ML22256A224

From kanterella
Jump to navigation Jump to search
NRC Staff Training on RG 1.21 and RG 8.34 - Public Version
ML22256A224
Person / Time
Issue date: 07/07/2022
From: Steven Garry
NRC/NRR/DRA/ARCB
To:
Garry S
References
Download: ML22256A224 (132)
v • d • e


Text

NRC Training RG 1.21 and RG 8.34 Revisions Steve Garry, CHP Sr. Health Physicist Radiation Protection and Consequence Branch Division of Risk Assessment Office of Nuclear Reactor Regulation July 7, 2022 1

RG 1.21 and RG 8.34 Revisions Session 1 RG 1.21, Rev. 3, Measuring, Evaluating, and Reporting Radioactive Material in Liquid and Gaseous Effluents and Solid Waste Session 2 RG 8.34, Rev. 1, Monitoring Criteria and Methods to Calculate Occupational Radiation Doses 2

Session 1 - RG 1.21, Rev. 3 3

RG 1.21 ~ Measuring Effluents

  • when to update long-term, annual average /Q and D/Q values
  • environmental monitoring for iodine-131 in drinking water
  • ODCM - making changes to effluent and environmental programs
  • calibration of accident-range radiation monitors 4

RG 1.21

/Q and D/Q values

  • Long-term annual-average /Q and D/Q should be based on 5 or more years of meteorological data
  • /Q and D/Q values should be reevaluated periodically (e.g., every 3-5 years).
  • If /Q and D/Q values are substantially nonconservative (e.g., higher by 20-30 percent or more), revise /Q and D/Q values used in dose assessment 5

RG 1.21 Environmental monitoring I -131 in drinking water

  • Some licensees believed they were required to perform unnecessary and expensive I-131 analyses in drinking water
  • Licensees have brought concerns to NRCs attention at REEW
  • HQ investigated the basis for NUREG-1301/1302 guidance
  • Some licensees have been misinterpreting NUREG-1301 guidance on the need for sampling for I-131 in drinking water
  • HQ determined that this guidance was applicable to setting up an environmental monitoring program
  • Licensees should perform an evaluation of the likelihood of doses exceeding 1 mrem/yr (max organ and max age group) per NUREG-1301/1302 6

NUREG-1301/1302 I -131 in drinking water 7

NUREG-1301/1302 - Environmental monitoring for I -131 in drinking water 8

NUREG-1301/1302 - Environmental monitoring for I -131 in drinking water

  • In order to get to 1 pCi/L LLD, labs need to do a resin column separation of iodine from water
  • A resin column separation is an expensive analysis
  • In order to get to 15 pCi/L LLD, labs can do a simple gamma scan 9

RG 1.21 I-131 in Drinking Water 10

RG 1.21 Environmental monitoring I -131 in drinking water

  • Licensees should perform a prospective dose evaluation to determine if the likely dose from I-131 in drinking water is

> than 1 mrem/yr

  • If likely dose is > 1 mrem/yr, perform I-131 sampling &

analysis using an LLD of 1 pCi/L (resin column separation and gamma scan)

  • If likely dose is < 1 mrem/yr, perform I-131 sampling and analysis using an LLD of 15 pCi/L (perform a gamma scan of drinking water) 11

RG 1.21 - Revising ODCMs

  • ODCMs need to be kept current
  • Plant changes affect ODCMs:

- Plant operating status (operating or decommissioning)

- Failed fuel or change to decommissioning status - noble gas and iodine have been eliminated

- Installation of new or removal of out-of-service radwaste processing equipment 12

10 CFR 50.59 Changes, Tests, and Experiments

- Exception (4) below 13

Tech Specs specify how ODCMs are to be revised

  • A specific change mechanism is specified in Technical Specs on how to make changes to the ODCM 14

Tech Specs specify how ODCMs are to be revised (Contd) 15

10 CFR 50.34a - Carbon-14 (C-14)

  • There is no waste processing equipment for removal of C-14 from gaseous effluents 16

10 CFR 50, Appendix I -

C-14 is not included as an organ dose criteria 17

Tech Spec requirements EPA 40 CFR 190.10(a) (excerpt below) 18

RG 1.21 C-14 Dose Assessments

  • C-14 is likely a principal radionuclide in operating reactor effluents
  • C-14 is likely the most significant organ dose radionuclide
  • Appendix I limits organ dose to radionuclides in iodine or particulate form (not C-14)
  • Note: Tech specs add tritium in the organ dose limit 19

RG 1.21 C-14 Source Term

- Scaling factors can be based on power generation, or

- NUREG-0016 GALE computer codes, or

- NCRP-81s C-14 report, or

- EPRI Report No. 1021106

- PWRs release ~ 6 curies, (~ 4.6 mrem max dose)

- BWRs release ~ 9 curies, (~4.7 mrem max dose) 20

RG 1.21 C-14 Dose Assessments

  • C-14 dose is primarily from eating vegetables from local gardens
  • Local gardens are identified in the land use census

