ML23242A313
| ML23242A313 | |
| Person / Time | |
|---|---|
| Site: | Washington State University |
| Issue date: | 08/30/2023 |
| From: | Washington State Univ |
| To: | Office of Nuclear Reactor Regulation |
| Shared Package | |
| ML23242A311 | List: |
| References | |
| Download: ML23242A313 (1) | |
Text
2023
ANNUAL OPERATIONS REPORT
WASHINGTON STATE UNIV ERSITY TRIGA REACTOR FACILITY LICENSE R-76 FOR THE REPORTING PERIOD JULY 1, 2022 TO JUNE 30, 2023
NUCLEAR SCIENCE CENTER l Washington State University, Pullman, WA 2023 Annual Report
TABLE OF CONTENTS
- 1. NARRATIVE
SUMMARY
OF OPERATION FOR FISCAL YEAR 2023................ 2
- 2. ENERGY AND CUMULATIVE OUTPUT............................................................... 3
- 3. EMERGENCY SHUTDOWNS AND INADVERTENT SCRAMS........................... 3
- 4. MAJOR MAINTENANCE....................................................................................... 4
- 5. CHANGES, TESTS, AND EXPERIMENTS PERFORMED UNDER 10 CFR 50.59 CRITERIA.............................................................................................................. 5
- 6. RADIOACTIVE EFFLUENT DISCHARGES.......................................................... 6
- 7. PERSONNEL AND VISITOR RADIATION DOSES.............................................. 7
- 8. REACTOR FACILITY RADIATION AND CONTAMINATION LEVELS................. 9
- 9. ENVIRONMENTAL MONITORING PROGRAM................................................. 12
Washington State University Nuclear Science Center 2023 Annual Report
- 1. Narrative Summary of Operation for Fiscal Year 2023
A. Operating Experience
Core 35A has accumulated 15,281 MWH from beginning of life (BOL) through June 30, 2023. During the reporting period of July 1, 2022 to June 30, 2023, a total of 1042 samples were irradiated, for 8609 user-hours. Additionally, 14 pulses greater than $1.00 of reactivity addition were performed during the reporting period. The quarterly operations summaries are shown in Table I.
B. Changes in Facility Design, Performance Characteristics, and Operating Procedures Related to Reactor Safety:
No changes in facility design, performance characteristics, or operating procedures related to reactor safety were made.
C. Results of Surveillance Tests and Requirements:
On October 14, 2022 while performing week 41 weekly maintenance the Reactor Supervisor discovered that the weekly security alarm operability test for week 40 during Q4 2022 had not been performed. Upon investigating it was determined the Reactor Supervisor was out the day it should have been performed and the Reactor Operator who was supposed to complete the maintenance stated they forgot about it and did not check if weekly maintenance had been completed before leaving for the day.
All other surveillance tests and requirements were performed and completed within the prescribed time period.
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- 2. Energy and Cumulative Output
The quarterly operations summaries are given in Table I. The cumulative energy output since the 1967 TRIGA fuel core was put into service is 1,867 megawatt days (MWD). The mixed Standard Fuel and 30/20 LEU Fuel Core 35A installed in 2008 has accumulated 637 MWD.
Table I. Fiscal Year 2023 Summary of Reactor Operation 1
Q3 2022 Q4 2022 Q1 2023 Q2 2023 Totals Hours of Operation 316 271 368 268 1223 Megawatt Hours 256 199 315 213 984 Sample Irradiations 81 40 63 69 253 Samples 332 149 168 333 982 External Irradiations 12 21 17 10 60 Pulses > $1.00 5 0 9 0 14 User Hours 1225 1864 3406 2115 8609
- 3. Emergency Shutdowns and Inadvertent Scrams
During the reporting period there was one emergency shutdown due to rupture of an in-core experiment.
On December 7, 2022 while operating at 1.0 MW, a loud pop was heard by the Reactor Supervisor who was at the controls and other personnel in the reactor pool room.
Exhaust Gas Monitor and Continuous Air Monitor readings spiked however stayed below their alarm setpoints. The operator shutdown the reactor and notified the facility Director of a suspected sample failure. A pool water sample was taken and ran on gamma spec to determine what isotopes were present in the water and help narrow down which sample failed. It was found that there were high levels of bromine in the water however no samples in core should have had bromine in them. While removing the C8 sample from the reactor core, bubbles rose to the pool surface along with the lid to a 2/5 dram vial. Using binoculars, the can containing the sample had ruptured open and was deformed. No samples in core should have contained vials and based on the information it was determined the sample which should have been antimony doped sand was actually KBr which is used in another project and is water soluble.
