E-07 -
< '1.75E-07 -
< 3.00E-07 -
St -125 < 6.08E-07 - < 1.20E-07 -
< 9.60E-08 -
< 2.97E-07 -
< 1.73E-07 -
H-3 < 2.16E-05 - < 2.17E-05 -
< 222E-05 -
< 2.33E-05 -
2.00E-04 2.90E-05 C-14 < 3.82E-05 - < 2.78E-05 - < 3.38E-05 - < 327E-05 - < 3.81 E-05 - Ni-59 < 6.86E& - < 3.36E-06 -
< 3.64E-06 - < 420E-06 -
< 1.02E-05 -
Ni-63 6.33E-05 9.64E-06 < 4.71 E-06 - < 4.88E-06 -
< 4.71E-06 - 8.39E-06 1.34E-06 Nb-94 < 9.61 E-08 - < 2.33E-08 - < 325E-08 -
< 1.94E-08 - < 5.18E-08 -
Tc-99 < 1.74E-06 - < 2.30E-06 - < 2.08E-06 - < 1.43E-06 - < 1.48E-06 - I-129 < 3.68E-06 - < 1.20E-06 - < 1.32E-06 -
< 2.12E-06 -
< 1.77E-06 -
Gross Alpha < 1.59E-06 -- < 1.68E-06 -
< 2.03E-06 - < 220E-06 - < 123E-06 -
NA NA NA NA NA Sr-90
\
Page 4 of 8
<- , m ,
GPU Saxton Results Summary Above 112 0 B&W NESI. Nuclear Environmental Laboratories 12/06/95 Area 6 - 5' Wall Above 1120' Area 6 Ceiling Area 6 Ceiling Area 6 Ceiling-SX822950147 NE32h50I48 all SX841950149 SX84ii,50150 SX841950151 NEL ID 9510024-21 9510024-22 9510024-23 9510024-24 9510024-25 NUCLIDE uCi/g +/- 2S uCilg +/- 2S uCilg +/- 2S uCi/g +/- 2S uCVg +/- 2S . Co-60 1.02E-07 2.74E-08 8.79E-07 1.29E-07 1.47E-06 2.12E-07 918E-07 1.38E-07 3.20E-07 5.68E-08 Ru-106 < 4.19E-07 -
< 6.87E-07 -
< 8.97E-07 -
< 7.17E-07 -
< 4.01E-07 -
Cs-134 2.89E-08 2.24E-08 < 8.15E-08 -
< 1.19E-07 -
< 9.23E-08 -
< 4.75E-08 -
Cs-137 3.21 E-05 4.48E-06 1.49E-05 2.10E-06 6.47E-06 9.14E-07 1.62E-06 2.54E-07 1.08E-06 1.98E-07 i Ce-144 < 2.00E-07 - < 3.34E-07 -
< 4.09E-07 -
< 3.35E-07 -
< 2.24E-07 - ,
Am-241 < 9.22E-08 -
< 2.64E-07 -
< 4.11 E-07 -
< 3.25E-07 -
< 2.00E -
Ag-108m < 4.25E-08 - < 6.72E-08 -
< 8.98E-08 -
< 7.40E-08 -
< 4.29E-08 -
Eu-152 3.28E-07 2.26E-07 7.57E'06 1.18E-06 1.07E-05 1.70E-06 9.05E-06 1.38E-06 3.30E-06 6.00E-07 Eu-154 < 8.32E-08 - 3.68E-07 2.00E-07 6.35E-07 2.38E-07 6.10E-07 2.34E-07 < 1.37E-07 - Eu-155 < 9.70E-08 -
< 1.33E-07 -
< 2.30E-07 -
< 2.00E-07 -
< 1.29E-07 ' -
l Sb-125 < 1.6SE-07 - < 2.29E-07 -
< 2.39E-07 -
< 1.86E-07 -
< 1.14E-07 - >
H-3 < 2.45E-05 - 7.54E-05 1.11 E-05 1.39E-04 2.06E-05 < 1.91 E-05 -
< 1.82E-05 -
C-14 < 2.84 E-05 -
< 3.21 E-05 -
< 3.17E-05 - < 2.25E-05 -
< 2.29E-05 - ,
Ni-59 < 5.07E-06 -
< 3.03E-06 -
< 3.80E-06 - < 3.89E-06 -
< 9.89E-06 -
< 5.09E-06 < 4.87E-06 < 8.11E-06 Ni-63 < 4.36E-06 - < 3.95E-06 - - - -
Nb-94 < 2.35E-09 - < 8.86E-08 - < 1.09E-07 -
< 9.80E-08 -
< 4.74E-08 -
l Tc-99 < 1.56E-06 - < 1.46E-06 - < 1.48E-06 -
< 1.43E-06 -
< 1.48E-06 -
I.129 < 1.28E-06 - < 2.60E-06 - < 4.65E-06 -
< 1.34E-06 - < 2.71E-06 -
Gross Alpha < 1.54E-06 - < 1.13E-06 -
< 2.58E-06 -
< 1.26E-06 - < 1.86E-06 -
Sr-90 NA NA NA NA NA r a Page 5 of 8
, (~ ,
(q,- ) i v
) (s,_j )
GPU Sarton Results Summary B 12/06/95 Area 6 Ceiling hh pin 8" NESI, Floor Nuclear N h" Environmental Floor Laboratories Area 6 S. Wall 5' Area 6 S. Wall 5' SX841950152 SX861950153 SX861950154 SX862950155 SX862950156 NEL ID 9510024-26 9510024-27 9510024-28 9510024-29 9510024-30 NUCLIDE uCilg +/- 2S uCi/g +/- 2S uCi/g +/- 2S uCi/g +/- 2S uCi/g +/- 2S Co-60 7.64E-08 2.18E-08 1.18E-05 1.24 E-06 1.27E-07 3.84 E-08 8.81 E-06 9.30E-07 1.25E-06 1.62E-07 Ru-106 < 2.12E-07 --
< 1.65E-06 -
< 2.90E-07 -
< 1.28E-06 -
< 8.95E-07 -
Cs-134 < 2.60E-08 -
< 9.88E-08 -
< 3.49E-08 -
< 1.08E-07 -
< 7.31E-08 -
Cs-137 1.89E-06 2.70E-07 5.13E-04 7.02E-05 3.66E-06 5.08E-07 1.19E-04 1.66E-05 4.87E-05 6.80E-06 Ce-144 < 1.04 E-07 -
< 1.00E-06 -
< 1.03E-07 -
< 6.05E-07 -
< 4.22E-07 -
Am-241 < 8.75E-08 -
< 9.13E-07 -
< 1.47E-07 -
< 6.64E-08 -
< 3.29E-07 -
Ag-108m < 2.31 E-08 -
< 1.70E-07 -
< 3.04E-08 -
< 1.31 E-07 -
< 9.19E-08 -
Eu-152 4.84 E-07 1.91 E-07 < 7.75E-07 -
< 2.58E-07 -
8.40E-06 1.33E-06 1.02E-05 1.28E-06 Eu-154 < 4.45E-08 -
< 1.32E-07 -
< 9.37E-08 -
5.93E-07 1.75E-07 3.66E-07 1.66E-07 Eu-155 < 1.05E-07 -
< 4.88E-07 -
< 4.36E-08 -
< 2.03E-07 -
< 2.24E-07 -
Sb-125 < 5.07E-08 -
< 8.34E-07 -
< 9.69E-08 -
< 4.91 E-07 -
< 3.32E-07 -
H-3 < 1.94E-05 -
< 2.17E-05 -
< 2.25E-05 -
3.12E-04 4.38E-05 1.35E-04 1.94E-05 C-14 < 2.58E-05 -
< 2.50E-05 -
< 3.09E-05 -
< 2.69E-05 -
< 2.14E-05 -
Ni-59 < 4.13E-06 -
< 4.94 E-06 -
< 3.00E-06 -
< 3.90E-06 -
< 3 62E-06 -
Ni-63 < 4.74 E-06 --- 6.62E-06 1.07E-06 < 4.73E-06 - 227E-05 3.56E-06 < 4.78E-06 - Nb-94 < 2.18E-08 -
< 9.03E-08 --
< 3.51 E-08 -
< 1.03E-07 -
< 728E-08 -
Tc-99 < 1.45E-06 -
< 1.82E-06 -
< 1.49E-06 -
< 1.91E-06 -
< 2.07E-06 -
l-129 < 1.39E-06 -
< 3.54E-06 -
< 2.08E-06 -
< 4.81 E-06 -
< 1.29E-06 -
Gross Alpha < 1.30E-06 -
< 3.55E-06 -
< 2.39E-06 -
< 2.65E-06 -
< 1.29E-06 -
Sr-90 NA NA NA NA NA Page 6 of 8
(g . "N (N k
%J GPU Saxton Results Summary B&W NESI, Nuclear Environmental Laboratories 12/06/95 Area 6 - CV Area 6 - CV Area 6-5' Wall Area 6-5' Wall outside Wall Outside Wall High Flux Area SX862950157 ~ SX862950158 SX862950159 SX862950160 SX862950161 NEL ID 9510024-31 9510024-32 9510024-33 9510024-34 9510024-35 NUCLIDE uCi/g +/- 2S uCi/g . +/- 2S uCilg +/- 2S uCi/g +/- 2S uCi/g +/- 2S Co-60 1.57E-05 1.68E-06 1.70E-06 1.84E-07 1.09E-06 1.16E-07 1.15E-07 3.20E-08 9.02E-06 9.34E-07 Ru-106 < 1.97E-06 -
< 3.50E-07 -
< 2.98E-07 -
< 2.47E-07 -
