WR-1 Reactor Decommissioning Update Reactor Characterization B. Wilcox, Director of Reactor Decommissioning J. Miller, Technical Lead 2019 November 12
Proposed In Situ Decommissioning
Multiple Barriers Protect the Environment
WM Main Campus (circa 2002) 4
WM Main Campus (planned 2026) 5
Current Activities Crawlspace and Ventilation System Characterization • Design of the Cap and Cover System • Updates to the Environmental Impact Statement • Updates to the Safety Assessment • Revisions to the Detailed Decommissioning Plan • Preparing submission packages for the Environmental • Assessment and licence Amendment Planning for lead removal and laboratory D&D •
Project timeline update
Characterization Update
Outline Why Characterize? • Historical Characterization Work • Summary of Data, Review and Gap Identification • Characterization Plan Development • Characterization Performed • Preliminary Results •
Why Characterize? • First of Many Steps Provides Confidence to Regulator • Supports Long Term Safety Case Modelling • Supports Worker Safety and Dose Estimates • Supports Strategic Decisions for Decommissioning •
Historical Work 1990 – Dose Rate Measurements taken in Fuel Channels • 1992 - Modelling Estimates of Inventory and Dose Rate • 1994 – Survey of Contamination in Restricted Access Area • Rooms (swipes) 1997 – Core Sample of Biological Shield analyzed • 2011 – Axial Dose Profiling of Fuel Channel • 2015 – Sample Collection from Moderator System • Various Routine Surveyor Logs of Building 100 and Reactor • Systems Ongoing Effluent Monitoring (Air and Liquid) •
Characterization Summary & 3 rd Party Review Summary document produced for historical characterization • work (WLDP-26100-041-000-0001). Submitted for 3 rd party review Oak Ridge Associated • Universities (ORAU), recognized experts in characterization work. Opportunities identified by ORAU • Suggested additional characterization samples be taken to • improve confidence.
Characterization Plan Developed by ORAU, Ranked Set Sample Methodology • Using 80% confidence intervals to determine number of • samples Survey 3 – Cut 1 • Separated by System • Primary Heat Transport Experimental Loops • • Moderator Fuel Transfer Systems • • Biological and Thermal Active and Process Drainage • • Shield Cooling
Characterization Work Performed • 2017-2018 121 coupon samples cut and analyzed – 363 Survey points • 4 Fuel Channel Scrape Samples from centre line of flux • Table Top Review of installed non-radiological contaminants • Lead Cadmium • • Chromium Boron • • Mercury Xylene • • PCB Potassium Hydroxide • • Field Characterization for Lead and PCBs •
Radiological Results • Coupon Results No unexpected results were observed. • Coupon results matched well with estimated radionuclide ‘fingerprint’. • Confirmed Primary Heat Transport System is main contributor to ‘out of core’ • inventory. Showed Historical Inventory Estimate conservative by factor of 5-20. •
Radiological Results • Coupon Results – Primary Heat Transport System No. of 99% UCL Mean @ 90% SD (Bq/cm 2 ) Nuclide Sample # >MDA (Bq) Confidence (Bq/cm 2 ) s Ag-108m 39 1 5.99E-02 2.84E-02 1.62E+06 Am-241 39 25 3.20E+01 2.43E+01 1.09E+09 C-14 39 35 3.31E+00 3.03E+00 1.26E+08 3/1 a Cm-243/244 3 4.38E-01 5.58E-02 8.78E+06 Co-60 39 26 1.53E+00 1.30E+00 5.55E+07 Cs-137 39 39 9.73E+02 5.90E+02 2.96E+10 Fe-55 3/1 2 2.82E+00 2.88E-01 5.40E+07 H-3 39 31 1.33E+01 6.44E+00 3.65E+08 Nb-94 39 13 2.19E-01 2.02E-01 8.36E+06 Ni-59 3/1 0 -8.59E-02 7.96E-01 2.73E+07 Ni-63 3/1 2 8.72E+00 8.60E-01 1.66E+08 Pu-238 3/1 4 8.75E+00 8.38E-01 1.65E+08 Pu-239/240 3/1 4 1.70E+01 1.27E+00 3.09E+08 Pu-241 3/1 4 1.60E+02 1.48E+01 3.00E+09 Sr-90 3/1 4 5.98E+02 9.47E+01 1.26E+10 U-235 39 6 7.77E-02 1.41E-01 4.67E+06 U-238 3/1 4 2.64E-02 4.66E-01 2.07E+07 Total Inventory 4.70E+10