- Licensees have problems with communities growing up around nuclear plants

- Licensees have problems locating/moving air samplers

  • Vegetables are grown year-round in the southern USA, but less than ~ 6 months in the northern USA
  • Vegetables absorb CO2 during photosynthesis (daylight hours) 21

RG 1.21 Compliance with EPA dose limits

  • EPA dose limits required by 10 CFR 20.1301(e) are 25 mrem/yr whole body and any organ (except thyroid)
  • C-14 is dominant source of organ dose (bone)

- maximum organ doses in 2020 (incl. C-14) were:

  • Assume dose from direct radiation is < 10 mrem/yr

- Total organ dose = effluent dose plus direct radiation

  • Total dose = 4.7 mrem + <10 mrem = <14.70 mrem
  • < 14.70 mrem is less than 25.00 mrem EPA 40 CFR 190 22

RIS-2008-03 Return/Reuse of Previously Discharged Radioactive Effluents

  • RIS 2008-03 states radioactive material properly released in gaseous or liquid effluent is not considered licensed material (if less than exempt concentrations)
  • The unlicensed material can be used and returned to the environment without being considered a new radioactive material effluent release
  • Licensees are responsible for evaluating any new on-site or off-site exposure pathways created by returned/recycled material that exceed 10% of total dose (per RG 1.109 guidance) 23

RG 1.21 Reporting Abnormal Discharges (leaks and spills)

  • RG 1.21, Section 9.5 Supplemental Information

- Discusses guidance on abnormal releases from plant equipment into onsite groundwater

- Reporting thresholds include:

  • Abnormal discharges to the unrestricted area

- Information should include:

  • Date, duration, volume, etc.
  • Doses to public 24

RG 1.21 Groundwater reference to RG 4.25 25

RG 4.25 Groundwater Discharges (GW)

  • RG 4.25 - guidance on calculating discharges from on-site groundwater to off-site groundwater
  • RG 4.25 is not a calculation of releases into on-site groundwater 26

RG 4.25 - On-site Discharges into On-site Ponds

  • Some licensees dispose of liquid effluents to on-site ponds
  • Releases are reported in ARERR as though releases were to the unrestricted area (10 CFR 50.36a)
  • Most on-site ponds leak into on-site groundwater
  • There is an important footnote in RG 4.25 that provides an exclusion for reporting leakage from on-site ponds into GW

- Leakage from bottom of lake or pond to groundwater does not need to be reported (again)

- However, the potential dose (through a new groundwater pathway) must be assessed if the off-site dose from leakage from onsite ponds is greater than 10% of all pathways combined (see RG 1.109) 27

List of Leaks and Spills (L&S)

  • 55 currently operating sites (2020 data)
  • 38 sites historically have had L&S of H-3 20,000 pCi/L reported
  • 7 sites currently have residual radioactive ground water with H-3 20,000 pCi/L 28

Remediation of Leaks and Spills

  • SRM-SECY-13-108, Remediation of Residual Radioactivity During Operations
  • Evaluate feasibility of prompt remediation
  • NRC Commission determined that prompt remediation is not a requirement 29

RG 4.13 Environmental Dosimetry &

Direct Radiation Dose Assessment

  • RG 4.13, Rev. 2 was revised in June 2019
  • NRC endorsed ANSI/HPS N13.37, ~ Environmental Dosimetry
  • RG 4.13 provides an NRC-approved method of determining facility-related dose (FRD) from direct radiation
  • RG 4.13 methods can be used in the demonstration of compliance with 10 CFR 20.1302 surveys requirements and EPA 40 CFR 190s dose limit of 25 mrem/yr 30

RG 4.13 - Data Analysis Method for Direct Radiation Using Environmental Dosimetry

  • At each location, using historical data, determine the baseline background dose rate and its standard deviation ()
  • Then, perform a 2-step quarterly data analysis process:

- At each location, determine if there is there a detectable increase greater than 3 above the baseline dose rate? (a yes/no question)

- If > 3, determine the facility-related dose (FRD)

  • Subtract current quarterly reading from baseline background dose rate
  • Note: Do not subtract the 3 value
  • Environmental dosimetry systems can measure FRD dose at:

~ 5 mrem/qtr, and ~ 10 mrem/yr 31

Decomm Planning Rule (DPR) 76 FR (2012) pp. 35512 - see NRC website at https://www.nrc.gov/reading-rm/doc-collections/fedreg/notices/

  • The DPR was first revised in 2007 for applicants to minimize contamination
  • The DPR was revised in 2012 for plant operations to minimize contamination [10 CFR 20.1406(c)]

- Licensees shall minimize residual radioactivity (contamination),

including subsurface (ground water) 32

10 CFR 20.1501 Radiological Surveys and Monitoring

  • § 20.1501 was revised (during the DPR - 2012) to require surveys of the subsurface (i.e., soil and ground water)
  • NEI groundwater guidance documents

- NEI 07-07 (~ Ground Water Protection Initiative (GPI))

- NEI 08-08 (~ FSAR template for minimizing contamination), and

- NEI 09-14 (~ Underground Pipes) are used as guidance for the ground water monitoring program defining:

  • how to minimize (prevent) leaks into ground water
  • how to survey subsurface (ground water) 33

Decommissioning Programs

  • Groundwater monitoring may need to be increased in support of license termination
  • Licensees must maintain and update 10 CFR 50.75(g) record keeping files to include leaks and spills
  • Decommissioning-related RGs

- RG 4.22, Decommissioning Planning During Operations

- RG 1.185, Standard Format and Content for Post-Shutdown Decommissioning Activities Report

- NUREG-1757, Rev. 1 (2006) and draft Rev. 2 (2020) Consolidated Decommissioning Guidance 34

RG 1.21 Accident-Range Gaseous Effluent Monitoring

  • RG 1.21 summarizes previously issued NRC requirements and guidance in:

- NUREG-0660 ~ TMI Action Plan

- NUREG-0737, Clarification of TMI Action Plan Requirements (ML051400209)

- HPPOS-001, Guidance on Calibration and Surveillance to meet Item II.F.1, Additional Accident-Monitoring Instrumentation 35

Accident-Range Gaseous Effluent Monitoring

  • Item II.F.1 is Additional Accident-Monitoring Instrumentation, requiring:

- Noble gas effluent monitoring (Item II.F.1-1)

- Iodine and particulate sampling and analysis (Item II.F.1-2)

- Containment high range radiation monitoring (II.F.1-3)

  • Specifications for radiation monitoring equipment are in Tables II.F.1-1, II.F.1-2, and II.F.1-3 36

Accident-Range Radiation Monitors

  • Three different instrument criteria to discuss:

- Instrument design criteria

- Instrument calibration criteria

- Instrument measurement criteria 37

Three different criteria:

  • Design criteria:

- RG 1.97 establishes a factor of 2 for design criteria, is not a calibration criteria

  • Calibration criteria:

- NUREG-0737 - sufficient to perform intended function

- ANSI N320-1979 and IEEE-497 - generally +/- 40% - +/- 50%

  • Measurement criteria:

- Effluent monitors should be able to measure fresh noble gas mixtures (0 - 10 days) within overall system accuracy factor of 2

- CHRMs should be able to measure within factor of 2 38

RG 1.97 footnotes Design Criteria for Effluent Monitors

  • BWRs, RG 1.97, Rev. 2, Table 1, footnote 9 10-day gas mixtures, overall system accuracy within a factor of 2
  • BWRs, RG 1.97, Rev. 3, Table 2, footnote 9 10-day gas mixtures, overall system accuracy within a factor of 2
  • PWRs, RG 1.97, Rev. 2, Table 2, footnote 8 10-day gas mixtures, overall system accuracy within a factor of 2
  • PWRs, RG 1.97, Rev. 3, Table 3, footnote 9 10-day gas mixtures, overall system accuracy within a factor of 2 39

RG 1.97, Accident Instrumentation Design Criteria for CHRMs

- 60 keV - 100 keV within factor of 2

- 100 keV - 3 MeV within +/-20%

- 60 keV - 100 keV within factor of 2

- 100 keV - 3 MeV within +/-20%

- 60 keV - 100 keV within factor of 2

- 100 keV - 3 MeV within +/-20%

- 60 keV - 3 MeV within factor of 2 40

Calibration of Accident-Range Radiation Monitoring Equipment

  • Actual guidance is in a letter from NRR to NRC Regional Administrators (ML103420044) 41

NUREG-0737 Item II.F.1-1 Noble Gas Effluent Monitoring

  • GM detector, scintillator or CdTe(Cl) detector output is in cpm or mR/hr
  • Manufacturer provides energy response characterization from low to high gamma energy (~81 keV to 3 MeV)
  • Manufacturer provides instrument response factor (efficiency factors) for Xe-133 (and Kr-85 for scintillators and CdTe(Cl) detectors)
  • Licensees perform periodic calibration checks with a solid source 42

NUREG-0737 Item II.F.1-2 Iodine and Particulate Monitoring

  • Real-time monitoring is not practical
  • Licensees must develop procedures for collection and analysis of samples
  • Iodine releases can be calculated based on partitioning (scaling) factors to noble gas 43

NUREG-0737 Item II.F.1-3 Containment High Range Monitor (CHRMs) 44

Instrument Calibration Process 45

Effluent Monitors Initial Vendor Calibrations

  • Vendors perform initial calibrations:

- perform dose-rate linearity check to high dose rates

- determine the detectors energy response characteristics

- determine the efficiency factor (instrument response factor) (cpm per µCi/cc or mR per hour per µCi/cc ) for a standard gas (Xe-133 or Kr-85)

- build a field calibrator with a Cs-137 source for licensees use for in-plant calibration checks

- determine and provide a Transfer Factor

(µCi/cc) / (cpm) or (µCi/cc) / (mR/hr) 46

In-plant calibration checks

  • Instrument and Control (I&C) technicians do a one-point radiological calibration check in the first scale/decade
  • I&C do electrical calibration checks for higher scales/decades
  • HP normally only provides radiological support (RWP, pre-job briefings, and job coverage)
  • Instrument response factors (efficiency factors) are normally NOT adjusted during calibration checks 47

Instrument Response Factors (Efficiency Factors)