< 1.13E-06 -
Cs-134 < 1.64 E-07 -
< 4.54E-08 -
< 2.04E-08 -
3.46E-08 2.36E-08 < 1.16E-07 - Cs-137 2.61 E-04 3.58E-05 3.94E-06 6.78E-07 3.74E-05 6.40E-06 1.39E-06 2.00E-07 2.33E-04 3.98E-05 Ce-144 < 1.11 E-06 -
< 1.89E-07 -
< 1.95E-07 -
< 1.28E-07 -
< 9.13E-07 -
Am-241 < 1.05E-06 -
< 1.56E-07 -
< 1.45E-07 -
< 1.27E-07 -
< 6.87E-07 -
Ag-108m < 2.07E-07 - < 3.58E-08 -
< 3.46E-08 -
< 2.78E-08 -
< 4.41 E-07 -
Eu-152 7.81 E-06 1.47E-06 1.06E-05 1.20E < 1.62E-07 -
< 2.34E-07 -
3.15E-05 3.50E-06 Eu-154 5.55E-07 2.58E-07 5.25E-07 1.12E-07 < 4.66E-08 -
< 7.00E-08 -
1.94E-06 3.12E-07 Eu-155 < 5.52E-07 -
< 7.14E-08 -
< 4.57E-08 -
< 5.47E-08 -
< 4.82E-07 -
Sb-125 < 9.17E-07 -
< 1.20E-07 -
< 1.37E-07 -
< 7.61 E-08 -
< 623E-07 -
H-3 1.37E-04 2.02E-05 1.11 E-04 1.65E-05 < 2.54E-05 - < 1.67E-05 - 6.93E-04 9.64E-05 C-14 < 2.31 E-05 -
< 3.40E-05 -
< 2.66E-05 -
< 2.11E-05 -
< 2.44E-05 -
Ni-59 < 3.61 E-05 -
< 4.86E-06 -
< 4.18E-06 -
< 4.90E-06 -
< 6.34E-06 -
Ni-63 1.44 E-04 2.22E-05 < 5.96E-06 -
< 627E-06 -
< 6.15E-06 -
<' 5.90E-06 -
Nb-94 < 1.55E-07 -
< 3.96E-08 -
9.01 E-08 2.54 E-08 < 2.60E-08 -
< 1.16E-07 -
Tc-99 < 1.97E-06 -
< 1.79E-06 -
< 1.90E-06 -
< 1.87E-06 -
< 1.86E-06 -
I-129 < 2.57E-06 -
< 6.74E-06
- < 1.42E-06 -
< 2.04E-06 -
< 5.15E-06 -
Gross Alpha < 2.42E-06 - < 3.36E-06 -
< 2.00E-06 - < 2.24E-06 -
< 2.99E-06 -
Sr-90 NA NA NA NA NA Page 7 of 8
- s n
.J -U GPU Saxton Results Summary B&W NESI, Nuclear Erwironmental Laboratories 12/06/95 High Flux Area High Flux Area High Flux Area NEL ID h)bf- 1 37 bf NUCLIDE uCilg +/- 2S uCilg . +/- 2S uCi/g - +I- 2S Co-60 7.85E-06 8.18E-07 4.42E-06 4.74E-07 2.59E-06 ' 2.82E-07 Ru-106 < 1.28E-06 -
< 1.08E-06 -
< 6.03E -
Cs-134 < 1.64E-07 --- < 1.4 4E-07 -
< . 8.08E-08 -
Cs-137 850E-06 1.47E-06 2.19E-06 4.06E-07 1.66E-06 2.48E-07 Ce-144 < 6.54E-07 -
< 5.83E-07 -
< 2.64E-07 -
Am-241 < 5.36E-07 -
< 3.43E-07 -
< 2.13E-07 -
Ag-108m < 127E-07 -
< 1.13E-07 -
< 521E-08 -
Eu-152 6.85E-05 7.24E-06 6.08E-05 6.42E-06 3.20E-05 . 328E-06 Eu-154 3.88E-06 5.72E-07 2.67E-06 3.82E-07 . 1.32E-06 2.30E-07 Eu-155 < 4.20E-07 -
< 6.96E-07 -
< 1.32E-07 -
- Sb-125 < 3.10E-07 -
< 2.64E-07 -
< 1.37E-07 -
H-3 2.10EA 3.10E-05 1.89E-03 2.50E-04 4.55E-04 628E-05 C-14 < 1.41E-05 -
< 2.60E-05 -
< 2.83E-05 -
Ni-59 < 4.76E-06 -
< 9.11E-06 -
< 5.90E-06 -
Ni-63 < 6.94E-06 - 2.30E-05 3.60E-06 < 720E-06 - Nb-94 < 1.75E-07 -
< 1.33E-07 -
< 9.0$E-OS -
Tc-99 < 1.47E-06 -
<- 1.80E-06 -
< 1.91 E-06 -
l-129 < 1.12E-06 -
< 1.97E-06 -
< 1.77E-06 -
Gross Alpha < 1.91 E-06 -
< ' 2.63E-06 -
< 2.82E-06 -
Sr-90 NA NA NA Slice Containing Rebar 4 Page 8 of 8 ,
~
; Total Grams 588.91! 1
, Hi-Flux Concrete Core-SX862950119 l Activity j ISOTOPES Sum (uCi's) % Of Total Average pCl/g l Ag-100m 1.82E-041 0.0266% 1 0.31 l Am-241 '
3.18E-041 0.0464% i 0.54I Am-243 0.00E+001 0.0000%1 0.00i Ba-137m 0.00E+ 001 0.0000%i 0.00
, Ce-144 . 6.33E-041 0.0924%! 1.08 Cm-242 0.00E+ 00 ! 0.0000%i 0.00 -
Cm-243 0.00E+00 i 0.0000%1 0.00 Cm-244 0.00E+00 i 0.0000%I 0.00i Co-60 1.68E-02 I 2.4443%! 28.46 Cs-134 1.46E-041 0.0213%I 0.25 - Cs-137 5.18E-021 7.5581 % i 88.001 C-14 1.60E-02 l 2.3263%i 27.08 l Eu-152 2.42E-02 i 3.5327%i 41.13I Eu-154 1.37E-03 ! 0.1996% ! 2.32i Eu-155 3.35E-04 i 0.0489%l 0.57I Fe-55 0.00E+00 ! 0.0000%i 0.001 Fe-59 ' O.00E+001 0.0000%! 0.00i H-3 4.35E-01 : 63.5044%; 739.37i O I-129 , 2.79E-031 0.4071%i 4.74! Q Mn-54 Nb-94 0.00E+001 1.40E-04 ! 0.0000%I 0.0204%I 0.00 0.24: Ni-59 - 4.83E-03 i 0.7042%! 8.20i Ni-63 1.28E-01 i 18.6422%i 217.05-Np-237 0.00E+ 00 ! 0.0000%' O.00 Pr-144 0.00E+00 ' O.0000% 0.00' Pr-144m 0.00E+00 ; 0.0000%i 0.00 Pu-238 0.00E +00 i 0.0000%! 0.00-Pu-239 0.00E+00 i 0.0000%1 0.00-Pu-240 ' O.005+ 001 0.0000%i 0.00; Pu-241 0.00E+00 l 0.0000%1 0.00 t Pu-242 0.00E+00 0.0000%; 0.00: Rh .106 0.00E+ 00 I 0.0000%1 0.00! Ru-106 1.06E-031 0.1540%i 1.79i Sb-125 , 3.47E-04 I O.0506%1 0.59! Sr-90 0.00E+ 001 0.0000% 0.00: Tc-99 - 1.51 E-03 l 0.2206%I 2.57! Te-125m 0.00E+001 0.0000% 0.00i U-234 0.00E + 00 : 0.0000% 0.00; U-235 0.00E+00 ! 0.0000%. 0.00; U-238 0.00E+ 00 i . 0.0000% 0.00 Y-90 0.00E+00 t 0.0000%' O.00 p Zn-65 0.00E+00 t 0.0000% 0.00:
/ TOTALS = > 6.86E-01 100.00%. 1164.28i j
J f% ; Total Grams 8
, SX.811950025,SX861950066' !
C ' 1372.13i . SX861950056,0057
! No-Flux Concrete Core-SX841950016,0017,0018 Activity ISOTOPES Sum (uCi's) % Of Total Average pCi/g Ag-108m 4.37E-04 l 0.0077 % ; 0.32 l
Am-241 8.26E-04 0.0145 % I 0.60 I Ain-243 0.00E + 00 0.0000 % I 0.001! Ba-137m 0.00E + 001 0.0000 % 0.00 Ce- 144 4.21 E-03 i 0.0738 % 3.07 {i i Cm-242 0.00E +00 i 0.0000 % I 0.00 t Cm-243 l 0.00E + 001 0.0000 % I 0.001 Cm -244 i 0.00E +00 l 0.0000 % i 0.00 l Co-60 i 1.47E-021 0.2585 % i 10.75 I . Cs-134 7.96E-04 i 0.0140 % I 0.58 Cs-137 ' 5.55E + 00 I 97.3334 % i 4047.07 C~14 i 3.60E - 02 ! 0.6302 % i 26.20 Eu-152 ' 2.24E-031 0.0392 % i 1.63 Eu- 154 ' 2.80E-041 0.0049% l 0.20 ; Eu-155 2.08E-031 0.0365 % i 1.52 j Fe-55 , 0.00E + 00 i 0.0000 % 0.00 i l Fe-59 0.00E +00 t 0.0000 % i 0.00i H-3 ' 3.11 E -02 i 0.5449 % i 22.66 l l-129 ! 2.76E-03 0.0484 % ! 2.01 Mn -54 0.00E +00 ! 0.0000 % 1 0.00!
