ML21076A471
| ML21076A471 | |
| Person / Time | |
|---|---|
| Site: | 07000925 |
| Issue date: | 02/26/2021 |
| From: | Environmental Properties Management, Enercon Services, Burns & McDonnell Engineering Co, Veolia Nuclear Solutions Federal Services |
| To: | Office of Nuclear Material Safety and Safeguards, Cimarron Environmental Response Trust, NRC Region 4 |
| Shared Package | |
| ML21076A479 | List:
|
| References | |
| Download: ML21076A471 (65) | |
Text
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env1ronme ta Radiation Protection Plan properties management. LlC Document No. RPP-001 Rev. 4 Effective date:
Section 10.0 Page 10-3
- Weekly, in office space located in areas surrounding or adjacent to RAs, other than Radioactive Materials Areas, where the potential exists for external radiation exposure or contamination spread.
- Weekly, in routinely occupied RAs, other than Radioactive Materials Areas.
- Monthly, or upon entry, if entries are less frequent than monthly, for Radioactive Materials Areas.
During routine contamination surveys, if contamination levels exceed action levels discussed in Table 12-1, the RSO or designee will determine the cause for the contamination and determine appropriate corrective actions, including decontamination, increasing the frequency of surveys, need for additional engineering or administrative controls, etc.
10.3 Job Coverage Surveys Job-coverage surveys are specified in Activity Plans and routine operations procedures. The types of radiological surveys (i.e., radiation, contamination, airborne radioactivity), frequency (e.g., number of times during a shift, at a specific step in the activity, etc.), and location are determined by the RSO or designee based on the radiological hazards associated with the work to be performed. Special survey requirements may be provided by the RSO or designee, when needed.
10.4 Investigative Surveys Investigative surveys shall be performed as soon as practicable following the discovery or indication of abnormal radiological conditions.
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Section 10.0 Page 10-4 10.5 Final Status Surveys Final status surveys will be required to support license termination. During the post-remediation monitoring period, a final status survey plan will be developed and submitted for approval by the NRC.
10.6 Personnel Contamination Monitoring Personnel shall routinely perform contamination monitoring (frisking) prior to exiting RAs that have the potential for spreading contamination or per Activity Plan or procedural requirement.
At a minimum, hands and feet shall be frisked when exiting these areas. Documentation of routine personnel frisking is maintained in field notes maintained by the HP Technician.
Notification of the RSO or designee is required when personnel contamination in excess of twice the background count rate is detected. RP procedures provide specific instructions approved by the RSO for performing, documenting, and reporting personnel contamination monitoring reports.
- 10. 7 Area Radiation Monitoring The RSO or designee will determine when and where area radiation monitoring is appropriate.
Area radiation monitoring may be performed using either passive devices, such as dosimeters (e.g., thermoluminescent or optically stimulated luminescent) or real-time radiation monitors.
Dosimeters are posted at the Cimarron Site to confirm that no occupational worker is likely to receive 100 mrem DDE in a year.
10.8 Air Monitoring Air monitoring is required whenever airborne radioactivity levels are expected to exceed lpercent of the DAC as listed in Appendix B, Table 1 "Occupational Values" of 10 CFR 20.
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environmenta Radiation Protection Plan properties management. LLC Document No. RPP-001 Rev.4 Effective date:
Section 10.0 Page 10-5 Considering the types of work activities described in the Decommissioning Plan, airborne suspension of licensed radioactive material is not anticipated to generate airborne radioactivity approaching I% of a DAC. However, the Decommissioning Plan requires that General Area (GA) air sampling, using either low or high-volume portable air samplers, will be performed throughout the resin unloading and packaging process for at least the first three resin exchanges.
Lapel samplers will be used in conjunction with the GA samplers to demonstrate that GA samplers are representative of the air breathed by workers. Following analysis of the air sampling results from each of these resin exchanges, the RSO will determine the need for and frequency of additional air sampling and types of air sampling to be performed (e.g., GA or lapel). In addition, the same air sampling regime will be conducted during the first three operations involving loading biomass from the filter press into the cart for disposal. Following analysis of the air sampling results from each of these biomass loading operations, the RSO will determine the need for and frequency of additional air sampling and types of air sampling to be performed (e.g., GA or lapel).
NOTE:
- A prospective evaluation of potential intake during groundwater processing operations was performed. The calculation supporting this evaluation is provided in Appendix A of the RPP. This calculation was based on the 60% design of the groundwater treatment system and supported the decision that internal monitoring (e.g., bioassay) and respiratory programs would not be needed at the Site. The evaluation also informed the development of the air sampling program described in Section 10.8. The supporting calculation will be reviewed at 90% design, updated, if necessary, and re-evaluated to determine if the RPP should be updated. In addition, periodically through groundwater processing, the supporting calculation will be reviewed to ensure it reflects operational experience and to determine if changes to the RPP are necessary.
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Cimarron Environmental Response Trust environmenta Radiation Protection Plan properties management. LLC Document No. RPP-001 Rev.4 Effective date:
Section 10.0 Page 10-6 The results of this evaluation indicate that continuous air monitors are not needed as the potential for an individual to be exposed to 40 DAC-hours in week does not exist at the Cimarron Site. Updates to this calculation are reviewed to ensure this conclusion remains applicable.
Selection of air samplers is based on the following criteria:
- GA air sampling will be accomplished by using portable air samplers, as discussed, above. Sampling heads will be placed within the breathing zone to ensure that the air sample is representative of the air breathed by the individual worker.
- GA air samplers typically sample at a rate of approximately 3-25 liters per minute (1pm)
(less than 1 cubic foot per minute (cfm)) for a low volume sampler to 1900 1pm (70 cfm) for a high-volume sampler. Based on the nature of the low enriched uranium encountered, the detection capability of the air sampling equipment and associated radiological analysis (e.g., sample counting) will be used to determine the total volume of air needed to be collected to ensure that 1% of the DAC. The enrichment of the uranium will be based on either the actual enrichment being collected on the resin or a conservative basis (i.e. 4%). This calculation will be documented in a Site procedure or technical basis document. As the actual enrichment of recovered uranium in each area changes (i.e., WA or BAI), the 1% DAC value may be recalculated Minimum collection times will be determined so adequate sensitivities are achieved for a given monitoring period.
- The need for air sampling will be prospectively determined based on the final process system design and potential for generation of airborne radioactivity. Due to the chemical and physical nature of the uranium-bearing media (e.g., water and moist ion exchange resin), minimal, if any airborne radioactivity is expected to be generated. Engineering Verify version is current prior to use
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Radiation Protection Plan Document No. RPP-001 Rev.4 Effective date:
Section 10.0 Page 10-7 and physical controls incorporated into the process equipment design will also be considered in determining the need for air monitoring.
- The frequency of calibration of the flow meters on the air samplers will be based on manufacturers' recommendations (typically annually).
- Specific action levels (i.e., specific projected or actual airborne radioactive material concentration levels) will be developed for assigning respiratory protection, collecting bioassay samples, and stopping work.
- Respiratory protection shall be considered if a worker' s intake is expected to exceed 40 DAC-hours in a week.
- A bioassay program must be implemented for any worker whose intake is expected to exceed 10% ALI or 40 DAC-hours in a week.
- Work shall be ceased if air sampling results show greater than 10% DAC is present.
The RSO shall evaluate the situation and provide recommendations for restarting work for the approval of the EPM PM.
Air samples will be counted on-site using existing laboratory bench top scalers (e.g., Ludlum Model 3030 or similar equipment). MDAs based on various sample count times will be calculated and used to determine the sample volume needed to detect less than 1% DAC for 4%
enriched uranium. This information will be documented and used to determine the minimum sampling time for GA and lapel air samplers.
10.9 Survey Training Surveys shall be performed by personnel who have been trained commensurate with the type of surveys to be performed. Training will address the following, as applicable:
- Appropriate instrumentation to be used,
- Operational and response checks for survey instrumentation, Verify version is current prior to use
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environme tal Radiation Protection Plan properties management. L Document No. RPP-001 Rev. 4 Effective date:
Section 10.0 Page 10-8
- Survey methods, recording of data,
- Calculations, data evaluation, and
- Action levels.
10.10 Survey Documentation Radiation, contamination, and airborne radioactivity surveys performed for compliance purposes, or radiation and contamination surveys performed to demonstrate that decommissioning criteria have been met, shall be documented and maintained in accordance with 10 CFR 20, Subpart L and the QAPP.
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Cimarron Environmental Response Trust environmentaf Radiation Protection Plan properties ma agement. C Document No. RPP-001 Rev. 4 Effective date:
Section 11.0 Page 11-1 11.0 RADIOACTIVE MATERIALS CONTROL 11.1 Section Overview This section addresses radioactive material controls employed at the Cimarron Site. This section includes requirements related to the following:
- Material Control and Accountability
- Receipt, Labelling, and Storage of Radioactive Material
- Shipment and Transfer of Radioactive Material
- Controls for Radioactive Sources.
- Theft or Loss of Radioactive Material 11.2 Material Control and Accountability The potential for a nuclear criticality event during the proposed decommissioning program at the Cimarron Site is extremely unlikely because both the concentration and the enrichment of uranium in material generated during decommissioning are low. Treatment of groundwater to remove the enriched uranium from groundwater will result in a more concentrated form of uranium on the ion-exchange resin. The accumulation of enriched uranium on resin has been evaluated by an analysis to demonstrate nuclear criticality safety.
The RSO is responsible for evaluating proposed changes to the groundwater treatment system and/or process in consultation with an individual with experience in nuclear criticality safety evaluation. The RSO will review and approve any changes made to the groundwater treatment system and will periodically conduct inspections of the system and operations to confirm that process and administrative controls assure that the license possession limits are not exceeded.
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Section 11.0 Page 11-2 All personnel responsible for the operation of the process systems will receive training on the potential for nuclear criticality and the need to comply with the controls established to maintain nuclear criticality safety during treatment and processing operations. The training will address:
- Awareness of the significance of exceeding the basic parameters necessary to stay within the Nuclear Criticality Safety analysis which are:
- Any measurement of an enrichment >7.33% U-235 ,
- Any measurement of the U-235 concentration on the resin >8g/kg,
- The need to assure that the U-235 concentration in packaged waste containers is
<0.5g U-235 per kg of waste,
- The need to assure that the process system inventory does not exceed 1,200 grams of U-235,
- The need to assure that the total Site inventory does not exceed 0.5 effective kilograms of U-23 5, and
- Any change in the storage of containers or process equipment that would result in a height > 7 feet.
- Awareness that any measurement of an enrichment >5% requires downgrading action in accordance with the license possession limits.