  • Noble gas effluent monitoring instruments are GM detectors, plastic scintillators, and CdTe(Cl) solid-state detectors, typically ~2 mm x 2 mm x 5 mm size
  • Each solid-state detector has its own counting efficiency
  • GM and ion chambers are typically calibrated to Xe-133; i.e., to low energy, 81 keV photons with low yield (~36%)
  • Plastic scintillators and solid-state detectors are calibrated to Xe-133 (gamma) and Kr-85 (beta) 48

Instrument Response Factors (Efficiency Factors)

  • Detector output is a count rate or a dose rate
  • Output is converted to a Xe-133 concentration, µCi/cc
  • Concentration (µCi/sec) times flow rate (cubic feet per sec)
  • µCi/cc x flow rate = release rate (µCi/sec) of Xe-133 49

Potential Errors in Use of Efficiency Factors (instrument response factors)

  • Licensees may be using wrong calibration geometry
  • Licensees may be using wrong efficiency factors; i.e.,

incorrectly:

- assume 1 efficiency factor fits all detectors

- replace detectors and do not update efficiency factors (particularly General Atomics CdTe(Cl) solid state detectors

- apply Xe-133 efficiency factor to the radionuclide mix 50

Radionuclide Mix

  • Gaseous effluent is not just Xe-133
  • Gaseous effluent is a mix of noble gases, and is very time dependent
  • Generally, short-lived noble gas nuclides have higher energy gammas than long-lived nuclides
  • Efficiency factors are 10 - 30 times than Xe-133 for higher energy gammas
  • A time-dependent efficiency factor (instrument response factor) is needed 51

Accident Source Term: 13 Noble Gases There are ~ 60 different gamma energies and gamma yields from 13 noble gas nuclides to consider 6 Kryptons 7 Xenons

  • 1. Kr-83m 7. Xe-131m
  • 2. Kr-85m 8. Xe-133m
  • 4. Kr-87 10. Xe-135m
  • 5. Kr-88 11. Xe-135
  • 6. Kr-89 12. Xe-137
13. Xe-138 52

~ 60 Gamma Energies 53

GM Detector Instrument Response Factors (based on calibration to Xe-133)

Relative Response Over-response compared to Xe-133 36.0 31.0 Core Melt 26.0 21.0 16.0 11.0 6.0 Gas Gap 1.0 0.1 1 2 4 8 12 24 48 168 720 54

Plant staff responsibilities

  • Plant staff:

- Some plants may not have rad-engineering expertise

- I&C may do calibration checks without knowledge of radiological response characteristics

  • Plant staff should know:

- what equipment is installed

- which department is in charge

- how equipment works (vendor manuals and calibrations)

- how calibration checks are performed

- the basis for efficiency factors and detector specific factors

- how monitor output interfaces with dose assessment codes 55

2018 REEW Presentation - Effluent Monitoring Accident-Range Gaseous Effluent Monitoring Calibration and Time-Dependent Instrument Response Factors ADAMS ML18171A035 Steve Garry, CHP Sr. Health Physicist Office of Nuclear Reactor Regulation U.S. Nuclear Regulatory Commission Radiological Effluents and Environmental Workshop June 27, 2018 New Orleans, LA 56

Iodine and Particulate (I&P) Monitoring

  • Real-time iodine and particulate monitoring is not required
  • However, licensees should have procedures for sample collection and analysis of hot iodine and particulate samples
  • Real-time dose assessment can be performed using scaling factors to noble gas 57

Containment High Range Monitors (CHRM)

  • CHRM measurements are used in Emergency Action Levels (EALs) and for core damage assessment
  • Core damage assessment methods are in NUREG-1940, Radiological Assessment System for Consequence Analysis (RASCAL) section 1.2.8 and NUREG-1940, Supplement 1, Section 2.6
  • Licensee staff perform a one-point radiological calibration check below 10 R/hr
  • Licensee staff perform an electronic calibration check for each decade above 10 R/hr 58

NRC staff training CHRMs

  • NRC developed, provided, and recorded training on CHRMs calibration in 2021
  • Training slides are publicly available at ML21327A271 59

Session 2 - RG 8.34, Rev. 1 60

RG 8.34 Calculating Occupational Dose

  • Reasons for revision:
  • to revise the definition of the total effective dose equivalent (TEDE) as the sum of the EDEX and the committed effective dose equivalent (CEDE)
  • to provide guidance on performing prospective dose evaluations to determine the need for required monitoring to meet the occupational dose monitoring requirements of 10 CFR 20.1502
  • to provide guidance on monitoring of unplanned, unintended doses when monitoring was not performed 61

RG 8.34 (Contd)

Calculating Occupational Dose

  • Reasons for Revision (continued):
  • to provide guidance on monitoring dose from hot particles or contamination on or near the skin
  • to define the term dosimetry processing and explain when there are requirements for processing by an accredited National Voluntary Laboratory Accreditation Program (NVLAP) processor
  • to provide guidance on assessing dose from intakes of radioactive material by wound injuries
  • to provide guidance on calculating soluble uranium intakes 62

Background

DDE vs EDEX

  • In 2007, 10 CFR 20 was revised to define TEDE as equal to EDE (for external dose) plus the CEDE
  • The term EDEX was coined as an acronym for EDE (for external exposure)
  • 10 CFR 20.1201 occupational dose requires NRC approval for methods of calculating EDEX (other than using DDE) 63

10 CFR 20.1201(c)

Occupational dose limits for adults

  • When the external exposure is determined by measurement with an external personal monitoring device

- the assigned deep dose equivalent must be for the part of the body receiving the highest exposure

- the deep dose equivalent must be used in place of the EDEX, unless the EDEX is determined by a dosimetry method approved by the NRC 64

RG 8.38 ~ HRAs and LHRAs 65

RG 8.38 HRAs and LHRAs

  • Accessibility is determined by whether an individual can reasonably occupy the area with a major portion of their whole body
  • An area into which an individual can only insert an extremity, or a portion of an extremity (e.g., a finger) is not accessible to individuals
  • However, the upper arm, the head, the eye, and the male gonads are considered to be major portions of the whole body 66

RG 8.40 ~ EDEX 67

RG 8.40 ~ EDEX

  • If the DDE is used as the EDEX, the EDEX can be overly conservative
  • Several methods of determining EDEX are acceptable to the NRC staff based on dose measurements on the surface of the whole body
  • Each of these methods generally involves the measurement of the DDE at one or more locations on the whole body.
  • The EDEX is then determined by applying a weighting factor to each dosimeter result 68

RG 8.40 - Compartment Factors

  • Method 1 - Using Compartment Factors

- The whole body is divided into 7 separate compartments 69

RG 8.40 - Compartment Factors

  • Method 1 - Using Compartment Factors

- Each compartment is monitored separately

- The (TLD/OSL) DDE dose measured for each compartment is then multiplied (weighted) by its compartment factor (see previous slide)

- The resulting weighted doses are then summed to determine the EDEX for the whole body 70

RG 8.40 - EPRI Method

  • Method 2 - Use Two Dosimeters - front and back

- a dosimeter worn on the front of the body (abdomen or thorax) is combined with a reading of a second dosimeter worn on the back of the body

- EDEX = 3/4 High dosimeter + 1/4 Low dosimeter 71

RG 8.40 - Other Methods

  • Method 3 - Medical scenarios wearing an apron during medical X-ray procedures

- Agreement States can develop their own EDEX calculational method, or

- One monitoring device worn on the neck outside the lead apron with calculational procedures

- Other specific methods as described in RG 8.40 72

RG 8.34 (Contd)

Prospective Dose Evaluations

  • 10 CFR 20.1502- licensees need to determine if the occupational dose is likely to exceed 10% of the dose limits
  • One method of determining likelihood is to do a prospective dose evaluation
  • What is a prospective dose evaluation?
  • Some licensees are unsure how to do a prospective dose evaluation 73

RG 8.34 (Contd)

Prospective Dose Evaluations

  • Section C.2 Determining the Need for Monitoring
  • 2.1 Establishing Categories of Workers for Consideration of the Need for Monitoring
  • Licensees should1 evaluate potential exposure scenarios to determine whether annual doses to individuals are likely to exceed monitoring criteria (i.e., by performing a prospective dose evaluation
  • 1 The term should denotes a recommendation and the term may denotes permission (neither a requirement nor a recommendation) 74

RG 8.34 Section C.2.2 Likely Exposures

  • Likely exposures do not include design-basis accidents
  • Precautionary monitoring for unlikely exposures is not required

RG 8.34 Section C.2.3 - Prospective Evaluation of Doses Not Likely to Exceed Monitoring Criteria

  • Potential exposure scenarios involving small doses:

- may be evaluated in the prospective dose evaluation

- may be determined as not likely to result in doses exceeding monitoring criteria

- therefore, are not subject to monitoring requirements 76

Examples of not likely to exceed monitoring criteria

  • If the prospective dose evaluation determined that:

- small, unplanned, unintended extremity exposures may occur but are not likely to exceed external monitoring thresholds, or

- that minor facial contamination or intakes may occur that are not likely to exceed internal monitoring thresholds, then

  • a follow up dose evaluation is not considered required monitoring 77

RG 8.34, Section C.2.5 Voluntary Monitoring and Reporting

  • The results of voluntary monitoring obtained when 10 CFR 20.1502 did not require monitoring are not subject to dose recording and reporting requirements 78

Section 2.6 Change in Exposure Conditions

  • If the radiation exposure conditions change during the year:
  • the need to provide individual monitoring should be reevaluated
  • if a new job assignment, a workers dose is likely to exceed 10 percent of the annual dose limit, then the licensee should provide monitoring
  • prior dose should be estimated, recorded, and reported 79

Section C.2.7 Detection Sensitivity

  • The monitoring criteria in 10 CFR 20.1502 are not required levels of detection sensitivity (e.g., the lower limit of detection)
  • For example, it may not be feasible to confirm intakes of 10 percent of the ALI
  • This is true particularly for bioassay measurements of some alpha-emitting radionuclides 80

Section C.3 Determination of External Doses

  • The DDE to the whole body is considered to be at a tissue depth of 1 centimeter (cm) (1,000 milligrams per square cm (mg/cm2)
  • The SDE to the skin or extremities is to be determined at 0.007 cm (7 mg/cm2)
  • The LDE is to be determined at 0.3 cm (300 mg/cm2)
  • In evaluating the SDE and LDE, it is acceptable to take credit for the shielding provided by gloves and protective lenses 81

RG 8.34 Section C.3.1 Placement of dosimeters

  • Licensees can use passive (TLD/OSL) or electronic dosimetry
  • If a portion of the whole body if receiving substantially more dose than rest of the whole body, then move dosimetry to highest exposed portion of the whole (substantial is a judgment call)
  • External dose not measured by dosimetry (e.g., from low energy gammas; e.g., Xe-133 at 81 keV) or radiation beams may be calculated per 10 CFR 20.1201(c)
  • If dosimetry was not moved and substantially higher dose was received to a part of the body, the DDE dose should be estimated by calculation 82

RG 8.34 Section C.3.2 ~ Multi-badging

  • Use of more than one dosimeter
  • Used for job-specific tasks to track dose to different parts of the body
  • Determine which dosimeter had the highest reading
  • Add that job-specific DDE to the whole-body DDE measured by a torso dosimeter (with subtraction) 83

RG 8.34 Section C.3.3 ~ Determining EDEX

  • When external exposure is measured with a dosimeter
  • Use DDE in place of EDEX, unless

- EDEX is calculated by a method approved by NRC

- RG 8.40 provides approved methods

- Licensees may apply for use of other weighting factors, see 10 CFR 20.1003, footnote 2 below 84

RG 8.34 Section C.3.4 ~ Determining SDE

  • When exposure is uniform, the SDE measured by a torso dosimeter is expected to be representative of the SDE
  • If SDE is expected to differ substantially from SDE measured by the torso dosimeter, the SDE should be monitored separately
  • Substantial is a judgment based on circumstances
  • If so, licensees should use an SDE dosimeter for some, but not all radiation exposures.
  • Extremity dosimeters may be worn under gloves
  • Note: Credit may be taken for protective equipment (gloves and safety glasses) 85

RG 8.34 Section C.3.4 ~ Determining SDE

  • For hot particles, SDE may be calculated using VARSKIN+1.0
  • VARSKIN+ 1.0 has added a wound model, eye dose model, and new alpha and neutron skin dose model
  • The HQ ARCB (HP) SharePoint site has a presentation on the chronology and overview of VARSKIN+ 1.0 at:

NRR Radiation Protection - Varskin + - All Documents (sharepoint.com)

RG 8.34 Section C.3.5 ~ Embryo/Fetus

  • RG 8.36 provides outdated guidance on calculating dose to the Embryo/Fetus
  • RG 8.36 relies mainly on the guidance provided by ICRP 30 (1992). However, this guidance is outdated
  • More up-to-date models are available at:

- ICRP Publication 73, Radiological Protection and Safety in Medicine (paragraphs 76 and 77), and

- ICRP Publication 75, General Principles for the Radiation Protection of Workers (paragraph 124)

- ICRP Publication 88 (2002), Doses to the Embryo and Fetus from Intakes of Radionuclides by the Mother 87

RG 8.34 Section C.3.6, EPA Federal Guidance Reports (FGR) 12 and 15

  • FGR-12 and FGR-15 are methods of dose assessment for external geometry from environmental contamination (airborne, soils, ground surface, water immersion)
  • The exposure geometry for environmental contamination is typically not representative of nuclear plant exposures
  • FGR-12 and FGR-15 methods should normally not be used in occupational dose assessments 88

Section C.3.6, EPA Federal Guidance Reports (FGR) 12 and 15

  • FGR-12 (1993 ) ~ Doses from External Environmental Contamination - EDEX, organ doses (not CDE), skin dose

- Note: By definition, CDE is from an internal source (not an external source) so FGR-12 calls it organ dose in lieu of committed dose

  • FGR-15 (2018) ~ EDEX from Environmental Contamination - organ doses from external radiation (not CDE from intakes), skin dose, and EDE from external radiation

- 6 age groups and 1,252 radionuclides

- Tissue weighting factors from ICRP-103 (not 10 CFR 20) 89

RG 8.34, Section C.3.7 Dosimetry Processing

  • 10 CFR 20.1501(d) - All personnel dosimeters (except for direct and indirect reading pocket ionization chambers ) that require processing must be processed and evaluated by an NVLAP accredited dosimetry processor
  • OGC and NMSS interpreted processing to mean a method, separate from and independent of the design of the dosimeter, that is required to extract dose information from the dosimeter
  • Alarming dosimeters and direct ion storage dosimeters do not require processing, so NVLAP accreditation is not required 90

RG 8.34 Section C.4 - Internal Dose Section C.4.1 - Assessing Intakes

  • Associated Regulatory Guides:

- RG 8.9 - determining intakes from bioassay results

- RG 8.22 - conducting bioassay programs at uranium mills

- RG 8.25 (not applicable to 10 CFR Part 50 licensees) - guidance on determining intakes from air sampling measurements