; Nb-94 ' 2.05E-041 0.0036 % i
( Ni-59 8.04E-03 i 0.1409% i 5.86 , Ni-63 3.31 E -021 0.5798 % ! 24.11 Np-237 0.00E + 00 ! 0.0000 % i 0.00t Pr- 144 0.00E + 00 ! 0.0000 % 1 0.00! Pr- 144m 0.00E + 001 0.0000 % i 0.00!: Pu -238 0.00E + 00 ! 0.0000 % i 0.00ii Pu-239 0.00E + 001 0.0000 % i 0.00 Pu -240 0.00E + 00 ! 0.0000 % 1 0.00: Pu -241 0.00E + 00 ! 0.0000 % i 0.00t! Pu-242 0.00E + 001 0.0000 % 1 0.00I j Rh-106 0.00E + 00 t 0.0000 % I 0.00i: Ru-106 6.69E-031 0.1173% l 4.88 l Sb-125 3.54E-031 0.0620 % I 2.58 i Sr-90 2.70E-03 i 0.0474 % ! 1.97 j Tc-99 2.46E-031 0.0431 % ! 1.79 : Te-125m t 0.00E +001 0.0000 % : 0.00 : j U-234 0.00E +001 0.0000 % i 0.00Ij U-235 ' O.00E + 00 i 0.0000 % i 0.00iI U-238 0.00E +00 i 0.0000 % 0.00 l l Y-90 0.00E + 00 i 0.0000 % ! 0.00!' __ Zn_ _65_ _ 0.00E+00 ! 0.0000% ; 0.00t' TOTALS = > 5.71 E + 00 i 100.00 % : 4157.80:' x i l i
e {~N . + ' Q/ \ /- Page 1 CORE BORE SUCES TO BE SENT FOR ANALYSIS 9/21/95 FISSION PRODUCT CONTAMINATED CORE BORES CORE BORE ' ore C Number SAMPLE -Length of Slices SAMPLES H EW" NUMBER (in) to be Sent Comments AREA TO BE SENT SAMPLE No. Slice No. 1l SX801950091 l 3l 21 TOP TWO SUCES l 1l 1 l SX801950127 l 1&2 l 2 SX811950028 6 2 TOP TWO SUCES 1 3 SX811950027 6 2 TOP TWO SUCES 1 1 l SX811950128 l 1 & 2 es cwe l 4 SX821950032 6 2 TOP TWO SUCES 2 5 SX821950033 6 2 TOP TWO SUCES 2 1 l SX821950129 l 1 & 2 ea cae l 6l SX832950021 l 46l Si TOP 2 EACH END & DEEPEST FROM SW l3 & 6l 3 SX832950130 1&2 SX832950131 19& 20 SX832950132 15 7 SX841950018 6 2 TOP TWO SUCES 4 8 SX841950017 6 2 TOP TWO SUCES 4 9 SX841950018 6 2 TOP 1WO SUCES 4 1 l SX841950133 l 1 & 2 em cue-l 10l SX852950103 l 9l 4l TOP 2 EACH END l 5l 1 l SX852950134 l 1, 2, S & 9 l 11 SX861950056 6 3 TOP TWO SUCES + DEEP SUCE 6 12 SX861950057 6 3 TOP TWO SUCES + DEEP SUCE 6 2 SX861950135 1 & 2 ea cae SX861950136 7&9 * [Totall to l O
l' age 2 ACTIVATION PRODUCT CONTAMINATED CORE BORES CORE BORE Core Number SAMPLE ~ Length of Slices SAMPLES 4fEW" NUMBER to be Sent (in) Comments AREA TO BE SENT SAMPLE No. Slice No. 1] SX811950025 l 16l 4l TOP THREE AND DEEPEST FROM RRF l 1l 4 SX811950137 1 , SX811950138 2 SX811950139 4' SX811950140 9 2l SX822950031 l 60l 6l TOP 2 EACH END & 2 DEEPEST FROM RC l2 & 6l 4 SX822950141 l 1&2 SX822950142 19 & 20 SX822950143 11 SX822950144 13 , 3l SX822950037 l 60l 61 TOP 2 EACH END & 2 DEEPEST FROM RC l 2 & 6l 4 SX822950145- 1&2 SX622950146 19 & 20 SX822950147 12 _ SXts22950148 16 4l SX841950125 l 60l 4 l CLOSEST TO SW & NEXT 3 TO DEPTH OF ~ 2* l 4 & 6 l 4 SX841950152 5 SX841950151 9 SX841950150 14 SX841950149 15 5l SX861950066 l 6l 3l TOP TWO AND DEEPEST l 6l 2 SX861950t53 1&2 SX861950154 6 6l SX862950062 l 6l 3l TOP TWO AND DEEPEST l 6l 2 SX862950155 1&2 SX862950156 7 7l SX862950063 l 6l 3] TOP TWO AND DEEPEST l 6l 2 SX862950157 1&2 ; SX862950158 5 8l SX862950065 l 61 3l TOP TWO AND DEEPEST l 6l 2 SX862950159 1&2 SX862950160 6 .
- 9) SX862950119 l 9l 51 TOP 2 AND 3 EVENLY SPACED TO DEEPEST l 61 4 SX862950161 1&2 SX862950162 6 SX862950163 9 SX862950164 13 l Totall 28 l I
____.__.__-________m_ _ _ _ _ _ . _ _ _ _ _ _ _ _ _ . _ _ _ _ _ _ _ _ _
O B&W NUCLEAR ENVIRONMENTAL SERVICES, INCORPORATED NUCLEAR ENVIRONMENTAL LABORATORIES DATA REPORT NUMBER 9511048 RADIOCHEMICAL ANALYSIS RESULTS Submitted to GPU Nuclear Corporation Saxton Site - Decommissioning Project December 22,1995 9 Released by: 6 WJ b . b k__ 11/z2./47 James L. Clark, Project Manager Nuclear Environmental Laboratories O
GPU Saxton Data Report December 22,1995 DATA REPORT
SUMMARY
) Laboratory Project Su'mmary- l e Plant Site: Saxton Decomminioning Project 4 e Shipment Number: 95-004 ! e Receipt Date: November 13,1995 l e NEL COC Number: SN-12973 i e NEL SDG Number: 9511048, 9510024 i l e NEL Proposal: NEL5-1258 e Contract Number: 0397355
\
e Analysis Service: Routine
.e Laboratory QC: Routine - Batch t
- I l
1 i
O GPU Saxton Data Report December 22,1995 t l Sample Delivery Group Summary 1. .l Customer ID NEL Sample ID Sample Type Description SX856950167-SD 9511048-01 Wet Sediment Area 5 Tunnel SX856950167-SD 9511048-01L Wet Sediment Liquid Phase SX865950167-SD 9511048-01S Wet Sediment - Solid Phase SX865950165-SM 9511048-02 Swab Sample RV Swab SX815950168-SD 9511048-03 Dry Sediment Tan'k Sludge SX862950162 9510024-36R Core Bore /1 Rebar Metal O Sample Physical Characteristid Summary; l NEL Sample ID Sample Matrix Mass / Volume Density 9511048-01 Wet Sediment 293 g 1.26 g/cc 9511048-01L Sedi / Liquid 73 mL 1.27 g/cc 951104'8-01S Sedi / Solids 201 g 0.85 g/cc 9511048-02 Swab / Metal 8.28 g (Note 1) Note 1 9511048-03 Dry Solids 191 g 1.36 g/cc 9510024-36R Rebar Metal 57.8 g -- Note 1: Sample could not be removed from swab matrix as requested. Sample Mass is total of sample and swab matrix. Density determination was not possible. O li 1 _ _ _ _ . _ . ~ _
_- = . . . . - -- .. - GPU Saxton Data Report December 22,1995 Sample: Anahrsis Summaryi ,
)
l Customer ID NEL Sample ID Sample Matrix Notes SX856950167-SD 9511048-01L Sedi / Liquid -- SX856950167-SD 9511048-01S Sedi / Solids -- SX865950167-SM 9511048-02 Swab / Metal (1) SX815950168-SD 9511048-03 Dry Solids --- SX862950162 9510024-36R Rebar Metal (2) Note 1: Sample could not be removed from swab matrix as requested. Entire matrix placed into solution for analysis. Activities in units of pCi/g where mass is that of entire sample matrix. Note 2: Rebar metal is that removed from NEL Sample 9510024-36, a Saxton Concrete Bore analyzed previously and reported in NEL Report Number 9510024. The entire piece of metal was dissolved and analyzed.