- Awareness that all individuals are required to implement an immediate "stop work" response if any of the above listed parameters are violated.
These necessary controls are addressed in the Material Control and Accountability procedures. If any of these parameters are exceeded, the Nuclear Criticality Safety analysis has been invalidated and it would be necessary to stop processing operations until either the analysis is redone and/or the situation corrected that led to the exceedance.
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Document No. RPP-001 Rev.4 Effective date:
Section 11.0 Page 11-3 Administrative controls are implemented in Material Control and Accountability procedures to ensure that the SNM inventory complies with the following requirements:
- Uranium mass determinations will be based on analytical measurements using the ICP-MS (EPA 200.8) method to report the U-235 and U-238 mass concentrations.
- The enrichment values will be calculated using analytical measurements of U-235 and U-238. The enrichment of the uranium is calculated (ignoring the U-234 mass contribution) by:
E = Mu-235 / (Mu-235 + Mu-238)
Where:
E = the Uranium enrichment level in wt. % U-235 Mi = the mass of the isotope in micrograms isotope per liter or gram of sample
- SNM mass contents of process lines transporting and influent tanks storing groundwater will be conservatively estimated based on reasonable concentrations and enrichments of the uranium. These components will contain minor quantities of SNM. These conservatively established mass values may be re-evaluated and revised as appropriate if information is obtained that the difference in calculated mass of U-235 is significant to the mass inventory value.
- The SNM contents of resin vessels will be established based on the total flow of groundwater through the vessel and the difference between the input and output uranium concentrations of the flow. The enrichment of the SNM will be initially based on conservatively established values until analytical results provide actual enrichment values. The following points describe in greater detail the process for establishing the total U-235 inventory for the resin material:
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environmenta Radiation Protection Plan properties managernen . LLC Document No. RPP-001 Rev.4 Effective date:
Section 11. 0 Page 11-4
- During the initial system startup phase the sample analysis tum-around time will be reduced to obtain data on an expedited basis
- Water samples for each treatment train will be taken from the influent and the effluent from each of the lead, lag and polish vessels
- The enrichment of the uranium for each train will initially be the assumed values presented in Appendix O of the DP and will be revised when analytical results are obtained from samples of processed resin
- The mass ofU-235 added to each vessel will be calculated based on the total flow of water processed through the vessel for each time-period between sampling events times the difference between the influent and effluent water concentration for that vessel. The total inventory of the vessel is the sum of all mass determinations over the entire period the vessel has been in the treatment process
- The total U-235 inventory of the resin material is the sum of the U-235 contained in:
o all the resin vessels in all three treatment systems o vessels containing spent resin if any are being stored while awaiting the blending process o resin in the blending equipment o packages of processed resin that have not been transferred to storage
- Once a vessel has been emptied of resin the inventory value for that vessel will be set to zero
- The minimum quantity of absorbent material to be blended with the resin to yield a fissile exempt, dry resin mixture will be calculated based on assumed enrichment and the uranium mass derived from process measurements (with safety factor)
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Section 11. 0 Page 11-5
- The resin from the lead vessel changeout will be blended with the absorbent and loaded into waste packages
- The sampling of the prepared resin waste will be performed in accordance with an approved sampling procedure
- Initially the SNM content of packages of waste will be based on process measurements. Upon receipt and validation of analytical data the SNM content of the packaged waste material will be finalized, the container inventory log updated, and the packages will be relocated to the "Ready-to-Ship" area of the Secure Storage Facility
- Waste containers which are awaiting final analytical data will be stored in an "In-Process" area within the Secure Storage Facility
- Once a package of waste has been determined to meet the transportation and waste disposal requirements, the SNM contents of the package will be removed from the "Mass Inventory Log" and the package and SNM contents will be entered on the "Container Inventory Log."
- When a package is shipped to disposal the package and SNM contents will be removed from the "Container Inventory Log."
11.3 Receipt, Labeling, and Storage of Radioactive Material All radioactive materials shall be received in accordance with radioactive material license possession limits and 10 CFR 70.19. The individual respo'nsible for radioactive material receipt shall ensure that all surveys required by 10 CFR 20.1906 are performed and review shipment paperwork to ensure compliance with 49 CFR.
Each container of radioactive material shall be labeled as required by 10 CFR 20.1904.
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environmenta Radiation Protection Plan properties management. LLC Document No. RPP-001 Rev. 4 Effective date:
Section 11.0 Page 11-6 Radioactive material shall be secured against unauthorized access or removal. Radioactive material storage areas shall be posted and controlled using appropriate barriers and radiological postings.
11.4 Shipment and Transfer of Radioactive Material Radioactive Materials shipments shall comply with NRC (10 CFR) and U.S. Department of Transportation (49 CFR) regulations. Low-level radioactive waste shipments transferred for disposal shall be accompanied by a shipment manifest prepared in accordance with 10 CFR 20.2006. Radioactive material shall only be transferred to authorized individuals in accordance with the appropriate regulations in 10 CFR 20, and 10 CFR 70.
11.5 Controls for Radioactive Sources The RSO shall approve all requisitions for radioactive sources and ensure that source inventories are performed on a quarterly basis. Radioactive sources shall be tested for leakage and/or contamination upon receipt and on a quarterly basis, except that any licensed sealed source is exempt from leak tests if the source contains less than 0.1 microcuries of plutonium or uranium, 100 microcuries of beta and/or gamma emitting radioactive material or 10 microcuries of other alpha emitting radioactive material. Leak testing and inventory of Exempt Quantity radioactive sources is not required; however, these sources should be stored in a secure area to prevent unauthorized removal or access.
Unless specifically authorized by the RSO, electroplated sources are not smear tested for leakage to prevent removal of radioactive material from the electroplating.
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Cimarron Environmental Response Trust environmental Radiation Protection Plan propertie managemen .
Document No. RPP-001 Rev. 4 Effective date:
Section 11.0 Page 11-7 The RSO shall approve locations for storage of radioactive sources. Radioactive source storage areas shall be secured against unauthorized removal or access of licensed radioactive material and posted per 10 CPR 20.1902.
11.6 Theft or Loss of Radioactive Material Any individual who discovers that radioactive material is lost, stolen, or missing shall immediately notify the RSO. The RSO shall evaluate the physical and radiological characteristics of the missing material and the potential hazards to workers and the general public, initiate an investigation to locate the material, and perform a root cause evaluation of the incident. The RSO shall determine the need for notifications to regulatory authorities and make notifications as necessary per 10 CPR 20.2201.
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environmenta Radiation Protection Plan properties management, L c Document No. RPP-001 Rev.4 Effective date:
Section 12.0 Page 12-1 12.0 CONTAMINATION CONTROL 12.1 Section Overview The purpose of contamination control is to prevent and/or minimize the spread of radioactive contamination to individuals, areas, and equipment. Control of radioactive surface contamination prevents or minimizes possible inhalation or ingestion of radioactive material by personnel, skin dose from small particles ofradioactivity, and the spread to or build-up of radioactive material in the facility or environment from decommissioning operations. Controls to prevent the spread of contamination shall be proposed by the Activity Leaders and approved by the RSO or designee prior to implementation.
12.2 Contaminated Buildings and Equipment Radioactive contamination of buildings and equipment located within an RA shall be maintained below the removable contamination limit of 1,000 dpm/100 cm2 alpha. In addition, Contaminated Area controls, including posting, shall be implemented whenever removable contamination in an Unrestricted Area exceeds 1,000 dpm/100 cm2 alpha or 1,000 dpm/100 cm2 beta-gamma. The Site incorporates the ALARA philosophy when selecting decontamination methods and practices.
As a general rule, decontamination is performed by working from areas of low contamination to areas of high contamination if possible. Decontamination materials should be limited to the minimum required for the task. All decontamination materials shall be collected, monitored, and properly dispositioned. Table 12-1 provides a summary of contamination action levels and associated actions to be taken.
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Section 12.0 Page 12-2 12.3 Contaminated Personnel Decontamination of personnel shall be performed under the guidance of health physics personnel and shall incorporate good health physics practices and ALARA principles. An individual whose skin or personal clothing is found contaminated above background shall not exit an RA without prior approval of the RSO. Appropriate surveys and monitoring shall be performed to evaluate dose to the individual resulting from contamination.
12.4 Spill of Radioactive Material A spill of radioactive material requires immediate actions which include:
- Stop the spill
- Wam other personnel
- Isolate the area
- Minimize radiation exposure
- Secure the area and stand guard (until otherwise direct by health physics personnel)
Supplementary actions should include the performance of radiological surveys in immediate and adjacent areas, including downwind.
12.5 Contamination Control During Groundwater Processing Contamination control during groundwater processing involves both process operations and activities necessary to supply groundwater to the processing facility. This section of the RPP is intended to implement contamination control commitments identified in the DP. Other contaminated materials that will be handled, including the filtered suspended solids and biodenitrification system waste, will have lower concentrations and are covered by the discussion of the resin material.
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Cimarron Environmental Response Trust environmenta Radiation Protection Plan properties management. LLC Document No. RPP-001 Rev.4 Effective date:
Section 12.0 Page 12-3 Subsurface soil will be brough to the surface during installation of injection and extraction trenches, monitoring wells, trenches for piping and utilities, etc. These soils have been previously released from license controls. Surveys shall be performed during these activities to determine if soil contamination is encountered. Survey requirements will be consistent with RP procedures and limits specified in associated Activity Plans to ensure compliance with license conditions.
Low-enriched uranium will be processed through ion exchange resins that will concentrate the uranium in the resins. The concentration of uranium on these resins represents a source of potential contamination. The following contamination controls ensure that contamination is contained and not spread throughout the processing facilities or across the Site.
- Influent piping contains low concentrations of uranium with little potential for generating contamination. Routine monitoring is performed during operations to ensure that contamination is controlled and not being spread at well heads where the groundwater is extracted. Connections to the water treatment systems are inspected and monitored to identify and repair leaks.
- Engineering controls are included in the design of the groundwater treatment system to contain contamination during ground water processing. Double walled tanks are used to contain the influent groundwater awaiting processing. Ion exchange resins are contained in stainless steel vessels. Spent resins are processed through a wet process that ensures airborne radioactivity is not generated. The spent resin is processed in an enclosed system. Spent resin is packaged as discussed in the DP. Procedures for contamination monitoring and air sampling are provided to demonstrate the effectiveness of these engineering controls.
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Section 12.0 Page 12-4
- Engineering controls are incorporated in the design to eliminate or minimize the potential for drips and leaks during sampling, resin vessel changeout, and spent resin processing.
Protective clothing shall be prescribed for maintenance activities involving potential exposure to spent resins.