Upon further investigation it was determined that the reactor operator who had prepared and installed the sample into core was rushing to prepare it and had assumed an unlabeled can sitting on the counter in 201 was the antimony.
The failed sample was removed from the pool and placed in a lead container to decay in storage. The radionuclide inventory of the pool water following the failure showed Br-80 was above 10 CFR 20 limits and therefore the reactor was inoperable until the
1 Number of samples and sample irradiations do not include exter nal client irradiations. User hours denotes the total user hours, including external client irradiations.
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water could be remediated. The primary cooling loop was used to mix the pool water in order to dilute the concentration and the purification loop cleaned the water at 12 gallons per minute. Once the radionuclide content was below 10 CFR 20 limits the reactor was operated in increasing increments to determine reactivation of the dissolved material. No notable reactivation of the KBr was seen and the reactor was deemed fully operational on December 16, 2022.
The dates and causes of the seven inadvertent scrams are listed in Table II. No scrams were due to exceeding the limiting safety system setting or safety limit.
Table II. Inadvertent Scrams
Date Description
7/15/2023 Pulse High Power SCRAM when firing the pulse rod for pulse no. 1426. Power levels and fuel temperature were monitored and did not exceed tech spec values. Connections were checked and the problem was found to be the connections on the mode selector switch 7/22/2023 The pulse rod air scrammed after an operator inadvertently bumped the mode selector switch to pulse while operating at 1.0MW. The Operator ran down after the air scram.
8/11/2023 Building power flickered due to a thunderstorm causing a loss of power to the console and a subsequent scram.
9/6/2023 Operator inadvertently took mode selector switch to test instead of rundown.
2/10/2023 Worker removing a sample from D8 knocked the electromagnet for blade 1, causing it to disengage and fall. The console operator ran down the remaining control elements.
3/6/2023 Linear High Power Scram while operating in steady state mode at 1.0MW. Through a check against the Pulse Power Channel measurements and fuel temperature pen trace the cause was determined to be a failure of the linear channel electronics.
5/15/2023 Building power flickered due to a thunderstorm causing a loss of power to the console and a subsequent air scram.
- 4. Major Maintenance
Although they are not part of routine preventative maintenance, the below listed items were performed.
7/18/2022: The gasket seal on the ion exchanger lid was replaced.
9/6/2022: Replaced a coil in a solenoid in the ventilation system controls preventing the ventilation system from resetting back to normal mode after a loss of power.
10/21/2022: The transient rod was not consistently engaging the rod down limit switch following pulse rod inspection. Cleaned, re-lubricated, and re-aligned the system, after which the rod and limit switches worked as intended.
10/31/2022: Replaced a float switch on the pool water level alarm with an original specification component. Verified the system operable.
11/21/2022 - 11/23/2022: During control element drop time measurements, the transient rod drop time averaged 2.9 seconds. The solenoid and 3-way valve were replaced however it did not resolve the problem. The transient rod drive was removed from the bridge structure for disassembly and cleaning. Extensive residue and buildup on the drive, piston shaft, and seals was determined to be the cause. All components
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were cleaned, re-assembled, and re-lubricated and returned to service. Measured drop time after maintenance was 0.567 seconds. After re-installation, the drive motor alignment was adjusted on 11/23/2022 and 11/28/2022.
3/15/2023: A fuse repeatedly blew while performing routine CAM maintenance.
Identified and replaced an erroneously instal led console indication light triggering the electrical fault. Re-soldered and applied heat shrink to cables and cable connections for cable management.
3/16/2023: Various alarms came in when lowering a control element during a checkout. Fuse IF4 then blew during troubleshooting. Insulation on multiple wires had been cut through where it passed through a hole in in the metal console frame, causing an electrical fault. Resolved the problem by covering wires in electrical tape and re-installing a plastic grommet to guard the wires from the sharp edges.
5/25/2023: Relays 5K1-3, 5-9,12-16, and 18 were replaced in accordance with the table in a previously approved 50.59 change. The relays are directly related to the operation of the pulse rod system which was tested and verified to be operational and functioning normally.