- Activity Results See Attached Report Sheet l
l l r ill
. . - - - . .= . - - - . - - . - . . - . . _ . . _ . - - - . - - - _ , . ~ - .. ~ .=-
GPU Saxton Results Summary B&W NESI, Nuclear Environmental Laboratories Revised 12/21/95 NELID 9511048-01L 9511048-01S 9511048-02 l 9511048-03 l 9510024-36R l NUCLIDE uCl/mL + /- 2S uCile + I- 2S ucile + I- 2S I uCi/a + I- 2S I uCl/a +l- 2S I Co-60 1.16E-07 1.09E-07 < 6.39E-06 -- 9.59E-02 8.92E-03 1.97E-04 6.40E-06 2.30E-04 2.22E-05 Ru-106 < 1.06E < 4.59E-05 -- < 6.05E < 2.87E < 8.74E Cs-134 < 1.23E 07 - < 5.36E-06 -- < 8.84E < 2.58E < 1.40E Cc-137 3.44E-07 1.26E-07 2.94E-04 9.44E-06 2.37E-02 3.18E-03 1.92E-03 1.14E-04 < 1.11E Co-144 < 5.11 E < 2.48E < 1.96E-04 -- < 2.01 E < 4.85E Am-241 < 1.11 E < 1.94E 1.71 E-02 1.82E-03 < 1.17E < 9.92E !' Ag-108m < 3.09E < 4.67E-06 -- < 5.11 E 05 - < 2.87E-06 -- < 9.77E Eu-152 < 8.89E < 4.03E < 7.70E < 2.15E < 1.23E ! Eu-154 < 3.42E < 1.35E 7.99E-04 1.58E-04 < 5.71 E < 2.48E Eu-155 < 1.96E < 1.21 E 2.19E-04 6.84E-05 < 8.53E < 1.72E Sb-125 < 2.64E < 1.31 E < 1.37E-04 -- < 1.04E < 2.39E ' H-3 < 4.44E-06 -- < 5.45E 4.87E-04 7.60E-05 <' 9.64E < 7.58E C-14 < 4.02E < 5.66E < 5.06E < 9.83E < 5.79E NI-59 < 3.59E < 5.99E 3.11 E-03 4.56E-04 < 5.85E 2.33E-05 2.06E-05 2.56E-01 3.74E-02 < 2.77E 2.12E-03 3.16E-04 i Ni-63 < 7.74E-07 -- < 1.29E Nb.94 < 1.07E < 4.38E < 8.26E 05 - < 2.37E < 1.31E l Tc-99 < 8.35E < 1.13E45 - < 1.90E < 3.14E < 9.73E ! l I-129 < 5.47E < 4.97E < 4.30E < 9.22E < 2.00E ! Gross Alpha 2.75E-07 6.76E-08 < 4.19E ! St-90 < 6.94E < 1.14E-OS - 2.35E-02 3.48E-04 2.55E-05 6.32E Fe-55 < 4.59E < 9.20E 7.54E-04 1.22E-04 l U-234 < 6.80E 4.60E-07 1.40E-07 1.22E-06 3.00E-07 U-235 < 6.80E < 2.OOE < 1.40E i U-238 < 2.10E 5.10E-07 1.60E-07 < 6.20E Np-237 < 2.40E < 9.68E-07 -- < 2.88E Pu-238 < 8.23E < 1.14E 3.15E-03 1.60E-04 [
- Pu-239/240 < 1.16E < 1.61 E 7.23E-03 3.68E-04 i Pu-241 < 6.97E < 2.24E 1.28E-01 8.08E-03 i Pu-242 < 1.72E < 4.64E-08 -- < 5.40E f Am-241 < 3.14E < 3.82E-06 -- 2.24E-02 1.11 E-03
- Am-243 < 6.48E < 2.52E-07 -- < 7.62E Cm-242 < 3.41 E < 1.07E 1.34E-04 7.26E-06 Cm-243/244 < 3.78E-08 -- < 7.13E 3.87E-04 1.9'/E-05 i
L Page 1 of 1 i
l B & W RESULTS 12/18/95 - Saxton Samples I
,r% \ SYSTEM = = = > Tunne/ Sediment - AREA 5 I Sexton i SA W I 0.00031 I Sample Number I Semote Aumber '
Total TRU Content (nCi/ml) ISX856950167-SDI 9511048-01L ' I 0.0000I i Sample Type i Sediment ll l Description ! Liquid Phase. I Results Normalized Nuclide Notes uCi/mi to 100 % Nuclides RATIOS 1 H-3 LESS THAN 4.44E-06 18.3144% '//######//###//#//###A%%%%%\W%%\%%M 2 C-14 LESS THAN 4.02E-06 16.5819% '//##############////a%\\%%\hW\\WWW\\) 3 Fe-55 LESS THAN 4.59E-07 1.8933 % Fe-55/Co-60 3.95690 4 Ni-59 LESS THAN 3.59E-07 1.4808% '////####//#/#/###//###n\\%%%\\\\\\\\\uB\% 5 Ni-63 LESS THAN 7.74E-07 3.1926% F//######//####/##/#//a\\%\W\%%%%%%\\% 6 St-89 Not Reported 0.00E+00 0.0000% '//#########//##/##/#A%%%%\%%%%%%%; 7 Sr-90 LESS THAN 6 94E-07 2.8627% St-90/Cs-137 2.01744 8 _ Nb-94 LESS THAN 1.07E-07 0.4414% '/####//#/####/#/##//#a%\hu\%%\WWu\wt 9 Tc-99 LESS THAN 8.35E-06 34.4426% ?//#//////##//####////#//#a\\%%%\%\\%%%%\B l 10 l- f 29 LESS THAN 5.47E-07 2.2563% j'//#/##/###/#//#///####a%%%%%\nW%%\%\; I 11 No-237 LESS THAN 2.40E -06 0.0990 %i'/#/#/#///#//#////#/#//#/#A%BW\\%%\%%%%%) 12 Pu-238 LESS THAN 8.23E-09 0.0339%l Pu-238/Am-241 0.07414 13 Pu-239/240 LESS THAN 1.16E-08 0.0478% F//#!/#/#//////#/##//##//#A\\\\%%%%%%%%%M 14 Pu-24 f LESS THAN 6.97E-06 0.2875%I Pu-241/Am-241 0.62793 15 Pu-242 LESS THAN 1.72E-09 C.0071% '#///#/#######////#/#/#A%%%%%%%%%%%M 16 Am-24 f LESS THAN 1.11 E-07 0.4579% Am-241/Pu-238 13.48724 gm 17 Am-243 LESS THAN 6.48E-09 0.0267% '#########/#/#/#/////#A\W\\h\\\\n\\u\%%M [ j 18 Cm-242 LESS THAN 3.41 E-06 0.1407% '//####//////######////#A%%%%%%%\%%\%M
\d 19 Cm-243/244 LESS THAN 3.78E-08 0.1559% '//##//########/####/A%%%%%%%%%%%%
20 f>-234 LESS THAN . 6.80E-09 0.0280% '//##//##//##//#####//#AB%%%%%%%%%%M 21 < >-235 LESS THAN 6.80E-09 0.0280 % '//####//##/#/####/#///A%%%%%%%%%%%M 22 f>-238 LESS THAN 2.10E-08 0.0866 % U-238/Am-241 0.18919 23 Co-60 1.16E-07 0.4785% Am-241/U-238 5.28571 24 Ru- 106 ! LESS THAN 1.06E-06 4.3724% '###//##//##//////#/##///A%%%%%%%%%%%% 25 An- 108m LESS THAN 3.09E-07 1.2746 % '////////#!//#/////#//##////#Ab%wh%%%%%%%% 26 Sb- f 25 LESS THAN 2.64E-07 1.0890% '//////////////##//////#//#//#a\\\\%%\\hB\\%%%M 27 Cs- f 34 LESS THAN 1.23E-07 0.5074 % Cs-134/Cs-137 0.35756 i 28 Cs-f37 3.44E-07 ' 1.4190% Co-60/Cs-137 0.33721
- 29. Ce- f 44 LESS THAN 5.11 E-07 2.1078% Cs-137/Co-60 2.96552 30! Eu- f 52 LESS THAN 8.89E-07 3.6670% '######//##///##////#///a%%%%%%%%%%%%
311 Eu-f 54 LESS THAN 3.42E-07 1.4107% Aw\W%%\%%\%%%%%\\%%%4%%%%\\%%%%%%% 321 Eu- f 55 LESS THAN 1.96E-07 0.8085 % '#/#/##//##//###//#/#/NAuk%%%%\n%M Totals = = > ll 2.42E-05L 100.00% l l 1 l l
\
l l I l r\
l . . p B & W RESULTS 12/18/96 - Saxton Samples
\ SYSTEM = = = > TunnelSediment - AREA 5 i Sexton i SAW I L_ o 0250ll I Semph Number i Semph Number I . Total TRU Content (nC1/g)
ISX856950167-SDI 9511048 -018 I I 0.00121 l Total Uranium content (nCilg) I Sample Type 11 Sediment l l Description # Solid Phase l Results Normalized Nuclide Notes uCi/n to 100 % Nuclides RATIOS 1 H-3 LESS THAN 5.45E-06 0.9776% '#/#//#/#####//#/#/###A%%\\%%%\W\\\\\%b , 2 C- 14 LESS THAN 5.66E-06 1.0152 % '###//#////#####////#//NA%%%\\WWMW\wu\b 3 Fe-55 LESS THAN 9.20E-06 1.6502 % Fe-55/Co-60 1.43975 4 Ni-59 LESS THAN 5.99E-06 1.0744 % '###/#//#/######//////#a%%%\\\\\\%%\\%%D 5 Ni-83 LESS THAN 1.29E-05 2.3139% '#####//#/#/##//#####A%%%\\%%%%%%%h 6 St-89 i Not Reported 0.00E +00 0.0000 % '#////####//##/#/#/#/##A%%%\\w%%%%%%b 7 - Sr-80 LESS THAN 1.14E-05 2.0448 % St-90/Cs-137 0.03878 8 Nb-94 LESS THAN 4.38E-06 0.7856 % f/#/##////######/##///#A%%%%%%%%%%%9 9 Tc-99 LESS THAN 1.13E-05 2.0269 % f//N##///////#/#///#/////#//d\\%%\\h\\\\%\\\\%B 10 l-f29 LESS THAN 4.97E-06 0.8915 % f/#/#///##//////#///#####A%w\%%%%%%%\%B 11 No-237 LESS THAN 9.68E-07 0.1736 % '####/#//##///////#/#///#A\%\\\\%%%\%%%\\9 12 Pu-238 LESS THAN 1.14E-07 0.0204 % Pu-238/Am-241 0.00588 l 13 Pu-23W240 LESS THAN 1.61 E-07 0.0289 % '###//###//######//////A%\\%\WWW\%%%%B l 14 Pu-24 f LESS THAN 2.24E-06 0.4018 % Pu-241/Am-241 0.11546 15 Pu-242 LESS THAN 4.64E-Os 0.0083 % '####/#/##///#/#####//a%%%%%%%%%%%B 16 Am-24f LESS THAN 1.94E-05 3.4796 % Am-241/Pu-238 170.17544 17 Am-243 LESS THAN 2.52E-07 0.0452 % f//#/#///#//////###//////////A\\\\\\\\\\\\\\\%\%%B l (Q 18 19 Cm-242 Cm-243/244 LESS THAN LESS THAN 1.07E-06 7.13E-07 0.1919% f#//##/#/######///#/#//A%%\\%%%%%%%%b 0.1279 % /###///#//#/##//##/#///#d%\\\\%%\\%%\%\hh 20 U-234 ' LESS THAN 4.60E-07 0.0625 % f####//##/#//#///####//A%%%%%\\%%%%%% 21 U-235 LESS THAN 2.00E-07 0.0359 % '##//#/#//#/#/#//#///#////d%%\n\%%%%%\%\% 22 U-238 LESS THAN 5.10E-07 0.0915 % U-238/Am-241 0.02629 23 Co-do 6.39E-06 1.1462 % Am-241/U-238 38.03922 24 Ru-106 LESS THAN 4.59E-05 8.2331 % f#//#/#/######//###//#d%%%%%%%%%%%V 25 Aa-106m _ LESS THAN 4.67E-06 0.8377 % I#/####/###//######//Aw%%%%%%%%%%V 26 Sb- 125 LESS THAN 1.31 E-05 2.3496 % '#//###//#//#/###///#////a%%%%%%%%%%%V 27 Cs-f34 LESS THAN 5.36E-06 0.9614 % Cs-134/Co-137 0.01823 28 Cs-137 2.9/E-04 52.7350 % Co-60/Cs-137 0.02173 29 Co- f 44 LESS THAN 2.48E-05 4.4484 % Cs-137/Co-60 46.00939 30 Eu-f52 LESS THAN 4.03E-05 7.2286 % '###/#///##//#//#//##//#A%%%%%%%%%%%b 31 Eu- 154 LESS THAN 1.35E-05 2.4215 % 6%%%%%%%%%\W%\W%%\%M%%%%%%%%%\\%t 32 Eu- f 55 LESS THAN 1.21 E-05 2.1704 % '##/###///////#//#////#/#/A%%%%%%%%%%%M i Totals = = > R 5.58E-04l 100.00_% l l I i i l
j i g B & W RESULTS 12/15/95 - Saxton Samples SYSTEM = = = > Bob's RV SWAB
%l f [_ Seurfort i BA W ll 161.304011 I Sample Number l Sample Aumber Total TRU Content (nCl/g)
I SX865950165-SW I 9511048 - 02 ll 0.00201 Total Uranium Content (nCl/g) I . Sample _ Type l GUN SWAB I I Description ll Rx Vessel i Results Normalized l Nuclide Notes Wl/a to 100 % Nuclides RATIOS 1 H-3 4.d$-04 0.0858 % '/###//#/###//#####//ku\%%%%%%%5%5 2 C- f 4 LESS THAN 5.06E-05 0.0089 % '##//##/#/#//######/Anu\%%\uu\\\%%W 1
- 3 Fe-55 1 7.54 E-04 0.1328 % Fe-55/Co-60 0.00786 1 4 Ni-59 3.11 E-03 0.5477 % '####/#/##########t\%%%%\\w\nunu\' l 5 Ni-63 2.56E-01 45.0849% '#####/###/##//#/##/tW\%%%\n\%\\\\\\\\'
6 St-69 Not Reported 0.00E+00 0.0000 % '#####/#/#/#/##//##/Annunu\%%%%%W 7 St-90 2.35E-02 4.1387 % Sr-90/Cs-137 0.99156 8 Nb-94 LESS THAN 8.26E-05 0.0145 % '##/#/#///#///##/#/#/#Au\%%%%%\unu\W 9 Tc-99 1.90E-05 0.0033 % '//#//########/#/###tu\%%%%%%%%%W 10 /-129 LESS THAN 4.30E-05 0.0076 % '#/#/###//#/////#/#/##A\%%%%%%\\wn%W l 11 No-237 LESS THAN 2.88E-06 0.0005 % '#####//#/#/#/#/###//L%\%%%%%%%%%W - 1 12 Fu-236 3.15E-03 0.5548 % Pu-238/Am-241 0.14063 l 13 Fu-239/240 7.23E-03 1.2733 % '##//#/#///#/#/#//#//#/Anu\W\\hW\n\%%W 14 Fu-24f 1.28 E-01 22.5424 % Pu-241/Am-241 5.71429 15 Fu-242 LESS THAN 5.40E-08 0.0000 % '####//########//#/h\n%%%%%%%%%W 16 Am-24 f 2.24E-02 3.9449 % Am-241/Pu-238 7.11111 e 17 Am-243 LESS THAN 7.62E-08 0.0000 % '#########///#/###/hu\%%%%%%%%%W
\\ 18 Cm-242 1.34E-04 0.0236 % '#/##/#//#//####/#/#/h%\%%%%%%%%%W 19 Cm-243/244 3.87E-04 0.0682 % '#/#///////#//###//#/#//h%\%%%%%%%%%W 20 U-234 1.22E-06 0.0002 % '####/#///####//#/#/#M\%%%%%%%%%W 21 U-235 LESS THAN 1.40E-07 0.0000 % '##////####//##/##/#/h%\%%%%%%%%%W 22 U-236 LESS THAN 6.20E-07 0.0001 % U-238/Am-241 0.00003 23 Co-60 9.59E-02 16.8892% Am-241/U-238 36129.03226 24 Ru-106 LESS THAN 6.05E-04 0.1065 % '###############/Au%%%\u%%%%%W 25 Aa- f 06m LESS THAN 5.11 E-05 0.0090 % '//##/#/##//####///#/#M%%%\%%%%%%W 26 Sb-125 LESS THAN 1.37E-04 0.0241 % '###/#///#//#####//##M\%%%%%%%%%W 27 Cs-f34 LESS THAN 8.84 E-05 0.0156 % Cs-134/Cs-137 0.00373 28' Cs-f37 2.37E-02 4.1739 % Co-60/Cs-137 4.04641 29 Ce-f44 i LESS THAN 1.96E-04 0.0345 % Cs-137/Co-60 0.24713 30 Ju-f52 LESS THAN 7.70E-04 0.1356 % '##////////##//#////#/##A\%%%%%%%%%%W 31 Ju- f 54 7.99E-04 0.1407 % '##///#///#//##//#/#////Ahu%%\%%%%%%W 32 Ju-155 2.19E-04 0.0386 % '#//##/#/#////##///#//#h%\W%%%%%%%%W ,
l Totals = = > R 5 68E-01 R 100.00% Il
,/
B & W RESULTS 12/15/96 - Sexton Samples 818 TEM ===> CondensatePump& TankSample
- ll 0.01171 ,
r Total THU content (nCi/g) 1 1 - 11 - ll 0.00001 Total Uranium content (nci/g) ement i Description 11 Solids !! Results Normalized Nuclide Notes uCi/a to 100 % ' Nuclides RATIOS 1 H-3 ESS THAN 9.64E-06 0.4105 % '#i/##############t\%%%%%%%%\%%\1 2 C- f 4 ESS THAN 9.83E-06 0.4186 % '##############//#i\%%%%%%%%%%%1 3 Fe ' 15 Not Reported 0.0000 % Fe-55/Co-80 0.00000 4 N/-; y WSS THAN 5.85E-06 0.2491 % '##//####//##//###/#h\\\\\%%%\\\B\h\%\' 5 Ni-< 3 WSS THAN 2.77E-05 1.1795 % '####/#/####//#####A%hh4W%%%%%W 6 Sr-i 19 Not Reported 0.0000 % '#//#####/#/#/#//#//#h\%%%%\\%\%\%%W 7 Sr-1 10 2.55E-05 1.0858 % Sr-90/Cs-137 0.01328 8 Nb .W WSS THAN 2.37E-07 0.0101 % '///#/########///##///A%%%%\\%%%%%W 9 Tc- 9 ESS THAN 3.14E-05 1.3370% '####//##//##//##////#Awww\\WuwWuhW 10 /- f; 9 ESS THAN 9.22E o 0.3026 % '#####//#######/##Au\%%%%%%%%%W 11 No-; 37 Not Reported 0.0000 % '#//###///##/##///#/#/huu\%\B\\W\%%%\\' 12 Pu-;36 Not Reported 0.0000% Pu-238/Am-241 0.00000 13 Pu-23 V240 Not Reported 0.0000 % r####//#####//####/hW\W%%%%%\uuu\' 14 Pu-; 4f Not Reported . 0.0000% Pu-241/Am-241 0.00000 15 Pu-242 Not Reported 0.0000 % '#//########//####/hWBuu\%%%%%%W 16 Am , Wf ESS THAN 1.17E-05 0.4982% Am-241/Pu-238 ERR
/ 17 Am-; WS Not Reported 0.0000% '#//#////#####//#####A%%%%\\W%%%%W 18 Cm-242 Not Reported 0.0000% '##//#//#######//##/h%%%\W%%%%%%W 19 Cm-244/244 Not Reported 0.0000% '##/##/##//#//#//###/h\%%%%%%%%%%W 20 11- 234 Not Reported 0.0000% '//######//##////##///h\WWWWh%%%%%W 21 4/~235 Not Reported 0.0000% '########//####/#//h\%%%%%%%%%%W 22 i/-238 Not Reported 0.0000% U-238/Am-241 0.00000 23 Go-60 1.97E-04 8.3885 % Am-241/U-2N ERR 24 Hu-f06 WSS THAN 2.87E-05 1.2221 % '#//#/#/######//##///h\%%%%%%%%%%W 25 An- f00m GSS THAN 2.87E-06 0.1222 % '######//########/h\WWWW\%\hW%%W 26 Sb- f 25 WSS THAN 1.04E-05 0.4428% '###/#/#######//##/hBW\WWWWWW%%W 27 Co- fad i GSS THAN 2.58E-00 0.1090 % Cs-134/Co-137 0.00134 28 Cs-f37 1.92E-03 81.7555 % Co-60/Cs-137 0.10260 29 Co-f44 MSS THAN 2.01 E-05 0.8550% Cs-137/Co-60 9.74619 30 < iu- f52 ESS THAN 2.15E-05 0.9155 % '#########//###//#/h\%%%%%%%%%%W 31 liu- f 54 GSS THAN 5.71 E-06 0.2431 % '##########!/###///huu\WWh%%%%%We 32 diu-f55 GSS THAN 8.53E-06 0.3632 %_ '###########///#/#/hW\uhWWWWWWWWl l Totals = = > 112.35E-03!! 100.00% l Gross AIDhn An6Vtv i 2. rue:-071 uC1/g f