- Effluent from treatment systems must contain uranium at concentrations below drinking water standards, as demonstrated by discharge sampling requirements specified in the discharge permit issued by the Oklahoma Department of Environmental Quality. Leaks or unintentional releases of effluent do not constitute contamination control concerns.
The QAPP requires that only appropriately trained workers (Section 2.3) are permitted access to Contaminated Areas. Work performed in Contamination Areas will be performed in accordance with procedures that require measures incorporated into the design to prevent or contain drips, leaks, etc. are correctly implemented.
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environmenta* Radiation Protection Plan properties management. LLC Document No. RPP-001 Rev. 4 Effective date:
Section 12.0 Page 12-5 Table 12-1 Contamination Action Levels Location/Type of Contamination Action Actions to be Taken Radiological Monitoring Contamination Level Unrestricted Area - 1,000 dpm/100 cm 2 1. Post area/restrict RSO or designee Removable beta/gamma or access. determines the need for Contamination 1,000 dpm/100 cm 2 alpha 2. Investigate for spread increased :frequency of and determine personnel contamination surveys.
affected.
- 3. Decontaminate the area and de-post, as appropriate.
- 4. Determine corrective I*
actions to prevent I recurrence, if necessary.
Unrestricted Area - Fixed 1,000 dpm/100 cm 2 I . Post area/restrict RSO or designee Contamination beta/gamma or access. determines the need for 1,000 dpm/100 cm 2 alpha 2. Investigate cause and increased frequency of corrective actions. contamination surveys.
- 3. Determine whether to decontaminate or I,
implement controls to prevent spreading contamination.
Restricted Area - 1,000 dpm/100 cm 2 1. Post area, if not RSO or designee Removable beta/gamma or already posted. determines the need for Contamination 1,000 dpm/100 cm2 alpha 2. Determine source of increased :frequency of contamination, if area was contamination surveys.
not posted and actions necessary to prevent further contamination spread.
- 3. Decontaminate, if appropriate.
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Cimarron Environmental Response Trust environm,enta Radiation Protection Plan properties management, LC Document No. RPP-001 Rev. 4 Effective date:
Section 12.0 Page 12-6 Location/Type of Contamination Action Actions to be Taken Radiological Monitoring Contamination Level Restricted Area- Fixed 1,000 dpm/100 cm 2 1. Post area, if not RSO or designee Contamination beta/gamma or posted. determines the need for 1,000 dpm/100 cm 2 alpha 2. Determine the source increased frequency of of contamination and contamination surveys.
necessary actions to prevent further contamination spread.
- 3. Decontaminate, if appropriate.
Release of materials for See Section 13 .3 1. Decontaminate or See Section 13 .3 unrestricted use dispose of as radioactive waste.
Personnel/clothing Detectable contamination 1. Decontaminate 1. See Section 12.3 contamination (e.g. 2 times background) personnel in accordance regarding personnel on clothing or skin with Section 12.3. contamination
- 2. Decontaminate or monitoring.
discard contaminated 2. RSO or designee personal clothing if authorizes release of unrestricted release personal clothing, if criteria cannot be satisfied appropriate.
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Cimarron Environmental Response Trust environmenta Radiation Protection Plan properties management. C Document No. RPP-001 Rev.4 Effective date:
Section 13.0 Page 13-1 13.0 RELEASE FOR UNRESTRICTED USE OF MATERIALS 13.1 Section Overview Site personnel are authorized to release tools, equipment, parts, and materials for unrestricted use provided that radiation levels and surface contamination levels do not exceed the limits in Condition 27(c) of the license. Such surveys will be performed and documented by qualified individuals.
Tools, equipment, parts, and material that do not come into contact with subsurface soil or groundwater containing radioactive material above the unrestricted release criteria do not require surveys prior to release from the Site.
13.2 Survey Instrumentation Instruments used to perform surveys for release for unrestricted use must be calibrated using NIST traceable, or equivalent, standards for energies and geometries similar to material being released. The energy dependence of the instruments to alpha, beta, and gamma radiation, as applicable, shall be known and documented.
13.3 Release Surveys of Materials As provided in License Condition 27(c), the Site uses the unrestricted release criteria listed in the August 1987 "Guidelines for Decontamination of Facilities and Equipment Prior to Release for Unrestricted Use or Termination of License for Byproduct, Source or Special Nuclear Material" for surfaces of buildings and equipment, and the October 23, 1981, BTP "Disposal or Onsite Storage of Thorium or Uranium Wastes from Past Operations," for soils or soil-like material.
The specific values are listed in paragraphs 13 .3 .1, 13 .3 .2, and 13 .3 .3.
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Section 13.0 Page 13-2 Release surveys will consist of direct (fixed + removable) and removable (smears) contamination monitoring. The Site is authorized to release materials provided that the direct and removable levels do not exceed the limits stated in the Trust license and summarized below. Such surveys will be performed and documented by qualified individuals.
Survey plans may be developed for the release of property or equipment associated with non-routine activities. Such survey plans will include the methods used to estimate uncertainty bounds for each type of instrument measurement.
13.3.1 Surfaces of Buildings, Equipment, and Outdoor Areas
- 5,000 dpm alpha/100 cm2 averaged over 1 m2 (direct)
- 5,000 dpm beta-gamma/I 00 cm2 averaged over 1 m2 ( direct)
- 15,000 dpm alpha/I 00 cm2 maximum over 1 m2 ( direct)
- 15,000 dpm beta-gamma/I 00 cm2 maximum over 1 m 2 ( direct)
- 1,000 dpm alpha/I 00 cm2 averaged over 1 m2 (removable)
- 1,000 dpm beta-gamma/I 00 cm2 averaged over 1 m2 (removable) 13.3.2 Soils
- Depleted Uranium - 35 pCi/g total uranium
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Cimarron Environmental Response Trust envi ronmenta Radiation Protection Plan properties management. LLC Document No. RPP-001 Rev. 4 Effective date:
Section 13.0 Page 13-3 13.3.3 Exposure Rates
- Surface of buildings and equipment
- 5 µR/hr - above background at 1 meter
- Soils
- 10 µR/hr - average above background at 1 meter
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Section 14.0 Page 14-1 14.0 RESPIRATORY PROTECTION 14.1 Section Overview The need for a respiratory protection for radiological work is not envisioned at the Cimarron Site. Work activities that could potentially expose workers to airborne radioactive material have been evaluated to determine the potential intakes during groundwater treatment and spent resin processing. The evaluation employed the methods discussed in Regulatory Guide 8.25, Rev. 1, "Air Sampling in the Workplace" and NUREG-1400, "Air Sampling in the Workplace." If processes or operations change, then a re-evaluation of potential intakes shall be performed to determine the potential intake that could result from these changes. If the potential intake determined from this evaluation is 2% ALI or greater, the RSO will perform an ALARA evaluation when it is not practical to apply engineering controls or procedures that demonstrates that he use ofrespiratory protection equipmentis ALARA. If the ALARA evaluation demonstrates that use of respiratory protection equipment is ALARA, then the RSO will implement the respiratory protection program described in this section.
Respiratory protection measures shall be employed when necessary to protect workers from airborne hazards. Groundwater treatment results in the generation of moist treatment media with little potential to generate airborne radioactivity. However, if future conditions change and the RSO or designee determines, through review of field conditions or anticipated work functions, that respiratory protection is required, procedures and controls will be instituted in accordance with the requirements found in 10 CFR 20, Subpart H, "'Respiratory Protection and Controls to Restrict Internal Exposure in Restricted Areas" for radiological hazards and the 29 CFR 1910.134 for non-radiological hazards. Section 14.2 provides specific requirements for the respiratory protection program, if needed.
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environmenta Radiation Protection Plan properties management. LC Document No. RPP-001 Rev.4 Effective date:
Section 14.0 Page 14-2 If a respiratory protection program is determined to be necessary, the program will be based on guidance provided in Regulatory Guide 8.15, Rev. 1, "Acceptable Programs for Respiratory Protection," and NUREG/CR-0041, Rev. 1, "Manual of Respiratory Protection Against Airborne Radioactive Material."
14.2 Respiratory Protection Program Respiratory protection will be required if work activities could potentially expose workers to 40 or more DAC-hours in a week. Respiratory protection will also be required for any areas where airborne radioactive material concentrations are expected to exceed 1 DAC. If either of these trigger levels are encountered, as required by 10 CFR 20.1703(c), the respiratory protection program will include:
- Air sampling sufficient to identify the potential hazard, permit proper equipment selection, and estimate doses;
- Surveys and bioassays, as necessary, to evaluate actual intakes;
- Testing ofrespirators for operability (user seal check for face sealing devices and functional check for others) immediately prior to each use;
- Written procedures regarding-
- Monitoring, including air sampling and bioassays;
- Supervision and training of respirator users;
- Fit testing;
- Respirator selection;
- Breathing air quality;
- Inventory and control; Verify version is current prior to use
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Section 14.0 Page 14-3
- Storage, issuance, maintenance, repair, testing, and quality assurance of respiratory protection equipment;
- Recordkeeping; and
- Limitations on periods of respirator use and relief from respirator use;
- Determination by a physician that the individual user is medically fit to use respiratory protection equipment:
- Before the initial fitting of a face sealing respirator;
- Before the first field use of non-face sealing respirators, and
- Either every 12 months thereafter, or periodically at a frequency determined by a physician.
- Fit testing, with fit factor > 10 times the APF for negative pressure devices, and a fit factor> 500 for any positive pressure, continuous flow, and pressure-demand devices, before the first field use of tight fitting, face-sealing respirators and periodically thereafter at a frequency not to exceed 1 year. Fit testing must be performed with the facepiece operating in the negative pressure mode.
Verify version is current prior to use
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environmenta Radiation Protection Plan properties management. LLC Document No. RPP-001 Rev. 4 Effective date:
Section 15.0 Page 15-1 15.0 ENVIRONMENTAL MONITORING 15.1 Section Overview Environmental monitoring shall be performed at various locations to monitor the migration of licensed material from former (now decommissioned) sources through environmental media.
Final surveys have demonstrated that buildings and soils have been decommissioned. Licensed material exceeds decommissioning criteria in groundwater in three areas: Burial Area #1, the Western Upland Area, and the Western Alluvial Area. The Licensee shall maintain an environmental monitoring program in these three areas until superseded by a groundwater remediation work plan.