6/14/2023: All the Control Elements were greased and the contacts for the connectors of the position indicators were cleaned with contact cleaner. The rod position indications were re-zeroed and upon reassembly each rod was cycled through its full travel to ensure that the position indications worked.
- 5. Changes, Tests, and Experiments Performed Under 10 CFR 50.59 Criteria
A total of one (1) proposed change to the facility was reviewed during the 2022-2023 year. The proposals was screened with the 10 CFR 50.59 requirements and found to screen out, thus allowing the change to be made. The following change was made to the facility under 50.59 criteria:
50.59-110-2023. The outlets on the reactor bridge were replaced with new ones and power supply was changed from the reactor co nsole to direct from building power as the outlets have no console related purpose and are unaffected by its operational state.
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- 6. Radioactive Effluent Discharges
A. Radioactive Liquid Effluent Releases
The liquid effluent releases for the facility during the reporting period are provided in Table III.
Table III. Monthly Liquid Effluent Releases
Month Volume (gallons)
July 2021 0 August 0 September 447 October 0 November 0 December 0 January 2022 0 February 0 March 0 April 0 May 0 June 0
Approximately 447 gallons of liquid effluents were released from the storage tank during the reporting period. Prior to discharge, a sample of the liquid in the tank was analyzed using gamma spectroscopy and liquid scintillation counting. The September gamma spectroscopy report indicated an activity of 1.54 µCi, which was the only liquid waste released in the reporting period.
B. Radioactive Gaseous Effluent Release
During the reporting period, no emission of a measurable quantity of gaseous or particulate material with a half-life greater than eight days was detected. The measured argon-41 out of the common exhaust stack did not exceed 20% of the effluent release limit. A total of 2.16 Ci of argon-41 was released, with an average argon-41 concentration of 1.10 x 10 -10 µCi/mL of air, after environmental dilution. The argon-41 release and the pool water analysis is used in the 2023 Annual Report for Radioactive Air Emission License (RAEL-004), stack number 7. Per COMPLY v1.7, the reactor facility (stack 7) complies at level 4 with an effective dose equivalent of 6.3 x 10 -3 mrem/yr. The monthly releases from Ar-41 are summarized in Table IV.
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Table IV. Monthly Argon-41 Releases2
Month Quantity (Ci) Conc. After % of DAC Dilution (µCi/mL) Limit July 2022 1.93E-01 1.19E-10 3.9E-03 August 2.02E-01 1.21E-10 4.0E-03 September 1.68E-01 1.04E-10 3.5E-03 October 5.98E-02 3.58E-11 1.2E-03 November 1.36E-01 8.42E-11 2.8E-03 December 1.99E-01 1.19E-10 4.0E-03 January 2023 1.99E-01 1.19E-10 4.0E-03 February 9.93E-02 6.35E-11 2.1E-03 March 1.25E-01 7.49E-11 2.5E-03 April 2.23E-01 1.38E-10 4.6E-03 May 2.44E-01 1.46E-10 4.9E-03 June 3.10E-01 1.91E-10 6.4E-03
C. Radioactive Solid Waste Disposal
During the reporting period, 19.9 µCi in 39 cubic feet of non-compacted solid waste was transferred to the WSU Radiation Safety Office for packaging and disposal.
- 7. Personnel and Visitor Radiation Doses
The quarterly doses of the WSU Nuclear Science Center reactor staff and experimenters are given in Table V. The maximum quarterly dose to a reactor staff member was 211 mrem, whole body. A total of 870 individual persons visited the Nuclear Science Center during the reporting period, of which 317 entered a controlled access area (CAA).3 A total of 28 group tours, consisting of 193 individuals, visited the center during the reporting period, also entering a CAA. None of these tour groups received measured exposures above 0.1 mR as determined by digital pocket dosimeters.
2 Quantity released based on 4500 CFM effluent of ventilation system in AUTO mode of operation. Concentration after dilution is based on 10 CFR 20 effluent release limit of 1.0 x 10 -8 µCi/mL for Ar-41 (Table 2, Col.1), and a dilution factor of 3.4 x 10-3 (WSU Technical Specifications 3.5.2). DAC limits are bas ed on 10 CFR 20 derived air concentration limit of 3.0 x 10 -6 µCi/mL for Ar-41 (Table 1, Col. 3) and a dilution factor of 3.4 x 10-3.
3 A controlled access area is an area in the building where radioac tive materials are used or stored and is a part of the licens ed reactor facility.
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Table V. Quarterly Reactor and Experimenter Staff Dose 4 (mrem)
Badge No. Q3 2021 Q4 2021 Q1 2022 Q2 2022 00706 124 70 51 72 01330 3 M -- --
03959 -- -- 36 60 10921 114 16 7 21 12176 199 29 16 --
01368 8 -- -- --
01360 8 -- -- --
00705 161 31 39 33 03752 -- M 2 --
01340 67 19 17 22 03654 -- 5 5 M 01428 4 M 2 --
01335 3 M -- --
03396 -- -- 10 46 05374 -- -- -- M 01327 4 M 7 3 03150 M M 1 --
03777 -- 1 3 --
03151 M -- -- --
03157 M M 3 --
01160 18 25 31 34 08141 2 M 3 1 03137 1 M 5 M 00698 211 56 -- --
03652 -- M 3 13 03768 -- 4 6 6 03145 -- M -- --
03653 -- 5 9 1 03776 -- M 1 --
03651 -- M 4 44 00704 61 17 43 35 03149 M -- -- --
01328 -- M 4 M 07117 -- -- -- 1 01329 3 M -- --
01328 5 M -- --
03788 -- 2 3 M
4 -- denotes data not available either due to departure from the facility or new personnel starting at the facility. An M de notes that the dosimeter reading was less than or equal to the background radiation level for that quarter.
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- 8. Reactor Facility Radiation and Contamination Levels
The limit of quantification (LOQ) for building removable contamination determination survey samples as measured by liquid scintillation assay is 8.80 x 10 -8 Ci/cm2; the survey sample data that was collected for removable contamination determination were averaged over one year. Routine building surveys showed average levels of removable activity to be less than the LOQ for all non-CAAs.
Table VI. Average Removable Contamination for Weekly Monitoring in CAAs and Non-CAAs5
Location Measured Activity Above LOQ
(µCi/cm2) 201 B M 201A M 201 RX BRIDGE STEPS M 210 SAMPLE DROP TUBE M 201 RX BRIDGE - SOUTH M 201 RX BRIDGE - NORTH M 201 EXPERIMENTER PLATFORM 2.4E-07 201 BENCHES M 201 FLOOR SOUTH M 201-C HEAT EXCHANGER FLOOR M 201 FLOOR NORTH M 106 ION EXCHANGER PIT M 101-A PURIFICATION PUMP PIT M 101 DOOR WAY M 101 SAMPLE PREP BENCH M 101 SAMPLE DROP HOOD #2 M 101 HOOD #1 M 101 HOOD #2 M RAM STORAGE SAFE M 101 ISLAND M 101 NORTH LAB BENCH M 101 HOOD #3 M 101 HOOD #4 M 101 SHIPPING BENCH M B21 PANORAMIC IRRADIATOR M B21 FLOOR M RM 2 SOUTH FLOOR M RM 2 THERMAL COLUMN M RM 2 THERMAL COLUMN FLOOR M
5 Bolded text indicates a non-CAA. Regular text indicates a CAA. M indicates the value is below the LOQ value of 8.80 x 10 -8 Ci/cm2.
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RM 2 NORTH FLOOR M RM 2 WEST CAVE FLOOR M RM 2 EAST CAVE FLOOR M
The results for the routine area radiation surveys of the building in CAAs and non-CAAs are given in Table VII. The highest average dose rate for a single location in a CAA was 42.85 mrem/hr, which occurred in Room 2 East Cave. This value is less than the limit for CAAs. The lowest average dose rate in a CAA was 0.04 mrem/hr (a level considered background), which occurred in Room 201B. The average dose rate in the radiochemistry sample hoods (a non-CAA) was 0.06 mrem/hr. The East and West cave are secured storage areas that are designed to house radioactive sources, and provide shielding. The space is posted as a high radiation area. Personnel do not typically work in this area and it is locked when not in use.
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Table VII. Average Radiation Dose Rates for Weekly Monitoring in CAAs and Non CAAs6
Location Average Dose Rate (mrem/hr)
ROOM 201 B 0.04 ROOM 201 A 0.05 ROOM 201 BRIDGE 0.32 ROOM 201 BENCHES 0.28 ROOM 201 SOUTH 0.21 ROOM 201 EAST 0.39 ROOM 201 C HEAT EXCHANGER 0.08 ROOM 201 NORTH 0.37 ROOM 106 ION EXCHANGER PIT 0.91 ROOM 101 A PURIFICATION PIT 0.62 ROOM 101 DOORWAY 0.04 ROOM 101 SAMPLE PREP BENCH 0.04 SAMPLE STORAGE 0.09 ROOM 101 SAMPLE DROP HOOD 0.04 ROOM 101 HOOD 1 0.09 ROOM 101 HOOD 2 0.06 101 NORTH LAB BENCH 0.08 101 HOOD #3 0.04 ROOM 101 HOOD 4 0.07 ROOM 101 SHIPPING BENCH 0.04 ROOM B21 PANORAMIC IRRADIATOR 0.04 ROOM 2 SOUTH 0.15 ROOM 2 THERMAL COLUMN 0.06 ROOM 2 NORTH 0.18 ROOM 2 WEST CAVE 5.28 ROOM 2 EAST CAVE 42.85
6 Bolded text indicates a non-CAA. Regular text indicates a CAA. M indicates the value is below the LOQ value of 9.09 x 10 -8 Ci/cm2.
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- 9. Environmental Monitoring Program
The environmental monitoring program is used to determine the offsite background radiation levels; thermoluminescent dosimeters (TLDs) are used to make the measurements. The offsite radiation monito ring program is required by the Technical Specifications. The TLDs that are used for offsite monitoring are designated as TLD numbers 3, 7, 9, 15 through 35, and 39 through 44. The average background radiation level is then compared to the nearest occupied dwelling.
Average quarterly dose rates for offsite locations are listed in Table VIII and are used to calculate the Technical Specification threshold of 20% above the background radiation level and compared to the limiting values which are listed in Table XI. The average environmental radiation levels for the closest offsite point of extended occupancy is listed in Table X. Table IX shows the quarterly environmental radiation levels for those TLDs located at onsite locations. The onsite locations are not required to be compared to background radiation levels.
The closest offsite points of extended occupancy are compared in Figure 1 to both the background radiation levels and the 20% above background radiation levels. The ALARA effluent release limits in Technical Specification 3.5.2(3) specify that annual radiation exposure due to reactor operation, at the closest offsite extended occupancy, shall not, on an annual basis, exceed the average offsite background radiation by more than 20%. For the reporting period, the average background radiation dose rate for off-site locations was 0.39 mrem/day, while the average radiation dose rate at the closest extended occupancy area 600 meters away was 0.38 mrem/day. This result indicates that no exposure level above normal background radiation were found, and that no dose levels exceeded Technical Specifications requirements for an offsite area of extended occupancy.
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Table VIII. Environmental Radiation Levels at Offsite Locations 7,8
Location Q3 Q4 Q1 Q2 Average 2021 2021 2022 2022 Fence E of NSC 0.33 0.52 0.29 0.34 0.37 Fence, N of Rad Waste Shed 0.38 0.54 0.34 0.39 0.41 Fence directly N Rad Waste Shed 0.46 0.80 0.55 0.55 0.59 S NSC, on parking lot fence 0.33 -- -- 0.38 0.36 Fence S Roundtop Dr, 10th pole W of pole C14 0.37 0.55 0.34 0.38 0.41 Telephone pole C12 0.38 0.55 0.31 0.37 0.40 Telephone pole near golf course gate 0.38 0.54 0.30 0.37 0.40 E across fairway on pine tree 0.32 0.51 0.28 0.35 0.37 Maple tree #54 along driving range 0.33 0.53 0.29 0.34 0.37 NW to fence uphill from driving range 0.35 0.55 0.33 0.36 0.40 Follow fence E to fence corner 0.35 0.58 0.34 0.37 0.41 S to lone spruce tree near water hazard -- 0.53 -- 0.39 0.46 Roundtop hill park, NW fence corner 0.32 0.51 0.30 0.36 0.37 Deciduous tree edge of 18th green 0.39 0.58 0.35 0.39 0.43 6ft pine tree, 3rd W down cart path from clubhouse 0.35 0.54 0.36 0.38 0.41 3rd to last tree after gap in same line of trees 0.34 0.47 0.30 0.33 0.36 SW to fence along path near 2nd to last tee box at bottom hill 0.37 0.54 0.33 0.36 0.40 Follow fence partway up hill after fence turns S 0.38 0.59 0.34 0.37 0.42 Follow fence, 15th pole E after fence turns W 0.35 0.54 0.30 0.38 0.39 Follow fence about halfway between last TLD and corner 0.37 0.55 0.33 0.39 0.41 Largest bush S of NSC 0.35 0.59 0.49 0.38 0.45 2nd fence S NSC, W end at gate 0.38 0.49 0.34 0.37 0.40 S Fairway Rd, 1st light post on right 0.33 0.52 0.30 0.36 0.38 S Fairway Rd, 2nd light post on right 0.33 0.55 0.31 0.36 0.39 Ellis Way and Hog Lane sign 0.34 0.53 0.30 0.35 0.38 Bottom of radio antenna hill, fence next to shrub left of gate -- 0.53 0.28 0.32 0.38 3rd fence S of NSC, SE corner, cow pasture 0.34 0.52 0.34 0.38 0.40 Airport fence W end runway at gate 0.37 0.55 0.35 0.37 0.41 Fence/entry bar E of Jewett Observatory 0.33 0.48 0.34 0.34 0.37 Railing at Terrell Mall / Library 0.28 0.51 0.31 0.32 0.36
7 Offsite defined by the Technical Specif ication 1.0 and 5.1.1 as any location, wh ich is outside the site boundary. The --
indicates a TLD which was missing.
8 Dose rate in mrem/day.
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Table IX. Environmental Radiation Levels at Onsite Locations 9,10
Location Q3 Q4 Q1 Q2 Average 2021 2021 2022 2022 E lower loading dock 0.31 0.57 0.33 0.37 0.40 Pool room truck door fence S end 0.51 1.01 0.50 0.51 0.63 Pool room truck door fence N end 0.75 0.95 0.70 0.62 0.76 E wall rad waste shed 0.45 0.63 -- 0.64 0.57 N wall rad waste shed 0.53 0.72 -- 0.69 0.65 Cooling tower fence, NE corner 0.60 0.80 0.52 0.62 0.64 Room 101 window 0.40 0.55 0.35 0.43 0.43 Railing next to upper liquid waste tank 0.37 0.63 0.36 0.37 0.43 Room 2 truck door fence 0.32 0.59 0.30 0.34 0.39 Transformer vault vent louvers 0.35 0.61 0.38 0.42 0.44 NSC main entrance, light fixture 0.38 0.63 0.40 0.41 0.46 NSC roof, pool room vent stack 0.38 0.57 0.41 0.28 0.41 NSC roof, guide wire E end of building 0.38 0.48 0.34 0.33 0.38 NSC roof, E pool room vent support leg 0.68 0.89 0.45 0.56 0.65 NSC roof, air conditioning support leg 0.38 0.57 0.35 0.32 0.41 NSC roof, W pool room vent support leg 0.78 0.98 0.55 0.59 0.73
Table X. Environmental Radiation Levels for the Closest Offsite Point of Extended Occupancy10
Location Q3 Q4 Q1 Q2 Average 2021 2021 2022 2022 Apt complex C, gas meter 0.31 0.52 0.29 0.34 0.37 Apt complex B, gas meter 0.34 0.52 0.28 0.32 0.37 1st fence S apt complex A -- 0.53 0.30 0.34 0.39
Table XI. Background Environmental Radiation Levels 10
Description Q3 Q4 Q1 Q2 Average 2021 2021 2022 2022 Background radiation levels 0.35 0.54 0.33 0.36 0.39 20% above background radiation levels 0.42 0.64 0.39 0.44 0.47
9 Onsite defined by the Technical Specification 1.0 and 5.1.1 as any location within the site boundary. The -- indicates a TLD which was missing.
10 Dose rate in mrem/day.
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0.70
0.60
0.50
0.40 Apartment Complex A Apartment Complex B 0.30 Apartment Complex C
Background
0.20 20% Above Background
0.10
0.00 Q3 2021 Q4 2021 Q1 2022 Q2 2022 Average Quarter
Figure I: Environmental radiation levels for the closest offsite point of extended occupancy radiation levels as compared to background radiation levels and 20% above background radiation levels.
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