Radiological Results • Coupon Results - Tritium Found little to no H-3 on moderator coupons. • Increase in Tritium measured in air effluent during couponing. • Indicated two possibilities: • 1. Tritium was released from sample during couponing (heat and vibration). • 2. Tritium on coupons is very low and effluent increases were due to trapped tritium • vapour released by opening the system. CNL explored several potential causes. •
Radiological Results • Coupon Results - Tritium Method Activity (Bq) 1.27x10 14 1% Remaining Estimate 1.11x10 9 Absorbed Tritium Analysis Amount of 3 H Released (2015) 2.47x10 15 Amount of 3 H Released (2017) 3.80 x10 14 Rate of 3 H Release (2015) 2.26 x10 15 Rate of 3 H Release (2017) 3.14 x10 14 5.31 x10 12 H Solubility Limit in SS
Radiological Results • Fuel Channel Scrape Samples Results
Radiological Results • Fuel Channel Scrape Samples Results
Radiological Results • Fuel Channel Scrape Samples Results
Radiological Results • Fuel Channel Scrape Samples Results
Radiological Results • Previous Estimate v. New Characterization Results System Pre-2018 (Bq) Post-2018 (Bq) Assessment Value (Bq) Bioshield 4.1E+09 N/A 4.1E+09 Core 1.1E+15 4.77E+14 1.1E+15 Out of Core 1.1E+12 8.45E+10 1.1E12 Total H-3 Out of Core 1.27E+14 2.47E+15 2.47E+15 Totals 1.18E+15 2.95E+15 3.53E+15
Non-Radiological Results • Desktop Review Contaminant Quantity Potassium Hydroxide 0.01 kg Boron 0.0009 kg Lead 40,800 kg Xylene 1.9 kg Palladium 15.5 kg Chromium 148 kg Cadmium 91.4 kg HB-40 (aka OS-84, Hydrogenated Terphenyl) 87,700 kg Mercury 0.74 kg
Non-Radiological Results • Lead Survey Not Removable 20,846 kg Potentially Removable 514 kg Removable 103,119 kg Within ISD Envelope 24,037 kg Outside of ISD Envelope 100,443 kg Grand Total 124,480 kg
Non-Radiological Results • Polychlorinated Biphenyls (PCBs) Survey of paint, caulking, other materials show no PCB’s within the ISD envelope • above exemption quantities (50 mg/kg). Some exterior window glazing found to contain PCB’s will be remediated prior to • demolition. Identified fluorescent light ballasts as possible PCB source. • All ballasts will be removed from the building prior to grouting and demolition. •
Uncertainty in Results • How is uncertainty dealt with? All results have an associated uncertainty • Confidence is improved through: • 1. Multiple methods of estimation 2. Conservative assumptions 3. Adherence to accepted best practices 4. Examination of effects of uncertainty through model sensitivity analysis
Conclusions
Conclusions • What do these results tell us? Confirm that most of the remaining radioactivity is in the reactor core • Confirm that estimates made on historical data were conservative • Provide additional evidence of the confidence in the inventory estimates • Allow us to compare WR-1 to natural analogues •
Conclusions • Natural Analogue
Jeff Miller, EIT (204)-753-2311 x63121 Jeffrey.Miller@cnl.ca
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