Effluent from the groundwater treatment process will be monitored to demonstrate that the concentrations of uranium complies with discharge permit limits and underground injection permits. Monitoring will be performed in accordance with permit requirements and the Sampling and Analysis Plan. The Sampling and Analysis Plan will address how background and baseline concentrations of radionuclides in environmental media have been established through appropriate sampling and analysis. The Sampling and Analysis Plan will include the following information:
- A description of known or expected concentrations of radionuclides in effluents;
- A description of the physical and chemical characteristics of radionuclides in effluents;
- A summary or diagram of all effluent locations;
- Justification that samples are representative of actual releases;
- A summary of the sample collection and analysis procedures, including the minimum detectable concentrations of radionuclides;
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Section 15.0 Page 15-2
- A summary of sample collection frequencies; and
- A description of environmental monitoring recording and reporting procedures.
Direct radiation from groundwater processing operations is monitored in the vicinity of the Western Area Treatment Facility and Burial Area #1 Treatment Facility as discussed in Section 15.3.
15.2 Surface and Groundwater Monitoring Surface and groundwater samples are collected annually and are analyzed for fluoride, nitrates/nitrites, gross alpha radioactivity, gross beta radioactivity, and uranium isotopes. The locations identified in Table 15-1 shall be sampled on an annual basis.
Upon approval of Decommissioning Plan, the in-process groundwater monitoring plan described in Section 8. 7 of the Decommissioning Plan will replace the environmental monitoring program described in the preceding paragraph.
15.2.1 Quality Control in Sampling Sample collection, preservation, shipping, and analysis shall be conducted in accordance with the Site-specific Sampling and Analysis Plan and associated procedures. Data review, reporting, and management will be conducted in accordance with Quality Assurance Implementing Procedure, QAIP-17.1, "Data Management Procedure."
15.2.2 Reporting Environmental monitoring results shall be reported to NRC within 30 days of the completion of data review.
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envi ronmenta I Radiation Protection Plan properties management. LLC Document No. RPP-001 Rev.4 Effective date:
Section 15.0 Page 15-3 15.3 Direct Radiation Dosimeters were deployed on October 1, 2019. These dosimeters collected background dose in the vicinity of the Western Area Treatment Facility and Burial Area #1 Treatment facility prior to construction of these facilities. These dosimeters are used to establish the baseline of background radiation levels prior to commencing decommissioning activities. Additionally, one dosimeter was deployed along the haul path between the facilities. Once decommissioning activities commence, these dosimeters will be used to determine radiation levels outside the RA from groundwater processing activities. Figures 15-1 and 15-2 depict dosimeter locations.
Table 15-2 provides a verbal description of the dosimeter locations. Dosimeters are changed on a quarterly basis.
NOTE: Dosimeter locations may be reevaluated during construction of these facilities and adjusted, if necessary, by the RSO. The rationale for any location adjustments shall be documented. Additionally, dosimeter locations will be periodically evaluated during decommissioning activities to determine if locations need to be adjusted or removed.
Justification for changes shall be documented and approved by the RSO.
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environmenta Radiation Protection Plan properties management. LC Document No. RPP-001 Rev. 4 Effective date:
Section 15.0 Page 15-4 Table 15-1 Surface and Groundwater Monitoring Locations BURIAL AREA #1 WESTERN UPLAND AREA 1314 1351 TMW-08 1352 TMW-09 1354 TMW-13 1356 02W06 02W08 02W09 WESTERN ALLUVIAL AREA 02W16 MWWA03 02W17 MWWA09 02W27 T-62 02W28 T-64 02W32 T-70R 02W35 T-76 02W42 T-77 02W43 T-79 02W44 T-82 SURFACE WATER 1201 Cimarron River Upstream 1202 Cimarron River Downstream
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Document No. RPP-001 Rev. 4 Effective date:
Section 15.0 Page 15-5 Table 15-2 Environmental Dosimeter Locations Location Description Designation Western Area Treatment Facility WATF-01 Northwest comer of Treatment Building.
WATF-02 Northeast comer of Treatment Building.
WATF-03 Eastern fence line in line with the northern wall of the Treatment Building.
WATF-04 Eastern fence line approximately center of the Secure Storage Facility.
WATF-05 Northwest comer of the Secure Storage Facility.
WATF-06 To the southeast of the Treatment Building along the eastern fence line where the fence runs to the southwest.
WATF-07 Southeast comer of Treatment Building.
WATF-08 South fence line at Treatment Building mid-point.
WATF-09 Southwest comer of Treatment Building.
WATF-10 Point directly west of the southwest comer of the Treatment Building on the western fence line.
WATF-11 Point directly west of the centerline of the Treatment Building on the western fence line.
Burial Area #1 Treatment Facility BAITF-01 Northwest comer offence line around BA #1 Treatment Facility.
BAI TF-02 Northeast comer of fence line around BA #1 Treatment Facility.
BAI TF-03 East fence line of BA# Treatment Facility at centerline of Uranium IX Treatment Skid.
BAITF-04 Southeast comer of fence line around BA #1 Treatment Facility.
BAITF-05 Southwest comer offence line around BA #1 Treatment Facility.
BAITF-06 West fence line of BA# Treatment Facility at centerline of Uranium IX Treatment Skid.
Roadway (Haul Path)
ROAD-01 Approximately half the distance between the WA TF and the BA 1TF; on the side of the road.
Figure 15-1 Western Area Treatment Facility
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Section 15.0 Page 15-6 Figure 15-1 Western Area Treatment Facility fRiJEC~ID lr1ACK P~V ll'J(~ SEC T CM .I!.
fRUEGRID PRO "AVING SECT ON 8 Figure 15-1 depicts approximate locations of 11 environmental dosimeter locations deployed for pre-operational (baseline) monitoring around the Western Area Treatment Facility.
A red border indicates the anticipated restricted area for the "Secure Storage Facility
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Section 15.0 Page 15-7 Figure 15-2 Burial Area #1 Treatment Facility
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Figure 15-2 Burial Area #1 Treatment Facility 13? ........ *-.
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? Y**.* ...~--: . Figur of . e 1.5-2 depicts environmental approx.imate dosimeter locations locations "~** .",c: --** , .-._ .n ,,*
\;;- _-:_ . ~.:-;<,;_~; ..;// *. *.** .. . deployed for pre-operation al (baseline)
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- :,:/ -_{ .. BA1TF-04 mon1tonng around the Bunal Area # 1
[:., ~ ~ ~;r.': 1~: Treatment Facility. A red border indicates llf '*
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Section 16.0 Page 16-1 16.0 DEFINITIONS Absorbed Dose: Energy imparted by ionizing radiation per unit mass of irradiated material. The units of absorbed dose are the rad and the gray (Gy). 1 Gy = 100 rad.
Adult: An individual 18 or more years of age.
Airborne Radioactive Material or Airborne Radioactivity: Radioactive material dispersed in the air in the form of dusts, fumes, particulates, mists, vapors, or gases.
Airborne Radioactive Material Area or Airborne Radioactivity Area: A room, enclosure, or area in which airborne radioactive materials, composed wholly or partly of licensed material, exists in concentrations:
(1) in excess of the derived air concentrations (DAC) specified in appendix B of 10 CFR 20.1001 - 20.2401, or (2) to such a degree that an individual present in the area without respiratory protection equipment could exceed, during the hours an individual is present in a week, an intake of 0.6% of the Annual Limit on Intake (ALI) or 12 DAC hours.
ALARA: An acronym for "As Low As is Reasonably Achievable". ALARA means making every reasonable effort to maintain exposures to radiation as far below the dose limits in 10 CFR 20 as is practical consistent with the purpose for which the licensed activity is undertaken, taking into account the state of technology, the economics of improvements in relation to state of technology, the economics of improvements in relation to benefits to the public health and safety, and other societal and socioeconomic considerations, and in relation to utilization of nuclear energy and licensed materials in the public interest.
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Section 16.0 Page 16-2 ALARA Committee: The Cimarron Site ALARA Committee that has responsibility for overall coordination of the ALARA Program. The Committee is composed of members as described in Section 4.0 of this RPP and meets on a regular basis (typically, quarterly) to review the status of the ALARA Program and to approve changes to the RPP and DP.
Alpha Particle: A positively charged particle ejected spontaneously from the nuclei of some radioactive elements. It is identical to a helium nucleus that has a mass number of 4 and an electrostatic charge of+2, i.e. two protons and two neutrons.
Annual Limit on Intake (ALI): The derived limit for the amount of radioactive material taken into the body of an adult worker by inhalation or ingestion in a year. ALI is the smaller value of intake of a given radionuclide in a year by the reference man that would result in a committed effective dose equivalent of 5 rems (0.05 Sv) or a committed dose equivalent of 50 rems (0.5 Sv) to any individual organ or tissue. (ALI values for intake by ingestion and by inhalation of selected radionuclides are given in Table 1, Columns 1 and 2, of appendix B to 10 CFR 20.1001 thru 20.2401 ).
Audit: An audit is an evidence gathering process. Audit evidence is used to evaluate how well audit criteria (procedures, requirements, policies) are being met. Audit evidence is used to determine how well policies are being implemented, how well procedures are being applied, and how well requirements are being met.
Atomic Number (Symbol Z): The number of protons in the nucleus of an atom.
Background:
Ambient signal response recorded by measurement instruments that is independent of radioactivity contributed by the radionuclide being measured in the person or sample.
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Section 16.0 Page 16-3 Background Radiation: Radiation from cosmic sources; naturally occurring radioactive material, including radon (except as a decay product of source or special nuclear material); and global fallout as it exists in the environment from the testing of nuclear explosive devices or from past nuclear accidents such as Chernobyl that contribute to background radiation and are not under the control of the Licensee. "Background radiation" does not include radiation from source, byproduct, or special nuclear materials regulated by the NRC.
Becquerel (Bq): The term used to describe one disintegration per second (dps ).
Beta Particle: Beta particles are emitted by the nucleus of an atom to attain stability. Beta particles are usually negatively charged, and are emitted from the nucleus of atoms with an excess of neutrons and serve to reduce the number of neutrons in the nucleus. Some beta particles are positively charged. These positively charged beta particles, known as positrons, are emitted from a nucleus and result in an increase in the number of neutrons in the nucleus.
Negatively charged beta particles and positively charged positrons have a mass equal to 1/1837 that of a proton. Beta particles are easily stopped by a thin sheet of metal or plastic.
Bioassay (radiobioassay): The determination of kinds, quantities or concentrations and, in some cases, the locations of radioactive material in the human body, whether by direct measurement or by analysis and evaluation of materials excreted or removed from the human body.
Breathing Zone: The breathing zone is that region adjacent to a worker's mouth and nostrils from which air is drawn into the lungs while he/she is performing assigned work.
Breathing Zone Air Sample: Air which is drawn through or into the sample media and is a fair representation of the workers "Breathing Zone."
