Internal Dosimetry Program - Entire Document (Very Large)
G-10 CFR 835/C1 - Rev. 1
NOVEMBER 1994
IMPLEMENTATION GUIDE
For Use With Title 10, Code of Federal Regulations, Part 835
OCCUPATIONAL RADIATION PROTECTION
INTERNAL DOSIMETRY PROGRAM
ASSISTANT SECRETARY for ENVIRONMENT, SAFETY and HEALTH
FINAL GUIDE - FOR UNLIMITED USE and DISTRIBUTION
U.S. Department of Energy IMPLEMENTATION GUIDE
G-10 CFR 835/C1 - Rev. 1
INTERNAL DOSIMETRY PROGRAM
CONTENTS Page
I. PURPOSE AND APPLICABILITY. . . . . . . . . . . . . . . 1
II. DEFINITIONS . . . . . . . . . . . . . . . . . . . . . . 2
III. DISCUSSION. . . . . . . . . . . . . . . . . . . . . . . 7
IV. IMPLEMENTATION GUIDANCE . . . . . . . . . . . . . . . . 10
A. Organization, Staffing, Training, and Facilities . . 11
1. Organization. . . . . . . . . . . . . . . . . . . 11
2. Staffing. . . . . . . . . . . . . . . . . . . . . 12
3. Training, Experience, and Continuing Education. . 12
4. Facilities and Resources. . . . . . . . . . . . . 13
B. Internal Dosimetry Technical Basis Document. . . . . 13
C. Procedures . . . . . . . . . . . . . . . . . . . . . 14
D. Design of the Bioassay Program . . . . . . . . . . . 14
1. General Guidance. . . . . . . . . . . . . . . . . 15
2. Investigation Level . . . . . . . . . . . . . . . 15
3. Derived Investigation Levels. . . . . . . . . . . 16
4. Factors Affecting the DIL . . . . . . . . . . . . 16
5. Methods of Measurement. . . . . . . . . . . . . . 16
6. Frequency of Bioassay Measurement . . . . . . . . 17
7. Supplementing Routine Bioassay Programs
(Where the DIL < the MDA). . . . . . . . . . . . 18
E. Participation in the Bioassay Program. . . . . . . . 19
1. Routine Bioassay Program. . . . . . . . . . . . . 19
2. Special Bioassay Program. . . . . . . . . . . . . 19
3. Exception to Routine Bioassay Requirement . . . . 20
4. Timely Receipt of Bioassay Results. . . . . . . . 20
F. Detection and Confirmation of Intakes. . . . . . . . 21
G. Internal Dose Evaluation . . . . . . . . . . . . . . 22
1. Guidance. . . . . . . . . . . . . . . . . . . . . 22
2. Interpretation of Bioassay Data . . . . . . . . . 23
3. Evaluation of Internal Dose from Bioassay Data. . 23
4. Periodic Reevaluation of Internal Dose. . . . . . 24
H. Internal Dose Management . . . . . . . . . . . . . . 24
1. Baseline Bioassay for New Employees or Workers
Initiating or Resuming Work with Radioactive
Materials . . . . . . . . . . . . . . . . . . . 25
2. Dose Limitation . . . . . . . . . . . . . . . . . 25
3. Lifetime Dose Control . . . . . . . . . . . . . . 26
4. Accidental Dose Control . . . . . . . . . . . . . 26
I. Recording Internal Doses and Related Information . . 26
1. Requirements. . . . . . . . . . . . . . . . . . . 26
2. Individual Information. . . . . . . . . . . . . . 28
3. Intake Records. . . . . . . . . . . . . . . . . . 28
4. Dose Evaluation Records . . . . . . . . . . . . . 28
J. Reporting Requirements . . . . . . . . . . . . . . . 29
K. Medical Response . . . . . . . . . . . . . . . . . . 30
L. Quality Assurance. . . . . . . . . . . . . . . . . . 30
1. General Requirements. . . . . . . . . . . . . . . 30
2. Independent Review. . . . . . . . . . . . . . . . 31
M. Guidance for Monitoring in the Workplace . . . . . . 31
1. Performance Requirements. . . . . . . . . . . . . 32
2. Allowance for Physical and Chemical Form. . . . . 32
3. Recourse for Technology Shortfall . . . . . . . . 33
V. REFERENCES. . . . . . . . . . . . . . . . . . . . . . . 33
VI. SUPPORTING DOCUMENTS. . . . . . . . . . . . . . . . . . 34
VII. APPENDIX 10 CFR 835, Implementation Guide, and DOE
Radiological Control Manual Cross-Reference. 36
Section I - Purpose and Applicability
I. PURPOSE AND APPLICABILITY
This Implementation Guide (IG) provides an acceptable methodology for
establishing and operating an internal dosimetry program that will
comply with U.S. Department of Energy (DOE) requirements specified in
Title 10 of the Code of Federal Regulations (CFR), Part 835,
"Occupational Radiation Protection" (DOE, 1993a); hereinafter referred to
as 10 CFR 835. For completeness, this IG also identifies applicable
requirements and recommendations contained in DOE Order 5480.11, as
amended, "Radiation Protection for Occupational Workers" (DOE, 1992a),
DOE's "Radiological Control Manual" (DOE, 1994), hereinafter referred to
as the RCM (with the associated numbers denoting the article numbers);
and secondary documents (American National Standards Institute (ANSI)
Standards, etc.) invoked by these and other primary documents. The
Appendix of this IG provides a cross reference of the applicable
material in 10 CFR 835, this IG, and the RCM.
This IG amplifies the regulatory requirements of 10 CFR 835, which are
enforceable under the provisions of Sections 223(c) and 234A of the
Atomic Energy Act of 1954, as amended (AEC, 1954). The requirements and
recommendations of the other DOE documents are enforceable through
contractual or administrative means. This IG also provides guidance for
the structure, function, and operations of an internal dosimetry
program. The criteria for internal dosimetry programs to serve
epidemiology, risk assessment, and litigation are not within the scope
of this IG.
Except for requirements mandated by a regulation, a contract, or by
administrative means, the provisions in this IG are DOE's views on
acceptable methods of program implementation and are not mandatory.
Conformance with this guide will, however, create an inference of
compliance with the related regulatory requirements. Alternate methods
that are demonstrated to provide an equivalent or better level of
protection are acceptable. Contractors are encouraged to go beyond the
minimum requirements and to pursue excellence in their programs.
The word "shall" is used in this IG to designate requirements from 10
CFR 835, DOE Orders, the RCM, and secondary documents invoked by them.
The requirements of 10 CFR 835 are mandatory except to the extent an
exemption has been granted pursuant to 10 CFR 820, "Procedural Rules for
DOE Nuclear Activities" (DOE, 1993b) and are identified by a bolded and
underlined "shall." Requirements taken from DOE Orders and the RCM are
mandatory to the extent they are invoked by a contract or through
administrative means.
Those facilities not subject to the requirements of 10 CFR 835 should
substitute the corresponding DOE 5480.11 requirements.
This IG is applicable to all DOE activities involving occupational
exposure to ionizing radiation of DOE employees and/or
DOE-contractor/subcontractor employees.
Section II - Definitions
II. DEFINITIONS
activity median aerodynamic diameter (AMAD): The diameter of a sphere
having a density of 1 g cm-3 and the same terminal settling velocity in
air as that of the aerosol particle whose activity is the median for the
entire aerosol.
administrative control level: A numerical dose constraint established
at a level below the regulatory limits to administratively control and
help reduce individual and collective radiation exposure.
air monitoring: Actions to detect and quantify airborne radiological
conditions by the collection of an air sample and the subsequent
analysis either in real-time or off- line laboratory analysis of the
amount and type of radioactive material present in the workplace
atmosphere.
alpha: The probability (not to be confused with an alpha particle)
of a Type I error or false positive. Also called the false positive
probability.
analyte: The material to be detected in a quantitative analysis.
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 either: 1) a given radionuclide in a year by the Reference Man (ICRP
Publication 23) that would result in a committed effective dose
equivalent of 5 rems (0.05 sievert); or 2) a committed dose equivalent
of 50 rems (0.5 sievert) to any individual organ or tissue. ALI values
for intake by ingestion and inhalation of selected radionuclides are
based on Table 1 of U.S. Environmental Protection Agency (EPA),
EPA-520/1-88-020, "Limiting Values of Radionuclide Intake and Air
Concentration and Dose Conversion Factors for Inhalation, Submersion,
and Ingestion," Federal Guidance Report No. 11 (EPA, 1988).
baseline bioassay: An appropriate bioassay measurement obtained from a
bioassay program participant prior to beginning or resuming work with
radioactive material.
beta: The probability (not to be confused with a beta particle) of
a Type II error or false negative. Also called the non-detection
probability.
bioassay: The determination of kinds, quantities, or concentrations,
and, in some cases, locations of radioactive material in the human body,
whether by direct (in-vivo) measurement or by analysis and evaluation of
radioactive materials excreted or removed from the human body
(in-vitro). Examples of bioassay are: (1) measurement of radionuclides
in urine, feces, hair, blood, or sputum; and (2) direct counting of the
whole body or portions of the body (chest, abdomen, neck, wound) using
external detectors.
biokinetic model: A series of often empirically determined mathematical
relationships formulated to describe the intake, deposition in
respiratory tract (if applicable), uptakes by the transfer compartment
from intake compartment(s), uptakes by tissues or organs from the
transfer compartment, translocation, retention, and elimination of a
radionuclide from the body.
censored data: Data that have been recorded as "less than" values
rather than the observed numerical values (whether positive, zero, or
negative).
committed dose equivalent (Ht,50) : The dose equivalent calculated to
be received by a tissue or organ over a 50-year period after the intake
of a radionuclide into the body. It does not include contributions from
radiation sources external to the body. Committed dose equivalent is
expressed in units of rem (or sievert).
committed effective dose equivalent (He,50): The sum of the committed
dose equivalents to various tissues or organs in the body (Ht,50), each
multiplied by the appropriate tissue weighting factor (wT)--that is,
He,50 = sum(wT*Ht,50). Committed effective dose equivalent is expressed in
units of rem (or sievert).
compartment: The smallest element in a biokinetic model for which a
mathematical representation of a retained quantity is given.