i . -
F l B & W RESULTS 12/15/96 - Saxton Samples
}n
\
. SYSTEM = = = > Concrete Re-Bar From Core Sample SX862950f f 9 -
Santon i SAW I 0.00101 _Sempie Number f Sample Number Total TRU content (nC1/g) SX862950162 i 9510024-S6R ll 0.000_0 ll Total Uranium Content (nCi/g) l Sample Type 11 Core Bore ll l Description jl Steel Re-Bar ll Results Normalized Nuclide Notes uCl/a to 100 % Nuclides RATIOS i l H-3 LESS THAN 7.58E-05 2.9437 % ,
'######/#////#/##/#/#6\%\u\\u\\\ mum \\W-2 C-14 WSS THAN 5.79E-05 2.2485 % '/#/####/////##/#/#/##i\\\\\\\\uu\\u\%i\W 3 Fe-55 Not Reported 0.00E+00 0.0000 % Fe-55/Co-60 0.00000 4 Ni-59 2.335 - 05 0.9049 % f####/##///##//####/A\%%\uu\\uuuu\W ,
5 Ni-83 2.12E-03 82.3301 % '/##/#///#///##/###/#/A\\unu\\\\\\u\\n\Wl 6 Sr-89 Not Reported 0.00E+00 0.0000 % '####/#///###/###/##6\\\%%\\%%%%\\%W 7 8r-90 Not Reported 0.00E+00 0.0000 % Sr-90/Cs-137 0.00000 8 Nb-94 ESS THAN 1.31 E-06 0.0509 % '#/#//#/##/##/###//#/A\%%%%\\%%\\MhW 9 _ Tc-99 WSS THAN 9.73E-06 0.3779% f/#/####////#////##//#/A\\n\\nu\\n\%%\\W 10 /-129 WSS THAN 2.00E-05 0.7707 % '####//##//#####//#/A\w\n\\uu\\\\\\\\u\' 11 Np-237 ' Not Reported 0.00E+00 0.0000 % '##//####/#/#//#####iuuuuuuuuuuuy 12 Pu-238 Not Reported 0.00E+00 0.0000 % Pu-238/Am-241 0.00000 13 Pu-239/240 Not Reported 0.00E+00 0.0000 % '#//###########/#/#Muuuuuu\W%%W 14 Pu-24f Not Reported 0.00E+00 0.0000 % Pu-241/Am-241 0.00000 15 Pu-242 Not Reported 0.00E+00 0.0000 % '####////#/g!#////####M\%%%%%%%%%W 16 Am-24f i WSS THAN 9.92E-07 0.0385 % Am-241/Pu-238 ERR 17 Am-243 Not Reported 0.00E+30 [mi 18 Cm-242 Not Reported 0.00E+00 0.0000 % '//##############/A\%uuuu%%%%%W 0.0000 % f///#/#/#//#/##//####/A\%%%%%%%%%%W C/ 19 20 Cm-243/244 U-234 Not Reported Not Reported 0.00E+00 0.00E+00 0.0000 % '#//##//##########/A\w%%%%%%%%%W 0.0000 % '##//#######/#/####Muuuuu\%%%\nv 21 U-235 Not Reported 0.00E+00 0.0000 % '###############//Muu\%%%%%%%W 22 U-238 I Not Reported 0.00E+00 0.0000 % U-23e/Am-241 0.00000 23 Co-80 l 2.30E-04 8.9320 % Am-241/U-238 ERR 24 Ru-106 I WSS THAN 8.74E-06 0.3394%l'####//##//#######/A\%%%%%%%%%%W 25 Aa- f 00m i ESS THAN 9.77E-07 0.0379% '/##/#/#///#####//#///A\h%%%%%%%%\\W 26 Sb-125 ! ESS THAN 2.39E-06 0.0928 % '####//##/#/###//##/A\w%%%%%%%%%W 27 Cs- f 34 ! WSS THAN 1.40E-06 0.0544 % Cs-134/Cs-137 1.26126 28 Cs-f37 WSS THAN 1.11 E-06 0.0431 % Co-60/Cs-137 207.20721 29 Co-f44 WSSTHAN 4.85E-06 0.1883 % Cs-137/Co-60 0.00483 30 Eu-f52 ESS THAN 1.23E-05 0.4777 % '//#######//######//Muu\%%%%%%%W , 31 Eu-f54 ESS THAN 2.48E-06 0.0963 % '###############/Au%%\%%%%%%%%\ 32 Eu-155 i WSS THAN 1.72E-06 0 0668 % '/#/#/##/#########/Anhu\u%%%%%%\n l Totals = = > ll 237E-031 100.00 % l l~ Gross4.1A/Dhs - luCVg ()
s TELE 0TNE SROWN ENGINEERINC ENVIRONMENTAL SERVICE 5 REPORT OF ANALi$l$ REVISED 09/26/95 RUN DATE 02/89/95 WORE ORDER NUNSER CUSTOMER P.O. kUNSER DATE RECElvf8 DELIVERT DATE PAGE 1 J PATRICE 00NNACHIE' JR ERL MANAGER 3-5243 GPU NUCLE AR CORPOR ATION 01/10/95 02/01/95 ' ,e PO 80X 404 . NIDOLETOWN PA 17951-9400 SWIPE 5 TELEDYNE COLLECTION-DATE SAMPLE CUS T ONE R' S STA START STOP MIO-COUNT NUNSER IDENT IFIC ATION ACTIVIff NUCL-UNIT-3 TIME NUN DATE TIME DATE TIME NUCLIOE WOLUME - UNITS I TCTAL_UCII U/M
- OATE TINE A$N-MGNI-5 e LAS.
73504 GR SATN-5NS-Al e ettet GR-A 4.4 +-0.6 E-04 GR-8 02/02 3 2 9 +-0 1 E-01 02/01 C-14 L.T. 3. E-05 3 NI-59 82/07 3 L.T. 3. E-04 02/02 SR-09 L.T. 1. E-84 3 .
$R-90 82/03 3 4.0 +-0.6 E-04 02/03 FE-55 3 L.T. J.- E-04 02/01 YC-99 3 Lei. 4. E-05 01/29 - L I-129 L.T. 5. E-05 02/95 SE-T 3 L.T. 7. E-03 01/19 E-40 4 L.T. 3. E-04 01/19 NN-54 4 L.T. I. E-44 01/19 4 CO '8 L.T. 1 E-44 01/19 FE-59 4 L.T. 3. E-04 41/19 4 CO-60 2.07+-0 29E-83 01/19 4 IN-65 L.T. 3. E-84 01/19 4 2R-95 'L.T. 3. E-04 01/19 4 NS-95 L.T. 2. E-04 01/19 4 20-103 L.T. T. E-04 01/19 4 RU-lee L.T. 3. E-03 81/19 4 I-831 L.T. 1. E-03 01/19 4 C$-134 L.T. 2. E-04 01/19 4 C5-137 3.56*-0 36E-01 08/19 4 84-140 L.T. 2. E-03 01/19 4 LA-140 L.T. 1. E-04 08/!9 4 CE-141 L.T. 5. E-04 01/19 4 CE-144 L.T. 2. E-03 01/19 4 RA-226 L.T. 7. E-03 01/19 4 TH-220 L.T. 6. E-04 08/19 4 H-3 L.T. 5. E-04 01/33 5 NI-63
' 1.2 +-c.1 E-05 02/00 5 PU-230 4.4 +-0.7 E-05 02/03 6 U-234 L.T. I.1 E-06 02/01 4
- 1 h a s. sample had high Cs-137 activity so it was counted for 5 minutes in order to keep the Cs-137-counts in the peak channel from "over flowing". A consequence was that there were no counts in the K-40 reElon. In order to have a non zero L.T. result, the K-40 calculation was performed with I (one) as the background count. The result is as reported above. qf .
N N . r TELE 0VME DROWN ENGINEERING ENVIRONMENTAL SERVICES REPORT OF ANALYSIS WORK ORDER NUMSER RUN DATE 02/09/95 CUSTOMER P.O. huMBER CATE RECEIVED DELIVERY DATE PAGE J PATRICE 00NNACP'" JR ERL MANAGER 3-5243 2 GPU NUCLE AR CORPL i! DN PO 80X 480 01/10/95 02/01/95 e MIOOLETOWN PA 17057-9440 - SWIPES TELEDTHE SAMPLE COLLECTION-0 ATE CUS T OME R' S STA START NUMBER STOP MIO-COUNT IDENTIF IC AT ION ACTIVITY NUCL-UNIT-3 TIME NUM OAYE TIME DATE TIME NUCLICE I TOTAL UCIB U/M
- WOLUME - UNITS 73504 GR SXTN-SMR-Al 0 OATE TIME ASH-WGNT-2
- LA8.
01/0R U-235 L.T. 1 1 E-06 02/01 U-238 L.T. 1 1 E-06 6 AM-241 1.8 +-0 2 E-04 02/01 6 CM-242 L.T. 02/01 6 1 3 E-06 02/01
- CM-244 L.T. 2 2 E-06 6 PU-239 1 0 +-0 1 E-04 '
02/01 6
'PU-241 6 1 +-1 6 E-04 02/01 6 e PU-242 L.T. 9.9 E-07 02/08 6 e 02/08 6 73505 GR S K TN- SMR -A 2 0 01/0R GR-A 3.4 +-0.5 E-04 GR-8 1 0 +-0.1 02/02 3 1 E-01 02/01 C-14 L.T. 2. E-05 3 '
NI-59 02/07 3 L.T. 3. E-04 02/02 SR-89 L.T. 2. E-04 3 SR-90 02/03 3-1.0 +-0.1 E-03 02/03 FE-55 L.T. 4. E-04 3 TC-99 02/01 3 L.T. 3. E-05' 01/29 I-129 L.T. 4. E-05 3 BE-7 02/04 3 L.T. 2. E-03 01/19-K-40 L.T. 2. E-04 4 MN-54 01/19 4 L.T. 3. E-05 01/19 4 CO-58 L.T. 4. E-05 FE-59 01/19 4 L.T. 9. E-05 01/19 . CO-60 4 8.31*-0.85E-04 01/11
- 4 2N-65 L.T. 8. E-05 ZR-95 , 01/19 4 L.T. 7. E-05 01/19 NS-95 4 L.T. 3. E-05 01/19 4 RU-103 L.T. 2. E-04 01/19 4 RU-106 L.T. 1. E-03 I-131 L.T.