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environmenta1 Radiation Protection Plan properties management. LLC Document No. RPP-001 Rev.4 Effective date:
Section 16.0 Page 16-4 Byproduct material:
(1) Any radioactive material (except special nuclear material) yielded in, or made radioactive by, exposure to the radiation incident to the process of producing or using special nuclear material; (2) The tailings or wastes produced by the extraction or concentration of uranium or thorium from ore processed primarily for its source material content, including discrete surface wastes resulting from uranium solution extraction processes. Underground ore bodies depleted by these solution extraction operations do not constitute "byproduct material" within this definition; (3) (i) Any discrete source of radium-226 that is produced, extracted, or converted after extraction, before, on, or after August 8, 2005, for use for a commercial, medical, or research activity; or (ii) Any material that-(A) Has been made radioactive by use of a particle accelerator; and (B) Is produced, extracted, or converted after extraction, before, on, or after August 8, 2005, for use for a commercial, medical, or research activity; and (4) Any discrete source of naturally occurring radioactive material, other than source material, that-(i) The NRC, in consultation with the Administrator of the Environmental Protection Agency, the Secretary of Energy, the Secretary of Homeland Security, and the head of any other appropriate Federal agency, determines would pose a threat similar to the threat posed by a discrete source of radium-226 to the public health and safety or the common defense and security; and
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Section 16.0 Page 16-5 (ii) Before, on, or after August 8, 2005, is extracted or converted after extraction for use in a commercial, medical, or research activity.
Calendar Quarter(s): First quarter - January 1 through March 31 Second quarter -April 1 through June 30 Third quarter - July 1 through September 30 Fourth quarter - October 1 through December 31 Calendar Year: From January 1 through December 31.
Calibrate: To adjust and/or determine:
(1) The response or reading of an instrument relative to a series of conventionally true values; or (2) The strength of a radiation source relative to a standard or conventionally true value.
Committed Dose Equivalent (CDE) (HT, so): Means the dose equivalent to organs or tissues of reference (T) that will be received from intake of radioactive material by an individual during the 50 year period following the intake.
Committed Effective Dose Equivalent (CEDE) (HE, so): The sum of the products of the weighting factors applicable to each of the body organs or tissues that are irradiated and the committed dose equivalent to these organs or tissues (HE, so = I:rWr, so).
Contact Dose Rate: A radiation dose rate as measured with the detector or instrument case within 1/2 inch of the surface being measured.
Contamination, Radioactive: Deposition of radioactive material in any place where it is not desired. Radioactive contamination may be removable (loose) or fixed.
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Section 16.0 Page 16-6 Contaminated Area: Any area that has radioactive contamination at levels greater than the radioactivity release limits for unrestricted use.
Continuous Air Sampling/Monitoring: A method of sampling used to measure airborne radioactivity levels in routinely occupied areas.
Controlled Area: An area outside of a Restricted Area but inside the Site boundary, where access can be limited by the Licensee for any reason.
Corrective Action(s): Action(s) taken to improve areas of performance or to eliminate causes of adverse trends in performance identified during Audits, Surveillances, and as a response to a Notice of Deficiency.
Counts Per Minute (cpm): The rate of ionizing event occurrence in one minute recorded by a radiation detection instrument designed to count ionizing events caused by radiation.
Curie (Ci): A measure of the amount of radioactive material present.
One curie equals 37 billion (3.7 E+ l0 or 3.7 x 10 10) becquerels (dps)or 2.2 trillion (2.2 E+12) radioactive disintegration's per minute (dpm).
A millicurie (mCi) is 2.2 billion (2.2 E+09) dpm A microcurie (µCi) is 2.2 million (2.2 E+06) dpm A nanocurie (nCi) is 2.2 thousand (2.2 E+03) dpm A picocurie (pCi) is 2.2 dpm.
Declared Pregnant Woman: A woman who has voluntarily informed the Licensee, in writing, of her pregnancy and the estimated date of conception. The declaration remains in effect until the declared pregnant woman withdraws the declaration in writing or is no longer pregnant.
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Section 16.0 Page 16-7 Decontamination: Means the process of removing or reducing the level of contamination on an item or individual.
Deep Dose Equivalent (Hd): The dose equivalent at a tissue depth of 1 cm (1000 mg/cm2)
Applies to external whole body exposure.
Derived Air Concentration (DAC): The concentration of a given radionuclide in air which, if breathed by the reference man for a working year of 2,000 hours0 days <br />0 hours <br />0 weeks <br />0 months <br /> under conditions of light work (inhalation rate 1.2 cubic meters of air per hour), results in an intake of one ALI. DAC values are given in Table 1, Column 3, of appendix B to 10 CFR 20.1001-2401.
Derived Air Concentration-hour (DAC-hour): The product of the concentration of radioactive material in air (expressed as a fraction or multiple of the derived air concentration for each radionuclide) and the time of exposure to that radionuclide, in hours. A Licensee may take 2,000 DAC-hours to represent one ALI, equivalent to a committed effective dose equivalent of 5 rems (0.05 Sv).
Detector: That portion of an instrument system sensitive to and used for the quantification of ionizing radiation.
Direct Contamination Survey: This method measures fixed and removable levels of surface contamination. A direct frisk is performed by scanning the survey location using a count rate meter.
Direct Reading Dosimeter (DRD): A monitoring device consisting of a collection chamber coupled with an optical lens and calibrated scale. DRD's can be used as a device to provide individuals with an immediate estimate of their external gamma radiation exposure.
Discrete Source: A radionuclide that has been processed so that its concentration within a material has been purposely increased for use for commercial, medical, or research activities.
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Section 16.0 Page 16-8 Disintegrations Per Minute (dpm): Refers to the number of nuclear transformations occurring per minute.
Disintegrations Per Second (dps): Refers to the number of nuclear transformations occurring per second.
Dose or Radiation Dose: A generic term that means absorbed dose, dose equivalent, effective dose equivalent, committed dose equivalent, committed effective dose equivalent, or total effective dose equivalent, as applicable to context and as defined in 10 CFR 20. The unit for absorbed dose is the rad. 100 rad = 1 Gy.
Dose Equivalent (HT): Means the product of the absorbed dose in tissue, quality factor, and all other necessary modifying factors at the location of interest. The units for dose equivalent rem.
100 rem = 1 Sv.
Dose Rate: The quantity of absorbed dose delivered per unit of time.
Dosimeter: Any of several types of devices used to measure radiation dose. Common types include TLD, OSL, film, and direct reading dosimeters.
Effective Dose Equivalent (HE): The sum of the products of the dose equivalent to the organ or tissue (HT) and the weighing factors (WT) applicable to each of the body organs or tissues that are irradiated (HE = "f.WTHT).
Effluent: Material discharged into the environment from licensed operations.
Embryo/Fetus: The developing human organism from conception until the time of birth.
Exposure: Means being exposed to ionizing radiation or to radioactive material. The unit of exposure is the roentgen.
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Section 16.0 Page 16-9 External Dose: That portion of the dose equivalent received from a source of radiation outside the body.
Extremity: Means hand, elbow, arm below the elbow, foot, knee, or leg below the knee.
Frisk: The performance of a direct survey for radioactive contamination.
Gamma Ray (Gamma Radiation): High-energy, short wavelength electromagnetic radiation (a packet of energy) emitted from the nucleus. Gamma radiation frequently accompanies alpha and beta emissions and always accompanies nuclear fission. Gamma rays are very penetrating and are best stopped or shielded against by dense materials, such as lead or uranium. Gamma rays are similar to x-rays, but are usually more energetic.
General Area Dose Rate: A radiation dose rate measured at 30 cm or more from a surface.
Gray (Gy): The SI unit for absorbed dose: 1 Gy = 1 Joule kg -t = 100 rad.
Groundwater: Water contained in pores or fractures in either the unsaturated or saturated zones below ground level.
In-Vitro Bioassay (indirect): The estimation of radioactivity in the human body based upon:
(1) the measurement ofradioactivity in excreta or other materials taken from the body, and (2) a biological model for the radionuclide movement in body tissues and organs.
In-Vivo Bioassay (direct): The measurement ofradioactivity in the human body using instrumentation which detects radiation emitted from radionuclides in the body.
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Section 16.0 Page 16-10 Individual Monitoring: The assessment of dose equivalent by use of devices designed to be worn by an individual; the assessment of committed effective dose equivalent by bioassay or by determination of the time-weighted air concentrations to which an individual has been exposed; or the assessment of dose equivalent by the use of survey data.
Individual Monitoring Devices: Devices designed to be worn by a single individual for the assessment of dose equivalent. Examples include thermoluminescence dosimeters (TLD's ),
optically stimulated luminescent (OSL) dosimeters, direct reading dosimeters, and lapel air samplers.
Instrument: A complete system designed to quantify one or more characteristics of ionizing radiation or radioactive material.
Intake: The amount of radioactive material taken into the body by inhalation, absorption through the skin, injection, ingestion, or through wounds.
Internal Dose: That portion of the dose equivalent received from radioactive material taken into the body.
Isotopes: Nuclides having the same number of protons in their nuclei but differing in the number of neutrons. Isotopes have the same atomic number and different mass numbers.
Lens Dose Equivalent (LOE): Dose equivalent due to external exposure to the lens of the eye.
It is taken as the dose equivalent at a tissue depth of 0.3 cm (300 mg/cm2).
Licensed Radioactive Material: Source material, special nuclear material, or byproduct material received, possessed, used, transferred or disposed of under a general or specific license issued by the NRC.
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Section 16.0 Page 16-11 License: Means the radioactive materials license issued by the NRC to the Trust to possess and/or use radioactive materials. Other licenses may be issued to the Trust by other state or federal agencies.
Licensee: The holder of the radioactive materials license (the Trust).
Limits (dose limits): The permissible upper bounds of radiation doses.
Low-Level Radioactive Waste (LLRW): Those low-level radioactive wastes containing source, special nuclear, or by-product material that are acceptable for disposal in a land disposal facility. Low-level waste has the same meaning as in the Low-Level Waste Policy Act; that is, radioactive waste not classified as high-level radioactive waste, transuranic waste, spent nuclear fuel, or by-product material as defined in paragraphs (2), (3 ), and (4) of the definition of Byproduct material set forth in 10 CFR 20.1003.
Member of the Public: An individual who is not receiving an occupational dose.
Micro: A prefix meaning "one millionth" (1 E-06), as in microcurie.
Milli: A prefix meaning "one thousandth" (1 E-03), as in millirem, millirad, or millicurie.
Minimum Detectable Activity: The smallest concentration of radioactivity in a sample that can be detected with a 5% probability of erroneously detecting radioactivity, when in fact none may be present (Type I error) and also, a 5% probability of not detecting radioactivity, when in fact it is present (Type II error). Often used interchangeably with Minimum Detectable Concentration, since the difference between the two terms is only one of units conversion.