Compartments may be organs (e.g., lung, liver), tissues (e.g., bone
marrow), or systemic (e.g., the transfer compartment).
confirmed intake: An intake confirmed by follow-up bioassay, by
association with a known incident, or by investigation.
continuous air monitor (CAM): An instrument that continuously samples
and measures the levels of airborne radioactive materials on a
"real-time" basis and has alarm capabilities at preset levels.
cumulative total effective dose equivalent: The sum of the total
effective dose equivalents recorded for an individual for each year of
employment at a DOE or DOE contractor site or facility, effective
January 1, 1989.
decision level (DL): The value of a net observation (result) at or
above which a decision is made that a positive quantity of the analyte
is present. The DL depends on the acceptable probability (alpha) of
incorrectly concluding that there is analyte present (a Type I Error);
(alpha) is usually taken as 0.05.
declared pregnant worker: A woman who has voluntarily declared to her
employer, in writing, her pregnancy for the purpose of being subject to
the occupational exposure limits to the embryo/fetus as provided in 10
CFR 835.206. This declaration may be revoked, in writing, at any time
by the declared pregnant worker.
decorporation: Accelerated removal of radionuclides from the body,
usually by medical or dietary intervention such as chelation, blocking,
excision, lavage, diuresis, increased fluid intake, etc.
derived air concentration (DAC): For the radionuclides listed in
Appendix A of 10 CFR 835, the air concentration that equals the ALI
divided by the volume of air breathed by an average worker for a working
year of 2000 hours (assuming a breathing volume of 2400 m3). For the
radionuclides listed in Appendix C of 10 CFR 835, the air immersion DACs
were calculated for a continuous, non-shielded exposure via immersion in
a semi-infinite atmospheric cloud. The value is based upon the derived
air concentration found in Table 1 of EPA-520/1-88-020.
derived investigation level (DIL): A value of a bioassay or air
monitoring measurement that triggers an investigation.
direct (in vivo) bioassay: The assessment of radionuclides in the body
by detection of radiations emitted using external detector and analyzer
systems.
dose equivalent (H): The product of absorbed dose (D) in rad (or gray)
in tissue, a quality factor (Q), and other modifying factors (N). Dose
equivalent is expressed in units of rem (or sievert) (1 rem = 0.01
sievert).
effective dose equivalent (He): The summation of the products of the
dose equivalent received by specified tissues of the body (Ht) and the
appropriate weighting factor (wT)--that is, Ht = sum(wT*HT). It includes the
dose from radiation sources internal and/or external to the body. The
effective dose equivalent is expressed in units of rem (or sievert).
elimination: The biological removal of a radionuclide from the body by
excretion, perspiration, exhalation, secretion (e.g., breast milk),
exfoliation (sloughing of dead tissue), or excision.
embryo/fetus: A developing human organism from conception until birth.
Same as unborn child.
evaluation: The process of arriving at a value for intake or dose that
uses, among other inputs, measurement results.
excretion: The biological removal of a radionuclide from the body via
one or more excretion pathways: urine and feces.
exposure: The general condition of being subjected to ionizing
radiation, such as by exposure to ionizing radiation from external
sources or to ionizing radiation sources inside the body. In this
document, exposure does not refer to the radiological physics concept of
charge liberated per unit mass of air.
false negative: A Type II (beta) error, that is, concluding that analyte
is not present when in fact it is.
false positive: A Type I (alpha) error, that is, concluding that there is
analyte present when it is not.
gastrointestinal (GI) tract model: A mathematical representation of the
behavior of radionuclides in the contents of the human gastrointestinal
tract.
general employee: An individual who is either a DOE or DOE contractor
employee; an employee of a subcontractor to a DOE contractor; or a
visitor who performs work for or in conjunction with DOE or utilizes DOE
facilities.
group of radionuclides: Two or more radionuclides that are contained in
a matrix such that an individual could not have an intake of one without
simultaneously having an intake of all of the radionuclides in that
matrix.
indirect (in vitro) bioassay: The measurement or analysis of
radionuclides in excreta or other biological samples removed from the
body.
intake: The amount of radionuclide taken into the body by inhalation,
absorption through intact skin, injection, ingestion, or through wounds.
Depending on the radionuclide involved, intakes may be reported in mass
(e.g., micro-g, mg) or activity (e.g., micro-Ci, Bq) units.
intake compartment: One of four compartments from which systemic uptake
can occur: the respiratory tract; the GI tract; a wound; or intact skin.
intake retention fraction: The fraction of an intake present in the
systemic body or excreta at some time after intake.
intake route: A pathway by which radioactive material enters the body.
The main intake routes are inhalation, ingestion, absorption through the
skin, and entry through injection or a cut or wound in the skin.
investigation level (IL): The value of the committed effective dose
equivalent from an intake(s) of a radioactive material by a worker at or
above which, for regulatory purposes, is regarded as sufficiently
important to justify further investigation.
lifetime control level: An administrative value used to limit a
worker's lifetime occupational radiation dose. The lifetime control
level is equal to N times 1 rem (N times 0.01 sievert), where N is the
age of the worker in years.
lifetime occupational dose: The sum of all occupational total effective
dose equivalent values for each year since January 1, 1989, plus the sum
of all occupational external dose equivalent (or deep dose equivalent)
values and occupational internal effective dose equivalent values prior
to January 1, 1989.
member of the public: An individual who is not occupationally exposed
to radiation or radioactive material. An individual is not a "member of
the public" during any period in which the individual receives
occupational exposure.
minimum detectable amount/activity (MDA): The smallest amount/activity
of a radionuclide in a sample that will yield a result above the
decision level with a beta probability of non-detection (Type II error)
while accepting an alpha probability of erroneously detecting that
radionuclide in an appropriate blank sample (Type I error). The MDA is
computed using the same value of alpha as used for the DL. The MDA depends
on both alpha and beta. Measurement results are compared to the DL, not the
MDA; the MDA is used to determine whether a program has adequate
detection capability. The MDA will be greater than or equal to the DL.
minimum detectable (effective) dose: The minimum detectable committed
(effective) dose equivalent associated with a bioassay program.
Formerly called "missed dose."
minor: An individual less than 18 years of age.
nonstochastic effects: Effects due to radiation exposure for which the
severity varies with the dose and for which a threshold normally exists
(e.g., radiation-induced opacities within the lens of the eye). Also
called deterministic effects.
occupational exposure: An individual's exposure to ionizing radiation
(external and internal) as a result of that individual's work
assignment. Occupational exposure does not include planned special
exposures, exposure received as a medical patient, background radiation,
or voluntary participation in medical research programs.
quality factor: The principal modifying factor used to calculate the
dose equivalent from the absorbed dose; the absorbed dose (expressed in
rad or gray) is multiplied by the appropriate quality factor (Q). The
quality factors to be used for determining dose equivalent in rem are
provided in 10 CFR 835.
radiological worker: A general employee whose job assignment involves
operation of radiation producing devices or working with radioactive
materials, or who is likely to be routinely occupationally exposed above
0.1 rem (0.001 Sv) per year total effective dose equivalent.
Reference Man: A reference human model with the anatomical and
physiological characteristics defined in the International Commission on
Radiological Protection (ICRP) Publication 23, "Report of the Task Group
on Reference Man," ICRP 23 (ICRP, 1975). The ICRP Reference Man includes
parameters for males and females of various ages.
respiratory tract model: A mathematical representation of the behavior
of particles and gases in the human respiratory tract.
retained quantity: The amount of material which, after being taken into
the body by inhalation, ingestion, entry through an open wound, or
absorption through the skin, exists in the whole body, a compartment, an
organ, or a tissue at a specified time.
routine bioassay monitoring: Any bioassay measurement made on a
predetermined, periodic schedule, to establish a worker's internal
exposure status relative to previous periods of time.
shall: Within the context of this Guide, the word "shall" is used to
designate requirements from 10 CFR 835, DOE Orders, the RCM, and
secondary documents invoked by them.
should and may: Within the context of this Guide, the words "should"
and "may" are used to represent optional program recommendations and
allowable alternatives, respectively. Deviations generally require no
specific approval or justification; however, exceptions or deviations to
"should" provisions referenced directly from the RCM require specific
justification and approval in accordance with Article 113.3 of that
manual (i.e., RCM 113.3).
special bioassay monitoring: Any bioassay measurement that is not
required for routine bioassay, but that is required for confirmation of
a suspected intake of radionuclides, or is required for follow-up
evaluation of confirmed intakes.
special control level: An individualized exposure control level invoked
for individuals with a lifetime occupational dose exceeding the lifetime
control level, and for workers with special concerns.
state-of-the-art: The most advanced technology that is commercially
available and successfully field tested.
stochastic effects: Malignant and hereditary diseases for which the
probability of an effect occurring, rather than its severity, is
regarded as a function of dose without a threshold for radiation
protection purposes.
termination bioassay: A bioassay measurement performed for the purpose
of documenting the retention of radioactive materials in the body due to
occupational exposure either upon termination of employment or upon the
cessation of potential exposure to a specific nuclide.
total effective dose equivalent (TEDE): The sum of the effective dose
equivalent (for external exposures) and the committed effective dose
equivalent (for internal exposures). For purposes of compliance, deep
dose equivalent to the whole body may be used as effective dose
equivalent for external exposures. ( Note that the TEDE does not
include the committed effective dose equivalent contributions from
intakes in prior years.)
translocation: Movement within the body of a radioactive material, such
as from bone to kidney.
Type I error: Incorrectly concluding from a result that there is
analyte present; the probability (alpha) of a Type I error is usually taken
as 0.05. The decision level is determined on the basis of an acceptable
level of Type I errors.
Type II error: Incorrectly concluding from a result that there is no
analyte present; its probability (beta) is usually taken as 0.05.
weighting factor (wT): The fraction of the overall health risk,
resulting from uniform, whole-body irradiation, attributable to a
specific tissue (T). The dose equivalent to tissue, T, is multiplied by
the appropriate weighting factor and summed to obtain the effective dose
equivalent to the whole body. The weighting factors are provided in 10
CFR 835.
whole body: For the purposes of external exposure, head, trunk
(including male gonads), arms above and including the elbow, or legs
above and including the knee.
workplace monitoring: The measurement of radioactive material and/or
direct radiation levels in areas that could be routinely occupied by
workers.
wound compartment: The compartment in a biokinetic model whose retained
quantity is the amount of radioactive material in a wound that has not
moved to the transfer compartment.
year: The period of time beginning on or near January 1 used to
determine compliance with the provisions of 10 CFR 835. The starting
date of the year used to determine compliance may be changed provided
that the change is made at the beginning of the year and that no day is
omitted or duplicated in consecutive years.
Section III - Discussion
III. DISCUSSION
Internal dosimetry is the analysis and measurement of radionuclides in
humans or bioassay samples and the evaluation of intakes and doses from
those measurements. It involves evaluation of bioassay data, evaluation
of the intake, distribution, retention, and elimination of
radionuclides, and evaluation of various absorbed doses and dose
equivalent quantities. Internal dosimetry is inherently indirect in
nature. It is not possible to determine the exact organ absorbed dose,
dose equivalent, or effective dose equivalent in a living human being
resulting from an intake of radioactive materials. Internal dose is
usually a derived or inferred quantity, obtained by evaluation of
indirect measurements and computational models. This is particularly
true for alpha- and beta-emitting radionuclides in the body which have
low photon emission abundances. Direct measurements of internalized
photon-emitting radionuclides in organs also may be difficult because of
attenuation and scattering by overlying tissues.
10 CFR 835 contains requirements affecting internal dosimetry programs
throughout DOE and contractor facilities. The salient features are:
-- Internal doses to workers "shall" be evaluated as specified in 10
CFR 835.402 (c).
-- internal doses are added to external doses. The total effective
dose equivalent during a year (TEDE) "shall" be determined by summing
the effective dose equivalent from external exposures (or deep dose
equivalent) and the committed effective dose equivalent from
intakes during the year (10 CFR 835.203(a)). This is done for the
purposes of demonstrating compliance with occupational dose limits
and keeping worker doses as low as reasonably achievable (ALARA).
The quantity committed effective dose equivalent is the mechanism
by which internal doses are added to external doses on the basis of
equal risk.