- RG 8.26 provides guidance on when bioassay programs are needed for those individuals subject to internal radiation exposure monitoring requirements

- RG 8.30 acceptable survey methods at uranium recovery facilities 91

Section C.4 Assessing Intakes Intakes can be determined two ways:

1. Air sampling (not a reliable method)
2. Whole body counting (WBC) (best method)

- WBC determines uptakes

- Uptakes need to be converted to Intakes

- NUREG/CR-4884 does that (see next slide) 92

NUREG/CR-4884 1988 93

NUREG/CR-4884 Retention of Elemental Cobalt 94

NUREG/CR-4884 Retention of Inhaled Co-60 95

NUREG/CR-4884 Retention of Inhaled Co-60 96

NUREG/CR-4884 - Co-60 fraction retained and excreted 97

RG 8.34 Section C.4.2 ~ CEDE from Inhalation

  • Dose from inhalation is the 50-year committed dose assigned to the year of intake
  • Licensees must use the 10 CFR 20 organ or tissue weighting factors 98

RG 8.34 Section C.4.2.1 FGR-11 (1988) 99

Comparison between 10 CFR 20 and FGR-11 for ALI and DACs 10 CFR 20

  • FGR-11 100

FGR-11 Description of Lung Models

  • FGR-11 provides a simplified description of the ICRP-2 lung model and ICRP-30 lung models 101

FGR-11 (pg 13)

Description of ICRP-2 lung model

  • ICRP-2 lung model for inhalation

- 25% is exhaled

- 25% is deep into lungs

- 50% is cleared from the throat into GI tract

- Considers material as either soluble or non-soluble 102

FGR-11 ICRP-30 Lung Model

  • A refined lung model

- Considers particle size AMAD (activity median aerodynamic diameter)

- AMAD is assumed to be 1 µm (diameter of a particle)

- Activity transferred to the GI tract is calculated based on linear differential equations

- Solubility (clearance from lung) is classified as days (D), weeks (W) or years (Y)

- ICRP-23 describes transit times through GI tract 103

RG 8.34 Section C.4.2.1 Use of FGR 11 to calculate CEDE

  • FGR-11 (~inhalation and ingestion)

- Uses the same organ weighting factors as 10 CFR 20

- Table 2.1 provides inhalation dose coefficients

- Table 2.2 provides ingestion dose coefficients

- Dose coefficients are expressed in Sv/Bq

- Convert to mrem/µCi by multiplying the listed Sv/Bq values by 3.7E9 104

FGR Inhalation CDE and CEDE values for Co-60

  • Table 2.1 (pg 124) - f1 is the fraction transfer from GI tract to blood

- Class W - f1 is 0.05 (5% transfer to blood)

- Class Y - f1 is 0.05 (5% transfer to blood) 105

Comparison between 10 CFR 20 and FGR-11 for inhalation dose

- Appendix B does not provide dose factors

- RG 1.109 uses ICRP-2 dose methods

  • FGR-11 106

RG 8.34 Section C.4.2.2 Use Stochastic ALIs to calculate CEDE

  • stochastic ALI (SALI)
  • nonstochastic ALI (NALI)
  • The more limiting ALI is listed first in App B

where 107

RG 8.45 Section C.4.2.3 Using DACs to calculate CEDE

  • Licensees perform air sampling to measure airborne concentrations and track exposure times (e.g., Rx Bldg entries)
  • Calculate CEDE (Hi,E) where 108

RG 8.34, Section C.4.2.4 Use of ICRP-30 to calculate CEDE

  • ICRP 30, Part 1, Supplement lists the CDE and the weighted CDE (the CDE multiplied by its weighting factor) 109

RG 8.34, Section C.4.2.5

~ Exceptions

- Physical properties (e.g., known AMAD values)

- Biochemical properties (chemical form)

- NRC approval is NOT necessary

- Caveats - must use:

RG 8.34, Section C.4.3 Calculating CEDE using 10 CFR 20, App B (Ingestion Doses)

RG 8.34, Section C.4.3.1 Use FGR-11, Table 2.2 Ingestion

  • Use of FGR-11, Table 2.2 (pg 157) - Ingestion
  • There is no Class D, W, or Y for ingestion
  • f1 is the fraction transfer from GI tract to the blood - f1 is 0.05 (5%) transferred
  • First (f1) value is for oxides, second (f1) value is organics 112

RG 8.34, Section C.4.3.2 Stochastic ALIs for CEDE Ingestion

  • Stochastic ALIs can be used to calculate CEDE =

113

RG 8.34, Section C.4.3.3 Use ICRP-30 for CEDE - Ingestion

  • ICRP-30, Part 1 supplement, lists the CDE and the weighted CDE per unit intake (Sv/Bq)
  • See next slide for Co-60 example 114

ICRP-30, Supplement to Part 1 Co-60 weighted CDE 115

ICRP publications

  • ICRP-2 (1959) ~ 5 (N-18) and 3 rem/quarter, MPCs, internal dose and use of critical organ concept
  • ICRP-26 (1977) ~ Recommendations

- Stochastic, non-stochastic, dose-equivalent, committed dose equivalent, ALIs, tissue weighting factors, planned special exposures, occupational dose of 5 rem/yr, public dose of 500 mrem/yr and ALARA concepts

  • ICRP-30 ~ Limits for intakes by workers

- New lung model (chapter 5), new GI Tract model (chapter 6), new bone dosimetry model, etc.

- New metabolic data to use in the models to calculate DACs and ALIs 116

RG 8.34, Section C.4.3.4 Ingestion CEDE using Individual or Material Specific Information

  • 10 CFR 20.1204(c) allows use of specific information on physical or biochemical properties
  • Individual or material specific information in not commonly available for nuclear power plants
  • Sometimes material specific information is used by fuel facilities 117

RG 8.34, Section C.4.4 CDE is calculated if CEDE > 1 rem

RG 8.34, Section C.4.4.1 Use FGR-11 to calculate CDE

  • FGR-11, Table 2.1 for inhalation
  • FGR-11, Table 2.2 for ingestion
  • CDE (rem) = µCi x Sv/Bq x 100 rem/Sv x 3.7E4 Bq/uCi 119

RG 8.34, Section C.4.4.2 Use of nonstochastic ALI to calculate CDE

  • Appendix B lists the nonstochastic ALIs

120

RG 8.34, Section C.4.4.3 Use of DACs to calculate CDE inhalation

121

RG 8.34, Section C.4.4.4 Use ICRP-30 to calculate CDE

  • ICRP-30 Supplement to Part 1 lists the CDE per Sv/Bq for both ingestion and inhalation 122

RG 8.34, Section C.4.4.5 Use Individual or Material Specific Values for CDE

  • Physical or biochemical properties may be taken into account 123

RG 8.34, Section C.4.5 Wound Dose

  • Decontamination and radiological assessment should not interfere with medical treatment
  • VARSKIN+ 1.0 has a wound model, neutron dosimetry and eye dosimetry models ) - see NUREG/CR-6918, Rev 4
  • Potential types of wound doses

- Tissue dose (volumetric)

- Skin dose

- Whole body dose from uptake

  • ALIs do not apply since wounds are not an ingestion or inhalation exposure 124

RG 8.34, Section C.4.5 NCRP-156 Wound Model

  • NCRP-156 ~ Wound Dosimetry & Assessment

- analytical methods and parameters may be used to assess dose if consistent with NRC regulations

- Chapter 5 provides dosimetry methods for skin dose and 10 mm (0.5 cm3) sphere

  • NCRP-156 recommends a default volume of 1 cm3 when exact volume is unknown
  • NCRP-156 also provides an extensive list of references 125

RG 8.34, Section C.4.5.1 Dose limits for wound intakes

  • 10 CFR 20 dose limits were not designed for wound dose exposures:

- TEDE = 5 rem

  • EDEX = 0 since external exposure
  • CEDE is approx. zero since unlikely to have soluble uptake from wounds

- TODE = 50 rem

  • DDE component = zero since DDE is external exposure
  • CDE component needs to be assessed
  • CDE component likely to be minimal as dose is averaged over the volume of the exposed organ 126

RG 8.34, Section C.4.5.2 Calculating Skin Dose from wounds

  • Assess dose to basal layer of skin at 7 mg/cm2 (70 µm) from embedded RAM
  • NCRP-130 provides estimated skin thickness for different parts of the body (e.g, calluses)
  • NCRP-156 describes external skin dosimetry models
  • NCRP-156, table 5.1 provides external skin dose coefficients to 1 cm2 surface area 127

RG 8.34, Section C.4.5.2 Calculating skin dose from wounds

  • Dose limits (Contd)

- SDE = 0, since SDE is an external exposure

- CDE to the skin is minimal since CDE is averaged over the volume of the skin

Conclusion:

- dose limits are not a major consideration in evaluation wound injuries

- However, dose assessments should be performed for medical purposes 128

RG 8.34, Section C.4.5.3 Calculating CDE to local tissues

  • Muscle tissue is most likely to be exposed
  • Muscle tissue has no real dose limit (due to dose averaging over volume of the tissue)
  • Dose assessment is primarily in support of medical diagnosis
  • Dose is primarily from beta and alpha (not gamma)
  • For comparison, NCRP-156, table 5.2 provides dose rates to a 10 mm (0.5 cm3) tissue sphere around a point source 129

RG 8.34, Section C.4.5.4 Calculating CEDE from wounds

  • Organ dose should be assessed if:

- wound source is near an organ (e.g., thyroid or lung)

- an uptake has occurred into an organ

- assessed as the 50-year committed dose

- dose is assessed over the entire organ or tissue per ICRP-26 130

RG 8.34, Section C.4.6 Dose from absorption through skin

  • Absorption of other RAM through the skin is negligible compared to intake through inhalation
  • Intakes through the skin should be considered when liquid solutions with RAM come into contact with the skin 131

Questions & Discussion

Retrieved from "https://kanterella.com/w/index.php?title=ML22256A224&oldid=3533514"