Minor: An individual less than 18 years of age.
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Section 16.0 Page 16-12 Monitoring (Radiation Monitoring): The measurement of radiation levels, concentrations, surface area concentrations, or quantities of radioactive material, and the use of the results of these measurements to evaluate potential exposures and doses.
Nano: A prefix meaning "one billionth" (1 E-09), as in nanocurie.
NRC: U.S. Nuclear Regulatory Commission or its duly appointed representatives.
Nuclide: Any one of the approximately 1800 isotopes of all the elements, whether radioactive or not. See radionuclide and isotope.
Occupational Dose: The dose received by an individual in the course of employment in which the individual's assigned duties involve exposure to radiation or to radioactive material from licensed and unlicensed sources of radiation, whether in the possession of the Licensee or other person. Occupational dose does not include doses received from background radiation, from any medical administration the individual has received from exposure to individuals administered radioactive material and released under 10 CFR 35.75, from voluntary participation in medical research programs, or as a member of the public.
Occupational Dose Limit: The maximum legally allowable dose to individuals during a specific time period, as defined by 10 CFR 20.
Particulate: Sometimes used to describe alpha and beta radiations, but most often used to mean dust or droplets containing radioactive material.
Pico: A prefix meaning one trillionth" (1 E-12), as in picocurie.
Planned Special Exposure: An infrequent exposure to radiation, separate from and in addition to the annual dose limits.
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Section 16.0 Page 16-13 Posting: A standardized sign or label which bears the standard trefoil radiation symbol in magenta or purple or black on a yellow background and information concerning a specific radiological hazard.
Protective Clothing: Clothing provided to reduce exposure and prevent the spread of contamination to personnel clothing or the body while performing work with radioactive materials.
Qualification: Certification of the fact that an individual possesses the knowledge, capabilities (e.g., physical) characteristics, or abilities gained through experience, training, or on-the-job training that an individual can perform a required task.
Qualified Escort: An individual that meets the Qualified Escort training requirements set forth in Radiation Protection Procedure RP-14, "Training".
Qualified Individual: An individual who has completed the training and/or testing requirements set forth by procedures or regulations, which in tum grants that individual permission to operate specific equipment or instrumentation, or to perform specific work duties.
Rad: The special unit of radiation dose. One rad is equal to an absorbed dose of 100 ergs/gram or 0.01 joule/kilogram (0.01 gray).
Radiation (Ionizing Radiation): Alpha particles, beta particles, gamma rays, x-rays, neutrons, high-speed electrons, high-speed protons, and other particles capable of producing ions.
Radiation, as used within the context of the Radiation Protection Program does not include non-ionizing radiation such as radio or microwaves and visible, infrared, or ultraviolet light.
Radiation Area: Defined as any accessible area where the dose equivalent to an individual could exceed 5 millirem (.05 mSv) in any one hour at 30 cm from the radiation source or surface that radiation penetrates.
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Section 16.0 Page 16-14 Radiation Safety Officer (RSO): The individual responsible for development and oversight of radiation protection program policies at the Cimarron Site. This individual shall meet the requirements set forth in NUREG-175 7, Section 17 .2. 3 .1.
Radiation Worker: An individual who has access to the Restricted Areas to perform work and has completed the training requirements listed in Radiation Protection Procedure RP-14, "Training".
Radioactive Material (49 CFR 173.403): For purposes of transportation, any material containing radionuclides where both the activity concentration and the total activity in the consignment exceed the values specified in the table in 49 CFR 173.436 or values derived according to the instructions in 49 CFR 173.433.
Radioactive Materials Area: Any area or room which is posted and is used to store or contains for use an amount of licensed material exceeding 10 times the quantity of such material as listed in Appendix C to 10 CFR 20.
Radioactivity: Rate of disintegration (transformation) or decay of radioactive material. The units of activity are the curie (Ci) and the becquerel (Bq). Bq = 1 (dps) disintegration per second; Ci = 3.7 x 10 10 dps.
Radionuclide: Any one of the radioactive nuclides.
Record: A document that provides evidence of the quality of services performed, demonstrates that actions were performed in accordance with radiation protection procedures, or demonstrates conformance of actions to regulatory requirements.
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Section 16.0 Page 16-15 Reference Man: A hypothetical aggregation of human physical and physiological characteristics arrived at by international consensus. These characteristics may be used by researchers and public health workers to standardize results of experiments and to relate biological insult to a common base.
Rem: The special unit for any of the quantities expressed as dose equivalent. The dose equivalent in rems is equal to the absorbed dose in rads multiplied by the quality factor ( 1 rem =
0.01 sievert).
Removable Contamination Survey: The method used to measure removable contamination.
Removable survey techniques are:
(1) Smear Surveys - A smear is obtained by using an absorbent filter disk to wipe with moderate pressure across the area or item to be evaluated. A smear is usually wiped over an area of 100 cm2 *
(2) Wipe Surveys - A wipe is obtained by wiping an absorbent pad or towel over a large area or the entire surface of the item being surveyed.
Respirator: An apparatus used to reduce the individual's intake of airborne radioactive materials.
Restricted Area: An area having access controlled by the Licensee for the purpose of protecting individuals against undue risk from exposure to radiation and radioactive materials.
Restricted Area does not include areas used as residential quarters, but separate rooms in a residential building may be set apart as a Restricted Area.
Sealed Source: Any by-product material that is encased in a capsule designed to prevent leakage or escape of the by-product material.
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environmenta , Radiation Protection Plan properties management. L c Document No. RPP-001 Rev. 4 Effective date:
Section 16.0 Page 16-16 Shallow Dose Equivalent (SDE): The dose equivalent at a tissue depth of 0.007cm (7 mg/cm2),
averaged over an area of one square centimeter. It applies to external exposure of the skin of the whole body or of an extremity.
Sievert (Sv): The SI unit of any of the quantities expressed as dose equivalent. The dose equivalent in sieverts is equal to the absorbed dose in grays multiplied by the quality factor.
1 Sv = 100 rem.
Site Boundary: The line beyond which the land or property is not owned, leased, or otherwise controlled by the Licensee.
Skin of the Whole Body: The skin of the whole body, exclusive of skin of the extremities.
Smear: A radiation survey technique which is used to determine levels of removable surface contamination. A medium (typically filter paper) is rubbed over a surface (typically of area 100 cm2), followed by a quantification of the activity on the medium. Also known as a swipe.
Source Material:
(1) Uranium or thorium or any combination of uranium and thorium in any physical or chemical form; or (2) Ores that contain, by weight, one-twentieth of 1 percent (0.05 percent), or more, of uranium, thorium, or any combination of uranium and thorium. Source material does not include special nuclear material.
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Cimarron Environmental Response Trust envi ronm enta Radiation Protection Plan properties management. LLC Document No. RPP-001 Rev. 4 Effective date:
Section 16.0 Page 16-17 Special Nuclear Material:
(1) Plutonium, uranium-233, uranium enriched in the isotope 233 or in the isotope 235, and any other material that the Commission, pursuant to the provisions of section 51 of the Act, determines to be special nuclear material, but does not include source material; or (2) Any material artificially enriched by any of the foregoing but does not include source material.
Stochastic Effects: Health effects that occur randomly and for which the probability of the effect occurring, rather than its severity, is assumed to be a linear function of dose without threshold. Hereditary effects and cancer are examples of stochastic effects.
Survey: An evaluation of the radiological conditions and potential hazards incident to the production, use, transfer, release, disposal, or presence of radioactive materials or other sources of radiation. When appropriate, such an evaluation includes a physical survey of the location of a source of radiation and measurements or calculations of levels of radiation or concentrations or quantities of radioactive material present.
Thermoluminescent Dosimeter (TLD): An integrating detector where radiation energy is absorbed (trapped) and can be read out later by thermal excitation of the detector.
Total Effective Dose Equivalent (TEDE): The sum of the deep dose equivalent (for external exposures) and the committed effective dose equivalent (for internal exposures).
Total Organ Dose Equivalent (TODE): The sum of the deep dose equivalent and the committed dose equivalent to the organ receiving the highest dose.
Unrestricted Area: Any area to which access is not limited or controlled for purposes of protection of individuals from exposure to radiation and radioactive materials.
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environmenta Radiation Protection Plan properties management. LLC Document No. RPP-001 Rev.4 Effective date:
Section 16.0 Page 16-18 Uptake: Quantity of a radionuclide taken up by the systematic circulation (e.g., by injection into the blood, by absorption from compartments in the respiratory or gastrointestinal tracts, or by absorption through the skin or through wounds in the skin).
Uranium (Natural, Depleted and Enriched):
Natural Uranium: Uranium found in nature. Natural uranium contains 0.71 weight percent U-235, 99.3 weight percent U-238, and a trace of U-234.
Depleted Uranium: Uranium in which the U-235 isotope represents less than 0.71 weight percent of the mass of the material. Depleted uranium is less radioactive than natural uramum.
Enriched Uranium: Uranium in which the U-235 isotope represents greater than 0.71 weight percent of the mass of the material. The alpha emission rate increases from 1.5 E3 dpm per mg at 0.71 weight percent enrichment to 1.4 E5 dpm per mg at 93%
enrichment.
Visitor: An individual who is not an employee or contractor of the Licensee.
Week: Seven consecutive days starting on Sunday.
Weighting Factor (WT): The proportion of risk of stochastic effects resulting from irradiation of the organ or tissue to the total risk of stochastic effects when the whole body is irradiated uniformly.
Whole Body (WB): Means, for purposes of whole body exposure, the head, trunk (including male gonads), arms above the elbow, or legs above the knee.
Year: The period of time beginning on January 1 and ending on December 31 that is used to determine compliance with the NRC.
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Section 16.0 Page 16-19 X-Ray: Penetrating electromagnetic radiation having a wavelength much shorter than that of visible light. X-rays are usually produced by a excitation of the electron field around certain nuclei. In nuclear reactions, it is customary to refer to photons originating in the electron field of the atom as X-rays.