-- exposures from background, therapeutic, and diagnostic medical
radiation, and voluntary participation in medical research programs
"shall" not be included in the assessment of compliance with the
occupational exposure limits or in dose records (10 CFR
835.202(c)).
10 CFR 835.202 requires that the occupational exposure of a general
employee to radiation or radioactive material resulting from routine DOE
activities "shall" not cause the following limits to be exceeded:
-- Stochastic effects. The annual limit for the TEDE received from
both internal and external sources is 5 rems (0.05 Sv); and
-- Nonstochastic effects. The annual limits are: (i) Lens of the
eye, dose equivalent of 15 rems (0.15 Sv); (ii) Extremity or skin,
shallow dose equivalent of 50 rems (0.5 Sv); and (iii) Any organ
or tissue, a total dose equivalent of 50 rems (0.5 Sv) (10 CFR
835.202(a) and RCM 213).
10 CFR 835.402(c) requires that monitoring of individual exposures to
internal radiation "shall" be performed in accordance with the following:
Internal dose evaluation programs (including routine bioassay programs)
"shall" be conducted for (1) radiological workers who, under typical
conditions, are likely to receive 0.1 rem (0.001 Sv) or more committed
effective dose equivalent and/or 5 rems (0.05 Sv) or more committed dose
equivalent to any organ or tissue, from all occupational radionuclide
intakes in the year; (2) declared pregnant workers likely to receive an
intake resulting in a dose equivalent to the embryo/fetus in excess of
0.05 rem (0.0005 Sv), (3) minors and members of the public who are
likely to receive, in one year, an intake resulting in a committed
effective dose equivalent in excess of 0.05 rem (0.0005 Sv) and (4)
internal dose evaluation programs "shall" be adequate to demonstrate
compliance with 10 CFR 835.202.
The RCM introduces:
-- Administrative Control Levels (RCM 211) that are below dose limits
of the RCM and 10 CFR 835;
-- the Lifetime Control Level (RCM 212) of N rems (N times 0.01 Sv),
where N is the individual's age in years; and
-- Special Control Levels for individualized exposure control (RCM
216).
The administrative control level values apply to TEDE, and the lifetime
control level applies to lifetime occupational dose.
For individuals with a lifetime occupational dose, in rems, exceeding
their age, in years, a Special Control Level of less than 1 rem (0.01
Sv) TEDE shall be established for the present and subsequent years of
the individual's employment (RCM 216). The Special Control Level should
allow the individual's lifetime occupational dose, in rems, to approach
their age, in years, as additional occupational exposure is received
over a period of time. Once established, the Special Control Level may
be discontinued after the criterion of lifetime occupational dose
exceeding an individual's age is no longer met (RCM 216.1).
Several limits besides those for general employees shall be considered
in the design of an internal dosimetry program. A dose equivalent limit
of 0.5 rem (0.005 Sv) from conception to birth is specified for the
embryo/fetus of a declared pregnant worker (10 CFR 835.206) and efforts
should be made to avoid exceeding 0.05 rem (0.0005 Sv) per month to the
declared pregnant worker (RCM 215 and RCM Table 2-1). Annual limits are
0.1 rem (0.001 Sv) TEDE for minors and students under age 18 (10 CFR
835.207 and RCM Table 2-1), members of the public entering a controlled
area (10 CFR 835.208 and RCM Table 2-1), and visitors (RCM Table 2-1 and
RCM 214).
Radiation protection programs for limiting internal exposures are based
on the DOE policy of controlling radioactive material at the source.
One key element in an effective program for minimizing internal
exposures is the control and minimization of contaminated equipment and
contaminated areas. 10 CFR 835.1002(c) requires that the design
objective "shall" be to avoid releases of airborne radioactive material to
the workplace atmosphere under normal conditions and, under any
situation, to control the inhalation of such materials ALARA. It is
nonetheless recognized that low-level, chronic occupational exposures to
some materials are difficult to avoid due to the types of material
handled or processed, their chemical or physical forms, and the nature
of operations, and that incidents may cause unplanned releases of
radioactive material. Either or both of these conditions necessitate an
internal dosimetry program at most DOE and DOE contractor facilities.
10 CFR 835 reflects many of the scientific recommendations of the
National Council on Radiation Protection and Measurements (NCRP) and the
International Commission on Radiological Protection (ICRP). 10 CFR 835
contains both primary and secondary limits for exposure of workers to
external and internal sources of ionizing radiation in the workplace.
The primary limits on internal exposures are expressed in terms of
committed effective dose equivalent. Secondary limits for radiation
protection include annual limits on intake (10 CFR 835.403(a)(1)) and
derived air concentrations (10 CFR 835, Appendices A and C). The ALI
and DAC values are similar to those recommended in ICRP Publication 30,
"Limits for Intakes of Radionuclides by Workers: Design and
Interpretation" (ICRP, 1979); ICRP Publication 48, "The Metabolism of
Plutonium and Related Elements" (ICRP, 1986); and ICRP Publication 54,
"Individual Monitoring for Intakes of Radionuclides by Workers: Design
and Interpretation" (ICRP, 1988). The values were chosen to provide a
safe working environment and to maintain exposures to within the primary
exposure limits.
Air monitoring is the primary method for demonstrating compliance with
workplace control limits. Bioassay measurements that indicate a failure
to control the workplace may indicate the need for improvements in
workplace air monitoring.
Internal dose evaluations, along with associated bioassay measurements,
are the primary methods for demonstrating compliance with dose limits
for protecting workers. 10 CFR 835 states that the estimation of
internal dose "shall" be based on bioassay data rather than air
concentration values unless bioassay data are unavailable, inadequate,
or internal dose estimates based on representative air concentration
values are demonstrated to be as or more accurate (10 CFR 835.209(c)
and RCM 521.2). This is discussed below under Section IV.D.7.
"Supplementing Routine Bioassay Program in Cases Where the DIL is Less
Than the MDA."
Both air monitoring and bioassay results "shall" be used by a facility's
radiation protection organization in managing worker exposures to
maintain them below the limits and as low as reasonably achievable
(ALARA; 10 CFR 835.2, 835.202, and 835.1001; RCM 111).
Section IV - Introduction
IV. IMPLEMENTATION GUIDANCE
Monitoring of individuals "shall" be performed to demonstrate compliance
with the requirements of 10 CFR 835 (10 CFR 835.401 (a)(1)). Internal
dose evaluation programs, including routine bioassay programs, "shall" be
conducted for certain radiological workers (10 CFR 835.402(c)(1)),
declared pregnant workers (10 CFR 835.402(c)(2)), and minors and members
of the public (10 CFR 835.402 (c)(3) and RCM 521). With few exceptions,
the estimation of internal dose "shall" be based on bioassay data rather
than air concentration values of radioactive material(s) (10 CFR
835.209(c) and RCM 521.2). The internal dose evaluation program "shall"
be adequate to demonstrate compliance with 10 CFR 835.202 (10 CFR
835.402(d) and RCM 722.1).
This section provides basic guidance for conducting internal dosimetry
programs for workers who have the potential for intakes of radioactive
materials. It includes guidance for design and implementation of the
bioassay program, and guidance for evaluating, recording, reporting, and
managing internal doses.
The essential elements of an acceptable internal dosimetry program are
as follow:
-- Adequate staff with appropriate technical training;
-- an Internal Dosimetry Technical Basis Document giving scientific
information and other rationale explaining each element of the
internal dosimetry programs to support dose evaluation methods (RCM
522.1);
-- written policies and procedures covering each step in the
activities used to determine worker internal dose;
-- defined criteria for identifying workers who need to participate in
the bioassay program;
-- appropriate bioassay measurement methods and frequencies;
-- adequate detection capability and quality of bioassay measurements;
-- defined criteria and actions for identifying individuals with
suspected intakes, based on workplace measurements and bioassay
measurements;
-- appropriate workplace monitoring programs, including air sampling;
-- appropriate action level guidelines;
-- methods for control, accountability, and safe handling of samples;
-- timely analysis of bioassay samples and measurements, transmission
of results, dose evaluation, and recommendations to operations
management;
-- appropriate dosimetric models and default parameters for evaluating
internal dose;
-- quality assurance program covering all steps in the activities that
determine worker internal dose;
-- defined program to report internal doses to workers, management,
and DOE;
-- historical records of bioassay measurement results and dose
evaluations; and
-- historical records of the program, and changes in the program over
time.
Section IV, Subsection A - Organization, Staffing, Training, and Facilities
A. Organization, Staffing, Training, and Facilities
Unless otherwise specified, the quantities used in the records required
by 10 CFR 835 "shall" be clearly indicated in special units of curie, rad,
or rem, including multiples and subdivisions of these units. The
System International (SI) units, becquerel (Bq), gray (Gy), and sievert
(Sv), are only provided parenthetically in 10 CFR 835 and this Guide for
reference with scientific standards (10 CFR 835.4).
1. Organization
The internal dosimetry program should be a function of the radiation
protection organization at each DOE and DOE contractor facility. The
manager of the radiation protection organization should have overall
responsibility for the internal dosimetry program. Each internal
dosimetry program should have a designated leader with demonstrated
expertise in internal dose evaluation.
When elements of the internal dosimetry program are performed by one or
more subcontractors, the radiation protection organization should ensure
that subcontractors meet all requirements pertaining to internal
dosimetry in 10 CFR 835, the RCM, applicable DOE Orders and performance
standards, and that subcontractors follow this IG and the internal
dosimetry technical basis document. A copy of all relevant subcontractor
procedures should be incorporated in the historical record files of the
DOE contractor.
Where one DOE contractor on a multiple-contractor site conducts the
internal dosimetry program, or parts thereof, letters of agreement
should detail the responsibilities, authority, and communication
requirements of the respective parties. A copy of this agreement should
be in the internal dosimetry technical basis document and the historical
file.
2. Staffing
The radiation protection organization management should ensure that the
internal dosimetry program is adequately staffed to carry out its
functions.
The analysis of workplace and bioassay measurement data and the
evaluation of internal dose involve complex evaluation and professional
judgment. Personnel with responsibility for internal dose evaluation
should have the necessary expertise and skill, based on appropriate
education and training in conjunction with practical experience, to
perform their assigned duties. It is important that internal dosimetry
specialists be capable of recognizing conditions warranting follow-up
bioassay and dose evaluation. Personnel should be familiar with the
relevant internal dosimetry literature and the recommendations of
national and international scientific organizations with regard to
internal dose evaluation.
3. Training, Experience, and Continuing Education
Management of the radiation protection organization should establish
minimum qualifications for those staff who evaluate internal doses. The
qualifications should include both experience and education
requirements. Educational background and formal training needed for
internal dosimetry programs are listed below. Members of the dose
evaluation staff should meet these requirements, or the staff should
have access to persons with the required background (perhaps through
interdepartmental agreements or contracted services). It is not
necessary for all personnel on the staff to have expertise in all of the
listed subject areas. The internal dosimetry organization should have
or have access to personnel who have education and formal training in
the following areas:
-- Advanced mathematical methods, such as calculus, differential
equations, and statistical analyses;
-- radiochemistry and radiometric methods and concepts;
-- computer technology and software used for dose evaluation;
-- anatomy and physiology of the human body and effects of ionizing
radiation on biological systems;
-- nuclear radiation physics;
-- radiochemical behavior of relevant radionuclides;
-- principles of radiation dosimetry including national and
international guidance;
-- operational health physics; and
-- technical writing.
New internal dose evaluators should undergo a period of apprenticeship
commensurate with their experience and education. In addition, other
radiation protection staff should be cross-trained in internal dose
evaluation to ensure adequate staffing during vacations, absences, and
vacancies.
The program should be supported by trained dosimetry technicians,
counting system operators, and radiochemistry staff, all of whom should
receive training commensurate with job requirements.
Management should establish continuing education requirements for all staff
performing internal dose evaluations. Retraining and/or continuing
education are essential for maintaining an adequate level of expertise and
familiarity with current concepts and requirements for internal dose
evaluation. The same subjects as listed above under the minimum
educational qualifications should be considered in establishing continuing
education requirements. Retraining and continuing education should include
changes in procedures, changes in systems or equipment, changes in federal
guidance and regulations, and significant operating events that occurred in
the facility or at other DOE or commercial facilities that are relevant to
internal dosimetry.
4. Facilities and Resources
Computational facilities and software tools used by internal dosimetry
personnel should be adequate for performing calculations required for
the evaluation of dose from radionuclides in the body.
A library of handbooks, reference materials, scientific publications,
and other resources pertaining to internal dosimetry should be readily
available.
Section IV, Subsection B - Internal Dosimetry Technical Basis Document
B. Internal Dosimetry Technical Basis Document
An internal dosimetry technical basis document (or other organized
collection of documents) shall be developed (RCM 522.1) and should give
the scientific and technical foundation for the internal dosimetry
program. The internal dosimetry technical basis document should provide
the approach to evaluating internal doses from bioassay data, and where
appropriate, from workplace monitoring data. It should describe: (1)
physical and chemical characteristics of radioactive materials
encountered in the workplace; (2) methods for calculating internal
doses; (3) methods for documenting calculations; (4) dose evaluation
quality assurance; (5) methods for evaluating dose equivalents from
specific radionuclides, mixtures of radionuclides, and materials of
differing chemical characteristics; (6) recording and reporting
practices for internal dosimetry; (7) selection of workers for
monitoring; and (8) establishment of the type and frequency of
measurements to be used (RCM 522.3).
The technical basis for evaluating dose from both routine and special
bioassay, and for evaluating data from personal air samplers and other
monitoring equipment should be included in the internal dosimetry
technical basis document. Biokinetic models, model parameters,
assumptions, and default parameters used in dosimetric modeling and
evaluation should be clearly identified. Statistical methods for
evaluating bioassay data, identifying bioassay results above
environmental background values, using appropriate blanks, and analyzing
trends should be described.
To preclude the "double counting" of intakes and resultant doses, the
methodology to account for the portion of a bioassay result that may be
due to one or more prior confirmed intakes should be described in the
internal dosimetry technical basis document. The derivation of decision
levels should also be documented in the internal dosimetry technical
basis document. Default trigger levels and preliminary actions to be
taken for exposures to the different radionuclides present at the
facility following suspected or confirmed intakes at various levels
should be described in the internal dosimetry technical basis document.
Additional topics to be addressed in the internal dosimetry technical
basis document are found throughout this IG.
The internal dosimetry technical basis document should be reviewed at
least once every two years to assure that the scientific bases are
current and updated, as necessary. The internal dosimetry technical
basis document should be a controlled document and retained as a
radiological protection program record with copies of all previous
revisions and changes retained for future program review. The
requirements for maintenance of the internal dosimetry technical basis
document should be specified, including responsibilities for authorship,
review, approval, and distribution.
Section IV, Subsection C - Procedures
C. Procedures
All elements of the internal dosimetry program should be specified in
written procedures. These procedures should be consistent with 10 CFR
835, the RCM, the relevant DOE Orders, this Guide, and the internal
dosimetry technical basis document. In summary, the methods and
requirements for measurement (bioassay) and evaluating and recording
internal dose should be specified. The procedures should specify
methods for consistent collection of workplace and personnel monitoring
data, its evaluation, documentation of results, and records maintenance.
The components of the internal dosimetry program and the organizational
structure to which it reports should be documented in procedures.
Responsibilities of line management and members of the dose evaluation
group should be described. Elements of the workplace and radiological
worker monitoring programs that are germane to internal dosimetry should
also be included. Guidelines for prompt follow-up of worker internal
exposures to radioactive materials should be carefully defined, and
appropriate follow-up response to intakes, including the medical
management of workers with excessive intakes, should be described.
The procedures should be reviewed at least once every two years and
updated as necessary. The requirements for maintenance of procedures
should be specified, including responsibilities for authorship, review,
approval, and distribution.
Section IV, Subsection D - Design of the Bioassay Program
D. Design of the Bioassay Program
The worker bioassay program should: (1) Provide for investigation of
suspected intakes; (2) provide data for evaluating internal dose; and
(3) provide results that are adequate to demonstrate compliance with the
radiation dose limits given in 10 CFR 835. The primary methods of
routine and special worker bioassay are in vivo counting (direct
bioassay) and in vitro excreta analyses (indirect bioassay).
1. General Guidance
Bioassay measurements should be of the appropriate type, frequency,
timeliness, and of sufficient accuracy, to demonstrate that dose limits
have not been exceeded, and that doses are maintained ALARA.
The Discussion Section (III) of this IG details the monitoring
requirements for individuals with potential internal exposures to
radiation as specified in 10 CFR 835.402(c).
RCM 521 contains detailed requirements for inclusion of individuals in a
bioassay program, including:
1. Personnel entering Radiological Buffer Areas who have the
potential to receive intakes resulting in a committed effective
dose equivalent of 0.1 rem (0.001 Sv) or more in a year;
2. declared pregnant workers likely to receive intakes resulting in a
dose equivalent to the embryo/fetus of 0.05 rem (0.0005 Sv) or more
during the gestation period; and
3. minors and students, visitors and members of the public likely to
receive intakes resulting in a committed effective dose equivalent
of 0.05 rem (0.0005 Sv) or more in a year.
RCM 521 also requires that personnel shall participate in follow-up
bioassay monitoring when their routine bioassay results indicate an
intake in the current year with a committed effective dose equivalent of
0.1 rem (0.001 Sv) or more, and that personnel whose routine duties may
involve exposure to surface or airborne contamination or to
radionuclides readily absorbed through the skin, such as tritium, should
be considered for participation in the bioassay program.
Further, RCM 521 states that personnel shall submit bioassay samples,
such as urine or fecal samples, and participate in bioassay monitoring,
such as whole body or lung counting, at the frequency required by the
bioassay program, and that personnel shall be notified promptly of
positive bioassay results and the results of dose assessment and
subsequent refinements. Dose evaluation results shall be provided to
the individual in units of rem or millirem.
2. Investigation Level
In this IG, DOE adopts an investigation level (IL) of 0.1 rem (0.001 Sv)
committed effective dose equivalent from intakes occurring in a year for
general employees. Each facility should evaluate the need for special
ILs for declared pregnant workers, minors, and visitors because their
dose limits are lower than the limit for general employees (see 10 CFR
835.402(c)(2) and (3) for dose levels requiring personnel monitoring).
Throughout this document, IL refers to the IL for the appropriate group
unless otherwise specified.
To ensure that all dose limitation and dose control requirements of 10
CFR 835 and the RCM are met, the internal dose evaluation program should
be capable of evaluating intakes of radioactive materials that occur in
a year and that deliver a committed effective dose equivalent at the IL.
3. Derived Investigation Levels
Derived investigation levels (DILs) are values of routine bioassay
results, such as organ or body contents, or excreta concentrations or
excretion rates, that indicate an intake resulting in a dose exceeding
an IL. Internal dosimetry programs should establish DILs for each
bioassay method applied for the analysis of all radionuclides to which
workers are likely to be exposed and document the derivation of such
DILs in the internal dosimetry technical basis document. The physical
and chemical characteristics of the radioactive material which may be
taken into the body should be taken into account in establishing DILs.
If an internal dosimetry program chooses to use Reference Man (ICRP
Publications 23 and 30) default parameters in conjunction with modeling
and assumptions recommended in ICRP Publications 30 and 54 in deriving a
DIL, these choices should be justified in the internal dosimetry
technical basis document.
4. Factors Affecting the DIL
Factors such as significant clearance of a radionuclide in less than a
year (e.g., tritium), the frequency of bioassay monitoring, and the
likelihood of multiple exposures during a year (or under chronic intake
conditions) should be considered in establishing a DIL. The DIL should
be established so that a committed effective dose equivalent of one IL
from all intakes in a year is likely to be detected by the monitoring
program, i.e., the minimum detectable dose should be less than one IL.
If a nonroutine or an unexpected intake of a radionuclide or group of
radionuclides occurs, the minimum detectable dose may be calculated
assuming a single intake that occurred on: the date of the intake, if
known; or the date that would result in the largest committed effective
dose equivalent. If intermittent or chronic intakes are expected, the
minimum detectable dose may be calculated assuming a chronic intake
during the sample period.
For nonroutine or unexpected intakes, the DIL for each independent
radionuclide or group of radionuclides should be based on the objective
that a committed effective dose equivalent of not more than one IL would
be missed in the year from intakes of that radionuclide or group.
If it is known or is likely that an individual has or could have intakes
during the year from different sources that could result in doses above
the IL, appropriately smaller DILs should be determined and the basis
for those DILs included in the internal dosimetry technical basis
document.
5. Methods of Measurement
The internal dosimetry program staff should determine the minimum
detectable amount/activity (MDA) for each bioassay method for each
radionuclide present in a facility to which workers are likely to be
exposed. In determining MDAs, the value of beta (non-detection
probability) should be chosen to be 5% or less. The value of alpha
(false positive probability) should be chosen considering the effect on
bioassay measurement time, the disruption and inconvenience of false
positive results, the costs of improved analytical technology, and
handling, analysis, and record-keeping costs associated with the
program. The MDAs should be documented in procedures and their
statistical bases given in the internal dosimetry technical basis
document. For the MDA to be valid, the false positive probability used
for setting the decision level for the bioassay method should be the
same as the one used in the calculation of the MDA.
Procedures should contain descriptions of the method(s) of bioassay
measurements (e.g., urinalysis, fecal analysis, or in vivo counting),
analytical methodology (e.g., chemical separation followed by alpha
counting), and measurement parameters (e.g., counting time or instrument
efficiency) to be used in each component of the bioassay program.
Several other factors affect the method of bioassay used and its
associated MDA. They include:
-- The possible need for improved detection capability to assess
worker dose during the special bioassay following an intake
requiring internal dose evaluation, due to diminishing amounts of
material in bioassay compartments as time goes on;
-- the need for improved precision and accuracy if residual retention
and excretion from prior intakes interferes with the detection of
additional intakes in subsequent years;
-- timeliness of results needed to manage workers and keep subsequent
intakes low enough to avoid exceeding dose limits;
-- convenience to the workers;
-- costs, including lost production time while workers are
participating in the bioassay program; and
-- the impact of the method of bioassay on the frequency of bioassay
measurements.
The method of bioassay, analytical methodology, and measurement
parameters should result in an MDA less than the corresponding DIL for
all radionuclides to which a worker might be exposed.
The methods of bioassay measurement, their MDAs, and accuracies should
be specified in the internal dosimetry technical basis document, along
with a rationale or justification for the methods chosen.
6. Frequency of Bioassay Measurement
The routine bioassay measurement frequency depends on the bioassay
measurement method and associated MDA. The frequency should be chosen
so that it is unlikely that intakes by a worker in a year will result in
doses exceeding one IL without detection. Other factors affect the
choice of routine bioassay measurement frequency. They include:
-- Expected frequency, duration, and magnitude of elevated airborne
radioactive material concentrations;
-- cost of bioassay measurements and the cost of lost production time
while workers are participating in the bioassay program;
-- magnitude of a worker's assessed internal dose resulting from prior
intakes;
-- convenience to the workers;
-- need to confirm an unexpected bioassay result at or above the DL
including time for sampling and analyses;
-- need to assess dose from chronic intakes; and
-- need to confirm the effectiveness of workplace air monitoring and
personal controls such as respiratory protection or limitation of
exposure time.
Recommendations for selecting the appropriate bioassay frequencies for
given work areas are available in Section 4 of National Council on
Radiation Protection and Measurements Report No. 87, "Use of Bioassay
Procedures for Assessment of Internal Radionuclide Deposition" (NCRP,
1987) and ICRP Publication 54. The frequency of the routine bioassay
program should be specified in procedures. Justification for the
bioassay monitoring frequencies should be specified in the internal
dosimetry technical basis document (RCM 522.3), along with an evaluation
of the largest internal dose (i.e., minimum detectable dose) from an
intake (acute or chronic) that could go undetected with the chosen
frequency.
7. Supplementing Routine Bioassay Programs (Where the DIL < the MDA)
It is recognized that DILs for reasonable and practical routine bioassay
programs may be significantly less than the achievable MDA for certain
radionuclides such as plutonium.
By definition, a technology shortfall exists when the bioassay program's
derived investigation level is less than the minimum detectable
amount/activity of the routine monitoring method (DIL less than MDA). A
technology shortfall occurs when a performance objective (expressed as a
DIL) cannot be achieved with current or state-of-the-art methods and
equipment.
10 CFR 835 states that the estimation of internal dose equivalent "shall"
be based on bioassay data rather than air concentration values. Air
concentration values may be used for this purpose only if bioassay data
are unavailable, inadequate, or result in a less accurate dose estimate
(10 CFR 835.209(c) and RCM 521.2).
In the case of a technology shortfall, the facility should:
-- Enhance workplace monitoring and the use of indicators (e.g.,
unexpected glove or surface contamination, increase in airborne
radioactive material contamination) to trigger early special
bioassay monitoring;
-- enhance personal contamination monitoring (e.g., clothing, skin,
nasal smears) to trigger special bioassay monitoring;
-- use the best practical state-of-the-art bioassay monitoring
methods;
-- implement enhanced design, operation, controls, and personnel
protection equipment and procedures to minimize internal exposures;
-- consider supplementary air monitoring; and
-- document and justify the planned supplementary approach in the
facility's internal dosimetry technical basis document in
accordance with Section IV.B of this IG.
When air monitoring data are used, each worker's stay times (in hours)
and the average concentration (in DACs) to which the worker is exposed
should be multiplied to yield exposures to airborne radioactive
materials in units of DAC-hours. Forty (40) DAC-hours corresponds to
0.1 rem (0.001 Sv) committed effective dose equivalent when the
stochastic DAC is used.
A technology shortfall is not sufficient cause for failing to place
workers on a minimum or best-available bioassay program.
Alternative approaches and assumptions used in dose calculations and the
level of intake or committed effective dose equivalent detection
achieved should be described and documented in the facility's internal
dosimetry technical basis document. If DAC-hour calculations are used
to assess exposures to airborne radioactive materials, any permitted
adjustment to such calculations to account for the use of respiratory
protection should be documented in the internal dosimetry technical
basis document.
Section IV, Subsection E - Participation in the Bioassay Program
E. Participation in the Bioassay Program
Workers may be selected to participate in either routine or special
bioassay programs. The routine bioassay program is generally used to
monitor workers to detect the occurrence of an intake of radioactive
materials, and special bioassay sampling is generally used to obtain
follow-up data from suspected or confirmed intakes.
1. Routine Bioassay Program
Routine bioassay monitoring "shall" be performed for radiological workers
who, under typical conditions, are likely to receive 0.1 rem (0.001 Sv)
committed effective dose equivalent, and/or 5 rems (0.05 Sv) or more
committed dose equivalent to any organ or tissue, from all occupational
intakes of radionuclides during a year. (10 CFR 835.402(c)(1) and RCM
522.1)
The routine bioassay program also should include baseline bioassay
measurements for workers (if appropriate based on work history) before
initiating a period of work assignment at a facility and a final
termination bioassay measurement upon termination of work at a facility.
Such measurements should be made before and after any potential for
exposure, respectively (RCM 522.2 and 522.4).
Workers should continue to participate in the bioassay program even when
the DIL is less than the MDA. While it may not be possible, in these
circumstances, to detect intakes resulting in one IL, it is important to
detect and evaluate larger intakes.
2. Special Bioassay Program
Workers shall participate in a special bioassay program to confirm or
rule out radionuclide intakes when routine bioassay program results are
unexpectedly above the appropriate DIL, or when workplace monitoring
program results, knowledge of facility operating conditions, or other
information indicate that it is likely that a worker may have had an
intake resulting in a dose in excess of an IL (RCM 522).
Special bioassay analyses shall be performed when any of the following
occur: (a) Facial or nasal contamination is detected that indicates a
potential for internal contamination; (b) airborne monitoring indicates
the potential for intakes leading to a committed effective dose
equivalent exceeding 0.1 rem (0.001 Sv); or (c) upon direction of the
radiation protection organization when an intake is suspected (RCM
522.5). Special bioassay analyses should also be performed when burns,
wounds, or punctures occur which may or may not result in intake of
radioactive material by tissue, or positive results from wound
monitoring are obtained indicating the presence of residual radioactive
contamination in the damaged tissue.
Reasons for suspecting an intake may include:
-- Detection of contamination on the head or neck, hands or forearms,
or inside of respirator;
-- detection of extensive or extended personal skin or personal
clothing contamination;
-- loss of containment;
-- failure of ventilation system or respiratory protection equipment;
or
-- elevated air sampling or contamination results in occupied areas.
Special bioassay sampling should be performed for workers following
exposure to radionuclides in air when the potential intake leads to a
dose that exceeds one IL during an incident or over a short period of
time, and for workers with confirmed intakes.
3. Exception to Routine Bioassay Requirement
A minimal internal dosimetry program should suffice if the probability
of a measurable intake of radioactive material at a facility is low or
negligible. The minimal program should consist of workplace monitoring
and periodic review of operations involving radioactive materials to
ensure that intake probability remains low. However, if an intake in a
year for any worker at a facility would result in a committed effective
dose equivalent greater than one IL projected from air monitoring
results, special bioassay and dose evaluation should be performed.
Facilities with the potential for internal exposures but with no routine
bioassay program of their own should have written contingency plans
detailing air monitoring result action level guidelines, bioassay sample
collection procedures, and arrangements with other qualified
organizations for in vivo counting, excreta measurements, and internal
dose assessment, as appropriate.
4. Timely Receipt of Bioassay Results
Results of bioassay measurements should be provided to dose evaluators
in a short enough time to provide an adequate degree of worker
protection. Consideration of timeliness should include the following:
-- The need to support decisions on implementing and/or continuing
medical intervention (RCM 522.6);
-- the need to support rapid reporting to the worker, management, and
DOE and subsequent follow-up for significant intakes;
-- the need to confirm a suspected intake based on a high routine
measurement before the detection capability is lost due to rapid
clearance from the bioassay compartment;
-- the need to support the ALARA program with timely information; and
-- 10 CFR 835.801(b) requires that records of exposures "shall" be
provided within 90 days of the termination of an employee, if so
requested by that employee.
The internal dosimetry program should establish with the bioassay
measurements laboratory an agreement of needed turnaround times, MDAs
for special and routine samples, and priorities for classification of
samples (e.g., routine, special, emergency). These arrangements should
be documented in procedures and the internal dosimetry technical basis
document.
Following suspected intakes, consideration should be given to performing
additional sampling while awaiting initial results to ensure an adequate
amount of data at early times after intake for dose evaluation purposes.
Additional sampling may include the evaluation of air sample media,
source terms, contamination surveys, respirator filters, nasal or mouth
swabs, irrigation fluids from personal decontamination, and wound
debris.
Section IV, Subsection F - Detection and Confirmation of Intakes
F. Detection and Confirmation of Intakes
The decision level (DL) should be set by considering the acceptable rate
of false positives, the cost and consequences of false positives, and
the dosimetric consequences of false negatives.
Bioassay results above the DL may be expected in the absence of a new
intake due to normal statistical fluctuations, non-occupational or
environmental sources, or prior confirmed intakes. The analytical
laboratory decision level should be based on a reagent blank. The
occupational intake decision level should be based on both the
analytical laboratory DL and considerations of expected levels of
activity in unexposed workers due to environmental exposures.
If a bioassay result above the DL is unexpectedly observed, follow-up
bioassay measurements should be promptly made to either confirm the
result as a true intake or identify it as a false positive result. An
intake should be considered confirmed when:
-- A bioassay result exceeding the DL is associated with a known
incident; or
-- a bioassay result exceeding the DL is shown not to be a false
positive by investigation or by appropriate statistical analysis of
follow-up measurements.
Investigations for the purpose of confirming an intake should consider
many factors, including evaluation of radionuclides detected versus
those expected (e.g., to rule out an unreported medical administration
of radioactive materials); evaluation of area survey results versus
radionuclide types and quantities detected; evaluation of skin or
clothing contamination contribution to bioassay indications; evaluation
of co-workers' bioassay results; and verification of results through
measurements of radionuclide transport within and out of the body.
In the absence of other confirming data, one acceptable decision rule
for confirming an intake is the observation that 2 out of the first 3
measurements in a bioassay series are above the DL. Follow-up bioassay
samples should be scheduled and obtained in response to an initial
positive bioassay result exceeding the DL.
If appropriate confirmatory follow-up measurements to an unexpected
bioassay measurement above the DL are not obtained, two options should
be considered depending on the magnitude of the bioassay measurements.
The first option is to simply presume an intake has occurred if the
committed effective dose equivalent (CEDE) from the intake is projected
to be less than one IL. This option minimizes costly and disruptive
investigations that would not be performed for comparable external
doses.
The second option is to perform an investigation if the CEDE from the
intake is projected to equal one IL or more. If an investigation is
performed, and fails to provide sufficient evidence to establish that an
intake did not occur, then an intake should be presumed to have
occurred. The basis for projecting a CEDE of one IL from bioassay
results should be documented in the internal dosimetry technical basis
document.
Both the need for promptness of the follow-up or confirmatory bioassay
measurement and the determination of the MDA used for the analysis
depend on a variety of factors. These include the clearance time of
the particular radionuclide, its chemical and physical form, the mode
of intake, the CEDE corresponding to the suspected intake, the
usefulness of the confirming measurement in assessing the internal
dose, the possibility of elevated bioassay results from
non-occupational sources (e.g., medical applications, diet, or radon
progeny), and the likelihood of the worker receiving additional
intakes between the first and second bioassay measurement. These
factors should be considered by the internal dosimetry staff in
determining the follow-up or confirmatory actions to be taken in
response to positive bioassay results.
These actions should be addressed in formal procedures. The internal
dosimetry technical basis document should contain the rationale for the
formal action procedures. The procedures should also address who will
establish confirmatory bioassay requirements in cases not covered by the
procedures.
Guidance given in Section IV.F of this IG should not be applied to
historical bioassay data prior to January 1, 1989, where follow-up
bioassay samples were not required on positive bioassay samples or where
documentation is lacking (counter efficiency, chemical recovery, minimum
detectable amount/activity, etc.). In these instances, the internal
dosimetry technical basis document shall describe the site policy for
confirming intakes.
Section IV, Subsection G - Internal Dose Evaluation
G. Internal Dose Evaluation
1. Guidance
Internal doses should be evaluated for all confirmed intakes, as defined
in Section IV.F of this guide. For intakes confirmed with bioassay
results below the DIL, no further investigation or follow-up bioassay
are indicated. For intakes confirmed with bioassay results above the
DIL, follow-up bioassay and investigation should be performed.
Bioassay data are the primary input for internal dose evaluations. The
extent of the investigation and the number and frequency of special
bioassay measurements following a suspected or confirmed intake should
be determined and documented on an individual, case-specific basis,
taking into account the potential magnitude of the intake, the effective
clearance half-time, the health of the worker, and the number of
measurements needed to evaluate the internal dose.
The schedule and frequency of long-term special bioassay measurements to
evaluate the CEDE to an individual who has had an intake resulting in a
dose in excess of one IL should depend on the expected magnitude of the
CEDE and the likelihood of the individual receiving additional intakes.
While the investigation should be tailored to the specific individual
and exposure circumstances, the trigger levels and preliminary actions
to be taken for exposures to the different radionuclides encountered at
the facility should be documented in the internal dosimetry technical
basis document and procedures,
Methods of evaluating the committed dose equivalent from internal
sources of radiation should be appropriate to the workplace conditions.
The methods should be consistent to the extent possible with EPA, NCRP,
and ICRP recommendations and DOE good practices.
2. Interpretation of Bioassay Data
Biokinetic models should be used to interpret bioassay data and assess
initial radionuclide intake. The particular biokinetic models used for
internal dose evaluation should relate well to the available bioassay
data and should account specifically (when possible and if known) for
the chemical and physical characteristics of the material taken into the
body. When the available data are lacking or are contradictory,
professional judgment will be needed to make a dose evaluation.
Since the evaluations of internal dose depend on knowing the intake
profile with respect to time, the dose evaluation staff should base the
time course of intake on known incidents, air monitoring data, records
of perturbations in facility operations, and/or discussions with the
worker(s) by radiation protection staff. If the time course of intake
cannot be plausibly established, then the procedure for evaluating doses
based on the internal dosimetry technical basis document should be used.
Evaluations of CEDE from a specific intake should account for expected
values of bioassay measurements from prior confirmed intakes.
3. Evaluation of Internal Dose from Bioassay Data
Internal dosimetry program staff should evaluate the CEDE from the
intake. The data necessary to calculate committed doses to tissues or
organs of concern should be maintained for possible future reevaluation.
Methods for evaluating the various doses from intakes should be
specified in the internal dosimetry technical basis document. The
methods should be based on recommendations given in ICRP Publications
30, 48, and 54, and other reports of the ICRP and NCRP which embody
improvements and updates of the science of internal dosimetry. Other
methods may be used provided they are documented and justified in the
procedures and/or internal dosimetry technical basis document.
In the calculation of internal doses less than one IL, default
parameters may be used. These parameters (e.g., intake date, deposition
probabilities, retention functions, organ masses, absorption fractions)
should be based on Appendix B of 10 CFR 835, the recommendations of the
ICRP, NCRP, or facility-specific factors as documented in the internal
dosimetry technical basis document. If the initial evaluation of an
intake indicates a worker dose in excess of 10 times an IL,
individual-specific and facility-specific factors should be used when
more appropriate parameters are expected to change the dose calculations
by a factor of 1.5 or more (ICRP, 1988, paragraph 74). The basis for
determining which individual-specific and facility-specific factors are
expected to change the dose calculations by a factor of 1.5 or more
should be documented in the internal dosimetry technical basis document.
Determination of individual retention patterns for a worker requires
participation in the special bioassay program and may require temporary
work restriction or reassignment to prevent subsequent intakes from
confounding the dose evaluation.
4. Periodic Reevaluation of Internal Dose
In the case of certain well-retained radionuclides (e.g., plutonium),
long-term follow-up and reevaluation of doses may be required. The
internal contribution to lifetime occupational dose should continue to
be reevaluated as further bioassay results and improved methods for
evaluating internal dose become available (RCM 212.2).
Evaluations for active workers with prior confirmed intakes should be
revised when information demonstrates a change in the currently
evaluated CEDE of 0.5 rem (0.005 Sv) or a factor of 1.5 of the
previously assigned dose for that intake, whichever is higher. In cases
where intakes are detected or confirmed in a year subsequent to the year
of the intake, the CEDE should be attributed to the known or assumed
year of the intake, and all records and reports for that year should be
amended as appropriate.
Section IV, Subsection H - Internal Dose Management
H. Internal Dose Management
DOE requires internal dose evaluation programs for evaluating internal
exposures to radionuclides and for maintaining adequate worker exposure
records.
Each site should have a plan that documents the dose management
practices. The plan should include procedures for managing workers with
retained radionuclides so that: (1) Monitoring is appropriate; (2)
additional exposures may be averted; (3) workers may receive adequate
medical care (including decorporation therapy), if necessary; (4)
internal doses can be appropriately evaluated and recorded; (5) total
dose (external and internal) may be assessed against appropriate annual
administrative control levels, dose limits, and lifetime control levels;
(6) workers are informed of the states of followup investigations and
dose evaluation; and (7) consideration is given to temporary work
restrictions to avoid exposures to radionuclides similar to those being
evaluated in the ongoing investigation.
Additional dose management criteria apply to the embryo/fetus of a
declared pregnant worker. Internal and external exposures to declared
pregnant workers "shall" be controlled to meet the following requirements:
-- 10 CFR 835.206(c) requires that if the dose equivalent to the
embryo/fetus is determined to have already exceeded 0.5 rem (0.005
Sv) by the time a worker declares her pregnancy, the declared
pregnant worker "shall" not be assigned to tasks where additional
occupational exposure is likely during the remaining gestation
period (also RCM 215); and
-- 10 CFR 835.206(b) also specifies that substantial variation above a
uniform exposure rate that would satisfy the limit of 0.5 rem
(0.005 Sv) from conception to birth for the embryo/fetus of a
declared pregnant worker "shall" be avoided. Article 215 of the RCM
states that efforts should be made to avoid exceeding 0.05 rem
(0.0005 Sv) per month to the declared pregnant worker.
1. Baseline Bioassay for New Employees or Workers Initiating or
Resuming Work with Radioactive Materials
Each new general employee should be evaluated for internally retained
radionuclides before the worker begins any work with radioactive
materials or resumes such work if he or she is likely to receive intakes
resulting in a committed effective dose equivalent greater than 0.1 rem
(0.001 Sv) (RCM 522.2). Similarly, students, minors, visitors, and
declared pregnant workers should receive a baseline bioassay before they
begin any work with radioactive materials or resume such work if they
are likely to receive intakes resulting in a committed effective dose
equivalent greater than one IL.
10 CFR 835.702(e) requires that efforts "shall" be made to obtain records
of prior years occupational internal and external exposure. Baseline
bioassay measurements should be requested if the worker has had previous
internal exposure. If a worker is going to work with radioactive
material for which the presence of naturally occurring radioactive
materials is detectable in bioassay measurements (e.g., uranium in
urine), baseline bioassay monitoring should be considered regardless of
prior occupational exposure.
If a worker has retained radioactive material from prior intakes, the
effect of those levels on the ability of the program to detect new
exposures must be assessed. Special monitoring procedures may be
required for such cases.
2. Dose Limitation
10 CFR 835.203 requires the combining of internal and external dose
equivalents from DOE activities. 10 CFR 835.203(a), requires that the
total effective dose equivalent "shall" be determined by summing the CEDE
from internally deposited radionuclides and the effective dose
equivalent (or deep dose equivalent) from external exposures. The sum
of the CEDE from intakes occuring during the year and the effective dose
equivalent from external sources received within the year "shall" be
compared to the 5 rems (0.05 Sv) annual limit (10 CFR 835.202(a)(1)).
In addition, 10 CFR 835.202(a)(2) limits occupational exposure to
general employees such that for any organ or tissue other than the lens
of the eye, the sum of the dose equivalent for external exposures and
the committed dose equivalent to that organ "shall" not exceed 50 rems
(0.5 Sv)(also RCM 213). 10 CFR 835.206 limits the dose equivalent to
the embryo/fetus from the period of conception to birth, as a result of
occupational exposure of a declared pregnant worker, to 0.5 rem (0.005
Sv)(also RCM 215). Any minor or member of the public exposed to
radiation or radioactive material during direct onsite access at a DOE
site or facility "shall" not exceed 0.1 rem (0.001 Sv) total effective
dose equivalent in a year (10 CFR 835.207 & .208 and RCM 213).
Committed dose equivalents and CEDEs should be calculated for intakes of
radioactive materials that take place during a year, and should not
include any contributions from intakes occuring in prior years. These
doses should be recorded and reported to the worker and management as
being assigned in the year of intake.
Determination of the effective dose equivalent "shall" be made using the
weighting factors provided in 10 CFR 835.2 (10 CFR 835.203 (b) and RCM
Appendix 2B).
3. Lifetime Dose Control
To administratively control a worker's lifetime occupational radiation
dose, a lifetime control level of N rems (N times 0.01 Sv) shall be
established where N is the age of the individual in years (RCM 212.1).
Special Control Levels (RCM 216) shall be established for individuals
who have doses exceeding N rems (N times 0.01 Sv). The lifetime
occupational dose is the sum of all TEDE values for each year since
January 1, 1989, plus the sum of external dose equivalent (or deep dose
equivalent) values recorded both prior to and after January 1, 1989.
Internal doses due to intakes prior to January 1, 1989, (i.e., prior to
DOE's adoption of the CEDE) do not need to be evaluated to determine
lifetime occupational dose (RCM 212.2).
For compliance purposes, the lifetime occupational dose is to be
compared to the Lifetime Control Level.
4. Accidental Dose Control
Action levels for administrative response to intakes of radionuclides by
workers should be detailed in the internal dosimetry technical basis
document and in internal dosimetry program procedures.