01/19 4,
- 3. E-04 01/19
*C5-134 4 L.T. 4.' E-05 01/19 C5-137 7.66+-0.77E-02 01/19 4
8A-140 4 L.T. 5. E-04 01/19 4 t
~ . . . . . . . . .. . . _ ,.
O . i i TELEOFNE SROWN ENGINEERING ENVIRONMENTAL SERVICES - REPORT OF ANALYSIS RUN DATE 02/09/95 WORK DRDER NUMSER CUSTOMER P.O. NUM8ER DATE RECEIVE 0 DELIVERT DATE PAGE 3 J PATRICK DONNACHIE JR ERL MANAGER 3-5243 ^ GPU NUCLEAR CORPOR ATION 01/10/95 02/01/95 PO SOK 480 * ,e MIDDLETOWN PA 17997-9480 SWIPES TELEDYNE SAMPLE COLLECTION-DATE CUS TOME R' S STA START STOP MID-COUNT NuneER IDENTIF IC AT ION ACTIVITY MUCL-UNIT-5 TIME NUM BATE TIME DATE TIME NUCLIDE f TOTAL UCII U/M
- VOLUME - UNITS
! DATE TIME ASH-WGNT-2
- LAG.
73505 CR SXTN-5MR-42 0 01/OR LA-140 L.T.- 2. E-05 01/19 CE-141 L.T. 1. E-04 4 CE-144 01/19 4 L.T. 5. E-04 01/19 RA-126 4 L.T. 2. E-03 01/19 4 TH-228 L.T. 2. E-04 01/19 H-3 4 L.T. 5. E-04 01/31 NI-63 5.4 +-0.5 5 E-04 02/04 PU-238 3.1 +-0.5 E-05 5 $ U-234 02/01 6 f L.T. 1 0 E-06 01/30 U-235 L.T. 6 1.0 E-06 01/30 U-238 L.T. & 1.0 E-06 01/30 AM-241 1.3 *-0.1 E-04 6 02/01 4 ' CM-242 2.6 +-1.T E-06 02/01 ' CM-244 4 L.T. 1.0 E-06 02/01 6' PU-239 s.3 *-0.9 E-05 02/01 ' PU-241 6 5.5 *-1.3 E-04 02/08
- PU-242 6 L.T. 1 2 E-06 02/08 4 73506 5NTN-SMR-A3&A40 01/0R GR-A 5.8 +-2.4 E-05 02/01 CR-B 8.3 +-0 2 E-03 3 :
C-14 02/01 3
- L . T .' 2. E-05 02/0T 3 NI-$9 L.T. 3. E-04 02/02 5R-89 3 L.T. 1. E-02 02/03 3
' SR-90 L.T. 3. E-05 02/03 FE-55' 3 L.T. 3. E-04 , 02/01 3 TC-99 L.T. 4. E-05 01/29 3 I-129 L.T. 7. . E-05 02/06 3 BE-T L.T. 2. E-04 01/19 4 K-40 L.T. 3. E-05 01/19 4, MN-54 L.T. 5. E-06 01/19 4 CO-58 L.T. 6. 'E-06 01/19 4 FE-59 L.T. 1. E-05 01/19 4 CO-60 2 59*-0.26E-04 01/19 4- f
.I
-.a. - - . _ _ _ . . . - - - - - - . - - _ -
9% N TELEDYNE 8ROWN ENGINEERING ENVIRONMENTAL SERVICES REPORT OP ANALYSIS RUN'0 ATE 02/09/95 WORK ORDER NUM8ER CUSTCMER P.O. kUMBER DATE RECEIVED DELIVERY OATE PACE 4 J PATRICK 00MMACNIE JR ERL MANAGER 3-5243 CPU NUCLE AR CORPOR ATION 01/10/95 . 02/01/95 e PO 80X 480 " MIDDLETOWN PA 17057-6680 SWIPES TELEDTNE COLLECTION-DATE 5 AMPLE CUS TOME R
- 5 STA START MIO-COUNT' STOP NUMBER IDENT IP IC AT ION ACTIVITV NUCL-UNIT-4 TIME NUM OATE TIME DATE TIME NUCLIOE I TOTAL UCII VOLUME - UNIT 5 U/M
- DATE TIME ASH-WGNT-2
- T3506 LA8.
SETN-SMR-A3&A40 Ol/OR ZN-65 L.T. 1. E-05 01/19 2R-95 L.T. 1. E-05 01/19 4 NB-95 4 L.T. 6. E-06 01/19 RU-103 4 L.T. 2. E-05 01/19
- RU-106 4 L.T. 9. E-05 01/19 4 I-131 L.T. 3. E-05 C5-134 01/19 4 L.T. 6. E-06 01/19 C5-137 4 6 26+-0.63E-03 01/19 4 +
'8A-140 L.T. 5. E-05 01/19 LA-140 4 L.T. 6. E-06' 01/19 CE-141 4 L.T. 1. E-05 01/19 4 CE-144 L. T. 4. E-05 01/19 4 RA-226 L.T. 1. E-04 01/19 TH-228 4 L.T. 1. E-05 01/19 4' H-3 L.T. 8. E-04 01/27 NI-63 5 8.9 +-5.0 E-05 02/08 5
PU-238 4.0 +-2 1 E-06 02/01 U-234 L.T. 1 1 E-06 6' 02/02 6 U-235 L.T. 1.1 E-06 02/02 U-238 6 L.T. 1 1 E-06 02/02 6 AM-241 1 2 +-0.3 E-05 02/03 6 8 CM-242 L.T. 1 3 E-06 02/03 CM-244 6 L.T. 9.5 E-07 02/03 6 PU-239 8.6 +-3.1 E-06 02/01
- 6 PU-241 L.T. 2.8 E-04 02/08 6
PU-242 L.T. 1.2 E-06 02/08 6 LAST PACE OP REPORT . APPROVED ST J. CUENTHER 02/09/95 SEND 1 COPIES TO MET 915 J PATRICK 00NNACHIE JR ERL MANAGER 2 - GA5 LAB. 3 - RADIO CHEMISTRY LA8. 4 - CEtLIl CAMMA SPEC LAB. 5 - TRITIUM 6ms/L.5. LAB. 6 - ALPNA SPEC LA8# t j [ _ _ _ . _ _ _ _ _ . _ _ _ _ _ _ _ _. -- - - - - - - - - - - - - - - - - - - - - --- ~ - ~~
TELE 0YME SR0bN ENGINEERING ENVIRONMENTAL SERVICES REPORT OF ANALYSIS ' R OATE O /2 /95 W OR K ORDER NUMSER CUSTOMER P.O. NUMSER DATE RECEIVE 0 DELIVERY OATE PAGE 1 J' PATRICK OONNACNIE'JR ERL MANAGER 3-1127 CPU NUCLE AR CORPOR ATION 06/27/95 07/26/95 , PO 804 400 MIDDLETOWN PA - 17057-0400 SWIPES TELEDYNE COLLECTION-DATE 5 AMPLE C US TONE R' S STA M10-COUNT START STCP ACTIVITY huMBER IDENTIFIC AT ION NUCL-UNIT-3 TIME VOLUME - UNIT 5 NUM OATE TIME DATE TIME NUCLIOE t TOTAL UCII U/M
- OATE TIME ASN-WGNT-2 *- LAs.
89209 5 X8 2195 0052-A R E A2 06/26 GR-A 2.4 +-0 2 E-03 GR-B 07/21 3 6.9 +-0.1 E-01 07/20 3 C-14 1 1 +-0.4 E-05 07/26 .3 NI-59 L.T. 1. E-03 07/19 3 SR-89 L.T. 1. E-03 ~ 10/04 3
'5R-90 1.2 +-0 1 E-02 07/26 3 FE-55 L.T. 5. E-04 07/20 3 TC-99 3.2 +-2.2 E-05 07/21 3 I-129 L.T. 2. E-05 07/26 3 SE-T L.T. 1. E-02 07/13 4 K-40 L.T. 9. E-04 07/13 4 CR-51 L.T. 8. E-03 07/13 4 MN-54 L.T. 3. E-04 07/13 4 CO-58 L.T. 3. E-04 07/13 4 FE-59 E-04 4.
L.T. 7. 07/13 CO-60 1.49*-C.15E-02 07/13 4 ZN-65 L.T. 7. E-04 07/13 4 NS-94 L.T. 3. E-04 07/13 4 2R-95 L.T. 5. E-04 07/13 4 NS-95 L.T. 3. E-04 07/13 4 M0-99 L.T. 2. E-02 07/13 4 RU-103 L.T. 1. E-03 07/13 4
, RJ-106 L.T. 5. E-03 07/13 4 AG-110M L.T. 4. E-04 07/13 4 59-124 L.T.- 1. E-04 07/13 4 58-125 L.T. 3. E-03 07/13 4 I-131 L.T. 3. E-03 47/13 4 C5-134 L.T. 3.. E-04 07/13 4 C5-137 5.63+-0.56E-01 07/13 -
4 BA-140 L.T. 4. E-03 07/13 4
- LA-140 L.T. 2. E-04 07/13 4 CE-141 L.T. 9. E 07/13 4 CE-144 L.T. 3. E-03 07/13 4 Reanalysis of these samples for Sr-89 did not confirm the original results, thus this revision. We surmise that in the original analysis there was incomplete separation of. the high Sr-90 activity from the Sr-89 planchet and consequently the remaining Sr-90 was interpreted as Sr-89. g l .
f.