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Document No. RPP-001 Rev.4 Effective date:
Section 17 .0 Page 17-1
17.0 REFERENCES
- 1. 10 CFR 19, "Notices, Instructions and Reports to Workers; Inspection and Investigations"
- 2. 10 CFR 20, "Standards for Protection Against Radiation"
- 3. 10 CFR 30, "Rules of General Applicability to Domestic Licensing of By-Product Material"
- 4. 10 CFR 61, "Licensing Requirements for Land Disposal of Radioactive Waste"
- 5. 10 CFR 70, "Domestic Licensing of Special Nuclear Material"
- 6. "Cimarron Facility Decommissioning Plan," Rev. 1
- 7. EPM0l 7-CALC-001 , "Dose Rate Near Uranium Treatment Train"
- 8. EPM028-CALC-00 1, "Potential Intake Calculation"
- 9. NCRP 87-1987, "Use ofBioassay Procedures for Assessment oflnternal Radionuclide Deposition"
- 10. NUREG/CR-0041, Rev. 1, "Manual of Respiratory Protection Against Airborne Radioactive Material"
- 11. NUREG-1400, "Air Sampling in the Workplace" NUREG 1757, "Decommissioning Process for Materials Licensees"
- 12. NUREG-1507, "Minimum Detectable Concentrations with Typical Radiation Survey Instruments for Various Contaminants and Field Conditions," June 1998
- 13. NUREG-15 56, Vol. 7, "Consolidated Guidance About Materials Licenses, Program-Specific Guidance About Academic, Research and Development, and Other Licenses of Limited Scope Including Gas Chromatographs and X-Ray Fluorescence Analyzers,"
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envi ronmenta, Radiation Protection Plan properties management. LLC Document No. RPP-001 Rev.4 Effective date:
Section 17.0 Page 17-2 Appendix H, "Sample Audit Program," 2018
- 14. Order Transferring License No. SNM-928 for the Cimarron Site
- 15. Regulatory Guide 4.15, Rev. 2, "Quality Assurance for Radiological Monitoring Programs (Normal Operations) - Effluent Streams and the Environment"
- 16. Regulatory Guide 8.4, Rev. 1, "Personnel Monitoring Device - Direct-Reading Pocket Dosimeters"
- 17. Regulatory Guide 8.9, Rev. 1, "Acceptable Concepts, Models, Equations, and Assumptions for a Bioassay Program"
- 18. Regulatory Guide 8.15, Rev. 1, "Acceptable Programs for Respiratory Protection"
- 19. Regulatory Guide 8.25, Rev. 1, "Air Sampling in the Workplace"
- 20. Regulatory Guide 8.28, Rev. 0, "Audible Alarm Dosimeters"
- 21. Regulatory Guide 8.34, Rev. 0, "Monitoring Criteria and Methods to Calculate Occupational Radiation Doses"
- 22. Regulatory Guide 8.36, Rev. 0, "Radiation Dose to the Embryo/Fetus"
- 23. The Cimarron Environmental Response Trust Special Nuclear Material License (SNM-928)
- 24. U.S. NRC, "Guidelines for Decontamination of Facilities and Equipment Prior to Release for Unrestricted Use or Termination of License for Byproduct, Source or Special Nuclear Material," August 1987
- 25. U.S. NRC, "Disposal or Onsite Storage of Thorium or Uranium Wastes from Past Operations," October 1981
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environmental Radiation Protection Plan properties management. LLC Document No. RPP-001 Rev. 4 Effective date:
Appendix A Page A-1 APPENDIX A POTENTIAL INTAKE CALCULATION (EPM028-CALC-001)
CALC NO. EPM028-CALC-001
() ENERCON CALCULATION COVER SHEET REV. 2 Ex.cet!ence-- h *ery p roject. Eve:y doy.
PAGE NO. 1 of 9 Cimarron Environmental Client:
Title:
Potential Intake Calculation Response Trust Project Identifier: EPM028 Item Cover Sheet Items Yes No 1 Does this calculation contain any open assumptions, including preliminary information, that require confirmation? (If YES, identify the assumptions.)
rgJ 2 Does this calculation serve as an "Alternate Calculation"? (If YES, identify the design verified calculation.}
rgJ Design Verified Calculation No.
3 Does this calculation supersede an existing Calculation? (If YES, identify the design verified calculation.)
[8J Superseded Calculation No. EPM028-CALC-001, Rev. 0 Scope of Revision:
Corrects calculation for determining potential mass intake of uranium through inhalation. Potential intake of Tc-99 through inhalation and ingestion were evaluated. A sensitivity analysis to determine potential intake for thorium and protactium daughters of U-235 and U-238 was peformed. Several editorial changes were made to correct page numbering, equation numbering, and improve clarity of the text.
Revision Impact on Results:
Conclusions related to chemical intakes of uranium do not change. Potential intakes of Tc-99 and uranium daughters are significantly less than 1% ALI.
Study Calculation Final Calculation ~
Safety-Related Non-Safety-Related ~
(Print Name and Sign)
Originator: Jay Maisler Jay Maisler, CHP Digitally signed by Jay Maisler, CHP Date: 12/12/2019 DJte: 2019)t .12 09:53:04_-05'00' A (\
Design Reviewer: A. Joseph Nardi j4 0?~1-~1~ }/~-- I v
~
Date:
11/l~ r1 Approver: Gerry Williams 'r,/r/in r
W~lr I lrlm<t
/ Digitally signed by Gerald Williams Date:
T '
V Date: 20 19.12.17 09:58:06 -05'00'
CALC NO. EPM028-CALC-001 ENERCON CALCULATION REV. 2 Excellence--Every project. Every day REVISION STATUS SHEET PAGE NO. 2 of 9 CALCULATION REVISION STATUS REVISION DATE DESCRIPTION 0 3/26/2019 Initial Issue 1 6/20/2019 Corrected mass intake from inhalation and made editorial corrections 2 11/22/2019 Added calculation of potential intake for Tc-99 and sensitivity analysis for potential intakes of U-235 and U-238 progeny with corresponding Attachments.
PAGE REVISION STATUS PAGE NO. REVISION PAGE NO. REVISION 1-13 2 APPENDIX/ATTACHMENT REVISION STATUS APPENDIX NO. NO.OF REVISION ATTACHMENT NO.OF REVISION PAGES NO. NO. PAGES NO.
Attachment A 2 2 Attachment B 2 2 Attachment C 2 2
CALCULATION TABLE OF CONTENTS ENERCON TABLE OF CALC NO. EPM028-CALC-001 Excellence-- Every project. Every day. CONTENTS REV. 2 Section Page No.
1.0 Purpose and Scope
4 2.0 Summary of Results and Conclusions 4
- 3. 0 References 4 4.0 Assumptions 4 5.0 Design Inputs 5 6.0 Methodology 6 7.0 Calculations 7 8.0 Computer Software 9 List of Attachments # of pages Attachment A - Potential Intake Calculation Spreadsheet - Uranium 2 Attachment B - Potential Intake Calculation Spreadsheet - Tc-99 2 Attachment C - Sensitivity Analysis 2
CALC NO. EPM028-CALC-001 ENERCON Potential Intake Calculation REV. 2 Excellence-Every pro;ect. Every day.
PAGE NO. 4 of 13
1.0 Purpose and Scope
The purpose of this calculation is to estimate potential intakes of uranium from groundwater processing and resin handling at the Cimarron Site. This calculation also provides an estimated intake of handling spent resin and bio-solids potentially contaminated with Tc-99. The methodology for potential intake and need for air sampling is based on the methodology provided in NUREG-1400 (Ref. 3.1).
As sensitivity analysis was performed to determine the contribution to potential intakes from airborne exposure to thorium and protactinium progeny of U-235 and U-238.
2.0 Summary of Results and Conclusions The highest anticipated potential intake from handling spent resin is less than 0.2% of the annual limit on intake (ALI) for uranium, U-235 progeny, U-238 progeny. The potential intake from handling spent resin and bio-solids is 0.22% of the ALI for Tc-99.
3.0 References 3.1 NUREG-1400, Air Sampling in the Workplace, September 1993.
3.2 49 CFR 173.434, Activity-mass relationships for uranium and natural thorium.
3.3 10 CFR 20, Appendix B, Annual Limits on Intake (ALis) and Derived Air Concentrations (DACs) of Radionuclides for Occupational Exposure; Effluent Concentrations; Concentrations for Release to Sewerage 3.4 Cimarron Radiation Protection Plan, draft Rev. 4 3.5 BA 1 Isotopic Data for Enercon.xlsx, Isotopic abundance of Uranium source provided as an Excel spreadsheet 3.6 Cimarron Facility Decommissioning Plan, Rev. 1 3.7 DOE-STD-1036-2009, Guide to Good Practices for Occupational Radiological Protection in Uranium Facilities, July 2009 4.0 Assumptions 4.1 The specific activity for uranium:
4.1.1 5% enrichment - 2.7 E-06 Ci/g (Ref. 3.2) 4.1.2 1.5% enrichment - 1.00 E-06 Ci/g (Ref. 3.2) 4.2 The maximum groundwater uranium concentration is assumed to be 5000 pCi/L.
This is conservative based on the highest reported groundwater concentration found in BA1 of less than 3000 µg/L. (Ref. 3.6, Figure 3-4)
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PAGE NO. 5 of 13 4.3 The average groundwater Tc-99 concentration is assumed to be 586 pCi/L. This is based on information obtained during groundwater samples analyzed in 2018 and 2019.
4.4 The maximum mass of U-235 assumed to be present in the spent resin being processed for disposal is 500 grams. This is higher than the maximum mass estimated in the treatability study. (Ref. 3.6, section 8.3.2) 4.5 The uranium in groundwater is assumed to be soluble.
4.6 Determining "potential intake" for uranium, based on the NUREG-1400 methodology, requires various assumptions. These assumptions are used in the sensitivity analysis:
4.6.1 Total activity processed is based on twelve spent resin bed vessels processed per year.
4.6.2 "Release Fraction," R, is based on "non-volatile powder" - 0.01.
4.6.3 "Confinement Factor," C, is equivalent to a well-vented hood - 0.1. This is based on equipment designed to confine resin within the process system.
4.6.4 "Dispersibility," D, is based on the contaminant being moist resin - 0.1 4.7 Determining "potential intake" for Tc-99, based on the NUREG-1400 methodology, requires various assumptions:
4.7.1 Total activity processed is based on groundwater flow of 130 gallons per minute processed for cleanup.
4.7.2 Total volume of groundwater processed in a year assumes continuous processing for a year - 365 days, 24 hours2.777778e-4 days <br />0.00667 hours <br />3.968254e-5 weeks <br />9.132e-6 months <br /> per day.
4.7.3 No credit for "Release Fraction," R, is taken - 1.
4.7.4 No credit for "Confinement Factor," C, is taken - 1.
4.7.5 "Dispersibility," D, is based on the contaminant being moist resin and bio-solids - 0.1 4.8 The ALI and DAC used for inhalation are for U-238, class Y. U-234, U-235, and U-238 have the same value for these ALis and DACs, therefore isotopic distribution is irrelevant to the dosimeteric calculation.