Section IV, Subsection I - Recording Internal Doses and Related Information
I. Recording Internal Doses and Related Information
1. Requirements
10 CFR 835 specifies the following internal dosimetry recordkeeping
requirements:
a. The results of individual external and internal dose monitoring
required by 10 CFR 835.402 "shall" be recorded (10 CFR 835.702(a) and
RCM 722.1).
b. The results of individual external and internal dose measurements
that are performed, but are not required by 10 CFR 835.402, "shall"
be recorded (10 CFR 835.702(b) and RCM 722.1).
c. The records required by 10 CFR 835.701 "shall" be sufficient to
evaluate compliance with 10 CFR 835.202; be sufficient to provide
dose information necessary to complete reports required by Subpart
I of 10 CFR 835 and by DOE Order 5000.3B, Occurence Reporting and
Processing of Operations Information (DOE, 1992b); and "shall"
include the following quantities for internal dose (10 CFR
835.702(c)(1), (2) & (4) and RCM 722.1 & 722.5):
(1) CEDE;
(2) committed dose equivalent to any organ or tissue of concern;
and
(3) estimated intake and identity of radionuclides.
d. The records required by 10 CFR 835.701 "shall" include the following
quantities for the summation of the external and internal dose (10
CFR 835.702(c)(5) and RCM 722.6, 722.7 & 722.9):
(1) TEDE in a year;
(2) for any organ or tissue assigned an internal dose during the
year, the sum of the dose equivalent from external exposure
and the committed dose equivalent to that organ or tissue; and
(3) cumulative TEDE received from external and internal sources
while employed at the site or facility, since January 1, 1989.
e. The records required by 10 CFR 835.701 "shall" also include the dose
equivalent to the embryo/fetus of a declared pregnant worker (10
CFR 835.702(c)(6) and RCM 722.8). The dose equivalent for the
embryo/fetus may be determined to be the summation of the deep dose
equivalent to the mother for external exposure and the dose
equivalent due to intakes of radioactive materials by the mother
which may be calculated using the methods described in NUREG/CR-
5631, "Contribution of Maternal Radionuclide Burdens to Prenatal
Radiation Doses" (Sikov, 1992).
f. For minors, students, and members of the public entering a
controlled area, records shall be kept, as applicable, of the total
effective dose equivalent (RCM 722.1).
g. Data necessary to allow at a later date the verification,
correction, or recalculation of recorded doses "shall" be generated
and recorded (10 CFR 835.702(g) and RCM 722.3). (Such data are
described in RCM 523 and in Section 4 of American National
Standards Institute Publication ANSI N13.6, "Practice for
Occupational Radiation Exposure Records System" (ANSI, 1989)).
h. 10 CFR 835.701(b) requires that records "shall" be retained until
final disposition is authorized by DOE unless otherwise specified
in 10 CFR 835 Subpart H.
i. Documentation of all occupational exposure received during the
current year "shall" be obtained when demonstrating compliance with
DOE occupational exposure limits. In the absence of formal records
of previous occupational exposure during the year, a written
estimate signed by the individual may be accepted (10 CFR
835.702(d)). Efforts "shall" be made to obtain records of prior
years occupational internal and external exposure (10 CFR
835.702(e) and RCM 213.2).
j. The records specified in 10 CFR 835 Subpart H that are identified
with a specific individual "shall" be readily available to that
individual (10 CFR 835.702(f) and RCM 712.4).
k. All records required by 10 CFR 835 Subpart H "shall" be transferred
to the DOE upon cessation of activities at the site that could
cause exposure to individuals (10 CFR 835.702(h) and RCM 774.1).
l. Results of surveys, measurements, and calculations used to
determine individual occupational exposure from external and
internal sources "shall" be documented and maintained (10 CFR
835.703(b)).
m. All required records shall be retained in accordance with DOE Order
1324.2A, "Records Disposition" (DOE, 1988)(RCM 712.3). When
dosimetry records are stored on easily corruptible media such as
magnetic discs or tape, a back-up system for data and computational
results should be available for recordkeeping.
n. Records shall be kept pursuant to DOE Order 5484.1, "Protection,
Safety, and Health Protection Information Reporting Systems" (DOE,
1987).
o. Records should be kept to document the appropriateness, quality,
and accuracy of monitoring methods, techniques, and procedures in
use during any given period pursuant to Section 6 of ANSI N13.6.
2. Individual Information
All records about individual workers should be identified by name and
Social Security Number or Passport Number and country. The following
personal identifiers should be retrievable along with individual
exposure data:
-- Full name and former names;
-- Social Security Number or Passport Number and country;
-- date of birth;
-- sex;
-- employment status;
-- occupation code (i.e., job title);
-- principal facility type and building number; and
-- organization code.
When the above personal identifiers change during the year, records
should be kept of the change and the date, where possible.
3. Intake Records
For each confirmed intake, the following information should be recorded:
-- Magnitude of intake in terms of activity or mass for each
radionuclide;
-- time course of intake including date(s) and time(s) and whether
known or assumed;
-- intake route (inhalation, ingestion, skin puncture, etc.) and
whether known or assumed;
-- radionuclides involved and their physical and chemical forms
whether known or assumed;
-- bioassay information pertinent to evaluation of the intake; and
-- methods and assumptions used for dose evaluation.
4. Dose Evaluation Records
All information that is necessary to review or recalculate each
evaluated dose should be recorded including uncensored bioassay data,
models, assumptions, parameters, and additional bioassay data as
appropriate. The names of the evaluator and reviewer and the outcome of
the review should be recorded.
Recording a bioassay result as "less than DL" rather than recording a
numerical value is called censoring data. No censoring of data should
be done, that is, actual numerical results should be recorded whether
negative, zero, positive below the DL, or positive at or above the DL.
The information contained in reports to individuals is discussed in
Section IV.J. of this IG.
Reevaluations of internal doses performed in accordance with this IG
(Section IV.G.4) should be documented such that a complete historical
record of preliminary and final CEDE estimates is retained.
Future refinements in radiation risk assessment and dosimetric modeling
may require reconsideration of the actual time course and organ
distribution of doses. The internal dosimetry records should therefore
include as much information as is available to reconstruct the organ or
tissue absorbed dose.
Section IV, Subsection J - Reporting Requirements
J. Reporting Requirements
The records specified in 10 CFR 835.702 that are identified with a
specific individual shall be readily available to that individual, if
the information is requested by that individual (10 CFR 835.702(f) &
.801(d) and RCM 712.4). On an annual basis, each individual monitored
in accordance with 10 CFR 835.402 "shall" be provided a radiation dose
report (10 CFR 835.801(c) and RCM 781.1).
10 CFR 835.801(a) specifies that exposure reports to individuals "shall"
include the data required under 10 CFR 835.702(c) and RCM 781.3). 10
CFR 835.702(c)(4) requires that the report include the following
quantities for internal dose resulting from intakes received during the
year:
-- Committed effective dose equivalent;
-- committed dose equivalent to any organ or tissue of concern; and
-- estimated intake and identity of radionuclides.
For situations in which there is no detectable internal dose or intake
of radionuclides, it is preferable to state in the report that there was
no internal dose component and that no radionuclides were detected as a
result of the internal dose monitoring program.
Individuals who terminate employment "shall" be provided a record of their
exposure if they so request (10 CFR 835.801(b)). The record of their
dose "shall" be provided to the terminating employees within 90 days of
termination of employment (also RCM 781.2). The termination report
shall include the TEDE for the year in which they terminate, the
cumulative TEDE, and the lifetime occupational dose (RCM 781.2 & 3). A
written estimate "shall" be provided at the time of termination if
requested (also RCM 781.2).
If an internal dose evaluation is still in progress at the 90-day limit,
the worker should be notified, and provided with an interim report and
later with the final dose record as soon as the evaluation is completed.
Other reporting requirements which are found in DOE Order 5484.1,
"Protection, Safety, and Health Protection Information Reporting System"
(DOE, 1987), DOE Order 5000.3B, and the RCM (performance indicator
program) are beyond the scope of this IG and therefore are not included.
Amended reports of CEDE by radionuclide or group of radionuclides for
intakes in prior years, if reevaluations are performed, should be
reported to the individual and entered in the individual's records as
well as reflected in the TEDE, cumulative TEDE, and lifetime
occupational dose reported under provisions of other Orders.
Section IV, Subsection K - Medical Response
K. Medical Response
Facilities with potential for intakes approaching dose limits should be
prepared to follow an action plan for medical response to any potential
or accidental intake of radioactive material. The plan should be
developed as a cooperative effort between medical and radiation
protection organizations and should include activation of key response
functions (internal dosimetry, analytical laboratory, in vivo counting,
medical assistance, etc.), training, and action levels for response.
The elements of this plan should include:
-- Action levels for medical response;
-- responsibilities of the affected worker, radiation protection
staff, internal dose evaluation staff, health physicist, medical
staff, and management;
-- guides for immediate medical care,
decontamination, monitoring, and long-term evaluation; and
-- provisions for periodically reviewing, updating, and rehearsing the
action plan.
Since there is no consensus on the decision levels for medical treatment
of workers, action levels should be established in the internal
dosimetry technical basis document based on decisions reached among
medical, management, and radiation protection staff (RCM 523.6).
Planning for such emergency actions should include the provision of
facilities and materials that will be required.
Section IV, Subsection L - Quality Assurance
L. Quality Assurance
1. General Requirements
The internal dose evaluation program "shall" be adequate to demonstrate
compliance with 10 CFR 835.202 (10 CFR 835.402(d)). Internal audits of
all functional elements of the radiation protection program "shall" be
conducted no less than every 3 years and "shall" include program content
and implementation (10 CFR 835.102 and RCM 134.1).
From the initial step (such as urine sample collection or reporting for
an in vivo count) through sample analysis and dose evaluation to
recording of the results, every step in an internal dosimetry program is
important in protecting workers and demonstrating compliance with 10 CFR
835, the RCM, and DOE Orders. All steps in the activities that control
or evaluate worker internal doses should be covered by written
procedures that provide appropriate quality control and quality
assurance. Quality assurance practices, such as having supervisors
ensure that bioassay samples are submitted on an appropriate frequency,
will enable corrective action when necessary. Quality control will
provide the needed documents and records for demonstrating compliance.
Computer software is normally used to perform internal dose evaluations.
Procedures for software quality assurance should be developed and
implemented to address:
-- Software documentation per ANSI/ANS 10.3, "Guidelines for the
Documentation of Digital Computer Programs" (ANSI, 1986);
-- validation and verification of models, data, assumptions, and
algorithms per ANSI/ANS 10.4, "Verification and Validation of
Scientific and Engineering Computer Programs for the Nuclear
Industry" (ANSI, 1987);
-- software security;
-- configuration management;
-- periodic testing to assure proper function; and
-- actions to be taken in the event that software errors are detected.
Hand calculations should be independently verified by a second qualified
internal dose evaluator. This review should be documented.
2. Independent Review
The internal dosimetry program should receive periodic assessment by the
site radiation protection organization to review technical basis
documentation, dose assessment procedures, instrumentation and
analytical methods, qualifications of personnel, quality assurance
program elements, and other elements of the program, as necessary to
ensure that the program maintains the capability to stay abreast of
scientific developments in internal dosimetry and provides a quality
radiation protection service to workers. External peer-review by
qualified individuals on a periodic basis is also recommended.
Appraisals of the internal dosimetry program should be included as part
of the contractor radiation protection program. These appraisals should
be conducted as often as necessary but no less frequently than every
three years (RCM 134).
Internal dosimetry program accreditation, when it becomes available,
should provide formal external review and testing of program
capabilities. Each site should work toward and plan for eventual
accreditation. Internal dosimetry program personnel should participate
in the conduct of intercomparison studies and should use the "DOE
Phantom Library" (RCM 522.9). Radiochemical laboratories and in-vivo
counting facilities whose measurements are used by internal dosimetry
programs are expected to have quality assurance programs, documented
regular equipment calibration programs, National Institute of Standards
and Technology traceable standards (RCM 522.8), and written procedures
that can be referenced by internal dosimetry programs.
Section IV, Subsection M - Guidance for Monitoring in the Workplace
M. Guidance for Monitoring in the Workplace
The objectives of the workplace monitoring program are to verify the
integrity of radioactive material containment, detect the release of
radioactive materials from some routine operations, detect inadvertent
releases of those materials in the workplace, evaluate and provide the
basis for modification to containment systems, and provide a basis for
design of bioassay programs.
1. Performance Requirements
10 CFR 835 specifies that area monitoring in the workplace "shall" be
routinely performed, as necessary, to identify and control potential
sources of personnel exposure to radiation and/or radioactive material
(10 CFR 835.401(b)). 10 CFR 835.403(a) specifies the uses of
measurement of radioactivity concentrations in the ambient air of the
workplace. Both air sampling and real-time air monitoring are
described. 10 CFR 835.209(a) states that the derived air concentration
"shall" be used in the control of occupational exposures to airborne
radioactive material. Appendix A to 10 CFR 835 states, for situations
where the particle size is known to differ significantly from 1 micro-m,
appropriate corrections can be made to both the estimated dose to
workers and the DACs.
For the purpose of workplace monitoring, air samplers positioned in the
breathing zone of workers may be used to complement fixed-station and
portable air samplers, as necessary to ensure that representative air
samples are obtained. Bioassay measurements may be used to help verify
the adequacy of the workplace air monitoring program, but should not
provide the primary basis for monitoring for loss of radionuclide
control in the workplace.
Real-time monitoring using continuous air monitors [CAMs] as defined in
10 CFR 835.2 "shall" be performed in normally occupied areas where an
individual is likely to be exposed to a concentration of radioactivity
in air exceeding one DAC as specified in Appendix A to 10 CFR 835, or
where there is a need to alert potentially exposed individuals to
unexpected increases in airborne radioactivity levels (10 CFR
835.403(a)(2) and RCM 555.3).
Air monitors shall be routinely calibrated and maintained on an
established frequency of at least once per year, and shall be capable of
measuring one DAC when averaged over 8 hours (8 DAC-hours) under
laboratory conditions (RCM 555.5). For the airborne radioactive
material that could be encountered, real-time air monitors "shall" have
alarm capability and sufficient sensitivity to alert potentially exposed
personnel that immediate action is necessary in order to minimize or
terminate inhalation exposures (10 CFR 835.403(a)(3) and RCM 555.6).
Should there be a need to use a higher alarm level in the actual
workplace, the need should be justified and documented. The
establishment of CAM alarm levels above 8 DAC-hours should be justified
and documented in the internal dosimetry technical basis document.
RCM 555 contains additional requirements for air monitoring.
2. Allowance for Physical and Chemical Form
The specific physical and chemical characteristics of the materials
potentially involved should be determined and taken into account in the
design of the monitoring program. These include radionuclide
composition, mode of intake, activity median aerodynamic diameter and
particle-size distribution, solubility and transportability from the
lung to other organs, and gastrointestinal absorption into the systemic
circulation. Accounting for physical and chemical characteristics may
necessitate a different set of secondary protection limits (ALIs and
DACs as specified in 10 CFR 835 Appendix A) applied to specific work
locations. The basis for revised secondary limits should be documented
in the internal dosimetry technical basis document.
3. Recourse for Technology Shortfall
The technology needed to perform workplace measurements for some
radioactive materials at levels indicative of a committed effective dose
equivalent of 0.1 rem (0.001 Sv) may not be available. If the
performance requirement cannot be achieved for this reason, the facility
should: (1) Continue to use the best practicable (state-of-the-art)
monitoring methods; (2) document the level of intake detection achieved;
and (3) implement enhanced design, operation, controls, and personnel
protection equipment and procedures to minimize internal exposures.
Section V - References
V. REFERENCES
(AEC, 1954) Atomic Energy Act of 1954, as amended. Public Law 83-703
(68 Stat. 919), Title 42 U.S.C. sec. 2011.
(ANSI, 1986) American National Standards Institute. 1986. "Guidelines
for the Documentation of Digital Computer Programs." ANSI/ANS 10.3-1986.
LaGrange Park, Illinois.
(ANSI, 1987) American National Standards Institute. 1987.
"Verification and Validation of Scientific and Engineering Computer
Programs for the Nuclear Industry." ANSI/ANS 10.4-1987. LaGrange Park,
Illinois.
(ANSI, 1989) American National Standards Institute. 1989. "Practice
for Occupational Radiation Exposure Records System." ANSI
N13.6-1966(R1989). New York, New York.
(DOE, 1987) U.S. Department of Energy. 1987. "Protection, Safety, and
Health Protection Information Reporting System." DOE Order 5484.1.
Washington, D.C.
(DOE, 1988) U.S. Department of Energy. 1988. "Records Disposition." DOE
Order 1324.2A. Washington, D.C.
(DOE, 1992a) U.S. Department of Energy. 1992. "Radiation Protection
for Occupational Workers." DOE Order 5480.11. Washington, D.C.
(DOE, 1992b) U.S. Department of Energy. 1992. "Occurrence Reporting and
Processing of Operations Information." DOE Order 5000.3B. Washington,
D.C.
(DOE, 1993a) U.S. Department of Energy. 1993. "Occupational Radiation
Protection." 10 CFR Part 835, 58 FR 65458, Federal Register, Vol. 58,
No. 236: December 14, 1993. Washington, D.C.
(DOE, 1993b) U.S. Department of Energy. 1993. "Procedural Rules for DOE
Nuclear Activities." 10 CFR 820, 58 FR 43680, Federal Register Vol. 58,
No. 157: August 17, 1993. Washington, D.C.
(DOE, 1994) U.S. Department of Energy. 1994. "Radiological Control
Manual." DOE/EH-0256T. Washington, D.C.
(EPA, 1988) U.S. Environmental Protection Agency. 1988. "Limiting
Values of Radionuclide Intake and Air Concentration and Dose Conversion
Factors for Inhalation, Submersion, and Ingestion." Federal Guidance
Report No. 11, EPA-520/1-88-020. Washington, D.C.
(ICRP, 1975) International Commission on Radiological Protection.
1975. "Report of the Task Group on Reference Man." ICRP Publication 23.
Pergamon Press. New York, New York.
(ICRP, 1979) International Commission on Radiological Protection.
1979. "Limits for Intakes of Radionuclides by Workers: Design and
Interpretation." ICRP Publication 30. Pergamon Press. New York, New
York.
(ICRP, 1986) International Commission on Radiological Protection.
1986. "The Metabolism of Plutonium and Related Elements." ICRP
Publication 48. Pergamon Press. New York, New York.
(ICRP, 1988) International Commission on Radiological Protection.
1988. "Individual Monitoring for Intakes of Radionuclides by Workers:
Design and Interpretation." ICRP Publication 54. Pergamon Press. New
York, New York.
(NCRP, 1987) National Council on Radiation Protection and
Measurements. 1987. "Use of Bioassay Procedures for Assessment of
Internal Radionuclide Deposition." NCRP Report No. 87. Bethesda,
Maryland.
(Sikov, 1992) Sikov, M. R., et al. 1992. "Contribution of Maternal
Radionuclide Burdens to Prenatal Radiation Doses." NUREG/CR-5631, Rev. 1.
U.S. Nuclear Regulatory Commission. Bethesda, MD.
Section VI - Supporting Documents
VI. SUPPORTING DOCUMENTS
American National Standards Institute. 1983. "Internal Dosimetry
Programs for Tritium Exposure--Minimum Requirements." ANSI N13.14-1983.
New York, New York.
Faust, L. G., et al. 1988. "Health Physics Manual of Good Practices
for Plutonium Facilities." PNL-6534. Pacific Northwest Laboratory.
Richland, Washington.
Johnson, J. R. and M. B. Carver. 1981. "A General Model for Use in
Internal Dosimetry." Health Physics, 40:341-348.
Johnson, J. R. and R. C. Myers. 1981. "Alkaline Earth Metabolism: A
Model Useful in Calculating Organ Burdens, Excretion Rates, and
Committed Effective Dose Equivalent Conversion Factors." Radiation
Protection Dosimetry, 1(2):87-95.
Lawrence, J. N. P. 1978. "A History of PUQFUA - Plutonium Body Burden
(Q) From Urine Assays. LA-7403-H." Los Alamos National Laboratory. Los
Alamos, New Mexico.
Leggett, R. W. and K. F. Eckerman. 1987. "Estimating Systemic Pu Burden
From Urinalyses." Health Physics, 52:337-346.
Lessard, E. T., et al. 1987. "Interpretation of Bioassay Measurements."
NUREG/CR-4884. Brookhaven National Laboratory. Upton, New York.
National Council on Radiation Protection and Measurements. 1985.
"General Concepts for the Dosimetry of Internally Deposited
Radionuclides." NCRP Report No. 84. Bethesda, Maryland.
Rich, B. L., et al. 1988. "Health Physics Manual of Good Practices for
Uranium Facilities." EGG-2530. EG&G Idaho, Inc. Idaho Falls, Idaho.
Watson, S. B. and M. R. Ford. 1980. "A User's Manual to the ICRP
Code--A Series of Computer Programs to Perform Dosimetric Calculations
for the ICRP Committee 2 Report." ORNL/TM-6980. Oak Ridge National
Laboratory. Oak Ridge, Tennessee.
Section VII - Appendix
Appendix
10 CFR 835, Implementation Guide, and DOE Radiological Control
Manual Cross-Reference
10 CFR 835 Implementation Guide Radiological Control Manual
--------------------------------------------------------------------------
835.2 II, III, IV.G & M Glossary
835.101 III 111, 124, 131, 133, 136 & 143
835.102 IV.L 134
835.202 III and IV.D, H, I, & L 211-216 and Table 2-1
835.203 III and IV.H Ch. 2 Part 1, 213
835.206 III and IV.H 215 and Table 2-1
835.207 III and IV.H 213
835.208 III and IV.H 213 and 214
835.209 III, and IV.D & IV.M 521, 522, 523 & 543
835.401 IV.M 551 and 555
835.402 III and IV.D 521, 551, 555 & 722
835.403 III and IV.M 555
835.701 IV.I 711
835.702 IV.I 213, 523, 712, 713,
722, 732 & 774
835.703 IV.I 712
835.801 IV.E & J 781
835.1001 III 111
835.1002 III 111
835 Appendix A II, III and IV.M. -
835 Appendix B IV.G. -
835 Appendix C II and III -
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