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>* TELEDTME SROWN ENGINEERING ENVIRONNENTAL SERVICES m REPORT OF ANALY$t$ RUN DATE 07/28/95 r >* h 0R E ORDEk NtfMtER CUSTOMER P.O. huMBER nATE RECEIVED DEttVERT CAIE PAGE 3 fu b
W J P'ilR ICK 0000NACHIE JA GAL paNAGE9 3-!!27 06/27/95 0F/26/15 .* Gru NUCLEAR CORPORAllON PC SOR 480 MIDOL E TCW88 PA 174S7-0400 $ SWIPES COLLECl80N-SAIt Mgg-CouMT TELEDYNE SAMPLE CUST OME R' S STA START STOP ACTIVITV NUCL-bNti-2 TIME WOLUME - UNIIS IDENTIF IC AT ION Nun DATE 11ME DA1E IIME NUCLICE t YOTAL UCl3 U/M
- SATE TIME ASN-WENT-2
- LAS.
teuntER 49280 SR4619$00SI-ARE A 6 64/26 MS-94 L.T. S. E-02 07/13 4 2R-95 L.T. 9. E-02 87/13 4
,3 4 >* MS-95 L.T. 4. E-82 07/13 .
MO-99 L.T. 2. E 80 01/13 4 Ru-103 L.T. S. E-92 87/13 4 2u-806 L.f. 4. E-et 07/13 4 AG-it0M L.T. 1. E-02 07/13 4 54-824 L.T. 2. E-02 07/13 4 h
>* $4-125 L.T. L. E-01 07/13 4 L.T. 2. E-el 07/13 4 "' t-131 C5-134 L.T. S. E-02 07/13 4 CS-Il1 9.03*-0.90E et 87/13 4 07/13 4 OA-140 L.T. 2. E-et E-02 07/13' 4 LA-140 L.T. 2.
4 CE-141 L.T. S. E-82 41/13 4 CE-144 L.T. 3. 5-05 07/t3 RA-226 L.T. 7. E -01 ' ~ 91/13 4 4 g
- TM-228 L.T. 6. E-02 81/13' 4
NP-237 L.T. 8. E-02 _07/13 07/20 - 5 M-3 2.0 +-0.4 E-03 5 hl-as I.3 +-0 1 g-st e7/25
- 47/22 e 3.1 +-0.2 E-SS PU-238 6 u-234 L.1. 9. E-07 07/22 07/22 6 u-235 L.T. 9. E-07 ,
07/22 6 U-238 1.T. 9. E-07 4 AM-24 1 3 +-0.9 E-02 07/22 e7/22 6 CM-ut 1.p,*-9,1 E-f9 6 CM-244 1.0 e-0.f t-ut 07/J2 07/22 6 . Pu-239 8 2 e-Ge5 E-03 4 P0-243 2.0 +-0.1 E-02 0F/25 ** 07/26 6 Pu-242 2.4 e-0.6 E-01 g
- y L A51 PAGE OF REPORT -
APPROVES SY . GufNTHER 07/28/95 p 1END 1 COPIES TO ME7915 J PATRICK 00NNACHIE JR ESL MANAGER 6- ALPHA SPEC LAS. 4 - GEEL!l GAMMA SPEC LAB. 5 - TRIlluM G45/L.5. L A8. i tas t AB. 1 - RA0lO CHEMI51RY LAS. Al!G b ~. : "
O O . R S APPENDlX E O l 0
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SAXTON COMBINED PIPING SECTIONS-LENGTHS (ft) & ESTIMATED WEIGHTS Total PIPING SIZE (in) = = > 1/2* 3/4* 1* 1 114* 1 t/2" _2* - 21/2* 3* 4* 8* 10* 12 3/4' 14* No. of feet Systone Percent __ insulated Footmeo Insadme=d_ Ausniary Steam _ Piping.JA. S) 0.0 l '.9 32.0 0.0 0.0 0.0 0.0 22.1 0.0 00 0.0 0.0 0.0 56.0 56.0 100.0% Schedule -- 40 40 -- -- -- -- 40 -- ~~ -- -- -- Material / Type -- CS CS -- -- -- -- CS -- -- -- -- -- infernal Surhee Area ((in ^2)/fg -- 31.1 39.5 -- -- -- -- 115.7 -- -- -- -- -- System kntemal_[n ^ 2) = = >I See0J1 Slow Down-2nd Side of SIGJ8L) 0.0 0.0 0.0 0.0 10.3 22.0 00 0.0 0.0 00 00 00 0.0 N/A 32.3 0.0% Schedule -- -- -- -- 80 80 -- -- -- -- -- -- -- Material / Type -- -- -- -- CS CS -- -- -- -- -- -- -- sneernas Surnace Ares ((in ^ 2)@ -- -- -- -- 56.5 73.1 -- -- -- -- -- -- --
. System entamal (in ^ 21= = >l 2190 42 ghemical Addition j CAJ 0.0 28.0 00 0.0 0.0 00 0.0 0.0 0.0 0.0 0.0 0.0 0.0 N/4 26.0 00%
Schedune -- 80 -- -- -- -- -- -- -- -- -- -- -- Material / Type -- CS -- -- -- -- -- -- -- -- -- -- -- oneernal Surhee Ares ((in *2)/fg -- 28 0 -- -- -- -- -- -- -- -- -- -- -- Svetem entomas tn ~ 2)= = >g 7e32s Component _p_ootengjC_C] 00 0,0 00 0,0 38.3 0.0 00 226.6 0.0 0.0 0.0 0.0 0.0 N/A 264.9 0.0% Scheduce -- -- -- -- 40 -- -- 40 -- -- -- -- -- Materint/ Type -- ~~ -- -- CS -- -- CS -- -- -- -- -- Anterna# Surmece Area f(in ^2)q -- -- -- -- 60.7 -- -- $ 15.7 -- -- -- -~ -- System _ intamal (in ^ 2)= = >l 28533 39 Condene. ate DrainlipingED) 0.0 4.0 73.9 4.3 50.6 13.5 0.0 00 00 00 0.0 0.0 0.0 146.3 146.3 100.0 % Schedufe -- 40 40 40 40 40 -- -- -- -- -- -- -- ManerialfType -- CS CS CS CS CS -- -- -- -- -- -- -- 39.5 52.0 60.7 77.9 '-- System _intemalin ^ 2} = = >I 7393 61 infernal Surhee Area ((in *2)/M -- 31.1 -- -- -- -- -- -- Drain Linge-LiquidjDL) 77.5 75.8 145.8 88.2 116.9 51.0 6.8 134.0 0.0 0.0 0.0 0.0 0.0 N/A 695.6 00% Schedude 40 40 40 40 40 do 40 40 -- -- -- -- -- afstusal/ Type CS CS CS CS CS CS CS CS -- -- -- -- -- internal Surtloce Area f(in ^2)/fy 23.4 31.1 39.5 52.0 60.7 77.9 93.1 115.7 -- -- -- -- -- System intemal Un ^ gn = = >l 41719.48 00 0.0 9.4 0.0 0.0 0.0 00 00 8.9 00 0.0 0.0 00 18.3 18.3 100.0 % Fe_ ed_ Wator (FW) Scheduce -- -- 60 -- -- -- -- -- 60 -- -- -- -- Material / Type -- -- CS -- -- -- - -- CS -- -- -- -- 36.1 -- -- -- -- - 144 2 -- -- -- -- System intemal (n ^ 2) = = >l 1622 84 infornet Suriloce Area ((in ^2J/M -- -- 100 0% Main Ccolant Pipino (taCJ 0.0 00 81.5 0.0 9.7 00 0.0 27.3 00 00 0.0 35.3 23.2 157.0 157.0 Schedufe -- - ISOS -- 160S -- -. ISOS -- -- -- ~ 120 ~120 Manerial/ Type -- -- SS -- SS ~~ -- SS -- -- --
~133 ~199 30.7 -- 50.4 -- -- 99.0 - -- -- 405 3 445.4 System entomajl in ^ 2)= = >I 29720.12 inkr_nal Surdece Area f(in ^ 2)_/fy -- --
Main Steam Pipine IMS) 0.0 00 0.0 0.0 0.0 0.0 0.0 0.0 00 11.9 0.0 0.0 0.0 11.9 11.9 100.0 % Scheduoe -- - " -- -- -- -- -- -- 60 -- -- -- Material / Type -- -- -- -- -- -- -- -- -- CS -- -- -- 217.2 -- -- -- System _ antemal [n ^ 2)= = >l 2564 So sneernas Surhee Area f(in ^ 2pg -- -- -- -- -- -- -- 10.0% 36.5 3.8 0.0 89.7 0.0 0.0 0.0 00 0.0 0.0 0.0 0.0 13.0 130 0 Make-up and Purtfication_JMU) 00 Scheduae -- 10s 10S -- 108 - -- -- -- -- -- -- -- Material / Type -- SS SS -- SS -- -- -~ -- -~ ~~ ~~ -- 41.4 63.4 -- -- -- Svetem intomat in ^ 2)= = >f 70s s .42 internal Surnace Area f(in "2)/M
-- 33.3 -- -- -- -- -- --
0.0 12.0 0.0 0.0 00 0.0 0.0 0.0 41.5 32.8 0.0 ' O.0 N/A 86.3 0.0% Plant Als System _.1P_A) 0.0 40 -- -- -- -- ~~ - 40 40 -- -- Schedufo -- - CS -- -- -- -- - CS CS -- -- , Materiat/ Type -- -- -- 39.5 -- - - - - 228 6 377.7 -- -- System intomal (in ^ 2)= = >I 22353.37 Internal SurAsce Area f(in ^ 2)/M . 55.0 0.0 0.0 0.0 0.0 10.0 164.9 4.1 % Pressuse Rettet Lines (PRJ 0.0 00 0.0 00 0.0 77 5 32.4 00
~~ -- -- 40 40 -- 40 -- -- -- --
Schedule -- ~~ Caferial/ Type -- - - -- -- CS CS -- CS -- -- --
-- -- -- -- 77.9 93.1 -- 151.8 -- -- -- -- Syesem en_temal lin
- 2} = =MO219 Interna # Surface Ares ((in *2)/ft) --
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