4.9 The solubility class for the Tc-99 ALI is W, the most restrictive inhalation ALI.
CALC NO. EPM028-CALC-001 ENERCON Potential Intake Calculation REV. 2 Excellence-Eve ry project. Every doy.
PAGE NO. 6 of 13 4.10 Isotopic distribution for U-235 progeny and U-238 progeny in the sensitivity analysis only consider isotopes in the decay chain that are in secular equilibrium with their uranium parents; Th-231, Th-234, Pa-234m.
5.0 Design Inputs 5.1 Annual Limit on Intake (ALI) for Inhalation The following properties are taken from Reference 3.3.
U-234, U-235, U-238: 4.00 E-02 µCi (Class Y)
Tc-99: 7.00 E+02 µCi Th-231: 6.00 E+03 µCi Th-234: 2.00 E+02 µCi Pa-234m: 7.00 E+03 µCi (Class Y) 5.2 Derived Air Concentration (DAC) for Inhalation (Uranium)
The following source characteristics are taken from Reference 3.3.
2.00 E-11 µCi/ml 5.3 ALI for Oral Ingestion The following material definitions are taken from Reference 3.3.
1.00 E01 µCi (class D, bone surface) (Uranium) 2.00 E01 µCi (class D) (Uranium) 4.00 E+03 µCi (class D) (Tc-99) 5.4 Resin Vessel Loading The first week's loading of the resin vessel is 3,706 grams of uranium. After the first full loading , the vessel contains 24,055 grams of uranium. These values are taken from Reference 3.5.
CALC NO. EPM028-CALC-001 ENERCON Potential Intake Calculation REV. 2 Excellence-Ev;:ry pro;ect. l:very doy.
PAGE NO. 7 of 13 6.0 Methodology 6.1 Oral Intake Consumption The amount of untreated groundwater that would need to be consumed for an individual to have an intake of 2% ALI and 1 ALI was calculated based on Assumption 4.2.
6.2 Spent Resin Loading The total amount of uranium that could be loaded onto a spent resin bed was calculated for 1.5% and 5% enriched uranium. These calculations provided for conservatism in the dose assessment performed.
6.3 NUREG-1400 Methodology NUREG-1400 provides calculational methods to support decisions related to the need to perform air sampling at a facility and determine the potential intake for workers.
The first calculation involves determining if the amount of unsealed radioactive materials handled in a year would indicate the need for performing air sampling.
Following this methodology, the amount uranium on a spent resin bed was determined. Then, as discussed in Assumption 4.6.1, the total amount of uranium handled in a year was determined. Other assumptions (4.6.2, 4.6.3, 4.6.4) were used to calculate "potential intake" from inhalation.
The total activity of Tc-99 handled in a year is determined as discussed in Assumption 4.7. Other assumptions (4.7.3, 4.7.4, 4.7.5) were used to calculate "potential intake" from inhalation.
6.4 Chemical Intake Intake of soluble uranium is limited to 10 mg per week as required by 10 CFR 20.1201 (e). Using the "potential intake" from inhalation, the mass that an individual could intake in a year was calculated. Because the limit is based on a weekly limit, consumption of contaminated water would be more limiting than inhalation. Based on Assumption 4.2, the amount of contaminated water that would need to be consumed during a week to ingest 10 mg of soluble uranium was calculated.
6.5 Sensitivity Analysis The sensitivity analysis considers the potential intake from inhalation of U-234, U-235 and its progeny (Th-231), and U-238 and its progeny (Th-234, Pa-234m). The
CALC NO. EPM028-CALC-001 ENERCON Potential Intake Calculation REV. 2 Excellence-Eve ry pro;ecl. Every day.
PAGE NO. 8 of 13 progeny for U-235 and U-238 are considered to be in secular equilibrium with their respective parent. Acvity distribution percentages is based on 5% enrichment, which provides an upper bound for dosimetric considerations.
7.0 Calculation 7 .1 Oral Intake Consumption The oral intake consumption is found using the Equation 7 .1.
/ 0 = -C- X 1£06 max Equation 7. 1 Where lo, in liters, is equal to annual limit on intake (ALI) µCi for U-238 or Tc-99 divided by the activity concentration (pCi/L) in groundwater multiplied by a 106 pCi/µCi.
The results from this calculation for uranium indicate that drinking 40 L of contaminated groundwater at maximum activity would result in an individual intake of 2% ALI; 2000 L would need to be consumed to have an intake of 1 ALI. Consuming groundwater at the site is prohibited.
The results of this calculation for Tc-99 indicate that drinking 1.37 E+05 L of contaminated groundwater would result in an individual intake of 2% ALI; 6.87 E+06 L would need to be consumed to have an intake of 1 ALI. The 2% ALI scenario would require consumption of 377 L per day for an entire year.
7.2 Spent Resin Loading The mass of total uranium in a fully spent resin bed is found using Equation 7.2.
M = 500 g E
Equation 7. 2 Where the total mass of uranium in spent resin vessel, M, equals the mass of U-235 on the spent resin bed (500 grams) divided by the enrichment.
Based on 1.5% enrichment, the uranium mass in one spent resin bed would be 33.3 kg.
CALC NO. EPM028-CALC-001 ENERCON Potential Intake Calculation REV. 2 Excellence-Every pm;ect. Every doy PAGE NO. 9 of 13 7.3 NUREG-1400 Methodology 7.3.1 Potential Intake for Uranium The total activity in spent resin processed in a year is based on four resin bed exchanges per year. The total activity handled during a year is calculated with Equation 7.3.
Q == SAenrich X M X 4 Equation 7.3 Where Q, the total activity handled in a year (Ci) is equal to the specific activity based on enrichment, SAenrich, Ci/g divided by the total mass of uranium on a fully spent resin bed from Equation 7 .2. The activity Q is compared to the ALI for uranium to determine if it exceeds 104 times the ALI. If it does, then air sampling should be considered.
The total activity assuming 1.5% enrichment would exceed 104 times the ALI. Air sampling should be considered and is addressed in the RPP (Ref. 3.4)
Potential intake from inhalation is determined from Equation 7.4 (Ref. 3.1, Equation 1.2).
Ip =QX R X C X D X 10- 6 Equation 7.4 Where Ip is the potential intake from inhalation in µCi, Q is the total activity from Equation 7.3 converted to µCi, R is the release fraction (0.01 for non-volatile powders),
C is the confinement factor (0.1 for well-ventilated hood), and D is dispersibility (0.1 for o-moist resin). 1 6 is an additional factor provided in Ref. 3.1 The potential intake from inhalation for 5% enrichment was more limiting than 1.5%
enrichment. The results for 5% enriched uranium was 2.59 E-05 µCi in a year, which is 0.2% ALI. At 1.5% enrichment, the potential intake was calculated to be 4.00 E-05 µCi in a year or 0.1 % ALI.
7.3.2 Potential Intake for Tc-99
- The total activity for Tc-99 handled in a year is calculated by multiplying the Tc-99 groundwater concentration (Assumption 4.3) times the amount of groundwater processed in a year (Assumption 4.7.1 times Assumption 4.7.2) .
Potential intake from inhalation is determined from Equation 7.4.
CALC NO. EPM028-CALC-001
]ENERCON Potential Intake Calculation REV. 2 Excellence-Every project. Every day.
PAGE NO. 10 of 13 Where Ip is the potential intake from inhalation in µCi, Q is the total activity as discussed converted to µCi, R is the release fraction (no credit is taken, a value of 1 is assigned),
C is the confinement factor (no credit is taken, a value of 1 is assigned), and Dis o-dispersibility (0.1 for moist resin and bio-solids). 1 6 is an additional factor provided in Ref. 3.1 The results for Tc-99 was 1.51 E-02 µCi in a year, which is 0.22% ALI.
7.4 Chemical Intake The amount of groundwater that would need to be consumed to have an intake of 10 mg of soluble uranium was determined using Equation 7.5.
10mg 3 V - 3000 µg/L x lO Equation 7. 5 Where V is the volume consumed in liters. 10 mg is the weekly soluble uranium intake limit (10 CFR 20.1201 (e)). 3000 µg/L is the maximum mass concentration of uranium in groundwater at BA 1.
The calculation resulted in a weekly intake exceeding 10 mg soluble uranium if an individual consumed 3.33 L of contaminated water. Consumption of groundwater at the Cimarron site is prohibited.
Chemical intake from inhalation was also calculated and compared to the potential intake from inhalation. Based on the results of the potential intake from inhalation, the potential mass intake would be 4E-2 mg in a year. The permissible weekly intake is 250 times the potential mass intake through inhalation in a year.
7.5 Sensitivity Analysis The sensitivity analysis was performed assuming that:
- Th-234 and Pa-234/Pa-234m are in secular equilibrium with U-238 (i.e., the activity of U-238 is equal to each of these progeny).
Total uranium activity was determined for the potential intake calculation. Isotopic activity was determined by calculating the activity fraction for each of the uranium isotopes (U-234, U-235, and U-238) and multiplying by the total uranium activity.
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PAGE NO. 11 of 13 Activity fractions were calculated using the equations provided in Example 3a of Ref.
3.7:
Enrichment
- SA 4 AF234 = - - - - - - - - - - - - - -234 - - - - - - - - - - -
Enrichment234
- SA 234 + Enrichment 235
- SA 235 + Enrichment 238
- SA 238 Equation 7. 6 Enrichment
- SA23s AF23s = - - - - - - - - - - - - - -235 -------------
Enrichment234
- SA 234 + Enrichment 235
- SA 235 + Enrichment 238
- SA 238 Equation 7. 7 AF23s = - - - - - - - - - -Enrichment - - - -238 - *-SA2 s
Enrichment234
- SA 234 + Enrichment 235
- SA 235 + Enrichment 238
- SA 238 Equation 7. 8 Where AF234 is the activity fraction for U-234; AF23s is the activity fraction for U-235; AF23a is the activity fraction for U-238. Enrichment is by percent weight. SA234 is the specific activity of U-234 (2.30 E+08 Bq/g); SA23s is the specific activity of U-235 (79312 Bq/g); SA23a is the specific activity of U-238 (12329 Bq/g).
Enrichment for U-234, based on 5% U-235 enrichment, is calculated by the following equation provided in Example 3a of Ref. 3. 7:
0.0055 natural fraction 234 U Enrichment 234 = - - ~
- 5.0
- 1.2
- 0. 72_natural_fraction_ 2 5 U Equation 7. 9 Enrichment 238 = 100 - Enrichment 234 - 5.0 Equation 7. 10 The NUREG-1400 methodology discussed in Section 7.3 was applied for each of the uranium istopes and the U-235 and U-238 progeny. The results of the analysis for each of the isotopes is provided below, as a percentage of ALI: